IR 05000443/1985025
| ML20136D882 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 11/12/1985 |
| From: | Beall J, Gallo R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20136D784 | List: |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 50-443-85-25, 50-444-85-02, 50-444-85-2, NUDOCS 8511210385 | |
| Download: ML20136D882 (22) | |
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U.S. NUCLEAR REGULATORY COMMISSION Region I 50-443/85-25 Report No.
50-444/85-02 50-443 Docket No.
50-444
'CPPR-135 License No. CPPR-136 Priority Category A/B
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Licensee:
Public Service Company of New Hampshire 1000 Elm Street Manchester, New Hampshire 03105 Facility Name: 'Seabrook Station, Units 1 and 2 Inspection at:
Seabrook, New Hampshire Inspection conducted: August 28 - October 18,1985 Inspectors:
A. C..Cerne, Sr. Resident Inspector D. G. Ruscitto, Resident Inspector R. S. Barkley, Resident Inspector E. H. Gray, Lead Reactor Engineer O!8!85~
Reviewed by:
h f. Beall, Pr6 Ject Engineer,0RP date signed Approved by: k b
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R. M. Gallo, Chief, Projects Section 2A, date signed
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Division of Reactor Projects, (DRP)
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Inspection Summary:
L I_nspection on August 28 - October 18,1985(Report Nos. 50-443/85-25 &
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50-444/85-02J Areas Inspected:, Routine inspection by the resident inspectors and one region-
based inspector of work activities, procedures, and records relative to pipe
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snubber and strut erection; the Reactor Vessel Level Indication System (RVLIS)
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Installation; a pressurizer power operated relief valve (PORV) field modifica-
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tion; installation of the spent fuel storage racks; the boron concentration and
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radiation monitoring systems; and startup test program activities to include
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l the witness of preoperational tests and the review of procedures for hot func-
ticnal testing. A region-based inspector examined the diesel generator brush
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holder assembly supports for the similarity in design to an assembly which had
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failed at another site. The inspectors also reviewed licensee action on' pre-i viously identified items, including 10CFR50.55(e) reports, and performed plant
inspection-tours.
The inspection involved 260 inspection-hours of Unit 1 and L
four inspection-hours of Unit 2 activities.
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Results: One deviation from FSAR commitments (paragraph 6) was identified in that the QA program requirements specified for tubing installation of the
RVLIS and RCS wide range pressure transmitters were not commensurate with Regulatory Guide (RG) 1.97 requirements.
The acceptability of the Seabrook
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diesel generator brush holder assembly support design requires further NRC review (paragraph 4) as the uncertainty in the determination of its adequacy appears to be greater than is currently recognized, given the potential conse-i
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quences of support failure while the diesel generator is in service.
It is noted that the Seabrook FSAR Consistency Review, which commenced in
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June,1985, has identified several items requiring FSAR clarification,
revision, or addition.
Conduct of such a program appears to ba valuable, not only from a document accuracy and licensing standpoint, but also from the per-i spective of consistency between design criteria and construction application.
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As evidenced by NRC closecut of open items related to FSAR accuracy (See para-graphs 3d & e), such consistency is expected.
Also, licensee corrective action to resolve the construction deficiency involving suspect weld NDE (paragraph 3h) appears to have been handled comprehensively from a quality standpoint; particularly since an evaluation, analysis, and repair program was tracked for all affected nonsafety, as well as, safety-related items.
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DETAILS 1.
Persons Contacted During the Course of Inspection 50-443/85-25 &
50-444/85-02 T.M. Cizauskas, PAPSCOTT Program Manager (YAEC)
W.B. Derrickson, Senior Vice President (NHY)
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DeVincentis, Project Engineering Manager (NHY)
R.E. Guillette, Ass't Construction QA Manager (YAEC)
G.A. Kann, Phase 2-6 Test Group Manager (NHY)
D.C. Lambert, Field Superintendent of QA (UE&C)
D.A. Maidrand, Assistant Project Manager (YAEC)
G.F. Mcdonald, Construction QA Manager (YAEC)
D.G. McLain, Startup Test Group Manager (NHY)
J.W. Singleton, Field QA Manager (YAEC)
R.J. Sherwin, Phase 1 Test Group Manager (NHY)
Interviews and discussions with other members of the licensee and contractors management and staff were also conducted relative to the inspection items documented in this report.
Licensee Personnel in Attendance at Management Meeting on October 2,1985 (See Paragraph 14b.)
R.Alexandra, Supervisory Engineer (EBASCO)
W.J.Daley, Jr., Senior Licensing Engineer (YAEC)
A.Dufault, Supervisory Engineer (UE&C)
D. Johnson, Project Engineer (YAEC)
P.McMahon, Project Engineer (Bechtel)
G.Rigamonti, Chief Power Engineer (UE&C)
J.Stacey, Project Engineering Manager (YAEC)
Licensee Personnel in Attendance at Manage.,ent Meeting on October 16,1985 (See Paragraph 14c.
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W.B.Derrickson, Senior Vice President (NHY)
J.DeVincentis, Director of Engirieering & Licensing (NHY)
W.P. Johnson, Vice President (NHY)
G.F. Mcdonald, Construction OA Manager (NHY)
G.S. Thomas, Vice President of Nuclear Production (NNY)
2.
Plant Inspection Tours (Units 1 & 2)
The inspectors observed work activities in progress, completed work and plant status during general inspections of the plant.
The inspectors examined work for any obvious defects or noncompliance with regulatory requirements or license conditions.
Particular note was taken of the presence of quality control inspectors and quality control evidence such as inspection records, material identification, nonconforming material
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identification, housekeeping and equipment preservation.
The inspectors interviewed craft, supervisory and quality inspection personnel as such personnel were available in the work areas.
Specifically, the inspector witnessed the unpacking, movement and installation of one of the power range neutron detectors. The inspector checked for proper packaging of the detector to ensure that it was not damaged during shipment and storage.
The inspector observed the placement of the detector in its designated storage cell location.
The inspector questioned the Startup Instrumentation and Controls (I&C) engineer installing the detector how he ensured proper vertical placement of the excore detector relative to the elevation of the nuclear core.
