ML20127F004
| ML20127F004 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 03/14/1985 |
| From: | Morgan R CAROLINA POWER & LIGHT CO. |
| To: | Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| Shared Package | |
| ML20127E982 | List: |
| References | |
| IEB-79-02, IEB-79-2, RNPD-85-471, NUDOCS 8505200289 | |
| Download: ML20127F004 (5) | |
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.f Carolina g wer & Light Company -
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' ROBIN NT l
POST. OFFICE BOX 790 f'7
'HARTSVILLE, SOUTH CAROLINA 29550 MAR'141985 l7 Robinson File No:
13510E Serial: RNPD/85 471
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Dr. J. N. Grace Regional Administrator
-U. S.' Nuclear Regulatory Commission'
< Region:II, 101 Marietta' Street, Suite 3100 Atlanta,. Georgia -30323 4
^H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 LICENSE NO.:DPR-23
' REGION II INSPECTION REPORT 85-06
Dear Dr.~ Grace:
' Carolina Power.and Light Company (CP&L) has received and reviewed the subject -
reporb and provides the=following response.
-A.-
Severity Level IV Violation (RII-85-06-01-SL4)
H. 'B. Robinson -Technical Specifications, paragraph 3 3 1, require that-
-piping associated with the safety injection pumps, residual heat -
cremoval pumps, and' residual heat exchangers.be operable for reactor-
' criticality..-NRC IE Bulletin 79-02, Revision 1, Supplement 1,. defines pipe supports with concrete expansion anchor safety factors less than 2 as inoperable.
Contrary to the above, prior to startup from the Steam Generator.
Replacement Outage, the licensee had erroneously used a concrete expansion anchor safety factor. of 1, considered the above noted systems -
operable, started up the reactor, and was critical at about 50 percent power.. Upon notification of the Bulletin criteria, the ' licensee.
determined the above'noted' piping and eight associated pipe supports to "be inoperable due to-concrete expansion anchor safety factors that-did 3
not comply with the above noted IEB'79-02 operability definition.- The reactor was subsequently shut down by the-licensee. Subsequent
. licensee evaluation resulted in 44 additional inoperable pipe supports on piping required by Technical Specifications to be operable for reactor criticality.
8505200289 850402 PDR ADOCK 0500026i G
- 1.Q*j#serigig/RNPD/85-471 Lettir to Dr. J.'N. Grded;'
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'ResDonse 1.-
Admission or Denial of the Alleged Violation CP&L: acknowledges the? alleged violation.
2.1 Reason 'for the Violation
- Closeout' work associhted'.with IE. Bulletin IEB 79-14 had been performed during'1983 and 1984. ' Analytical work associated with these closeout-activities identified 662 seismic restraint points t'
which required corrective action (modification) to meet the
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requirements of IEB.79-14 and-associated. IE Bulletins of which IEB 79-02 was one.
Corrective action-associated with the 662 seismic restraint
. points was started during the 1984 Steam Generator Replacement
-. Outage (SGRO).. Due to ' the schedule impact of these modifications on the SGRO, criteria were established which allowed interim operation with ' seismic restraints whose designs did not fully meet the IEB 79-02 requirements for expansion anchor bolts.. The-4 criteria were based on the expansion anchor. bolts and structural'
- steel having a minimum safety factor of 1.0 compared to)the '
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-ultimate capacities as-allowed by. codes and manufacturer's allowables. ' Based on the criteria _ established, 132 structures were declared operable for interim operation', and modifications
- were not performed on those structures prior to the end of the
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SGRO.
M On January 15,.1985, after discussions ~ with NRC personnel, 'it 'was realized that the-interim criteria established did not include the requirements for restraint operability included in IEB 79-02, f'
Revision 1, Supplement 1,' dated August 20, 1979. - This was p
omitted due to an oversite by-the personnel establishing the interim operation criteria. Review of the 132 structures against r
the IE Bulletin 79-02, Revision.1,-Supplement 1 criteria revealed that-a number of structures did not meet this criteria and were
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i-p therefore,'by definition, inoperable..The inoperable structures-
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resulted in the inability to analytically' qualify the associated piping under the seismic criteria of the FSAR.
3 Corrective Steps Which Have Been Taken The unit was shut down on January 16, 1985.
F Interim operation criteria were reestablished which included the l
requirements of the IE Bulletin 79-02, Revision 1,
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Supplement-1. Each of the 132 structures not modified was reviewed against the new criteria. Results of this review j.;
indicated that 95 of the 132 structures did not meet the criteria.
