IR 05000324/1987018
| ML20236D572 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 07/21/1987 |
| From: | Blake J, Chou R, Coley J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20236D562 | List: |
| References | |
| 50-324-87-18, 50-325-87-18, IEB-79-02, IEB-79-14, IEB-79-2, IEB-83-07, IEB-83-7, IEIN-83-01, IEIN-83-1, NUDOCS 8707300529 | |
| Download: ML20236D572 (13) | |
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- UNITED CTATES
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NUCLEAR REGULATORY COMMISSION.
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REGION 11 q,
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101 MARIETTA STREET, N.W.
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. Report Nos.: 50-325r37-18 and 50-324/87-18 s;
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Licensee:
Carolina Power and Light Company
P. O. Box 1551.
.E.+g Raleigh, NC 27602
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Docket Nos.:
50-325 and 50-324 License Nos.:
Facility Name:
Brunswick 1 and 2 Inspection Conducted: June 8-12,'1987 i
Inspectors:_ ?A. M. k
. atgSidned
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J.t% Cole Q
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LL /z m7 R 'Cg Chou
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e31g'ned Accompanying rs n 1:
J. J. Blake Approved by:
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7 A/ 67 J. J./B1'ake, Section Chief Dite Signed I
g fieering Branch iv)isionofReactorSafety SUMMARY Scopei This routir.e, announced inspection was in the areas of review of visual inspection procedures, review of inservice inspection (ISI) data and evaluations, visual inspection-observation of completed work, previous enforcement matters, licensee event reports and I.E. Bulletins.
Results:
No violations or deviations were 16ntitled.
8707300529 870723 PDR ADOCK 05000324L G
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REPORIDETAILS
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h,ersonsCentacted
'O LicenseeEmpNyees C.,R. Dietz, General Manager j
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- E. 4. Bishop Manager, Ope. rations
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'"tj,' A* Smith, Director,' Administration Support
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- K.S+k. Eckstein, Manager, Technical Support
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E. Enzor, Director, Regulatory Compliance
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- L. E. Jones, Director, QA/QC
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- R! M. Poulk, Senior Regulatory Compliance Specialist
- B.,Altman, Principle Engineer, Maintenance
- J.'M. Brown. Resident Engineer
- R. J. Grocyn', Project Construction Manager
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"L. Wheathay, Project Engineer, ISI M. North, Principal Engineer
B. Nonroe, Structural Principal Engineer l
'R. Fronereth, Senior Engineer
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W. Niermans, Structural Engineer Other licensee emp'oyees contacted included construction craftsmen, engineers, technicians, mechanics, security force members, and office personnel.
NRC Resident Inspector
- L. Garner
- Attended exit interview 2.
Exit. Interview The inspection scope and findings were summarized on June 12, 1987, with those persons indicated in paragraph 1 above. The inspectors described the areas inspected and discussed in detail the inspection findings.
No dissenting com.nents were received from the licensee.
The following new items were identified.
(0 pen) Unresolved Item 325, 324/87-18-01, Corrective Action on Violation 324/85-19-02.
(0 pen') Unresolved Item 325, 324/87-18-02, Approval of Overlap Modeling Techniques.
The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection.
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3.
Licensee Action on Previous Enforcement Matters a.
(Closed) Violation (324/80-29-02):
Failure to Provide Adequate-Procedures.
CP&L's letter of response dated October 1, 19o0 and subsequent follow-up response letters have been reviewed and determined to be acceptable by - Region II.
The inspector examined the corrective actions as stated in the response. The system review program was not available for review because it was kept in the Philadelphia Office of United Engineers of Constructors, Inc. (UE&C).
The inspectors evaluation of the licensee response was b::ed on other evidence provided by'the licensee. The inspector concluded that CP&L had determined the full extent of the subject violations, performed the necessary survey and follow-up actions to correct the subject conditions, and developed the necessary corrective actions intended to preclude the recurrence of similar circumstances. The corrective actions identified in the letter of response were implemented.