The engineer stated that proper vertical orientation was ensured by design through the use of a specially designed positioning device attached to the detector and the position of the detector storage cell.
The inspector also observed the installation of some of the BARCO ball joints which connect the discharge of each of the main steam safety relief valves to their individual discharge stacks.
The inspector reviewed the Design Change Notice (DCN 67/0029E) which addressed the ball joint installation. The inspector has no questions concerning the design or installation of the ball joint assemblies.
l The inspector reviewed Nonconformance Report (NCR 82/763A) concerning work l
performed on the ASME Class 1 Bottom Mounted Instrumentation (BMI) tubes which were treated as non-safety related.
The inspector questioned the Startup Test Department Quality Control (STD-QC) supervisor about the NCR
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and its disposition and identified no unresolved safety concerns. The l
inspector also raised questions concerning the licensee's seismic analysis of the wheel restraints for the trolley unit of the flux mapping support system.
The question was raised due to concerns over the possible seismic l
interactions between components of the non-nuclear safety flux mapping system and the ASME Class 1 pressure boundary portions of the seal table / Bottom Mounted Instrumentation.
The issue was brought to the attention of the licensee by a letter from Westinghouse to YAEC, dated June 6,1985.
The licensee stated that the seismic adequacy of the wheel restraints was being handled by the licensee's TP-4 review program, which
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encompasses, the handling of nonsafety-related equipment in safety-related
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buildings.
The analysis will be completed prior to the turnover of con-tainment from Construction to the Station Staff.
The inspector has no l
further questions on this issue at this time.
l The inspector toured the Westinghouse Testing Analysis Mobile Unit (TAMU)
with the licensee's Startup Test Phase 2-6 Group Manager.
The TAMU unit is designed to measure the thermal and dynamic movement of portions of the plant during the upcoming Hot Functional Test (PT-40) and record the data collected. Analysis of the data is part of a requirement of Regulatory Guide 1.68 to which the licensee is committed.
In the plant the inspector observed two main steam line sensors, which will measure lateral movement of the lines due to thermal growth, and examined the wiring running from
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the sensor to the TAMU trailer.
The Group Manager discussed how the data from the sensors would be recorded and handled by the TAMU computer.
The inspector questioned the responsible Startup Test Engineer (STE) as to the number, type and location of the sensors and to the variables they will record.
The STE referred the inspector to the Preoperational Test (PT)
governing the TAMU's operation (PT 4.1 and 4.2) and indicated that over 300 sensors were installed. Since that time, the inspector has observed several other sensor installations on safety-related piping and compon-ents.
In conjunction with the TAMU, Startup QA involvement in the installation and calibration of the TAMU equipment was examined; plans for the subject PT activities involving thermal expansion measurements were discussed with both Startup and QA personnel; and adequate QA attention to the TAMU set up prerequisite to the actual conduct of the testing was verified.
The inspector had no further questions at that time concerning the installations observed.
Selected construction work, both in progress and completed, was examined during inspections of several different areas of the plant.
Post weld heat treatment operations on weld repairs to the pressurizer support ring were checked in process. Certain as-installed I&C components were spot-checked for conformance to FSAR commitments on diversity, redundancy of power supply, and IE safety classification.
The inspector also veri-fled that the use of Ke11um's grips as support mechanisms on a sample of vertical cable runs conformed to the requirements of UE&C Specification 48-2.
Civil work external to the plant buildings was observed as the inspector noted drainage piping and manholes being assembled and the trenches being backfilled. Particular attention was paid to trench backfill compaction and manhole joint grouting as well as installation of cathodic protection in nearby earthwork.
Unit 2 structures were toured with particular emphasis placed on licensee efforts to preserve and protect equipment.
Specifically, the inspector verified that all installed Unit 2 pumps and diesel generators had weathertight housings with operable motor heaters.
On August 30, 1985 approximately 30,000 gallons of salt water spilled from an open pipe into the Primary Auxiliary Building (PAB).
The licensee attributed the spill to 1) piping disconnected for if ning repair activi-ties, 2) a valve lef t open to provide ventilation for the repair work, and 3) the failure of another valve to open during testing of the thermal back-flushing operations of the ocean intake tunnel.
The inspectors con-ducted a walkdown of the areas of the PAB that were flooded, noting that major adverse impact was minimized by the absence of a substantial amount j
of electrical equipment in the spray path.
The inspectors spot-checked l
clean-up activities as they progressed and confirmed that demineralized water and spray cleaners were used only where authorized. Nonconformance Reports (NCR) 82/762 A&B were reviewed for the adequacy of scope and disposition, record of the chloride contamination levels, and engineering evaluation of any potential damage to the stainless steel components.
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After cleanup operations were completed, the inspector reexamined specific components wetted by the saltwater (eg: conduit for valve, SW-V-34) to verify acceptable conditions.
Licensee corrective action to the salt water spill was immediate and appeared adequate to prevent any long-term adverse effects on safety-related components. The use of the permanent plant drains, sumps, and pumps served to confirm some of the assumptions made in the UE&C Moderate Energy Line Break (MELB) study for a water spill in the PAB. On September 18,1985, this event was reported to the NRC as a potential construction deficiency under 10CFR50.55(e).
Further NRC inspection follow-up will be provided after receipt of the licensee's written report on this accidental spill (reference: CDR 85-00-15).
With regard to all of the above independent inspection and plant inspection-tour items, no violations or unresolved safety concerns were identified.
3.
Licensee Action on Previously Identified Items a.
(Closed) Unresolved item (443/82-03-07 & 444/82-03-06): Question on the consideration of shear cone influence in the Hilti Kwik-bolt testing data.
Combined NRC Inspection Report 50-443/82-11 &
50-444/82-06 expanded the question regarding the 1/2" Hilti bolt spacing to include consideration of interacting cones of influence for other bolt sizes at corners and the opposite faces of walls.
With respect to the original question, tests were performed at the Seabrook site on June 9,1982 which confirmed the validity of using full design load allowances for 1/2" expansion anchors spaced at a ten diameter center-to-center distance. Additional testing at the Hilti Test Facility in Tui a, Oklahoma was condu::ted for Hilti anchors used on corners and on the opposite faces of walls and slabs.
