IR 05000498/1986025

Insp Repts 50-498/86-25 & 50-499/86-23 on 860811-870313.No Violations or Deviations Noted.Major Areas Inspected: Licensee Reported Const Deficiencies & IE Bulletin & Info Notice Responses

ML20214G507
Person / Time
Site:	South Texas
Issue date:	05/13/1987
From:	Hunnicutt D, Ireland R, Tapia J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20214G433	List: IR 05000498/1986025 ML20214G430
References
50-498-86-25, 50-499-86-23, IEIN-79-02, IEIN-79-2, IEIN-83-31, IEIN-83-80, NUDOCS 8705270087
Download: ML20214G507 (17)

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-APPENDIX U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

NRC Inspection Report: 50-498/86-25' Construction Permits: CPPR-128 50-499/86-23 CPPR-129 Dockets: 50-498 Category: A2 50-499 Licensee: Houston Lighting & Power Company P. O. Box 1700 Houston, Texas 77001 Facility Name: South Texas Project, Units 1 and 2 Inspection At: South Texas Project, Mata0orda County, Texas, and Houston Lighting & Power Engineering Offices in Houston, Texas )

Inspection Conducted: August 11, 1986, through Marcn 13, 1987 Inspector: n7 6fT BT J/ y. Tapia, Reaq)or Inspector, Operations Dat'e Mection, React # Safety Branch Reviewed: abht otE u'#f- 7!87 D. M. Hunnicutt, Chief, Operations Sections D#te /

Reactor Safety Branch

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Approved: y 3 p M. E. Ireland, Acting Chief, Reactor Safety Date

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u Branch Inspection Summary Inspection Conducted August 11, 1986, through March 13, 1987 .

(Report 50-498/86-15); 50-499/86-23 Areas Inspected: Special, announced inspection of licensee reported significant construction deficiencies; followup of deficiencies reported by the 8705270007 070518 PDR ADOCK 05000498 G PDR

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~ Construction Appraisal Team' Inspection (NRC Inspection Report 50-498/85-21;

50-499/85-19); inspection of an' allegation; and responses to NRC I&E Bulletin

~and'Information Notice , Results: Within' the areas inspected,-no violations or deviations were

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DETAILS 1. Persons Contacted HL&P Personnel M. Wisenburg, Manager, Nuclear Licensing E. Dotson, Manager, Engineering A. Sharon, Senior Mechanical Engineer R. M. Attar, Special Programs, Engineering R. C. Munter, Consulting Engineer G. D. Purdon, P.E. Programs S. Head, Project Compliance W. Trujillo, Supervisor, Project QA Analysis B. Poole, Lead Engineer, Piping and Pipe Supports C. L. Harvey, Quality Systems and Administration Supervisor Bechtel R. Parik, As-built Reconcilation Group S. A. Bokhari, Pipe Support Stress Group Supervisor A. Frano, Stress Group Leader M. Nanjappa, Stress Engineer C. W. Humes, Site Project Engineer A. Lopez, Civil / Structural Group Supervisor N. Shah, Deputy Engineering Structural Group Supervisor W. Miller, Structural Group Leader M. Hsu, Deputy Structural Group Leader R. Singh, Deputy Piping and Supports Group Supervisor C. C. Chen, Responsible Engineer for Penetrations L. Young, Soils Supervisor R. Talmadge, Engineering Geologist E. Cvikl, Civil / Structural Engineering, Group Leader P. Glover, Project Geotechnical Services Manager 2. Review of IE Information Notice and Bulletin Followup .

