IR 05000498/1986042
| ML20212K896 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 02/20/1987 |
| From: | Imbro E, Parkhill R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| Shared Package | |
| ML20212K853 | List: |
| References | |
| 50-498-86-42, 50-499-86-42, NUDOCS 8703090448 | |
| Download: ML20212K896 (15) | |
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US NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Division of Quality Assurance, Vendor, and Technical Training Center Programs . Report Nos.: 50-498/86-42, 50-499/86-42 , Docket Nos.: 50-498, 50-499 Licensee: Houston Lighting and Power Company Facility Name: South Texas Project, Unit 1 Inspection At: Bechtel Energy Corp, Houston, Texas South Texas Project Site, Bay City, Texas (12/18/86) Inspection ' Conducted: December 15 to 19, 1986 Inspection Team Members: Team Leader: R. Parkhill, Inspection Specialist, IE Mechanical Systems: T. DelGaizo, Consultant, WESTEC Services Electrical Power: S. Athavale, Electrical Engineer, IE Mechanical Components: J. Blackman, Consultant, WESTEC Services
- E. Imbro, Section Chief, Quality Assurance Branch,IE Prepared by:
$w&lWbw A-/f -H Ronald Parkhill Inspection Specialist, IE
Team Leader
Approved by: Lfpy $ mj . i ' 2-2n-{} Eugene V. Imbro, Chief Licensing Section . Quality Assurance Branch
- Attended exit meeting and Site Walkdown on 12/18/86
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. SOUTH TEXAS PROJECT - UNIT I , INSPECTION OF AUDIT RESULTS: TASKS 84-4, 85-4, 85-5, AND 86-1 December 15, 1986 Through December 19, 1986 1.
Backaround On March 1, 1984,' representatives of Houston Lighting and Power Company (HL&P) presented to the NRC the details of the Engineering Assurance Program , (EAP) being conducted on the South Texas Project (STP). As a result of this ! meeting and additional information provided by HL&P, the NRC determined that this program, if properly implemented, could provide the additional assurances of design adequacy normally provided by an Independent Design ! Verification Program (IDVP). Formal acceptance of the EAP as a substitute for an IDVP was provided via letter to HL&P dated August 20, 1984.
The NRC monitored the STP Engineering Assurance Program in three phases for Task 85-1, Control Room HVAC, Task 85-2, Component Cooling Water Systems, and i Task 85-3, AC Power Supply Systems: (1) implementation of program plan and procedures, (2) review and evaluation of audit results and (3) follow-up of corrective actions. The first phase of inspection was accomplished at the f headquarters of the Stone and Webster Engineering Co. (SWEC) in Boston during i the week of April 23, 1985. The report of this inspection (No. 50-498/85-09) was forwarded to HL&P on July 12, 1985. The second phase of inspection for Tasks 85-1 and 85-2 was accomplished at SWEC's headquarters in Boston from . July 22, 1985 through July 26, 1985 and a report of this inspection (No.
50-498/85-14) was forwarded to HL&P on August 28, 1985. The second phase of , inspection for Task 85-3 was accomplished at SWEC's headquarters in Boston from October 15, 1985 through October 17, 1985. The report of that inspection (No. 50-498/85-22) was forwarded to HL&P on November 18, 1985. The third phase of inspection was conducted at Bechtel Energy Corporation (BEC) offices in Houston, Texas from August 25, 1986 to August 29, 1986. The report of this inspection (No. 50-498/86-39) was forwarded to HL&P on December 12, 1986.
. In addition, the NRC inspected action item resolution and corrective actions associated with Tasks 83-1, 83-3, 84-1, 84-2, and 84-3.
These are contained In Report 50-498/85-26 dated March 25, 1986.
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Purpose
The purpose of this inspection was to verify the adequacy of the ' corrective action associated with the following design activities: Task R4-4, Separation and Fire Protection Criteria; 85-4, High Energy Line Break Analysis; 85-5, Class 1 Piping Stress Analysis; and 86-1, Field Walkdown. The , - methodology utilized by the NRC Inspection Team involved reviewing the initial , action item identified by SWEC and then reviewing the documentation trail ' between HL&P Engineering Assurance, HL&P Project Engineering and Bechtel Energy Corporation (BEC) to ensure that the resulting corrective action was ' responsive to the action item.
