IR 05000327/1993049
| ML20059K483 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 11/02/1993 |
| From: | Blake J, Chou R, Lenahan J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20059K471 | List: |
| References | |
| 50-327-93-49, 50-328-93-49, NUDOCS 9311160075 | |
| Download: ML20059K483 (20) | |
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UNITED STATES
,p#nn aro%,
NUCLEAR REGULATORY COMMISSION
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-. -' t REGION 111
- 6L
%
101 MARIETTA STREET, N.W., sVITE1900
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- j ATLANTA, GEORGIA 30323-0199 k.....,/
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i Report Nos.:
50-327/93-49 and 50-328/93-49
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Licensee: Tennessee Valley Authority
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6N 38A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Docket Nos.:
50-327 and 50-328 License Nos.: DPR-77 and DPR-79
Facility Name: Sequoyah 1 and 2 Inspection Conducted: September 29 to October 4, 1993 Inspectors:
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J. J. Lenahan W
Date Signed
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/h/9 3 R. C. Chou Date Signed Accompanying Personnel:
R. W. Wright (October 3-4,1993)
J. J. Blake (October 3-4,1993)
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Approved by:
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J. J.
ahd, Chief D' ate Signed Mater alf and Processes Section Engi eering Branch Division of Reactor Safety SUMMARY
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Scope:
This special, announced inspection was conducted in the areas of operability.
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reviews of civil / structural items which will remain open after restart.
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Results:
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In the areas inspected, violations or deviations were not identified.
Weaknesses-were identified in preparation, review, and approval of procedures,
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(Paragraph 2); documentation of operability assessment and justifications for
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continued operation, (Paragraph 3.2); and in the failure to adequately verify vendor supplied information. (paragraph 3.2.12.)
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9311160075 931104 l
PDR ADDCK 05000327-
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REPORT DETAILS
1.0
. Persons Contacted Licensee Employees
- J. Bassaszewski, Licensing Engineer
- M. Burzynski, Site Engineering Manager
- R. Cutsinger, Lead Civil Engineer
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- R. Driscoll, Site Quality Manager
- R. Eythison, Vice-President, Nuclear Operations
- R. Fenech, Sequoyah Site Vice-President
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- K. flouse, Civil, Piping Analysis Supervisor
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D. Lundy, Technical Support Manager
S. Patel, Civil - Pipe Support Analysis Superviscr
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- K. Powers, Plant Manager
- W. Roberts, Civil Engineer, Component Qualification
- R. Shell, Site Licensing Engineer
- R. Thompson, Compliance Licensing Manager
- J. Ward, Engineering and Modification Manager Other licensee employees contacted during this inspection included engineers, techniciar.s, and administrative personnel.
l Other Organization
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- J. Lockaby, Lead Civil Engineer, Stone and Webster
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NRC Employees
- W. Holland, Senior Resident Inspector A. Long, Resident Inspector.
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S. Shaeffer, Resident Inspector
- P. Kellogg, Section Chief, Region II, Division of Reactor Project-
- Attended exit interview l
2.0 Review of Civil Structural Operability Criteria
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The inspectors reviewed TVA procedures which specify requirements for
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operability criteria. The procedures were as follows:
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Site Standard Practice SSP-3.4, Revision 8, Corrective
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Civil Engineering Instruction SQN-CI-90.02, Revision 0,
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Piping, Pipe supports and Equipment Operability Criteria for SQN 1 and 2
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Appendix H of SSP-3.4 provides guidance for determining and formally documenting operability of identified nonconforming'or degraded conditions. Appendix H also provides guidance for development and use of continued safe operation (CS0) determination, justification for.
continued operation (JCO) and requests for discretionary enforcement.
Paragraph 3.2.2 of Appendix H refers to NRC Generic Letter 91-18 for use in operability determinations. However discussions with licensee personnel disclosed that the licensee was confused regarding the requirements of GL 91-18.
Paragraph 3.3.1.B of Appendix H also references Generic Letter 87-02. However the inspectors questioned the meaning of this reference since Sequoyah has been excluded from the requirements of Generic Letter 87-02.
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The inspectors noted that Appendix H was not user friendly in that
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references were not clear and titles of some paragraphs / sections were
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confusing.
For example: Section 3.3.1 is titled " Operable (TS and Non-
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TS Equipment)";
Section 3.3.1.8 references Section 3.3.2 for criteria
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to be used to write a JC0, but the title of Section 3.3.2 is " Inoperable
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(non-TS Equipment)" and the thrust of the discussion in Section 3.3.2 is
toward the requirements for a CSO.