The inspector reviewed licensee documentation of the subject test results in response to a QA request for follow-up action to NRC inspections (ie: Blue Sheet No.48).
From the opposite face tests, it was determined that the average ultimate tensile values (AUTV) of the tested bolts were consistent with the Seabrook/Hilti baseline data.
For the corner tests, which used representative site concrete, the test results Indicated that no reduction in the credit taken for AUTV is required for anchors installed at greater than three diameters distance. On outside corners where the Hilti anchors are installed at less than three diameters distance, a 10% reduction in AUTV was noted, but when evaluated in light of the inherent conservatisms of expansion anchor design this reduction was found to be acceptable.
The inspector reviewed the assumptions made by engineering in ultimately concluding the acceptability of expansion anchor proxi-mity, based upon the test results and further analysis.
The inspector examined summaries of the wall and corner test data and
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determined that the licensee had adequately investigated the concern on Hilti Kwik-bolt overlapping cones of influence. No inconsistan-cies in the engineering conclusions were identified. The inspector has no further questions on this issue and considers this unresolved item closed.
b.
(Closed) Deviation (443/83-17-01): ' Adequacy of vibration collar installation on the diesel generation SKv bus duct.
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reviewed the following documents to verify that a new diesel l
generator bus duct collar had been designed, procured, and installed such that an adequate expansion / vibration joint is in place, per Specificution 144-1.
Engineering Change Authorization (ECA) 54/2869C
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Startup Work Requests (WR) DGG-0184 & 0188
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Fischbach Inspection Report (IR)47-133
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UE&C Change Order No.35 to Purchase Order 144-1
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The inspector examined the new collar installation on the Unit 1
"A" diesel generator and found the configuration to differ from the-sketch provided by ECA 54/2869C with regard to the position of certain collar components.
UE&C nonconformance report (NCR) 54/5868A was then written to document the fact that certain bus duct sections had been installed upside down from the supplier's (Brown-Boveri)
illustrated design sketches (ie: Foreign Prints, FP 34115 & 34116). A bus duct / terminal box misalignment condition prevented installation as detailed, resulting in an installation with certain sections inverted.
The inspector reviewed the disposition to NCR 54/5868A which accepted the installed conditions as-is and which justified acceptance based upon measured gaps, providing for both the vertical and horizontal movement of the bus duct enclosure.
Since the allowable vertical /-
horizontal deflection of the collar with bus duct enclosure was determined to be acceptable as installed, no rework was necessary.
Design conditions per Specification 144-1 have now been achieved.
The inspector has no further questions on this item and considcrs it closed.
c.
(Closed) Unresolved item (443/83-20-02):
Electrical interconnections between redundant divisions. Deviations from FSAR commitments with regard to physical separation requirements for electrical trains /
components had already been documented by the licensee in letters (reference: SBN-587 & 671) to the NRC Division of Licensing (NRR),
dated December 1,1983 and June 20,1984, respectively. On September 24-26, 1985, personnel form NRR Power Systems Branch (PSB) conducted a Seabrook site audit. Among the items reviewed and discussed were the documented electrical separation deviations, which PSB is evaluating on a case-by-case basis.
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Since NRR is tracking the subject deviations as a licensing issue in conjunction with other SER confirmatory action items in the electrical area, the inspector considers this issue closed from an inspection perspective.
Routine inspection of this area will confirm any additional licensee commitments for corrective action, if required.
d.
(Closed) Violation (443/84-07-06):
Inadequate quality control of certain component installations. As documented in NRC inspection report 443/84-20, corrective action in the form of a procedure revision was found to adequately address the concern regarding the mishandling of QC controls for safety-related equipment installation.
However, the FSAR had not yet been amended to reflect an updated list of the safety-related equipment which would be affected by such procedural controls.
' Amendment 55 to the FSAR which was submitted to the NRC in July,1985, contains amendments to FSAR Tables 3.2-1 and 3.2-2 for the seismic and safety categorization of systems and components. Additionally, the licensee FSAR Consistency Review has identified other changes and additions the subject FSAR Tables, which will be included in the Amendment 56, currently in process.
Based upon previously documented procedural changes and licensee efforts to keep the affected FSAR Tables current with the proper classification of the Seabrook components and systems, the inspector determined that corrective action to the violation has been adequate and considers this item closed, e.
(Closed) Unresolved item (443/84-17-03): Discrepancy between safety injection (SI) system description and the FSAR/ procedural requirements for operator action during SI recirculation mode.
Specifically, the inspector questioned the lineup of the two valves in the Residual Heat Removal (RHR) cross-connected discharge piping (ie: RH-V21 & RH-V22) when the transfer is made from cold leg recirculation to hot leg recirculation following the SI phase of Emergency Core Cooling System (ECSS) actuation.
Based on correspondence from Westinghouse, it is equally acceptable to open either or both of the subject valves during het leg recirculation.
The licensee chose to plan on opening both valves which was the lineup already delineated by the Operations Procedure (0S1313) for switchover to hot leg recirculation. The licensee modified Section 6.3.2.5.c and Figure 6.3-2 of the FSAR in Amendment 55 to agree with this lineup.
The inspector identified no problems with the technical justification for this valve lineup.
The licensee also submitted a response to Region I, (reference: PSNH Letter, SBN-789, dated April 8,1985) as required by the NRC transmittal letter for combined inspection Nos. 50-443/84-20,
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50-444/84-08, explaining that the FSAR discrepancy was caused by an isolated human error.
The licensee also stated that a task team was-set up to review the interface between Westinghouse and UE&C and to make recommendations to management for improvement of that interface.
Based on the licensee's corrective actions, the inspector considers this item closed.
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(Closed) Unresolved item (443/84-17-04):
Inconsistencies in sealant specification criteria and Mechanical Energy Line Break (MELB) study commitments. The inspector verified that the UE&C Specification 249-7 for penetration sealant work which had been revised (Rev. 5) on July 24,1985, now delineates specific criteria for seals performing a fire barrier function and for those required to withstand a hydro-static pressure head. Also, Revision 4 to the MELB study, issued in March,1985, corrected the description of the plate seals welded to the spare penetration sleeves in the common wall between the "A" and
"B" train equipment vaults.