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(Closed) Violation 324, 325/85-19-01:
Deficiencies in Pipe Support and Anchor Bolt Installation and Inspection.
l CP&L's letters of response dated August 16, 1985 and October 25, 1985 have been reviewed and determined to be acceptable by Region II.
The inspector examined the corrective actions as stated in the response.
The inspector concluded that CP&L had determined the full extent of the subject violation, performed the necessary survey and follow-up actions to correct the subject conditions, and developed the necesscry corrective actions intended to preclude the recurrence of similar circumstances.
The corrective actions identified in the letter of response were implemented.
c.
(Closed) Violation 324/85-19-02:
Failure to Meet Code Requirements in Pipe Support Weld Design. This violation involved the licensee's failure to meet AISC and/or AWS D1.1 Code requirements for the minimum weld size in pipe support weld connection design.
Support No. PS-3695, Rev.
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Unit 2 and Support No. 2E11-127-PG-1002, Rev. A, Unit 2 should have the minimum fillet weld size of 1/4" and 5/16" to meet code requirements since the thicker metals are 3/4" and 1" respectively.
The detail drawing for Support No. PS-3695 and 2E11-127-PG-1002 specified the fillet weld size as 1/8" and 1/4" which failed to meet the minimum fillet weld size in accordance with code requirements.
The inspectors discussed the above probler.is with the licensee's engineers and reviewed the documents records, corrective actions and the licensee's response dated August 16, 1985.
The detail drawing for Support No. PS-3695 and 2E11-127-PG1002 were revised to meet Code requirements, fhe licensee also took two other corrective actions
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to prevent it from recurrence.
A training meeting was held on July 3,1985 with the design engineers to stress the necessity for meeting all applicable code requirements when sizing weld.
The Structural Design Guide (SDG-2) was revised to include the minimum weld size table, table 7.1A, to provide a more complete procedure'to the design engineers.
During the inspector's review of the above coructive actions, the l
licensee's engineers were questioned as to whether they had similar minimum weld size and dimension problems on the identical supports for Unit 1.
The licensee's engineers presented support drawing Mark No, pS-3293, Unit 1 (similar location to PS-3695 on Unit 2) and PS-7361, Unit 1 (similar location to PS-7384 on Unit 2) which had dimensions interchanged problems (see violation 324, 325/85-19-01)
and stated that the two supports had dif ferent assemblies with no identical minimum weld size or dimension problem.
The inspectors reviewed two supports provided and found that both supports had min' mum weld size problems.
Support No. 3293, Section F-F, Rev. C, Unit 1 exhibited two places with 1/4" fillet weld at connections between tube TS 3" x 3" x 5/16" and plate 1 1/4" x l'-6" x l'-10" which ' violated the 5/16" minimum weld size requirement.
Support No. PS-7361, Section B-8, Rev. A, Unit I had one place with a 1/4" fillet weld for connection between tube TS 3" x 3" x 5/16" and base plate 1" x l'-41/2" x l'-2" which violated 5/16" minimum weld size requirements.
Per Paragraph II, of the licensee response to Response, Violation No. 2, dated August 16, 1985, the licensee stated that d'These are believed to be isolated cases in that no other code violation were identified during the review of numerous other drawing."
It was not mentioned by the licensee that how many drawings were reviewed. Since the three places of the minimum weld size problems are found from the two supports presented, the inspectors acknowledge that there is a generic problem existing in the minimum weld size for both units which the welds do not meet the AISC and/or AWS D1.1 code requirements. The licensee is requested to develop a program to review and correct the minimum weld size problem on all safety-related piping systems for both units.
Violation 324/85-19-02 is considered closed.
The licensee's failure to identify and correct the minimum weld size problem will be addressed by Region II in as Unresolved Item 324, 325/87-18-01, Corrective Action on Violation No. 324/85-19-02.
d.
(Closed) Unresolved Items 324/82-33-01 and 325/82-31-01:
Overlap Modelling Technique Used in Seismic Analysis.
This Unresolved Item identified the Overlap Modelling Technique in the UE&C Procedure DEDP-2607, Revision 1.