(The inspectors questioned whether I
the criteria for a JC0 is the same as for a CSO.) Also, it is not clear
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that the reference to Section 3.3.2 means Section 3.3.2 of Appendix H, or'could be referring to Section 3.3.2 of the main body of the precedure, or to another Appendix. Overall, the inspectors concluded.
that SSP 3.4 is very lengthy and confusing. Another observation by the inspectors is that the licensee's corrective action program has been
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changed frequently over the years. The name for Nonconformance Reports i
has also been frequently changed.
Instruction SQN-CI-90.02 establishes the civil design criteria to use when performing interim operability evaluations for equipment with nonconforming conditions. The criteria was based on TVA Civil.
i Engineering Branch Instruction CEB-CI-21.89. CEB-CI-21.89 was approved
by the NRC Office of Special Projects in a Safety Evaluation Report attached to an NRC letter dated February 23, 1988, Subject: Non-Nuclear Heatup for Sequoyah Unit 2 Prior to Restart (TAC R00253). The.
inspectors identified several editorial errors in Instruction SQN-CI-90.02.
These include references throughout the body of the report to various " Tables", when in fact the references should be to " Figures".
There are three separate lists of references.in the instruction, (See Paragraphs 6.9.8, 7.6, and 8.2.3.9).
The. numbering system for each list of references starts with number 1, therefore-there are three different reference number ones in the same procedure, three reference number twos, etc. These types of editorial errors cause confusion and could lead to errors when using the instructions.
The above examples are indicative of weaknesses in preparation, review
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and approval of procedures.
Violations or deviations were not identified.
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3.0 Review of Post-Restart Modifications 3.1 Background This special inspection was performed to assess issues which were previously identified by the licensee and which will remain open after
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restart of Unit 2.
The issues included open problem evaluation reports
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generic items, and other issues. The inspectors reviewed the licensee's
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justification for continued operation (JCO) and/or technical evaluation which the licensee performed for the open issues randomly selected by-the inspectors. Acceptance criteria utilized by the inspectors are those procedures listed in paragraph 2 above.
'I 3.2 Post-Restart Issues Review i
3.2.1 Leak in Fire Protection Piping i
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Issue A pin hole leak was identified in a portion of the high pressure fire
protection piping located in the Auxiliary Building. The cause of the
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leak was attributed to microbiological 1y induced corrosion (MIC). The licensee completed a temporary repair to the leak and scheduled completion of the permanent repair for the next refueling outage. The licensee completed a safety assessment and determined that the temporary
repair was acceptable for restart.
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Discussion The inspector reviewed work request C-223323 which documented the
problem, installation of the temporary repair, and the post maintenance
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testing. The leak was repaired by installing a sleeve made of a rubber gasket material over the leak. The rubber gasket material is held in
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place by hose clamps. After the repair was completed the pipe was pressurized to normal operating pressure and inspected for leaks.
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order to determine the extent of the corrosion problem, pipe thickness measurements were made using ultrasonic testing (UT) techniques. The j
inspector reviewed the UT pipe thickness measurements made in the area i
where the pin hole leak occurred. These meammnents showed the leak
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was a pin hole, and that no wall thinning trends exist in the area of the leak.
The licensee's safety assessment indicates that the UT measurement results were reviewed and found acceptable by Nuclear Engineering.
However, the acceptance criteria and basis for accepting the UT measurements were not documented in the safety assessment package. The licensee provided additional information to the inspector, (after the inspector requested it.) which documented the technical justification for the licensee's position. The additional information included a reference to Calculation number MIN WALL-UlV2-050193-CWF B87 930901001, which developed the minimum wall thickness for the fire protection
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system piping. A summary of the minimum wall thicknesses developed in
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this calculation showed that the values of wall thickness measured by UT exceeded the calculated minimum wall thickness and the licensee's acceptance criteria.
Conclusion The repairs to the fire protection pipe are acceptable for restart. The
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pressure boundary integrity of the fire protection pipe was not degraded by the selected repair method. The licensee has a long term project to replace portions of the fire protection system affected by corrosion. A similar evaluation will be performed if other pin hole leaks develop in the fire protection system after restart.
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A weakne>.s was identified regarding the lack of documentation in the licensee's safety assessment.
3.2.2 Main Steam line Operability Evaluation Issue This non-safety related portion of the Main Steam piping system for Unit 2 was reanalyzed for a steam hammer ever.t due to an employee concern regarding several inconsistencies in the record analysis contained in Report CEB-76-10 and the modifications of DCN M1977A which were not implemented during the last refueling outage. The piping consists of two 36-inch diameter pipes running from the flued head anchors in the East Steam Valve Room and two others running from the flued head anchors in the West Steam Valve Room to a common manifold in the Turbine Building. The steam hammer event results from fast closure of the Main Turbine Stop Valves. When a turbine trip occurs the stop valves will
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immediately " slam" closed. The " slamming" of the stop valves cause a
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shock wave to be propagated throughout the Main Steam System.