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Personnel Contacted . The following is a brief list of key personnel contacted during this inspection: Name Position E. W. Dotson HL&P Manager of Engineering R. A. Frazar HL&P EA Program Manager . S. R.-Basu HL&P EA Program Staff M. K. Chakravorty HL&P EA Program Staff R. D. Reed HL&P EA Program Staff J. E. Woods HL&P EA Program Staff D. E. Root HL&P Engineering C. R. Allen HL&P Engineering R. C. Munter-HL&P Engineering W. L. Schuler Bechtel Engineering K. J. Romano Bechtel Engineering A. L. Matiuk Bechtel Engineering J. Hurley Bechtel Engineering R. L. Beck Bechtel Engineering D. Ashton Bechtel Engineering W. Watson Bechtel Engineering 4.
General Conclusions The NRC inspection team reviewed a large number of calculations, diagrams, specifications, and other design documents during the five day inspection, as further detailed in this report.
In addition, the NRC inspection team visited the plant site and walked down areas and systems associated with the EAP audits. Based on this review, the NRC team concluded that the resolution of action items and the verification of the STP Project commitments by Engineering Assurance personnel were supported by sufficient technical justification.
The NRC team concluded that Engineering Assurance Tasks 84-4, 85-4, 85-5, and 86-1 reflect in-depth analyses of the state of design of the South Texas Project. Details of the inspection for each of the four technical audits are contained in Attachment 1 to this report.
, Attachment I also identifies certain action items which were unresolved at the time of the NRC inspection or for which corrective action is continuing. Attach-ment 1 identifies Action Items 84-4-17/11C, 84-4-10a, 86-1-22, and 86-1-29 as open and identifies action necessary to close them.
Attachment 2 lists specific items questioned by the NRC inspection team during the site walkdown of 12-18-86, along with the resolution of these items and identified observations #4, #7, and #9a as being open, requiring further information.
Minor inconsistencies noted in certain documents during this inspection had no impact on plant design and were of the type which would be corrected during the STP Design Finalization and Document Turnover Program. The NRC team noted that HL&P is relying on the STP Design Finalization and Document Turnover Program to address the generic impact of the documentation inconsistencies and recognizes the inherent value of such a progra _ _ .. .
a- ., ... - ' Attachment 1 1.0 Independent Technical Assessment 84-4, Separation and Fire Protection Criteria HL&P authorized ~SWEC to perform an independent technical assessment of the adequacy of Separation and Fire Protection Criteria for the STP, Task 84-4.
The objective of the review was two-fold. First, SWEC verified that the fire protection and fire detection features adequately protected the capability of the plant to achieve and maintain safe shutdown under fire conditions. Secondly, SWEC verified, through assessment of selected areas of plant design, that fluid system layouts and separation of redundant electrical raceways permit availability of the components and systems required to safely shut down the plant under fire conditions. The basis of SWEC's evaluation was the draft Amendment 4 of the Fire Hazards Analysis Report (FHAR) and an Impell report regarding sections III. G. and III. L. of 10 CFR 50 Appendix R.
The resolution and verification of the following action items were reviewed by the NRC inspection team: Action Items 84-4-2, 10a, 10b, 10c, 10d, lla, 11b, lle, 17, 18, 22, 23, 26, 30a, 31a, 31b, and 31c. The following observations pertain to the more significant of these items. Action items considered of lesser signifi-cance are not addressed in this report.
Action Item 84-4-2 dealt with heat tracing associated with the boric acid storage tanks. There was a concern that this heat tracing was not on the list of essential equipment and therefore would be unavailable if off-site power is lost. Bechtel Energy Corporation (BEC) prepared calculation MC 6045 which showed that tank temperature remained above the boric acid crystalli-zation temperature (55' F for a 7000 ppm solution) both for the 8 hours necessary to inject the boron and for the 24 hours that power is assumed to be lost. During the inspection at BEC on December 17, 1986, the NRC inspection team reviewed calculation MC 6045 and determined that the methodology of the s calculation was proper and that the design input information was consistent with plant parameters. _ Based on its review, the NRC inspection team concurs that the concern has been adequately addressed. This item is closed.
Action Item 84-4-10a raised a concern regarding lack of indepth rev aw by Impell, for effects of fire related electrical failures (e.g., short circuits or shorts e to ground) of instrument circuit CIRCTE0474. The team reviewed the BEC response which stated that Impell had performed a detailed review of this concern for fire area 63 zone 220. This analysis, however, did not address spurious actuation of the ESF system. Bechtel has been asked by EA to perform this analysis considering the potential for spurious actuation of the ESF system. This analysis is expected to be completed by February 1987. This item will remain open pending confirmation finm the applicant that the analysis has been completed and any associated corrective action implemented.