The procedure was used for computerized Piping Analysis in Brunswick Stress Analysis Problems.
The UE&C procedure did not meet the modeling criteria recommended by NUREG/CR-1980 BNL-NUREG-51357.
The previous inspector expected the licensee to settle this difference with NRC's Office of Nuclear Reactor Regulations (NRR).
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.The inspectors held discussions with the licensee's cognizant engineer 'and. the senior regulatory compliance specialist.
The licensee's engineer stated that there were about 20 stress analyses which used the UE&C Overlap Mndelling.: Technique.. The ' licensee's regulatory compliance specialist stated that the licensee-misunderstood.the request for' settling this issue between licensee
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and NRR.
Therefore, no information was provided during this
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The original Unresolved Items.324/82-33-01 and 325/82-31-01 are considered closed.
This item is now identified as
. Unresolved Item 324, 325/87-18-02,. Approval of Overlap Modelling Technique. The licensee should take one of the following actions to resolve the Overlap Modelling Technique Problem.
(1) Submit'a complete report to NRR for approval. The report should include theory, development, process or procedure, two complete examples for comparison with output using NUREG/CR-1980 BNL-NUREG-51357 and explanation of results.
(2) Recalculate the stress analysis based on the NUREG for all lines where questionable Overlap Modelling Technique developed by U&EC was used.
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(3) Show that UE&C Overlap Modelling Technique was yproved by NRR for another licensee and provide reference.
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(0 pen) Unresolved Item 324, 325/83-01-01:
Plant Unique Analysis Report - Column Loa is Verification.
This matter concerned the fact that 69 kips was used for the column design at each vent header described in Section 1.12 of Plant Unique Analysis. Report (PUAR).
This was inconsistent with the load shown in figure 1.12-1 where.90 kips was shown on the same column. The licensee agreed to find out the cause of difference and provided information in a supplement for q
review. The inspectors held discussions with licensee's engineer and q
reviewed the information provided.
The load units shown in Table 3.7.4-1, qualification of columns for pool swell and thrust loads,
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and load units (kips) used in the column calculations are J
inconsistent.
The unbraced length used in the column design j
calculation should be 8.4 ft (whole length) instead of 5.8 ft (length j
for smaller section) for the pin connection at two ends when no
lateral loads are applied.
Only a single bending curvature for the J
total compression length will exist between the pin connection at the
.two ends. This is a total unbraced compression length which is to be used for column design.
When the inspectors addressed the above i
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questions in the PUAR and the design calculations the licensee's j
engineers and UE&C engineers replied that the load units for KSI in
PUAR should be kips to be consistent with the calculations and the i
unbraced length of 5.8 ft was based on their interpretation of the AISC Code. The error in the KSI unit appeared to be the result of a clerical mistake. Per telephone conversation between the ir.spectors and. the licensee's engineer and UE&C engineer, the inspectors fo' nd that the information provided in the PUAR and column design
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i calculation had been revised. Therefore, the inspectors requested licensee to submit the updated PUAR on column' load summary, column j
design calculation and column detail drawing for future review. This.
l Unresolved Item.will remain open due to incomplete information, i
outdated information provided and inadequate explanation from UE&C.
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(0 pen) Unresolved Items 324, 325/83-01-03:
Plant.
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Unique Analysis
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Report - Vacuum Breaker Verification. This matter concerned the fa'ct
that neither GE nor the licensee has performed a techn.ical review
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with regards to the evaluation of the Mark I vacuum breaker. The evaluation was performed by Continum Dynamics, Inc. (CDI) on purchase Order'No. 205-XJ102.for the General Electrical Company, dated August
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1982. 'The inspectors held discussions with the licensee's engineer
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who explained why GE would not take responsibility for the evaluation of the. vacuum breaker that had been performed by.CDI. The reason GE.
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would not accept responsibility was because the licensee thru. UE&C
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.provided all input data to CDI for vacuum breaker analysis and evaluation under the requirement of Plant Unique Anaiysis Program which.is an own e r '. s group program used by. approximately ten utilities. The analysis and evaluation from this program could be different from GE's program. Therefore, GE in the letter to licensee claimed no, responsibility for the evaluation done. by CDI.