Discussion The inspectors discussed the problems with the licensee's engineers and reviewed the pipe stress calculation N2-1-20T, Rev.1, and the pipe l
support calculations.
The licensee reanalyzed the system based on the current condition I
without the two supports which are to be added by DCN M1977A (which was not implemented). The TPIPE computer program ran four cases as shown below:
Case 1A:
Run made without thermal stops and all current supports included (the current downward load supports only were included).
Case IB:
Run made with thermal stops and all current supports included.
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Case 2A:
Run made without thermal stops and supports at 12
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Case 2B:
Run made with thermal stops and supports at 12 nodes with downward load supports only, t
The piping was qualified for the long term design requirements without
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any modification. The pipe support design group reviewed the pipe
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suphort calculations using the maximum design loads from the above four i
cases. The supports were categorized into three groups which were as l
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Qualified Supports:
the current support based on the design drawings were qualified for the long term design requirements.
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Modification Supports:
the supports required modification
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to meet the long term design requirements and the l
modifications had been completed.
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Operability Supports:
the supports required modifications _
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to meet the long term design requirements, but supports had
been evaluated to meet the " operability" requirements
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without modifications before restart.
l The inspectors reviewed the Qualified Supports and Operability Supports which are the critical supports for restart. The support calculations
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reviewed are listed below:
j Calculations Rev.
Support Qualification Discrepancies /
Number Number Number Comments / Licensee
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Action 47A40001001
47A400-1-1 Long Term Note (1)
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47A40001002
47A400-1-2 Long Term
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47A400010ll
47A400-1-Il Support deleted 47A400010ll
47A400-1-12 Support deleted
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2MSH0505
2MSH-505 Long Term
2MSH0507
2-H1-507 Long Term
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2MSH0508
2-H1-508 Long Term:
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2-MSH-510 Long Term Note (2)
2MSH008)
N/A 2-MSH-081 Short Term-
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2MSH0082 N/A 2-MSH-082 Short Term i
2MSH0502
2-MSH-502 Short Term Note (3)
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2MSH506
'2 2-MSH-506 Short Term Note (4)
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2MSH509
2-MSH-509 Short Term Note (5)
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Notes l
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Stress allowables for the stanchion and weld were reduced based on the high line temperature. The previous uplift modification was not required and was not deleted from the calculation.
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(2)
U-bolt stress allowables should be reduced based on the high line i
temperature. The existing snubbers were Pre-NF which have a 35 kip capacity. The calculations used 50 kips capacity to qualify
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the actual load of 32.2 kips.
The configurations of base plates in the calculation for Section D-D and E-E were different from the drawings. The licensee i
revised the calculation and issued a DCN to revise the drawings.
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(3)
The computer model for the supporting frame was doubtful because it used concrete as a support for tension and compression even i
though the frame had no anchorage to the concrete. The actual load i
for the snubber was 34 percent more than the snubber allowable l
permitted in the short term operability criteria. The licensee i
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modified this support during the inspection.
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(4)
Calculation did not have qualifications for members 2 and 3.
Dimension of 42" instead of 52" should be used to check the concrete pullout capacity for the four anchor bolts considered as a block. 0.42 Su (ultimate stress) stress allowables instead of l
0.7 S should be used for checking the ratio of interaction for i
u the weld for operability evaluation.
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The existing snubber was verified to be a pre-NF model with a 35
kips capacity. The applied load of 78.4 kips resulted in an over
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capacity of 124 percent. The calculation used the snubber capacity i
of 50 kips without verification in field. The calculation also
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did not use the four anchor bolts as a block to calculate.the
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concrete pullout capacity due to the anchor bolt spacing
violation.
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The licensee modified the support during this inspection after the l
inspectors questioned the short term operability evaluation.
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long Term - The support meets the long term design criteria
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(FSAR).
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Short Term - The support meets the short term operability criteria.
The above calculations were reviewed for completeness, accuracy, j
adherence to design criteria and procedural requirements, acceptability of calculation methods with American Institute of Steel Construction I
(AISC) code criteria, and good engineering practices. The design criteria used for reviewing calculations are:
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Design Criteria No. SQN-DC-V-24.2, Supports for Rigorously
and Alternately Analyzed Category I Piping, Rev. 4 l
Civil Engineering Instruction SQN-CI-90.02 i
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During the calculation review, the inspectors found that Calculation
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Nos. 2MSH0502 and 0509 for Support Nos. 2-MSH-502 and 509 did not meet
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the operability criteria for short term operation. The main problems
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The licensee's engineers did not verify the actual snubber
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model in the field to distinguish the Pre-NF and NF models.
t The capacity for a Pre-NF snubber, model PSA 35, is 35 kips
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versus the 50 kips capacity for an NF snubber, model PSA 35.