Consistency between the MELB and the-design /as-constructed details has thereby been established.
Based upon licensee actions to resolve the subject specification /
drawing and MELB ' study inconsistencies and to better clarify and quantify the technical requirements of the penetration sealant work, this item is considered closed.
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(Closed) Construction Deficiency Report (CDR 82-00-13): Cold pulling
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of pipe. This CDR had been closed in NRC inspection report (IR)
443/84-13 and subsequently reopened in IR 443/85-17 based upon the determination that certain piping lines may have been inadvertently omitted from the original cold pull review. A Final 10CFR50.55(e)
Report (SBN 869), dated September 10, 1985, from the licensee to Region I, documented completion of the ASME Class I piping review.
The licensee indicated that only the pressurizer surge line (RC-49)
had been omitted from the previous analysis of cold pull.
The inspector examined a sample of the piping packages used by
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engineering in its review for potential cold pull applicability.
The inspector verified the proper use of the dates of fit-up and tack welding as the times when cold pull could have occurred and confirmed consideration of the fact that closure welds were the most likely I
candidates for cold pull. Cold pull of the pressurizer surge line,
-beyond the limits of available piping alignment clamps, had already been investigated by the licensee and evaluated by the NRC in the follow-up of an allegation documented in IR 443/84-17. The allegation was not substantiated.
The cold pull of piping within the limits of alignment clamps in use at Seabrook has been analyzed for several ASME Class 2 & 3 piping systems and will also be considered in the As-Built Reconciliation i
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Program for. Class 1 systems.
Thus, licensee corrective action on this CDR has not only provided a detailed review of systems for potential' cold pull, but also an analysis of the effects of cold pull within the limits of the pipe alignment clamps available to the craft. This analysis has yielded no evidence of detrimental effects.
The inspector has no further questions on licensee corrective action and considers this cold pull deficiency closed.
h.
(Closed) Construction Deficiency Report (COR 83-00-08): Suspect NDE examinations. The Final 10CFR50.55(e) Report (SBN 603) on this deficiency, dated December 21, 1983, documents the status of all 2,399 suspect items. The inspector reviewed this status against the detailed disposition of each item per Pullman NCR 4490, with all its supplements. A question developed on the status of two particular welds which had been originally categ'orized as unaccessible and accepted by engineering evaluation. A subsequent disposition to these welds indicated that they had indeed been reinspected and accepted on that basis. Thus, the status of all 2,399 items, only about a third of which are safety-related, appears to have been accurately documented and-transmitted to the NRC.
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NRC inspection has monitored the progress of licensee reinspection, repair,.and engineering evaluation over the course of the licensee's implementation of its corrective measure program.
Such NRC follow-up is documented in the following inspection / meeting reports:
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Combined Inspection Report Nos. 443/83-06 & 444/83-06,
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Combined Meeting Report Nos. 443/83-10 & 444/83-07, paragraph 4
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Combined Inspection Report Nos. 443/83-18 & 444/83-14, paragraph 4 and Attachment 1 Inspection Report No. 443/85-19, paragraphs 2.2 & 5 and
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Attachment 1 NRC independent reinspection of a sample of both those welds reexamined by the licensee and accepted without repair and those welds repaired by the licensee revealed no deficiencies or-inadequacies in the licensee reinspection program.
Generic licensee corrective actions have included 1) the evaluation of weld NDE performed by Pullman technicians other than the one involved in the suspect welds and 2) generation of new procedures for licensee / contractor verification of an employee's background and technical qualifications.
YAEC third party verification activities have been established for contractor / supplier NDE operations and record F'
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Overall, NRC inspection of the licensee corrective actions, both specific to the suspect welds and generic to the adequacy of the NDE controls, has determined that the licensee program to address the concerns raised by this CDR, to correct all hardware deficiencies, and to prevent a future recurrence of this problem has been responsive and complete.
This construction deficiency is closed.
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(Closed) Construction Deficiency Report (COR 84-00-19): Main Steam l
Safety Valve Ring Setting Deficiency. The licensee stated in their final report to the NRC, Region I (SBN 863, dated August 27,1985)
that the optimum ring settings for the main steam safety valves were determined by testing at Wyle Laboratories to be (-25) notches for the lower (nozzle) ring and (+25) notches for the upper (guide) ring.
These settings assured that the valve would pass design steam flow over the full range of discharge back pressures. The inspector reviewed the Crosby Valve and Gage Company (manufacturer of the
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safety valves) Field Service Report (F.0.No. J50139) and confirmed that the ring locations were adjusted to these settings.
The inspector randomly selected three of the safety valves (Main Steam Safety Valves MS-V7, V8 & V9) and verified that they were properly installed, the new nozzle and guide ring settings were stamped on the valve body, and a tamper proof seal was installed on the nozzle and l
guide ring adjustment screws.
Based on the Wyle test results, which confirmed that the present ring settings allow the valves to relieve t
rated steam flow, and the overall licensee's evaluation and
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corrective action, the inspector considers this CDR closed.
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(Closed) Construction Deficiency Report (CDR 85-00-08): Questionable grease found in fan motor bearings. This item was initially reported
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as a potential 10CFR50.55(e) report because the color (brown) of l
the grease found in certain safety-related fan motor bearings
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implied that the grease was lithium-based. The subject Class IE motors had been qualified with a polyurea-based grease (normally green in color) of a type recommended by the manufacturer.
Subse-quently, the licensee sent samples of the questioned grease to Chevron Research for an infrared spectra analysis.
The results revealed that the subject grease was in fact a polyurea-based grease, as requried. Thus, no deficiency exists and the licensee indicated to the NRC by letter (SBN 875) dated September 30,1985 that this issue was not reportable per 10CFR50.55(e).
The inspector examined documents from Chevron U.S.A., Inc. confirming the adequacy of the subject grease.
This CDR is closed.
4.