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'1 licensee did not provide any further information on this matter during this inspection.
The licensee agreed to to review the CDI evaluations, g.
(Closed) Unresolved Item 324, 325/85-06-01:
Documentation of Nonconforming Condition Corrective Action.
This matter dealt with the concerned that no documentation was available for review on procedures to prevent potential water hammer problems or on the repairs to the residual heat removal (RHR) service water pipe support j
problems which incurred due to vibration in the system, l
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the procedures submitted.
The licensee's engineer stated that no water hammer problems have been experienced since the event reported in LER-01-84-23 Periodic Test Procedure 07.2.4a was also reviewed by
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the inspectors.
This procedure was for the core spray system and presently requires that an operator be stationed to monitor system piping for excessive motion when the core spray pump is started.
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Similar procedures were also developed to monitor potential water hammer problems for other systems.
The inspector also reviewed CP&L's Engineering Evaluation Procedure, ENP-12.
This procedure requires the responsible engineer and QA engineer to verify each
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action item or surveillance assumed as part of the I
I evaluation / disposition to ensure that the effects of any repairs or L
nonconforn'ance -on the system, or related components are evaluated.
ENP-3, Plant Modification Procedure and ENP 3.1, Direct Replacement Procedure, contain provisions for updating design documents to reflect modifications to the plant equipment and components.
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The above procedures provide adequate assurance that-modifications of q
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-Unresolved Item,is considered closed.
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. U' 'nresolved Items!
! Unresolved ' items are matters about which more information is required to determine whether they are acceptable or may involve. violations or
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deviations..Two unresolved. items identified during this inspection are discussed in paragraphs-3.c and 3.d.
L5.
Inservice Inspection (ISI)
Duringlthe ' spring 1987 outage for Brunswick Unit 1, Carolina Power.and
Light had (CP&L) divided the scope of the ISI work among the following l
vendors and. CP&L had organizations:.(1) General Electric (GE) performed code and L augmented. ultrasonic examinations, (2) CP&L's Corporate Quality Assurance 1 Department (Material and Quality Section) performed surface
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examinations, (3) Site' Quality Control (QC) performed visual examinations and, (4) CP&L performed the overlay weld repairs using Power Cutting Inc.
welding operators that.had been qualified ~to CP&L's welding procedures.
During the week of June 8-12, 1987, CP&L ms attempting to achieve criticality. 'The inspectors reviewed documents, examined completed work activities (which included walk down visual inspections of safety-related systems o' tside the' reactor drywell area) and reviewed the licensee's evaluations of reported ultrasonic indications to determine whether ISI activities had been conducted-in accordance with the applicable procedure,.
regulatory requirements and licensee commitments. The applicible code for the second interval examinations is the ASME B&PV Code, Sect.on XI,1980 edi tion, winter '1981 addenda.
However, since CP&L started the second
. interval early in order to have both units utilizing the same Code edition, 1st interval examinations completed during this outage were performed in accordance with ASME B&PV Code,Section XI,1977 edition, with addenda through summer 1978, a.
Review of Procedures (73052) Units 1 & 2 The inspectors reviewed CP&L's visual examination procedures to determine whether the procedures used by the licensee met applicable ASME Code, regulatory, specification and contract requirement.
The procedures were also reviewed in the area of procedure approval, l
requirements for qualification of nondestructive examination (NDE)
personnel, and compilation of required records.
l The following ISI examination procedures were reviewed:
Procedure Identification Number Title Periodic Test Procedure:
PT-91.0.52 VT-3/VT-4 Examination of i
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Periodic Test Procedure:
PT-91.0 VT-3/VT-4 Examination of Component Supports Following Maintenance
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The inspectors also reviewed the visual examination procedures to determine whether they contained sufficient instructions to assure that the following parameters were specified and controlled within the limits permitted by the applicable code, standard, or any specificat1on requirement for: method; how the visual inspection is to be performed and what specific attributes are to be ' examined; special illumination; instruments or equipment to be used; if any; sequence of performing examination; when applicable; data to be tabulated; if any; acceptance criteria specified and consistent with the applicable code section or controlling specification; and report form completion.