Licensee engineers just assumed the higher capacity of 50
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kips in the calculations until the inspectors raised
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questions about the actual capacity of the snubbers in the field.
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The licensee engineers used the capacities for the snubbers in the calculations which are higher than published in the vendor's catalogs.
The licensee immediately implemented modifications for Support Nos. 2-MSH-502 and 509 during the inspection. All the other calculations were
determined to be acceptable except for the discrepancies noted. The
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licensee agreed to revise calculations and drawings for the
discrepancies noted above, walk down the system to generate as-built
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drawings for post-restart work, and revise calculations if necessary since this is a high energy line. The item will be identified as
Inspection Followup Item 327,328/93-49-01, Post-Restart Walkdown and Evaluation of Main Steam Line.
Conclusion l
The non-safety related Main Steam Line Piping and supports will be I
acceptable for restart after the completion of modifications for Support
Nos. 2-MSH-502 and 509.
3.2.3 Flued Head Pipe Anchor Loading, Penetrations X48A & X48B (PER SQ 930564)
Issue
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Problem Evaluation Report (PER) SQ 930564 was written because of a modeling problem for penetrations X48A and X48B in the stress calculation. Accurate modeling of flued head piping penetrations i
requires that the anchor point be loca+ed at the interface of the
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penetration sleeve and the steel containment vessel _ (SCV) structure.
The subject 12" diameter pipe with 16" diameter penetration sleeve was modeled as an in-line anchor located on the 12" diameter pipe at the imaginary projection of the SCV shell without consideration of the 16"
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diameter sleeve. This resulted in a misrepresentation of the
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configuration and stiffness on both sides of the penetration which was unconservative. Approximately 5 feet of piping was omitted from-the inside piping model and approximately 2.5 feet of piping was modeled as the wrong diameter for the outside model. This modeling did not meet the requirements of Section 5.4, SQN-RAH-214, Rigorous Analysis Handbook.
Discussion The inspector partially reviewed Stress Calculation 0600154-01-01 Rev. 2 for this line and discussed the problem with the licensee's engineers.
This calculation also incorporated vender supplied weight and center of gravity (CG) information for valves 2-FCV-72-2 and -39 per DCN-M08710-A.
i All the pipe stresses from the new analysis were within the design
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allowable stresses. Revised accelerations for valve 2-FCV-72-39, -2 and 2-72-502 were accepted by the Component Qualification Group. All the pipe supports were evaluated base on the new support loads and determined to be acceptable and within the operability allowable loads, with most meeting the long term design criteria.
Conclusion The stress calculation was acceptable for restart.
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3.2.4 Containment Vessel Loading, Penetratioa X48A and X48B (PER SQ 930565)
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PER SQ 930565 was written to report that an improper allowable stress intensity value, S,,,, for the Steel Containment Vessel (SCV) was used to
qualify the penetration (or support) X48.A and X48B.
This violated the i
design requirement of ASME Boiler and Pressure Vessel Code,Section III, Subsection B.
The correct S-value (or stress allowable) for the material, SA-516 Gr. 60, for the class B Nuclear Containment Vessel, is
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15 ksi. This should be used as allowable stress S,,,;
instead, the S,,,-
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value used in the original calculation was 19.38 ksi.
Discussion l
The stress calculation 0600154-01-01, Rev. 2 was a rerun of the stress calculation based on the correct model'due to PER SQ 9309564 and the correct valve weights and operator moment arm due to DCN-M08710-A as stated in the above item.
The support loads for Penetrations X48A and
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X48B from the revised stress calculations were found to increase. The
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loads provided by the stress calculations are in radial and tangential i
directions of the spherical shell. The penetrations were qualified by four load combinations and used the following stress allowables for SCV:
Membrane - S 15000 psi - 15 ksi
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- 35,,,= 45000 psi - 45 ksi The inspector reviewed support (penetration) calculation SCG2S593039, Rev. O and discussed the problem with the licensee's engineers. The results showed that the design margin for X48A and X48B are 1.06 and 1.03 respectively by using the faulted loads compared with the normal allowables which is the conservative approach. Therefore, penetrations X48A and X48B were qualified to the long term design criteria and were acceptable.
Conclusion Penetrations X48A and X48B were qualified to the long term design criteria and were determined to be acceptable for restart.
3.2.5 Hydrogen Collection Line and Conduit 2CR397 Attached to Interior of SCV (PER SQ 930292)
Issue These two items were part of DCN M09913A. The 12" diameter Hydrogen Collection Line is located inside the Steel Containment Vessel (SCV)
approximately at azimuth 250*-30' at radius of 56'-11".
The line is a duct, but is designed and analyzed as piping.
It runs between elevation 847'-11%" and 725'-10 /3,", and is entirely supported by the SCV and is attached to pads on the plate of the SCV. A one-inch diameter conduit (2CR397) is attached to the supports for the hydrogen collection line.