(0 pen Item 86-42-1).
i . Action Item 84-4-10b identified that the fire area analysis for the load sequencer failed to address some circuits fed from the load sequencer. Bechtel responded that the effects of a fire on these components will be evaluated and i the safe shutdown evaluation will be updated accordingly. The NRC inspection i team noted that this activity is being tracked by EA per punch list item j IFP-01-057, therefore, this item is closed with regard to this inspection ' report.
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- .. _ - _ _ - _ .- - - . . - . - .. ' . Action Item 84-4-10e identified that the evaluation of fire area 7, zone 071 . assumed that removal of power from the cables associated with the sequencers would immunize the system from fire damage, but the evaluation failed to address other fire related failures such as loss of power due to a blown fuse.
Bechtel responded that this evaluation will be revised to address fire in the panels and related failures. Since this activity is being tracked by HL&P i Engineering Assurance via punchlist item IFP-01-064, this item is. closed.
Action Item 84-4-17 and lle noted that operator actions to be taken in the case of a failed open steam generator PORV (as result of fire damage) were not listed on the operator action list. Provisions were subsequently made to de-energize the PORV to prevent possible post-fire spurious operation. Westinghouse, however, noted that it may be necessary to relieve pressure in the isolated steam generator since a substantial differential temperature between steam generators could impede the natural circulation flow in the primary loop.
Westinghouse recommended that provisions be made to reduce pressure in the isolated steam generator. At the time of the NRC inspection BEC was in the process of finalizing its procedure for relieving pressure in the isolated steam generator. This item remains open pending confirmation by the licensee that an acceptable procedure has been established and that all corrective actions associated with this item have been completed.
(Open Item 86-42-2) Action Item 84-4-26 identified two inconsistencies between Impell's calculations ), and the actual plant configuration for some fire area locations and cable routings. First, the actual cable routing of cable B1AF01C25B is between fire ' area 3 zone 042 and fire area 31 zone 047, but Impell Calculation No. 0230-040-008 i page 139 shows cable location entirely in the fire area 3 zone 042.
BEC responded ' that this was a typographical error and will be corrected in the next revision.
Secondly, cable DIRCIOCDXAW as identified in Impell Calculation No. 0230-040-008 to be in fire area 63 zone 220, whereas Impell Calculation No. 0230-040-005 shows , this cable in fire area 63 zone 224. BEC responded that Impell Calculation No ! 0230-040-008 was correct and Impell Calculation No. 0230-040-005 will be revised ! to correct the discrepancy. These were considered to be isolated errors, therefore, ! this item is closed.
f 2.0 Independent Technical Assessment 85-4, High Energy Line Break ! . ' The resolution and verification of the following action items were reviewed: Action Items 85-4-3a, 4a-d, 10a, 12a, 15c, 16a, 24a-f, 26b, 27a, 28f, 41c, 43a, M1-2, M1-3, and M3-1.
The following observations pertain to the more { significant of these items. Action items considered to be of lesser l significance are not addressed in this report.
Action Items 85-4-24a-f, 85-4-26b and 85-4-27a expressed EA concerns relative to documentation of the plant design to mitigate the consequences of high-energy line break events. These concerns are as follows: Action Item Concern 85-4-24a-f Lack of discussion of (1) the effect of breaks, (2) postulation of worst case single failure, and (3) ability to go to a safe shutdown in Facility Response Analyses (FRA).
The FRA contained vague statements as to protection requirements; and lacked documentation explaining why impactees are unacceptable.
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Action Item Concern
85-4-26b HELBA work logic procedures contained no steps to transmit calculated impingement forces to the per-sonnel preparing hazards analysis calculations.
85-4-27a Documents transmitting zone-of-influence informa-tion between disciplines were not clear and lacked necessary justification for protection requirements.
In reply to these EA concerns, Bechtel stated that several changes have occurred in the licensing arena within the past 18 months which have now been reflected in the STP HELBA program. Not all of the following revisions have been formally accepted by the NRC licensing staff, however, they are being tracked and the impact of each item is being closely monitored by Bechtel.
(a) RCS Loop Breaks - Using Leak-Before-Break (LBB) technology the NRC has accepted the complete elimination of all mechanistic breaks in the primary coolant loop.