The inspectors conducted several pre-critically walk down visual inspections of safety-related systems outside the drywell.
These inspections are documented in detail in paragraph 5.b. of this report. The result of these inspections concluded that CP&L's ISI visual examination program for pipe supports and components had achieved a high level of quality and the systems re-examined by the inspectors were mechanically, structurally and operationally ready for criticality.
The inspectors however, did find several supports where gap allowances between the support base plate and the concrete structure appeared to be excessive.
Each support was evaluated by the licensee and found to be acceptable.
The inspectors noted also, that CP&L's ISI visual procedures did not have a specific inspection attribute to address gap allowances of base plates. Discussions with the licensee indicated that CP&L felt this inspection attribute related primarily to installation variations and had been addressed during visual examinations at that point.
The inspectors agreed that gap allowances were primarily the result of installation variations, however, they could also indicate degradation of a support resulting from thermal stress movement of a system during heat up and cool down and should be ve ri fi ed.
The licensee agreed to add an iaspection attribute for gap allowance of the base plate to the ISI visual inspection procedures so that a base line could be established on this condition for future reference. On June 26, 1987 the inspectors were notified by telephone that CP&L had implemented this procedure change in both of the above visual examination procedures.
b.
Visual Examination of Components and Supports The inspectors assisted the resident inspector in performing pre-criticality walk down visual inspections of safety-related piping and components in the following areas:
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The minus 17 foot level; portions of division 1 & 2 Residual Heat Removal (RHR), high pressure core injection (HPCI) service water and reactor core isolation (RCIC) coolant syste _ _ _ _ _ _ _ _ _ _ _
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Control. rod drive (CRD) and Hydraulic Control Units (HCU).
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Division 1 core spray in the penetration area.
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General areas of the reactor building.
The above areas represented approximately 40 percent of the safety-related piping outside of the drywell.
The inspectors re-examination revealed that the licensees maintenance and inspection practice appeared to be very good.
No significant item was l
identified by the inspectors.
Several minor items were handled by the licensee in an expeditious manner.
The inspectors examined the supports and components using the following inspection attributes:
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Deformation / structural degradation g
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Missing, detached, lossed support items W
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Machined /slidifig surface interference
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Improper settings or clearances
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Insufficient thread engagement
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Loose bolting or missing handwheels
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Locking device degradation Missing ID tags
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Other discrepancies not specifically identified above.
c.
Data Review and Evaluation - Unit 1 (73755)
The inspectors reviewed the licensees ISI data files to ascertain whether the licensees disposition of adverse findings were consistent with regulato ry requirements.
Data reviewed by the inspectors included ultrasonic examination reports, photographs of A-Scan presentations of the indications, GE's Indication Notification / Resolution Report, GE's Indication Resolution Sheet, Nutech's evaluation of the indications and data resulting from CP&L's review.
Evaluations of the following ultrasonic indications were reviewed by the inspectors:
Indication Notification /
Resolution Report (INRR) No.
Weld Number 87-001 1B32-RR-28" A8 87-002 IB32-RR-22" AM-3 87-003 IB32-RR-28" A4 (Overlay Weld)87-004 1B32-RR-12"-AR-A3 (Overlay Weld)87-005 1832-RR-28"-A14 (Overlay Weld)87-006 1832-RR-12"-BR-G3 (Overlay Weld)87-007 1832-RR-12"-BR-K3 (Overlay Weld)87-008 1832-RR-12"-BR-J2 87-009 1832-RR-12"-BR-H2 (Overlay Weld)
Within the areas examined, no violation or deviation was identified.
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Onsite Follo'wup of Licensee Event Reports (LER's) ' Units 1 & 2 (92700)
a.
(Closed).LER 02-83-97, During the performance of testing required by
'IE Bulletin 83-02,. it was determined that crack indications. existed in 19 of '131 welds examined en the recirculation, reactor water cleanup-and residual heat removal. systems.