Both lines are non-safety related systems.
During the resolution of the bicklog civil calculations which contained Unverified Assumptions (UVAs), four calculations were evaluated which contained UVAs which stated that modifications were required to eliminate the overstressed condition in the pad plate or shell of steel containment vessel. The modifications were deferred in 1988 when the uni.ts were restarted, based on QIR SQPSQN 88209 which stated that the modifications could be eliminated if a refined analysis with removal of conservatism would be performed.
I Discussion i
The licensee performed further evaluations on the conduit and 12 inch diameter hydrogen collection lines in stress calculations and pipe (actually duct or conduit) support calculations to qualify the pad plates and shell of the SCV.
Support calculations 47A91504012 and 47A91504013 were partially reviewed and determined to be acceptable. After reanalyzing and evaluating the stresses in the hydrogen line and conduit, the associated supports, and pad plates and shell of the SCV, the licensee concluded that the pad plates and shell of the SCV were acceptable and did not require modification. Modifications were required 'cr two pipe and two conduit
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supports and had been completed by the licensee under DCN M09913A, Unit 2 restart item #539.
Conclusion This item had been reviewed and determined to be acceptable with the modifications completed. All the stresses for the lines, supports, pad plates, and shell of the SCV had been determined by the licensee to meet -
the long term design criteria.
3.2.6 Piping and Pipe Supports Inside ERCW Pipe Tunnel (CAQR SQP890484)
Issue This CAQR concerned the use of seismic response spectra in the pipe stress analyses and pipe support designs for the Essential Raw Cooling Water (ERCW) pipe tunnel that may be nonconservative.
Discussion The licensee developed the seismic response spectra for.the ERCW.
In some cases, the new spectra were higher than those used to originally analyze the safety related piping in the tunnel.
Seven rigorously analyzed stress problems that were affected by this change were reanalyzed using the new spectra and other criteria updates.
The seven stress problems were: N2-67-1A; N2-67-2A; N2-67-3A1; N2-63,72,74-1A, 2A; N2-72-3A; N2-63-9A; and N2-72-9A. All pipe stress reanalyses have been completed and the stress calculations have been issued.
The detailed review for the 683 pipe supports affected by the new response spectra had been deferred due to the plant shutdown. However, the pipe supports were evaluated for new loads and/or movements using a conservative load ratio technique. Using this method, 85 percent (581)
of the pipe supports were determined to meet design basis. The other 15 percent.(102) required a more detailed evaluation. Detailed evaluations of 28 of the 102 supports had been completed. Thirteen of the 28 supports meet design basis; the other 15 were determined to meet the operability criteria, but modifications were required to meet FSAR design criteria.
The licensee concluded, without performing a detailed review, that the remaining 11 percent (74) will also meet operability criteria.
The inspectors disagreed with the licensee's conclusion that all the l
pipe supports meet the operability criteria without the benefit of a detailed review. After questions by the inspectors, the licensee performed a detailed review for the remaining 74 supports to determine l
if they met operability criteria. The licensee concluded that all but
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one support passed the operability criteria.
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The licensee concluded it was necessary to modify this support, Support No.1-ERCWH-165. in order to meet the operability criteria. The support
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modification was issued under DCN M04244B.
The modification (DCN) will be completed before restart.
Conclusion This item was determined to be acceptable for restart after the modification for Support No. IERAWH-165 is completed.
However, the licensee plans to perform the detailed review for all 683 supports affected by the new stress analyze and spectra after restart.
3.2.7 Missile Protection Issbe
This item initially concy ned the discovery of 8 fuel oil vent lines from the 7-day fuel oil :;torage tank in the Diesel Generator Building (DGB) that were unprotected from impact by tornado missiles. A subsequent walkdown using extremely conservative criteria without regard for the significance of the finding was performed to identify any other items which were either comphtely or partially unprotected from impact by tornado missiles.
Discussion The inspector reviewed Problem Evaluation Report (PER) SQ910041 which documents the problem and its status. The above walkdown disclosed no items that were totally unprotected, which confirmed the overa'il compliance to Sequoyah Nuclear Plant's (SNP) tornado design criteria.
Of the 31 possible openings, or targets of possible impact trajectories, identified by the walkdown, 23 were determined very unlikely since many openings were extremely small or could only be p: cetrated by very limited flight paths.
TVA's evaluation determined the only credible missile is a vertically descending 1-inch diameter rod that could pass through the wall of the DGB exhaust stack creating a small hole and exiting into the DGB exhaust room. The inspector examined the licensee's Operability Assessment for Equipment Subject to Tornado Generated Missiles.