(b) Arbitrary Intermediate Breaks - The NRC has concurred with the elimination of arbitrary intermediate breaks (including breaks in the feedwater system) as previously defined in MEB 3-1.
(c) Cumulative Usage Factor = 0.4 - The Project has advised the NRC by letter that breaks based on cumulative usage factor will only be postulated when this factor is 0.4 or higher.
(d) 10 Pipe Diameter Zone of Influence - The Project has further advised the NRC by letter that the effects of jets of steam or flashing liquids will only be considered for a distance, from the break, of 10 times the diameter of the broken pipe.
(e) Coupled Branch Lines - The criteria for consideration of terminal end < breaks at branch lines have been narrowed to allow a greater range of branch line sizes to be included with consideration of the main run,
thereby reducing the number of terminal end breaks assumed to occur.
Hechtel further responded to the EA concerns that the cumulative effect of these initiatives has been a substantial reduction in the number of pipe breaks to he considered. This has further led to a consolidation of analytical efforts . elated to design for pipe breaks.
Previously, these efforts were multi-discipli-nary with a single full time coordinator and weekly meetings to provide transfer of information, interpretation of program requirements, and resolution of problem These efforts have now been incorporated into a single group represen-areas.
ting all disciplines and having a full-time group supervisor. The group has a dedicated staff capable of assessing all impacts and analytical results. Analyses which previously were performed in part by several disciplines are now performed within this group. Potential targets are identified and simultaneously evaluated for essentiality. The NRC team recognizes that the formation of this group has resulted in a significant decrease in the amount of interdisciplinary coordi-nation required and a greatly simplified HELBA process.
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. , - .. - . ' To ensure that BEC was postulating break locations in accordance with the piping stress analysis results, the NRC inspection team compared Calculation No. 4N229MC5318 Rev. 4 " Facility Response Analysis No. CU-01" to the piping stress summary sheets A2, A3, A4, and A5 from Calculation No. RC9818 Rev. 1.
From this review it was noted that nodes 101, 106, 110, 115, 127, and 133 in the piping stress summary required postulation of circumferential and longitudinal breaks since the calculated stress at these nodes exceeded 2.4 Sm.
However, no breaks were postulated in Calculation No. MC 5318. Bechtel responded that a new stress calculation (i.e. RC 5306) was in preparation, which demonstrated that the associated stresses at the aforementioned nodes were below the break postulation limits. The NRC team reviewed this new preliminary calculation and verified that the stresses were below the break postulation limits.
In addition, the NRC team noted BEC had correctly omitted these break locations and correctly documented that the basis for omission was " preliminary," which recognizes the need for confirmatory action when available. Therefore, the design process appears to be functioning adequately.
During its inspection at BEC and the site from December 15-19, 1986, the NRC inspection team reviewed the following calculations which reflect the revised STP HELBA procedure described above:
Calculation Pipe Break Nodes MC 5318 4"-CV-1002-5A-C 4"-RC-1320-142C , 2"-RC-1321-151C MC 5311 12"-RC-1312-498 12"-RC-1312-471C MC 5322 6"-AF-1008-005C 6"-AF-1008-125C i In addition to reviewing the zones of influence and the Facility Response Analyses contained in these calculations, on December 18, 1986, the NRC team physically walked-down these areas to verify the appropriate target identifications. During the site walkdown, the NRC team identified one line, 8" RH-1204, which appeared to be in the zone of influence
of postulated break 12" RC-1312-471C. Further review, however, indicated that because 12" RC-1312 was restrained near the break, the zone of influence was such that 8" RH-1204 is not impacted by the jet. This review verifies the detailed nature of the BEC analysis.
, As a result of this review, the NRC inspection team concludes that the concerns
identified above have been satisfactorily addressed and that an effective HELBA program is in place. Therefore, the items discussed above are closed.
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Action Items M1-2, M1-3, M3-1 resulting from the moderate-energy-line-crack analysis (MELCA) walkdown performed by Engineering Assurance as part of ITA 85-4, identified the following concerns: Action Item Concern i M1-2 Spare penetrations relied upon in flooding ' calculations are not appropriately identified in drawings.
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Action Item Concern ... M1-3 Spent fuel pool cooling pumps were not noted on impactee sheets and this was not identified in the preliminary walkdown by BEC.