DieselGeneratorBrushHolderAssemblySupport(Units 1&2)
As documented in the NRC inspection report 443/85-20, the licensee had initiated evaluation of the generic applicability to Seabrook of a metallurgical fracture in the support assembly holding the electrical
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contact brushes on the diesel generator exciter ring, which was identified at the Millstone Unit 3 plant. This problem was reviewed at Millstone by a Region I specialist inspector.
The NRC inspector concluded that the configuration (or design) of the support mounting hexagonal nut component was a significant contributor to the failure. This component has an internal 5/8" diameter threaded hole overlapping the base of the mounting threaded extension, such that the fatigue life is significantly reduced below that of a 5/8" diameter steel rod. The hexagonal nut configuration of the Millstone and Seabrook diesels were found to be almost identical because they were supplied by the same manufacturer using a similar design.
On September 11,1985 the NRC specialist examined this area of the Seabrook Unit 1, "B" diesel.
The brush holder assembly support was observed and the status of startup testing, vibration measurements and hours of operation were reviewed.
The NRC specialist concluded that the hexagonal nut component of the Seabrook Diesel Generators A & B brush holder assemblies may have an indeterminate time to failure. This is a result of the uncertain stress level in the area of the base of the drilled hole and the threaded extension. The details of the failure and component configuration were reviewed with YAEC Engineering.
Subsequently,'the licensee contacted Louis-Allis, the generator manufacturer, to determine the minimum requried metal thickness in the area of interest of the-hexagonal nut.
Louis-Allis has calculated this dimension to be 1/8".
Using radiographic examination, the licensee has determined the as-built minimum metal thickness for this area in question on the Seabrook Unit I diesel generators to be 9/64".
Pending further NRC review of the accuracy of the radiographic examination and the adequacy of a measured margin of safety of 1/64", this issue remains unresolved (443/85-25-01 & 444/85-02-01).
5.
Pipe Supports - Snubbers & Struts (Unit 1)
The inspector examined the in process and as-built condition of certain snubber and strut supports, checking the adequacy of procedural controls and QC coverage for the in process work, and the consistency of the as-built work to the design details for the completed supports. The following supports were inspected on a sample basis:
1304-RM-9
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15-SG-6
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30-SG-6
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44-SG-6 59-SG-6
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Multiple Support (MS) 45-SG-3/45-RM-4 Steam Generator (SG) upper lateral support snubbers
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The inspector noted that for the multiple support,
'nstallation of the snubber (45-RM-4) followed completion and as-built 1. Section of the guide (45-SG-3). However, the actual SG-3 support point was affected by the RM-4 installation and was approximately 1/2" off center-from the design drawing detail.
The inspector discussed this situation with licensee engineering personnel and calculations were subsequently performed to confirm the adequacy of the as-constructed condition. The licensee has also documented this 45-SG-3/RM-4 interface as an example of a load point moving off its frame and a special attention item in UE&C Technical Guideline, TG-1, for.the Visual Inspection of Hot Piping for Hot Functional Testing (HFT).
TG-1 training of the engineers involved in the cold and hot pipe walkdowns was also conducted.
The inspector also questioned the acceptability of loading the wall brackets for the steam generator upper lateral support snubbers during HFT without grouting of the brackets and anchor bolts having been completed. A review of the snubber bracket installation field instructions (Pullman FI 120 &
436) and discussion with engineering personnel revealed that the completion of grouting after HFT would result in no adverse impact on either the snubber components or the test measurements.
The inspector confirmed that Westinghouse structural engineering personnel concurred in this position regarding the lack of grout.
On certain strut installations, particularly those utilizing ITT Grinnell Figure 215 pipe clamps, the inspector noted sufficient clearance for the strut paddle to move laterally in the pipe clamp bracket, apparently because the washers installed on either side were not thick enough to serve as spacers.
Since Pullman Procedure JS-IX-6 requires spacers in strut installations to ensure the ball bushing remains centered on its load pin, the inspector questioned the noted use of the thin washers as spacers in the subject strut asemblies.
It was also noted that NRC Circular 81-05 specifically addresses the problem of too large a gap between the strut paddle and the clamp bracket, which may allow the paddle to slide over and become disengaged from its bushing.
In response to NRC concerns in the area of these strut spacer
installations, UE&C engineering personnel have requested ITT Grinnell by l
memorandum (SM 13300), dated October 7,1985 to identify the function of the
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washers, to quantify a permissable gap, and to address any Circular 81-05 l.
or eccentricity concerns.
Pending review of the ITT Grinnell response and l
evaluation as to whether rework is required on those struts currently with
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large gap clearances, this item remains unresolved (443/85-25-02).
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The inspector noted adequate construction control for the protection of the snubbers installed in the field, thru the use of padlocked wooden enclosures around each snubber and the issuance of a Project Notice to all
construction personnel cautioning them to not apply any external loads to installed snubbers and to excercise care when working around the snubber
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assemblies. The inspector also checked the adequacy of procurement control and records for certain snubber and component standard support material by questioning the authorization to use ASTM A434, Grade BD bar stock in light of ASME Section III, Subsection NF material recuirements. YAEC Blue Sheet 094 was written to document the NRC questions in this area and was responded to by engineering personnel.
ASME Code Cases 1644 and N71 have been invoked by UE&C Specification and authorize use of the questioned material. Also, in line with RG 1.85 guidance concerning use of materials exhibiting an ultimate tensile strength greater than 170 Ksi, the licensee indicated that thru specification and control of material hardness requirements (ie: Brinell Hardness Numbers), tensile strengths under 170 Ksi are ensured.
With regard to the pipe support inspections conducted during the course of
this report period, no violations were identified.
6.
Reactor Vessel level Indication System (RVLIS) - Unit 1
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a.
The inspector conducted a walkdown of selected RVLIS lines -- both from the reactor pressure vessel head branching to the "A" & "B" train low pressure lines thru their respective bellows, capillary tubing, containment penetrations to the level transmitters and from
"A" & "B" train high pressure lines coming off two spare in-core instrument tubes at the seal table to the same level transmitters (LT-1.311 & 1312 for "A" train and LT-1321 & 1322 for "B" train). The inspector noted that the wide range pressure transmitters (PT-405 for
"A" train and PT-403 for "B" train), used for certain safety-related valve interlocks discussed in Section 7.6 of the FSAR, tapped off their respective trains of high pressure RVLIS lines and were located near the LTs in the lower electrical penetration area.