Data evaluation determined.these indications to ranged from 5'to 22.0 percent of the pipe' wall thickness with lengths that ranged from.5" to 11.0".
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indications were attributed to IGSCC.
Eight welds were repaired
using acceptable. weld overlay techniques.
The remaining indications were evaluated and determined 'to be acceptable for continued operation' until April 30, 1984.
The inspectors had performed surveillance of licensees work activities when the examinations ~ were
.in process.. In addition completed examination reports and crack indication evaluation data were also reviewed.
This item is considered closed.
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(Closed).LER 01-85-026, During the Unit 1 1985 refuel / maintenance
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outage, visual and/or ultrasonic examinations.of reactor-recirculation piping following induction heat stress improvement -
revealed the existence of 92 indications within the heat-affected -
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zones of 23.of the 79 welds tested.
This-included pinhole leaks at
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five of the welds. 77 of the 92 indications were axial. Additional inspection of the welds using the General Electric automated SMART Ultrasonic Testing System yielded'a positive correlation between the'
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manual and automated inspections made on circumferential indications.
-l The cause of the indications ' and pinholes was. attributed to Intergranular Stress Corrosion Cracking (IGSCC) of the class 1, type j
304 stai'less steel piping material. Corrective action consisted of.
applyina acceptable weld overlays over the weld joints. 0ne weld was evaluated and found acceptable without overlay welding.
The inspectors conducted surveillance of the licensee's inspection
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activities during the reporting period and reviewed with licensee's
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completed UT data and evaluations. The inspectors also monitored the j
overlay repair welding.
This items is considered closed.
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(Closed) LER 01-87-008, During the Unit 1 1987 refueling / maintenance outage, ultrasonic examinations of reactor recirculation piping
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System revealed the existence of a circumferential indication, 3.7" in length and 30% through wall depth, within the heat-affected zone of Piping Weld No.1-832-RR-22"-AM3.
These ultrasonic examinations were performed in accordance with the requirements of Generic Letter 84-11.
Analysis determined the indication is the result of integranular stress corrosion cracking of the Class 1, type 304, stainless steel piping material.
Action to correct the subject indication involved the use of approved weld overlay techriiques and a full structural weld overlay.
A subsequent penetrant and SMART system ultrasonic examination determined this weld overlay to be acceptable.
The inspectors conducted surveillance of the licensees
activities during the reporting period and reviewed the licensee's l
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completed UT data and evaluations. The inspectors also monitored the overlay repair. welding.
This item is considered closed.
Within the areas examined, no violation or deviation was identified.
7.
IE Bulletins (IEB) - Unit 1 & 2 (92703)
a.
(0 pen) IE Bulletin 83-07, Apparently Fraudulent Products Sold By Ray Miller, Inc.
Power reactor facilities were informed in January 1983 by Information Notice 83-01 that fraudulent products may have been sold to nuclear industry companies by Ray Miller, Inc. An updated and comprehensive list of Ray Miller, Inc. customers for the years 1975-1979 was provided to power reactor licensees and to selected fuel cycle and Category B, Priority I material licensees in Supplement 1 to IN 83-01 that was issued' on April 15, 1983.
Information became available regarding specific purchase orders for which materials were apparently substituted, licensees were requested to determine where suspect material had been installed in plants, evaluate its safety significance, and tag or dispose of the suspect material not installed.
The specific information on apparently fraudulent material covered a five year period and pertained to orders filled by the Charleston office only. Although the Charleston office was apparently the major offender, the other Ray Miller, Inc. branch offices may have also supplied fraudulent materials, in some cases.
However, the NRC was unable to locate any records for review from these other branch offices.
In an effort to aid in determining the scope of the f raudulent materials problem, licensees were requested to examinc and test materials from other Ray Miller, Inc. branch offices that they have been able to identify, but were not included in the
' NRC-identified list of apparently fraudulent materials.
CP&L's letters of response to IEB-83-07 dated; March 23, 1984, June 1,1984 and November 9,1985 were reviewed by the inspectors.