M3-1 3"-CH-1430 and 1"-BA-1022 were not included in the ,- flooding calculation.
' In 'ddition to correcting the specific items, BEC indicated that all walkdowns a to date have been preliminary. BEC has committed to additional training of
personnel prior to the final walkdowns to ensure that discrepancies such as those listed above are identified. The NRC team considera the items identified during the EA review of the MELCA Walkdown to have provided examples of where
additional training is required prior to final confirmatory walkdowns by BEC.
i The NRC team agrees that additional training is an appropriate corrective action.
. These items are closed.
Action Item 85-4-41c questioned the use of area post-accident temperatures for ' , purposes of equipment environmental qualification rather than localized tempera-ture effects from postulated pipe cracks. BEC responded that bulk temperatures envelope local temperature effects since the bulk temperatures are based upon <
the most limiting pipe failure which are considerably more severe than pipe cracks. Further, the BEC indicated that this matter had been resolved with the Office of Nuclear Reactor Regulation (NRR) through FSAR question 210.35N.
This answer was acceptable to the EA team. The NRC team reviewed both the i ' answer to FSAR question 210.35 and Project Engineering Directive PED-016 which provided implementation information to the project. As a result of this review, the NRC team concurs that this item is closed.
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Action Item 85-4-43a identified an inconsistency in the criteria used to select , pipe whip restraint locations. BEC instruction PED-016, Section D.IV.B.1.b(1)
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specified that the ultimate resisting moment (Mu) is to be used.
While licensing commitments per FSAR Section 3.6.2.3.4.1.1 as well as ANSI /ANS 58.2-1980 specified that a value of 0.8 Mu was to be used. BEC responded by revising PED-016 to reflect the use of 0.8 Mu and performed a review of all pertinent calculations to determine if the calculated moment exceeded 0.8 Mu.
, It was determined that the calculated moments were in all cases less than > 0.5 Mu.
Based upon a review of the pertinent documentation and the fact that the " allowable" moment criteria was not exceeded, this item is closed.
3.0 Independent Technical Assessment 85-5, Class 1 Piping Stress Analysis
The resolution and verification of the following action items were reviewed: , 85-5-1, 2, 11, 5, 15, 17, 18, 19, 21, 22, 23, 25, 30, 31, 35, and 36. The , i following observations pertain to the more significant of these items. Action items considered to be of lesser significance are not addressed in this report.
Action Items 85-5-1, 85-5-17 and 85-5-18 identified omissions and inconsisten-cies between the requirements of the Westinghouse Systems Standard Design, , ! Calculation SR169MC5768, Rev.0 which provides mechanical system design require- ' ments and Class 1 pipe stress Calculation No. RC 5113 Rev. 1.
The piping in question is the RHR system piping from the hot leg of loop 3 to the suction j intake of RHR pump 1C.
Specifically, the inconsistencies were: i ,
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Action Item 85-5-1 Sixty (60) starts were associated with the inadvertent / spurious initiation of the safety injection system (upset mode) as described in the Westinghouse System Standard Design but the ,. mechanical discipline system design requirements (Calculation No. MC 5768, Rev. 0) indicated that system doesn't operate in , { the upset mode.
' 85-5-17 Calculation No. MC 5768, Rev. O provided process design data for the RHR system but not histograms which are required to perform Class 1 analysis.
, 85-5-18 Calculation No. MC 5768, Rev.0 indicates that the ntaber of operating cycles need not be addressed, however the process data in Attachment A of the calculation includes cyclic information for Class I lines.
The Bechtel Project's response was that, due to schedule constraints, pertinent
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design information was forwarded to the piping analysis group by inter-office memo without revising Calculation No. MC 5768, Rev.0 or was neglected with the knowledge that it would be incorporated into the calculation at a later date.
i , However, to resolve the matter, BEC revised mechanical discipline criteria document MC.5768 to reflect all required RHR design information as well as the
applicable criteria documents for the Safety Injection System (SIS), Reactor ! Coolant System (RCS) and the Chemical and Volume Control System (CVCS). All required revisions to the Class 1 piping analysis packages were made. The NRC team reviewed the pertinent documentation and confirmed that the required revisions have been made. The matter is considered closed and properly resolved.
While the formal design process was not adhered to during the initial plant design when the subject calculation was developed, existing project procedures
as well as the stress reconciliation program preclude a reoccurrence of the problem.