Per FSAR Section 7.6.2.1 and NRR Request for Additional Information (RAI)
420.57, the inspector confirmed that PT-403 & 405 were " diverse" in that different manufacturers supplied each of the two transmitters.
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As illustrated on the RVLIS Control Loop Diagrams (M506645 & M506646),
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both PT's and all the RVLIS LTs are designed as Class IE instruments.
The inspector noted that work in progress on the RVLIS lines in the seal table area was being QA controlled by the UE&C QAS-5 program, a quality assurance program scoped for nonsafety-related components used in seismic applications. Questioning by the inspector revealed that because the capillary tubing was installed to ANSI B31.1, and not ASME Code, requirements, a graded QA program typical of other seismic B31.1 installations was being applied.
The inspector noted, however, that the licensee in Section 1.8 of the FSAR had committed to USNRC Regulatory Guide (RG) 1.97 and had documented in a letter to the NRC (SBN 864), dated August 30,1985 that the process signals provided by PT-403 & 405 and LT-1311,1312,1321 & 1322 were all RG 1.97 Design Category 1 variables, for which the full nuclear quality (10CFR50, Appendix B) QA program must be applied.
This is consistent with the design of PT/LT instruments as Class 1E
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and with the safety related application of both the RVLIS and wide range pressure signal usage, but was inconsistent with the application of the QAS-5 program for their QA/ installation.
The QAS-5 program meets the QA requirements of a graded program per Design Category 2 of RG 1.97, but fails to encompass all Design Category 1 QA standards.
Therefore, the inspector informed licensee management personnel on October 18,1985, that a deviation existed in the licensee's failure to meet QA commitments to RG 1.97 with regard to the RVLIS installation. An NRC unresolved item had previously existed in the area of RG 1.97 application and commitment. While the inspector's review of the SBN 864 letter serves to confirm the licensee's position
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on compliance with RG 1.97 sufficiently to close the open item (443/84-13-05), a new item is hereby opened as a result of the deviation (443/85-25-03).
b.
During inspection of the RVLIS lines, the inspector checked the tubing containment penetration assemblies and noted that fillet seal-welds were designed (reference: UE&C drawing SL-X77-01) as containment penetration boundary welds in a sleeve-to-socket fitting configuration.
However, it appeared that the welds were undersized per the requirements of ASME Section III, Subsection NE, Figure 4244(e)-2. Discussion with UE&C-engineering personnel revealed that while the design generally met the requirements of Class MC, Category D joints, the subject fillet weld size had been erroneously based on the thinner dimension of the RVLIS tubing wall instead of the thicker wall dimension of the socket fitting.
Further review of similar containment penetration designs by engineering personnel indicated that the welds on 13 tubes i
in six different penetrations did not meet ASME code criteria.
However, calculations were done which verified the stru"tural adequacy of these welds.
The inspector reviewed-documentation of licensee plans to redesign the affected welds to meet ASME Code requirements.
Based upon this corrective action, the previous determination of the structural adequacy of the existing welds, and the identification of no other problems on the RVLIS containment penetrations, the inspector has no further questions at this time.
c.
Additionally, the inspector witnessed hydrostatic testing of a section of RVLIS tubing.
This section of tubing is part of the reactor coolant-pressure boundary and was not tested during the reactor coolant system hydrostatic test of April 5,1985.
The inspector verified that the
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(about 3100 psi) and for the required time length (10 minutes). The inspector verified proper quality control and Kemper American Nuclear
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Inspector (ANI) coverage. The inspector has no questions concerning the test.
With regard to the RVLIS inspection, no violations were identified; however, a deviation from FSAR commitments, as documented above, was note,
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7.
Power Operated Relief Valve (PORV) Modification (Unit 1)
The inspector witnessed an in process modification to both PORV's in the piping lines off the head of the pressurizer and verified authorization per ECA 08/109569C and control of the work under the direction of Westinghouse and Crosby engineers per the rules of ASME Section XI. As allowed by the ASME Code, welding was accomplished utilizing site procedures qualified to ASME Section III criteria.
The modification entailed the installation of tubing from the solenoid to the valve bonnet to equalize pressure on both sides of the valve seat during valve closure.
Also, gaskets on this tubing at the valve body / bonnet interface were being replaced by welds to seal off a potential RCS leak path. The inspector confirmed QA coverage of the subject modification.
Hydrostatic test requirements for the above work were delineated in ECA 08/109569C.
However, the listed valve set pressure of 2385 psig was 50 psi higher than listed as the PORV setpoint in Table 5.4-10 of the FSAR.
Upon further review, the inspector determined that the Westinghouse Precautions, Limitations & Setpoints (PLS) document (Revision 2, June 1985) had changed the setpoint to the higher value, but that the FSAR had not yet been amended.
UE&C engineering subsequently issued ECA 08/110849 to document the required revision to the FSAR, based upon the current PLS setting.
The inspector had no further questions on the PORV modification work. No violations were identified.
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8.
Walkdown of Piping Systems (Unit 1)
The inspector conducted a walkdown of three piping systems, evaluating as-constructed conditions against design details.
Residual Heat Removal (RHR) Train "A" was checked against design drawings from the point where the suction line enters the equipment vault to where the various discharge lines exit the vault downstream of the heat exhanger.
Particular note was taken of instrumentation and small-bore piping runs.
The Startup Feed Pump (SUFP) discharge line was traced for seismic support installation.
The Service Water (SW) cooling tower piping and instrumentation were checked within the tower structure to verify proper as-built configura-tion.
For the SW lines, discussion was held with a startup QA inspector on support flange welds recently completed on some motor operated butterfly valves. One concern identified during this inspection is discussed below.