This assessment which utilized site-specific meteorological data concluded the frequency of the probability of a tornado occurrence striking within a 30-mile radius of SNP to be 1.635
X 10 per year. A combined probability calculation which considered both the tornado occurrerme frequency and the very limited openings / trajectories associated with the targets mentioned above would conclude a very small probability of actual impact.
Consequently, the licensee feels the safe operation of the plant is not degraded by allowing the plant to operate through Cycle 6 refuelling outage.
Conclusions The inspector concurs that the probability of a tornado missile striking safety-related equipment by penetrating any of the openings identified
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l is extremely remote, and sufficient 1.y low enough such that safe
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operation of the plant is not adversely affected while corrective
actions are being undertaken. SNP is currently performing it's i
Individual Plant Examination for External Events (IPEEE) to further
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assess plant tornado hazards and the results of this study areL scheduled.
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to be available by. June 1994.
3.2.8 Seismic Impact of IE Cauinets and Attachments Closely Spaced to Other.
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Commodities Issue
This problem involved two issues.
Various auxiliary, control, diesel i
generator, and ERCW pump station building closely spaced (IE) cabinets i
were identified as being susceptible to seismic impact against adjacent, cabinets or other plant features since these cabinets and equipment-
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contained within were not qualified for impact loads. Secondly, rigid -
conduits attached to the top of some cabinets were identified during the
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ongoing IPEEE walkdowns for having the potential for unqualified impact
with closely spaced plant features.
j Li Discussion
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The inspector reviewed PER SQ910196, DCN M-09365 (Unit 1), and.DCN M-
09365 (Unit 2), which investigate, evaluate, and provide the required modifications to the susceptible cabinets to insure _ no seismic impact-a could occur. The subject Unit 2 modifications have already reportedly
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been implemented, and the Unit 1 DCN is to be worked prior to the unit'.,
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restart. Analysis performed by the licensee-(Calculation 'SCG-4M-00861),
i demonstrates that the energy impacted due to seismic impact of the identified conduits is negligible. However, to assure satisfaction of
industry and IPEEE interaction concerns the licensee h<
committed'to~
reposition / shim all affected conduits to preclude any pusible seismic
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interaction durin' the next outage.
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Conclusion
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All issues identified in the PER have been resolved, and all necessary corrective actions associated with the Unit 2 restart have been implemented, therefore this item is closed.
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3.2.9 Commercial Grade Valves Used for Containment Isolation J
Issue Sixteen isolation valves for the 2-inch chill water lines (4-penetrations per unit, both inboard and outboard valves) penetrating the-steel containment vessel were procured as Class B, without seismic qualification.
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Discussion The licensee performed a post-dedication of the subject' valves, in that an engineering evaluation and justification for continued operation was
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performed and documented in PER SQ920366 which was examined by the-inspector. This evaluation provides details showing the valve body type and mounting bracket to have been qualified previously to loads exceeding the loadings expected for these valves. Vendor data on the-design features of the actuator were reviewed and it was determined that:
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the actuator was similar to actuators previously qualified to higher acceleration levels than expected for these valves.
Conclusion The component qualification group has demonstrated that the subject valves in question are similar enough to qualified components, therefore-there is no problem from an operability standpoint. The problem with a less than adequate procurement specification has been properly addressed.
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3.2.10 Conduit Overspans Issues
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DCN M06484A was issued by the licensee to reduce conduit spans. All conduit support modifications except for two Unit 2 supports have been completed.
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Discussion The licensee performed an evaluation of the conduits affected by.the two new supports which had not yet been installed.
Four conduits, one 1",
two 11", and one 2" diameter, have existing spans of approximately twenty feet. The addition of the new supports will reduce the span to within the typical conduit spans used for new construction / installation work. The conduits are non-safety related. The licensee reviewed the
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existing conduit spans for the four conduits and concluded that they comply with the design basis. The only function the conduits are i
required to perform is to retain their position, i.e., not fail and impact any safety related equipment, during a seismic event.
j Conclusion The exi-ting conduit spans are acceptable for restart.
3.2.11 Binding Condition of Snubber Rear Bracket
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The upper snubber on reactor coolant pipe support number 2-RCH-114 has'a seven degree misalignment between the paddle and rear bracket. This is two degrees greater than the five degrees allowable value for misalignment.
Discussion l
This problem was documented by the licensee on drawing deviation (DD)
report 92 D06355. The licensee performed an evaluation and concluded
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that this item was acceptable for restart based on discussion of their
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observations regarding operating condition affecting the snubber.
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However, the licensee had not completed any calculations to support r
their position. After being questioned by the inspectors, the licensee
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initiated a calculation, number 0600 154-13-33, to justify their
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position.
Conclusion
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The above referenced calculation demonstrated that the support snubber
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was. acceptable for restart.