Action Item 85-5-11 identified a sign error in Calculation No. RC 5113 where
thermal displacements, applied at the intersection of the reactor coolant loop and the RHR system piping, were applied in the wrong direction.
In addition,
. the rotational displacements were not addressed in the calculation. Resolution of the deficiency by BEC involved rerunning the analysis and evaluating the
effects of the change on the piping, supports and other components. The only significant changes due to the reanalyses were a 32% increasr in one pipe support loading and increases, beyond vendor allowables, to valve and reactor toolant loop nozzle loads. The pipe support was reevaluated and found to be within allowable load limits. The equipment vendor, Westinghouse, was sent the revised loading and reevaluated the affected hardware to be within acceptable limits.
In addition, the BEC reviewed all other similar calculations (where
L l Westinghouse supplied displacement interface loads were considered) and found that in one calculation, one displacement was applied incorrectly and in eight i calculations, rotational displacements were not evaluated. The affected < calculations were rerun and vendor approval was obtained for all cases when , nozzle loads exceeded vendor allowables. No hardware changes resulted. The NRC team reviewed several of the affected calculations and resulting correspond-ence and concludes that the original problem and corrective measures adequately address all relevant concerns. The item is closed.
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_ _. _ ~ . . . . - .. ' . Action Item 85-5-21 addressed neglecting pipe support attachment weight in ' piping analyses RC-5113 Rev. I without providing substantiating documentation.
The potential significance of this finding is that the additional mass, if appreciable, can have a substantial effect on the dynamic excitation of the piping when subjected to seismic or hydrodynamic loadings. BEC responded by ' indicating that, for Calculation No. RC-5113, Rev. 1, the attachment weight is equivalent to less than one-half foot of pipe which is insignificant if con-sidered in the calculation.
In addition, a review of five Class 1 calculations was performed by BEC and it was found that additional pipe support weight had been - considered. Further, an evaluation of ASME III Class 2 and 3 lines indicated that while pipe support attachment weights weren't considered, the effect on lines with significant hydrodynamic loading was not appreciable. The NRC team reviewed the substantiating calculations and studies and considers this item closed since BEC has adequately addressed all potential significant ] concerns and has shown them to be inconsequential. This item is closed.
Action Item 85-5-25 identified an error in Calculation No. RC 5113 where individual stress levels from the appropriate loading conditions (which included jet and LOCA loadings) were manually combined. The intent of the analyst was to determine the maximum combined stress so that they could be compared to allowables. To accomplish this, node points were selected where the individual
maximum stresses occur for each loading. However, the analyst failed to con-sider other node points where the same load combinations may result in greater stress levels. BEC responded by revising the subject calculation to remedy the deficiency and reviewing all other similar calculations.
For all other calcu- , lations, the load combinations for each node were developed within the computer analysis program, NE 101. This approach precluded the problem from being repeated in other calculations.
In addition, PED-023, Rev. 3 was revised to
include acceptable methods for determining maximums when manual load combina- ' tions are performed. The NRC team reviewed the calculations and revised j documentation. Since the corrective actions address the scope of the problem identified, this item is considered closed.
Action Item 85-5-30 addressed an inconsistency in evaluating the thermal ratcheting equation of the ASME III Section NB-3653.7. BEC initially ased time phased AT, values corresponding to the time when the Ta-Tb value is maximum. Since the thermal ratcheting equation used in the Code is not a specifically a function of time, the maximum AT should be greater than the , value initially used. Accordingly, BEC reevaluated the ratcheting effect using a value of Sy at the average temperature for thermal transients.
Acceptable fatigue levels resulted. The NRC team reviewed the substantiating documentation and concurs with the resolution of the item since it addresses the technical concern raised by SWEC and the adequacy of the piping to resist i thermal ratcheting. This item is closed.
Action Item 85-5-31 identified a flaw in the manner in which pipe sleeve clearance verification information was handled.
If the radial expansion of the pipe was greater than 3/16 inch, PED-023 required that the verification
of sleeve clearance be placed on the open items list (OIL). However, no further evaluation as to how to disposition the item was provided. The BEC responded by revising PED-023 to include a specific methodology for evaluating the effects of the radial expansion and revising other relevant documents to q provide a mechanism for resolving potential clearance problems. Implementation j of the corrective action was verified by a review of sample calculations which i -- - - - - - - - - - -.. - - -. - -. - - _ .. - _ - - - - -
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.. .' ' contained sleeves. The NRC team has reviewed the relevant documentation and concurs that the corrective action implemented by BEC satisfactorily resolves the concern. This item is closed.