A review of piping and Instrumentation Diagrams (P&ID) F202076 and F202079
' indicated that the startup feed pump and steam generator wet layup recircu-lation pump discharge piping lines are Non-Nuclear Safety (NNS), and not seismic Category I (NNS-I). However, the Seabrook FSAR Chapter 6.8 indicates that this piping is seismically supporte _
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The inspector conducted a walkdown inspection of the piping in the turbine building, main steam and feedwater pipe chase and the emergency feed pump buildings and verified that the supports are seismically designed.
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addition, the contractor Piping Support Detail Drawings, Expansion Anchor Field Process Sheets & Support Weld History Records were examined and verified to be NNS-I. The inspector interviewed licensee Quality Assurance and Engineering personnel and determined that a program has
.been initiated to update drawings of ANSI B 31.1 piping based on an NRC inspector question concerning the handling of ANSI B31.1 work packages in December 1984.
Implementation of this program is underway and will identify and correct drawing discrepancies such as those differences noted by the inspector to exist between the FSAR and P& ids.
Since the support hardware has been properly designed and installed under the correct program and since corrective steps are already in progress to correct the subject P&ID's, the inspector has no further questions on this issue. No violations were identified.
9.
Installation of High Density Spent Fuel Storage Racks (Unit 1)
The inspector witnessed installation of three of the six high density spent fuel storage racks in the spent fuel storage pool. Adequate QC coverage of the work and adherence to the installation procedure was observed.
The inspector confirmed 1) that all of the funnel plates at the entry of the storage cells along the perimeter of rack number three (3)
which were bent during shipping and installation were bent back into position and 2) that supervisory personnel were aware that each of the racks had a unique location in the pool due to the positioning of the
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neutron poison material on the perimeter storage cells.
Proper leveling was verified during installation in spite of problems encountered with the installation procedure leveling instructions and was confirmed by visual examination and by review of the results of a dummy fuel assembly drag test on the storage cells.
The inspector reviewed the Westinghouse Quality Release, as well as the design specifications (P.O.NSS-229) for the spent fuel racks.
No problems were identified.
The inspector questioned the removal schedule for the Boraflex surveillance capsules installed on the side of the storage racks.
The capsules contain samples of the neutron poison material and must be periodically removed and analyzed to ensure that the poison material has not leached out of the binder material which encapsulates it.
The integrity of the poison material is essential to ensure subcriticality of-the nuclear fuel when stored in the spent fuel pool. The inspector was
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informed that the surveillance program had not yet been developed and will l
not be completed until prior to the end of the first cycle of reactor l
operation.
This is acceptable due to the fact that the poison material l
can not leach out of its binder material until it is immersed in water.
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The spent fuel pool will not be flooded until the first refueling outage.
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Based upon awareness on the part of Reactor Engineering personnel of licensee commitments regarding the surveillance capsules, the inspector was satisfied that a surveillance program would be developed at the appro-priate time. The inspector had no further questions.
No violations were identified.
'10. Boron Concentration and Radiation Monitoring (Unit 1)
The inspector verified the design versus as-built configuration of the Boron Concentration Measuring System.
The Boron Concentration Measuring System determines the boron concentration of the reactor coolant system letdown flow using a four curie americium-beryllium neutron source and a neutron detector.
Specifically, the inspector verified that the piping and valve layout to and from the measuring system agreed with the design drawings. The inspector checked the electrical wiring to the solenoids of two of the air-operated fail-closed ASME Section III, Class 2 isolation valves (V541 & B478) on the inlet lines to the monitor.
The inspector also found the wiring tags to agree with the UE&C Conduit and Cable Sche-dule Routing Program (CASP) which addresses the safety classification and routing of these cables.
The inspector raised questions concerning the neutron radiation fields in the' area of the unshielded monitoring unit.
The licensee's Radiation
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Protection Manager (RPM) informed the inspector that the neutron radiation fields around the monitoring unit were negligible because the neutron source is shielded by a pool of water in the unit (the water shield is also an integral part of the operation of the monitor).
Furthermore, the RPM stated that even if the water was removed (and there are design features to prevent this), the neutron radiation fields will still remain low.
The inspector had no further questions on this item.
The inspector also verified the design versus as-built configuration of the Failed Fuel Radiation Monitoring Unit (RM-SKO-88) and confirmed the installation and proper tagging of the two ASME Class 2 isolation valves on the monitor's inlet line.
l The inspector examined the installation arrangement of the radiation monitor, which is a non-nuclear grade electrical component (ie: non-class-1E).
The inspector questioned how intense the radiation fields around the unshielded inlet and discharge lines of the monitor would be since they will contain highly contaminated reactor coolant. This situation was discussed with a Health Physics supervisor who agreed to calculate the radiation fields in the area of the inlet and discharge < lines with reactor coolant radioactivity levels up to the Technical Specification (TS)
limits. The calculations showed that even at TS limits, the radiation levels should remain below the level of a high radiation area (100 millirem / hour). Given that the TS limit on reactor coolant activity can
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only be reached if there is significant (up to 1%) fuel cladding damage, the radiation levels in the area of the monitor inlet and discharge lines should normally be significantly lower than this level.
With regard to all of the above inspection items and issues, no violations or unresolved safety questions were identified.
11. Preoperational Testing (Unit 1)
The inspector reviewed the following preoperational test (PT) procedures for conformance with FSAR Section 14, other pertinent technical sections of the FSAR, and the applicable regulatory guidance:
1-PT-40 (Revision 0) - Hot Functional Testing
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1-PT-40.1 (Revision 0) - Pressurizer Level and Pressure Control
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1-PT-41 (Revision 0) - Integrated Plant Heatup for Hot Functional Testing Sections of the procedures were sampled with respect to FSAR system descriptions and setpoints in light of the actual status of plant components listed in the procedure as required for test conduct.
Low Temperature Overpressurization Protection (LTOP) procedural steps were evaluated against specific design requirements for the required equipment, including the PORV's and the wide range RCS pressure and temperature transmitters. The inspector reviewed the prerequisites for hot functional heatup and testing to ensure proper consideration of necessary valve lineups, lifted lead and jumper status, and controls over the physical protection of installed systems and components.