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_ Additionally, _ the reviews performed to determine compliance with
- the newly established criteria revealed that-a number of r
structures did not meet the criteria for interim operation applied prior.to the end of the Steam Generator Replacement Outage.- Specifically, 90 structures (a subset of the 95.
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identified not meeting the new criteria) did not have safety _
factors greater than or equal to11.0-or.did not meet all the-criteria for_being a' seismic restraint. This deviation was
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Jdetermined to be caused by the -use of engineering judgements based on_ limited analytical information which, in most cases, were non-conservative.- Lack of.information was due to the failure of the engineering organization to complete necessary analyses required to support the judgements made.
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- One additional structure previously identified as-not requiring
-modification'was reevaluated'as requiring modification, bringing-the total number of structures not meeting at least the ' interim operational criteria to 96.
i Modifications-were performed on the-96 structures not meeting the interim criteria (78 were modified to seet actual design values, 18 were modified to meet the ' interim criteria). Prior to' returning to power, all but 39 of the structures met _ the original
~ design criteria. These 39 structures' meet the' interim criteria but will require additional modification. Some structures are-being modified now where operations permit.-
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. Corrective Steps Which Will be Taken Modifications to the remaining 39 structures, which are not
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- permitted by plant operation, will be completed during the next refueling outage.
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Date'When Full-Compliance Will be Achieved The structures met operability requirements.or interim operability requirements prior to returning the unit to service on February 10,' 1985. The remaining structures will be modified -
prior to power operation following:the next refueling oute.ge.
B.
Severity Level V Violation (RII-85-06-02-SL5) 10CFR50,: Appendix B, _ Criterion V requires that activities affecting quality be ' accomplished in accordance with ' instructions, procedures or-drawings.
1.
EBASCO Services, Incorporated Procedure No. 79-14/C-3, Revision 3, Seismic Restraint Analysis and Design, paragraph 6.6 requires that "new loads" documented by the Mechanical. Stress Analysis Department be used for_ restraint' modification design.
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Sericli lRNPD/85-471-Page'4-2.
EBASCO SSrvices,. Incorporated " Procedure for Inspection and Testing of Existing Concrete Expansion Anchor Bolts," Revision 4,
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. Table 4 requires reduction of the' capacity of 1 1/4" diameter concrete expansion anchors that'are less than 7-1/2" from a concrete edge.
-Contrary.to the above:
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Maximus' new loads were not used for the baseplate and concrete expansion anchor design for safety-related seismic restraint MS-1C-1062 modification.
'2.
A reduction:of the capacity of'1 -1/4" hiameter concrete expansion
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anchors that were ' designed to be less than 7 1/2"' from a concrete edge was not documented on the design calculation for safety-related seismic restraint SI-20-2310.-
Response
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Admission or Denial of' the Alleged Violation 1CP&L' acknowledges the' alleged violr. tion.
2.
Reason for Alleged Violation The cause of the violation was failure to follow procedures by
--the EBASCO Services personnel performing the respective calculations.
Item 1 of the violation relates to-the incorrect load being used in the. design of a seismic restraint MS-1C-1062. Item 2 of the violation was due.to the failure of the analyst / engineer to adequately document,the conservative approach used in the design.
calculation associated with restraint SI-20-2310.
3 Corrective Steps Which Have Been Taken The calculations associated with restraints MS-1C-1062 and SI-20-2310 were repeated using the correct loads and criteria.
Deth calculations revealed 'that the restraints were technically adequate.
A technical review and documentation audit were performed by the -
CP&L Nuclear Engineering and Licensing Department and Corporate Quality Assurance Department. Results of their sampling indicate that there are numerous documentation inadequacies; however, the calculations are technically adequate.
y iLett;r to Dr. J' N.'Grtos.
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Serin 1: RNPD/85-471-Pago-5 4.
' Corrective Steps Which Will be Taken Each final pipe stress analysis and restraint analysis will be checked.for adequacy of documentation.-
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Date When Full Compliance Will be Achieved The above corrective action will be completed prior to Bulletin closecut which will be completed prior to power operation following the next refueling.
If you have any questions concerning this response, please contact Mr. David C. Stadler at (803) 383-4524, Extension 363 Very truly yours, J 8A./
R. E. Morgan General Manager H. B. Robinson S. E. Plant CLW:tk/C-111
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