4.0 Independent Technical Assessment 86-1, Field Walkdown HL&P had SWEC perform a field walkdown with two general objectives. First, SWEC was to assess the as-built plant configuration with regard to seismic interaction and electrical separation. Secondly, SWEC was to verify that the as-built plant configuration was in agreement with the design documents and licensing commitments.
The resolution and verification of the following action items were reviewed by the NRC inspection team: Action items 86-1-9, 10, 14, 15, 19, 22b, 25, 26, 27, 29, and 30.
The following observations pertain to the more significant of these items. Action items considered of lesser significance are not addressed in this report.
Action Item 86-1-9 involved flexible conduit jacketed with PVC. The initial concern involved questions of combustibility and Standard Review Plan (SRP) requirements to limit use of PVC within the plant. Those questions were lar8ely resolved when the STP project demonstrated that the PVC was used only when vendor supplied and the PVC represented a small fraction of combustible matter in the plant. However, a separation issue exists which is currently subject of testing by Wyle Labs for free air cable in contact with the flex conduit. This matter is being resolved with the Office of Nuclear Reactor Regulation (NRR). The applicant has committed to provide results of the Wyle testing to NRR by January 1987 (ST-HL-AE-1810, Nov. 19, 1986). Since this matter is an open issue between the applicant and NRR, it is closed for pur-poses of this inspection report.
Action Item 86-1-14 identified that NSSS supplied cabinets wiring violated separation requirements of R.G. 1.75.
Upon further investigation it was found that NSSS supplier's design criteria took exceptions to commitments of R.G.
1.75 as listed in Chapter 7 of the STP FSAR. HL&P has informed NRR of these exceptions via letter dated December 5,1986.
Since NRR is reviewing this issue, this item is considered closed for this inspection report.
- Action Item 86-1-15 identified drafting errors regarding location of the sample nozzles for radiation monitors RA-RT-8033 and RT-8034. BEC committed to revise the affected drawings (SM-15-9Z-00136(6), SV-11-9-V-0173(7) and SM-15-92, FCR IW-01977 was issued to implement this corrective action. This item is closed.
Action Item 86-1-22 identified that non qualified instruments were connected in parallel with the safety related instruments.
In such cases, although the sensing lines are seismically supported, a failure in the non qualified instrumentation could affect the readings of parallel connected class IE instruments and create loss of process fluids. BEC intends to correct this problem by establishing adequacy of ur., alified instruments, either by analysis or by re qualification. This effort is scheduled to be completed by February 1987. This item will remain open pending confirmation from the applicant that all associated instruments have been satisfactorily qualified.
(open Item 86-42-3)
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.- Action Item 86-1-29 addressed the cause of observed structural damage to the
end rung of a cantilevered cable. tray from which cables exit to the equipment below. The concern was that additional load induced due to seismic conditions might jeopardize the structural integrity of the tray. Upon examination of the Electrical Cable Tray Notes and Details-General Notes, drawing No. 5-E-20-0-E-0104 Sheet 2 Revision 23, Note 12, it was observed that dropouts (to prevent violation of minimum cable bend radius requirements) are required at the end of the tray and were missing from several installations.
In order to resolve this ites, BEC ' performed a calculation which showed that maximum allowable cable weight permitted in the tray is the design loading. However, no commitment was made to install dropouts which are missing from the as-built installation.
The NRC inspection team has reviewed the approach used in performing the required calculation and other associated documentation and concurs that the cable tray structural integrity was not violated. However, since the necessity of using a dropout is to maintain the proper cable bend radius, this item will remain open pending justification, if _ dropouts are not used, or confirmation that dropouts have been installed.
(0 pen-Item 86-42-4) Action Item 86-1-30 concerned situations where cables had been looped, coiled or retrained in the safety-related and non-safety cable trays in a manner which potentially imposed a greater loading than was originally assumed in the design. As a consequence, the cable tray could be overloaded and thus, reduce the structural integrity of the cable tray support system.
To address this item, BEC performed walkdowns of the plant to identify all instances where the situa-tion existed. The actual cable fill weight was determined at these locations and compared to the original design loadings.
In most instances, the total fill weight was less than the design fill weight. In the remaining instances, further evaluation of the cable tray support system integrity was performed and found to be adequate.