The inspector also witnessed preoperational testing of the thermal barrier cooling system (PT-16.2) which included demonstrations of pump performance and system controls to include indications and interlocks locally, in the control room and at the remote safe shutdown panels (RSSP). Discussions were held with the licensed control room operators, auxiliary operators, the system test engineer (STE) and the startup QA inspector. Adherence to the test procedure was confirmed and the STE/QA/ Shift Crew interface was noted. The inspector witnessed the test briefing. Calibration and range of test gages were checked as satisfactory.
Following a review of the Startup Feed Pump Acceptance Test Procedure (AT-1.3), various portions of the test were witnessed.
Pretest pump runs as well as electrical portions of the test were observed locally in the turbine building and in the essential switchgear room.
Instrumentation and controls (I&C) testing relative to AT-1.3 was observed in the control room.
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Additionally, the inspector witnessed several individual system and integrated plant operations in support of testing. These included Emergency Diesel Generator (EDG) instrumentation and vibration test runs, service water cooling tower performance runs, pre-Hot Functional Test (HFT) filling and venting of the Reactor Coolant System (RCS) and Reactor Coolant Pump (RCP) venting. Additionally, service water (SW) and Circulating Water (CW) system runs were observed including use of the heat treatment discharge flow path and intake structure recirculation following chlorine treatment.
The inspector observed overall control room operations, including shift turnovers, routine and test-related operations and general shift crew performance.
Discussions were held with several shift crew personnel, shift test directors (STD) and STEs.
Since the performance of the shift crew in coordinating and controlling plant operations has become more complex in support of the diverse preoperational testing activities in progress, four different shift crews were closely monitored and found to be working well together.
Turnover records, control room operating and test procedures and drawings were reviewed periodically. The unit shift supervisor (USS) was observed to be in charge of the control room and access to the main control board appeared to be adequately controlled.
Communications, noise levels, and coordination between the operations shift crew and the Startup Test Department were observed, evaluated, and judged to be adequate during this phase of testing activity.
The inspector observed a sample of control room procedural operations, including solid plant pressure control with RHR flow through PCV-131, plant makeup, thermal barrier cooling system operations, EDG operations, SUFP operations, main condenser vacuum and condensate operations, RCS filling and venting and RCP operations.
Primary Component Cooling Water (PCCW) and Service Water (SW) system controls were also observed.
With regard to all of the above preoperational testing procedures reviewed, activities witnessed and control room operations observed, no violations were identified.
12. Unit 2 Construction Affecting Unit 1 Completion Seabrook Unit 2 remains in an " indeterminate" status with preservation, protection and maintenance of equipment / structures being the primary construction activity currently in progress.
However, certain construc-tion work must be completed on the Unit 2 side of the plant to support Unit I construction and testing completion. As an example, Unit 2 piping and supports in the common service water cooling tower were designed to be completed to a certain point consistent with the structural and opera-tional adequacy of the connected Unit 1 piping.
This was confirmed during a previous inspection.
During this inspection, the inspector reviewed the design of the west air intake and associated piping for the Control Building ventilation system.
Physically located in a Unit 2 area, UE&C drawing F624260 requires a continuous security fence around and over the air intake enclosure for
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protection outside the Unit 1 protected area. Also, the air line piping
.in Unit 2, which ultimately is designed for routing to the upper elevations of the Unit 2 Diesel Generator building before connecting to the Unit I crossover line, will receive a jumper spool piece tapping the west air intake directly into the Unit I system.
Since this run of piping parallels the south end of the Unit 2 Turbine Building and penetrates a retaining wall between Unit 1 & 2 which was not seismically designed or constructed, the question of how this safety-related piping, necessary to Unit 1 operation, would be protected was raised to licensee engineering personnel.
The licensee is evaluating two options to address this concern.
The first would involve backfilling the area locally and thus meeting FSAR commit-ments. The second option would provide for design of a buttress type support for the Unit 2 turbine building in the area of the affected pipe.
Since a decision as to which option will be chosen has not yet been made and work has thus not started, the inspector questioned how QA controls would track this item to resolution.
Subsequently, this issue was entered into the YAEC Future Verification Item List (item no.276), ensuring future follow-up by QA personnel.
Based upon QA tracking of this item and L
engineering recognition of the need for resolution prior to Unit 1 operation, the inspector has no further questions on this issue at this time.
With respect to the above Unit 2 construction activities, no violations were identified.
13. Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations or deviations. Unresolved items disclosed during the inspection are discussed in Paragraphs 4 & 5.
Management Meetings At periodic intervals during the course of this inspection, meetings were held with senior plant management to discuss the scope and findings of this inspection. An exit meeting was conducted on October 18,1985 to discuss the inspection findings during the period. During this inspection, the NRC inspectors received no comments from the licensee that any of their inspection-items or issues contained proprietary information.
No written material was provided to the licensee during this inspection.
On October 2,1985, a meeting was held in the Region I office in King of Prussia, Pennsylvania by mutual licensee / Region I agreement to discuss the conduct and status of cable tray testing being performed by the licensee L
to evaluate the feasibility of implementing design changes with respect to
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the axial tray bracing.
Licensee representatives (see paragraph ib)
l discussed the favorable results achieved on two phases of tests already
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21 performed and the plans for a third phase of testing. Questions by both Region I and NRR personnel as to the criteria used in modeling the tests after general plant cable tray configurations and as to the relationship of this testing to other testing done for the purpose of increasing allowed seismic damping values were discussed.
The NRC personnel indicated that inspection and licensing follow-up of these testing acti-vities and results would be conducted routinely in the future.
The licensee committed to providing the test results to the interested NRC parties.
On October 16,1985, a meeting was held in the Region I office in King of Prussia, Pennsylvania by mutual licensee / Region I agreement to discuss the Seabrook project status and schedule.
New Hampshire Yankee Management personnel (see paragraph 1c.) presented information on startup activities, operational readiness, and current status of the Employee Allegation Resolution (EAR) program activities, follow-up,of NRC open items,and the licensee conduct of a FSAR Consistency Review.
NRC questions related to long-range planning, plant staffing, and new management initiatives for
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Unit 1 completion and 1 ensing. Both the licensee and NRC management agreed that such meetins, were beneficial on a periodic basis to provide for information interchange and for a consistent understanding of both regulatory developments and current project status.
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