3.2.12 Modify Supports to Meet Design Requirements i
Issue I
During review of pipe support calculations, the licensee discovered that the calculated faulted load exceeded the capacity for the snubber
support number 2-MSH-440 by 17 percent.
Discussion The licensee issued PER number SQN PER 920062 to document and disposition this problem.
The licensee reviewed other pipe support calculations to determine if the snubbers met the allowable load capacity tables supplied by' vendors. This review disclosed that the actual faulted loads exceeded the snubber capacity for ten snubbers on
]j nine supports. The NF allowable faulted values were those established by the snubber vendor. The licensee determined that these snubbers meet
the operability requirements of SQN CI-90.02. The inspector questioned licensee engineers regarding the basis for the NF allowable faulted loads listed in the PER.
Licensee engineers provided a letter dated
_l-June 2, 1986 from the vendor, Pacific Scientific, which. listed the i
fau'Ited load capacity for NF PSA snubbers to be approximately 50 percent
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greater than the normal load rating value stamped on the snubber name plate. This same letter states that Pre-NF snubbers, which are discussed in paragraph 3.2.2, above, have a one-time load rating of twice the load indicated on the snubber name plate. The inspectors questioned licensee's engineers regarding their verification of these allowable load ratings.
Licensee engineers stated that they did not perform any independent design verification of the vendor's load ratings.
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Conclusions
The pipe supports / snubbers are operable for restart.
The licensee will implement long term corrective actions to replace the " overloaded" i
snubbers during the next refueling outage.
The inspectors expressed concern that the licensee accepted vendor data without performing some independent verification of the data. The inspectors discussed past examples of cases where the vendor data was incorrect. An example of
this type of problem involved the foam concrete under the floor of the reactor building ice condenser bays. The vendor who supplied the foam
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concrete stated ti.at the foam concrete was impermeable and would not
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absorb water when the problem with the ice-condenser door binding was
found. The licensee performed some independent testing on the foam
'i concrete and found it absorbed approximately 35 percent of its weight in water. The failure of the licensee to evaluate and verify vendor supplied data was identified as a weakness.
- 3.2.13 Increase in Seal Water Outlet Temperature
Issue The original piping design analysis used a maximum temperature of 140' F i
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for design of the piping in the seal bypass and leakoff lines.
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Westinghouse has determined that this temperature can reach 225' F.
Discussion
The licensee issued PER number SQ PER 920353 to document and disposition this problem. The licensee reviewed the piping isometrics and
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identified the Unit 1, number 1 seal leakoff lines as the most i
sus'ceptible to temperature increase. A bounding calculation was performed on this line using the TPIPE computer program. The piping stresses met long term requirements.. Review of the pipe support calculation showed that ten supports met long term conditions while the remaining 15 meet short term operability criteria. Two other lines were also analyzed. These were found to be acceptable for long term
conditions with modification.
Conclusion
The seal leakoff lines meet operability criteria for restart.
3.2.14 Nonfunctional Support Issue Pipe support 2-CSH-2 was found not to be qualified for " local effects" at the interface of support number 2-CVCH-559, which was attached to this support.
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Discussion The licensee issued PER number SQ PER 930006 to document and disposition this problem. The inspector reviewed calculation N2-62-12A, Rev. 2, which was completed to assess the operability of these supports. The calculation showed that pipe support 2 CHS-2 meets operability criteria at the interface with 2-CVCH-599.
Conclusion The piping / support is operable and acceptable for restart of Unit 2.
3.2.15 Pipe Clamp Overtorques Issue Employee concern 22106-SQN-01 identified a potential problem that excessive torques inay have been applied to tieback pipe clamps.
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Discussion All tieback support pipe clamp problems identified in the employee concern were. resolved except for two. The remaining two were found to have been excessively torqued. However the clamps and bolts were found to be in good condition, i.e., not distorted. Testing was performed on a similar clamp configuration and showed that the overtorqued condition did not effect clamp operability. The corresponding Unit 1 supports were inspected and were also found to be overtorqued. An evaluation of the piping showed that these supports could be deleted. Since the piping was qualified with or without the subject supports, it was concluded that this item was acceptable for restart.
Conclusion The pipe clarnps are operable and acceptable for restart.
3.2.16 Improper Pipe Support Hardware Thread Engagement Issue During an NRC inspection, documented in NRC Report Number 50-327/92-12 and 50-328/92-12, an unresolved item was identified regarding improper thread engagement for a strut on Support 2-SIH-22 and missing spacers in the rear clevis.
Discussion The licensee completed calculation 2 SIH002L to evaluate these concerns.
The conclusion of the calculation was that the support was acceptable for restart.
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.J Conclusion
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The licensee's engineering evaluation showed the support was_ operable
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and acceptable for restart.
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3.2.17 One Inch Diameter Glycol Piping Installed without Proper Supports
Issue During a walkdown inspection, the licensee identified a run of approximately thirty feet of pipe existing in the field which had not been shown on the plant drawings.