In addition, construction procedure SSP-27 was revised to require documentation of all instances where looping, coiling and retraining is used so that structural integrity may be evaluated and permanent plant records noted to reflect these situations. The NRC team has reviewed correc-tive measures taken and associated documentation and has concluded that the matter was adequately addressed. This item is closed.
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- . . . ~ s= . - .. ' ... -. NRC Site Walkdown Items The following questions or items were raised by the NRC inspection team during the site walkdown on December 18, 1986. The STP project response to each item is also provided. All items have been resolved and are closed, except for items
- 4, #7, and #9(A). The licensee has committed to providing additional infor-nation for these items.
i Observation I f There is no curbing on CCW pump room C.
The flooding calculation takes credit for a curb.
STP Response Curbing is provided for this room / area. The curbing is shown on archi-tectural drawing 9M131A1030, and the design details are shown on civil drawings 9C34015 and 9C34953. The curb is at the double doors between columns E-26 and E-27.
Some curbing has not yet been installed. This item is closed.
Observation 2 l Are the traveling hoists above CCW pumps and charging pumps seismically
supported?
' STP Response Yes, the hoist rails are seismically supported. Calculation #SC086-2A is available in Houston. This item is closed.
Observation 3 Are the fan blades of the room cooler fans in the CCW and charging pump , rooms considered as internally generated missiles? The screens over ! the blades do not appear to be very substantial.
. STP Response Calculation MC5554 shows that the fan blades will be contained within the air handling unit (AHU) housing. The calculation is based on data furnished by the vendor per submittal 14926-001-4099-0019BA. This item
is closed.
Observation 4 Is the room cooler in charging pump room "B" seismically supported? STP Response ! The AHUs are seismically supported. See drawing 9C4189 and calculation CC6044, Rev. 1.
(Note that the drawing does not reflect the as-built con-
figuration of the AHU support and hence raises a question as to the validity of its seismic design. Hence this item remains open pending clarification from the applicant. Open Item 86-42-5) i r -- . -,. .- ,. -.... -,. _. - -.. -. -. - - - - -,,.- -..- . ~, -, -.....,. -.. -. -.
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, Observation 5 , There is a block wall being put in at the "A" charging pump room. Should this wall be leak tight at the floor for ALARA purposes since the doorway to the room has curbing and any contamination would be restrained by the curbing but not by the block wall.
. STP Response The block wall is a removable shield wall and will be appropriately coated.
It is not required to be water or air tight. This item is closed.
Observation 6 Are the penetrations in the AFW pump rooms to be sealed below the flood level? Is a water tight door being installed at the 10' elevation? STP Response Yes, the penetration sealing will be as shown on architectural drawing 9A1108. Yes, it is a water tight door. This item is closed.
Observation 7 There are train A (red) conduit and train B (blue) conduit in AFW pump room A02 below the flood level. How is this being handled? STP Response The train A and train B circuits are individually fused. Further infor-mation will be provided.
(0 pen Item 86-42-6) Observation 8(A) Separation violations were found on EL. 21, Trays AIXE2BTYAJ and NIXE2BTOBL Trays AIXE2BTSAJ and NIXE2BTTBL have horizontal separation distance less than one foot.
, STP Response: Project is conducting walkdown to identify separation violations. Walkdown of this area has not been completed. STP will fix this problem by covering trays. This item is closed.
Observation 8(B) Separation violation was found on EL. 10 where vertical separation distance between Tray AIXM1BTYAA and NIXM1BTJAB was found to be less than three feet.
STP Response: Project has completed walkdown for separation violations for this area and is aware of this violation. This will be fixed by covering the trays.
This item is closed.
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e , ' Observation 9(A) The color of color coded class IE cables is fading away. This may create confusion in the future since the ID tags are attached only in the ends of run and verification may be difficult in the middle of run.
STP Response: ' STP will wrap cables with appropriately colored tape where color of cable is fading.
(Confirmation with regard to the completion of this corrective action is required.
Open Item 86-42-7) Observation 9(B) Spare wires inside the class IE control cabinets have not been properly identified.
STP Response: All spare wires have been identified at the cable entrance either at the top of cabinet or at the bottom of cabinet. This item is closed.
Observation 10 Steam driven Auxiliary Feedwater Pump is class 1E but a local control panel is powered from non-1E power.
STP Response: Panel is a non-essential, local instrument mounting panel for alarm purposes.
It has no control function and is supplied as part of the turbine package. This item is closed.
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