In addition, other CP&L and vendor correspondence dealing with this issue was reviewed.
CP&L concluded that they had not received any fraudulent materials from Ray Miller Inc.
The inspectors started a sample review of Brown and Root Purchese Orders for the years 1975-1979 to insure the licensee's review was complete. Although the inspector's review has not been completed no indication of fraudulent Ray Miller materials were identified.
This sample audit will be continued during a subsequent inspectio _
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(0 pen) Pipe Support Baseplate Designs Using Concrete Expansion:
Anchors (IEB 79-02) and Seismic Analysis for As-Built Safety-Related Piping.(IEB 79-14).
File No. 507 2-252, BESU-87392, dated May.4, 1987 was submitted by L
the licensee for NRC review, the_ licensee. completed all.-IEB 79-02
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related work and all modifications in the field on IEB 79-14 for both units of the Brunswick Steam Electric. Plant.
UE&C is. currently.
performing reconciliation"of pipe support calculations and drawings
.for as-built conditions for IEB 79-14. The licensee is currently in-the process of. receiving, reviewing and accepting all UE&C '.s safety-related ipiping and pipe ^ support computer runs, drawings,
calculations and as-built records. To close:out the requirements for IEBs 79-02 and 79.14, the. licensee. was requested to submit a final summary report in response to ' all* requirements stated in IEB 79-02 i
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and 79-14 by end of.1987. 'In order to - veri fy the. licensee's '
performance on supports modifications finished in compliance' with l
IEBs 79-02 an.d 79-14 during -1986, the inspectors selected the
.1 following 27 restraints in Unit 1 in the area of dynamic pipe q
supports. and component supports structures t!.at had been. QC final inspected to see.if. they c'omplied with -the bulletin requirements.
The following restraints were reinspected.with the assistance of licensee QC and design engineers:
Support Mark No.
-Corresponse Calc. Reviewed Remarks 1CAC-32PG117 80-133-03 1CAC-32PG118 80-133-04 No cross reference 1RNA-261A143
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IRNA-265PG94 84-195-56 As-built drwg.
should be part of calc.
,
1RNA-265PG110 84-195-12'
-As-built drwg.
.
should be part of calc.
1RNA-266PG-1 1RNA-266PG-2 4'
1RNA-266PG-3 1RNA-267PG-4 1RNA-267PG-5
,
'
1RNA-271PG-6 1RNA-271PG-7 1RNA-273PG-8 1RNA-273PG-9
,
-
.
-
_ _ - _ -
- _ _ - - - _ _ _ _ _.
]
g.
.
q
<
,,
12 l
a
'
i Support' Mark No.
Corresponse Calc. Reviewedt Remarks (cont'd)
')
ij 1RNA-271PG27 BP-19027
.1RNA-271PG28 BP-19027
)
,
l 1RNA-274PG64'
!
'1RNA-274PG70 84-195-17 As-built drwg.
should be part of.
' calc.
!
PS-4343 82-219P-25 i
PS-4344 82-219P-25 PS-4345
- 82-219P-25 t
The'above restraints were partially reinspected against,their detail-
. drawings for configuration, identification, fastener /anchort
',
installation member. size, welds,~nd damage / protection. In general,
,
.a the: restraints were installed in'accordance with-design drawings with
. good workmanship.
No discrepancies were identified.
The inspectors also performed.the review of.seven. support calculations as shown above.
In general, all calculations. are j
acceptable with a minor deficiency on weld connection checks. All
-
the calculations in weld connection check' used weld metal allowable
,,
of 21000 psi for E70xx electrodes. Per Table 1.5.3, Allowable Stress on Welds, ' AISC, Eight Edition, the allowable stress on ' welds should be the lower value of 0.3 x nominal tensile strength of weld metal or allowable in the base metal.
The licensee's responsible ' structural engineer agreed to revise the Structural Design Guide (SDG-2) to meet-the AISC requirement.
Within the areas ' examined, no violation or deviation was identified.
.
_ _ _ _ _