Discussion This problem was documented by PER number SQ 930535 PER. An engineering evaluation was performed using the TPIPE computer program. The piping analysis showed that the piping met operability requirements, One vertical hanger will be required to meet long term requirements.
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Conclusion The piping meets operability requirements and is acceptable for restart.
3.2.18 EGTS Damper Seismic Qualifications
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Issue j
Isolation dampers (butterfly valves) for the emergency gas treatment system (EGTS) were purchased as seismic category 2 equipment. The design criteria for this system requires that the damper be seismic category 1.
Discussion
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These dampers are located in the reactor building annulus area. During normal plant operation, the dampers are closed tof solate the flow path
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to the shield building vent stack and to the lower annulus area.
In the event of a LOCA, the dampers are required to be operable, the position, open or closed, dependent upon plant conditions. These valves are not
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required to be operable following a seismic event. The licensee based this position on the position in Generic Letter 87-02 which states that a safe shutdown earthquake and a LOCA can be treated as independent events.
The operability of the dampers is demonstrated every 18 months, during the EGTS Annulus Vacuum Drawdown Test.
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Conclusion s
The EGTS dampers can perform their safety function during a LOCA. This problem is primarily one involving lack of documentation.
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i 3.2.19 FSAR Verification - Civil Engineering
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Some information in the FSAR does not have backup documentation.
Discussiqn The licensee has determined that some backup documentation was incomplete.
This information included lack of documentation for'various
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civil design criteria for foundation design values, wind design (velocity data) etc. This is not an operability problem; but lack of documentation results in difficulties in cross-referencing design criteria to FSAR requirements.
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Conclusion Completion of the backup documentation will be a benefit in improved
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references of information sources. This is not an operability issue.
t 3.2.20 Increase in thermal mode for CCS heat exchangers.
_LssJte SQPER920230 reported that an error had been made in determining the l
required ERCW flow rates for CCS plate heat exchangers.
Discussion i
t One of the concerns raised by SQPER920230 was that increased thermal loads from changes to Essential Raw Cooling Water (ERCW) and Component
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Cooling System (CCS) operating modes could cause failure of piping or supports. Changes to the CCS heat exchangers had increased the ERCP t
maximum temperature from 83*F to 84.5*F.
The inspectors reviewed the
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calculations provided to support the JC0 for the effected piping and i
piping supports.
t Conclusion
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lne inspectors agreed that the calculations showed that the increase in j
thermal loads on the piping and supports were minimal and would not
invalidate previous operability determinations for these systems.
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3.2.21 Oil collection system for Reactor Coolant Pumps (RCPs)
Issue The oil collection systems had not been properly restored to drawing requirements after maintenance activities.
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19 Discussion
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The purpose of the oil collection system on the RCPs is to reduce the fire hazard by preventing oil spills in the event of a seal failure.
SQ930430PER documented an instance where the systems engineer discovered problems with the restored collection systems after maintenance. As a result of thT PER, Temporary Alteration Control Form (TACF) -2-93-0036-040 was written to restore the configuration of all of the Unit 2 RCP t
oil collection systems. Three of the oil collection system were fully restored to full siphon capability. The collection system on RCP 3 still lacked capability for the full pipe condition required for.
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siphoning at full dam condition.
Even without siphoning capability, the
RCP 3 collection system drain rate would exceed the maximum possible
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fill rate of a seal leakoff line failure; therefore, the RCP 3 collection system was declared operable. The licensee has concluded
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that the RCP oil. collection system needs to be redesigned to provide
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easier access by maintenance.
Conclusion The inspectors reviewed the engineering A s:ification for the operability of the RCP 3 oil collection system and agreed that the dam
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configuration would provide collection and drainage of oil from a seal leakoff line failure and that full siphon capability is not required for the system to reduce the fire hazard in this area.
3.3 Summary and Conclusions The inspectors concluded that the items which will remain open after
restart of Unit 2 comply with the licensee's operability criteria.
Deficiencies were identified in the licensee's operability assessments
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for the ERCW and main steam (MS) system pipe supports. These i
deficiencies necessitate modifications to one ERCW and two MS supports i
to meet operability requirements. The inspectors also found that the
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licensee's safety assessments were sometimes inadequately documented.
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However, in most cases the licer.see was able to furnish additional information to justify their operability determinations.
j Violations or deviations were not identified.
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4.0 Exit Interview The inspection scope and results were summarized on October 4, 1993, with those persons indicated in paragraph 1.
The inspectors described the areas inspected and discussed in detail the inspection results listed below.
Praprietary information is not contained in this report.
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IFI 327,328/93-49-01, Post-Restart Walkdown and Evaluation of Main Steam Line.
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