An inspection was conducted of the licensee's response to Information Notices issued by the NRC in order to determine the adequacy of the information review along with any actions taken. The following items were reviewed during this inspection: Error in the ADLPIPE Computer Program On May 19, 1983, the USNRC Office of Inspection and Enforcement issued Information Notice No. 83-31 dealing with an error in a computer program used for piping analyses. The error in the ADLPIPE computer code involved the use of the run pipe section modulus for calculating stresses at branch pipe terminations. The resulting stress calculation at reduced outlet tee connections is

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underestimated by an amoE3t depending on the ratio of run pipe to branch pipe section modul6s. The error is present in calculations performed using the Class 2 component option for the 1972 version of the ASME B&PV Cod The error is not present in calculations using the options for the 1974 and 1977 editions of the ASME B&PV Cod Calculations for Class 2 piping systems at STP were performed in accordance with the requirements of ASME B&PV Code,Section III, Subsection NC-1974. The version of ADLPIPE used in the calculations, therefore, does not contain the erro This matter is considered close b. Use of Specialized " Stiff" Pipe Clamps On November 23, 1983, the USNRC Office of Inspection and Enforcement issued Information Notice No. 83-80, which identified three concerns with the use of stiff pipe clamps. The three concerns were:

excessive bolt preload induced stresses in the pipe, localized overstress due to small contact bearing areas, and the effect of a clamp on elbow stress indices.Section III of the ASME B&PV Code does not provide rules for evaluating stresses due to loadings from nonintegral attachments such as " stiff" pipe clamps. However, stresses induced by the torquing of the " stiff" pipe clamp are required to be considered in order to satisfy the intent of Section III of the ASME B&PV Cod Methods consistent with the intent of the Code were developed to address the concerns of Information Notice No. 83-80. These analytical techniques and their results were. reviewed by the NRC inspecto There are four stiff pipe clamps used at STP, three located on the RHR/SI system and one on the pressurizer surge lin None of the stiff clamps used at STP were installed on an elbo The procedure utilized for the evaluation of the effects of stiff pipe clamps involved the addition of the primary membrane and primary bending stresses caused by the load from the clamp to the stresses caused by internal pressure and bending as calculated per ASME B&PV Code,Section III, Subsection NB 365 Clamp-induced stresses caused by the constraint of the expansion of the pipe due to internal pressure were added to other secondary and peak stresses. The fatigue usage from clamp-induced plus pressure, temperature and support loading stresses was also calculated. The results of the calculations were found to show that the primary stress intensities for all operating conditions, the primary plus secondary stress intensities, and the cumulative usage factors all met the ASME B&PV Code requirement This matter is considered close c. Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts On March 8, 1979, the USNRC Office of Inspection and Enforcement issued Bulletin No. 79-02 which identified concerns with the use of concrete expansion anchor bolts in pipe support base plate design Subsequently, Revision 1 of Bulletin No. 79-02 was issued on June 21,

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1979,. Supplement 1 to Revision 1.was issued on August 20, 1979, and Revision 2 to the Bulletin No. 79-02 was issued on November 8, 197 Bulletin No. 79-02 required a response which was to include consideration of the effect of base plate flexibility on calculated expansion anchor loads, verification that all expansion anchors have a factor of safety of four or five (for wedge and shell type anchors, respectively) as compared with the manufacturer's recommended values for ultimate anchor capacities, a description of design requirements for expansion anchors subjected to cyclic loads, and verification by documentation that such design requirements have been me The third and final revised response to Bulletin No. 79-02 was submitted on July 30, 1984. The revised submittal reflects the design methods employed by Bechtel and the result of their review of the design and installation of expansion anchor bolts as documented by Brown & Root specifications and procedures. The NRC inspector reviewed the supporting documentation and technical basis for the revised submitta '

The analyses of baseplates which utilize expansion anchors are performed either by computer or manual calculations. The computer code utilized is CDC, CE035, BASEPLATE 2. This program utilizes finite element analysis to consider the flexibility of the plate interacting with the stiffness of the expansion anchors and the bearing concrete under the plate. The code includes the effects of prying action in calculating expansion anchor tensions. Manual techniques for calculating expansion anchor tensions due to externally applied moments considers the flexibility of th'e plate by prescribing that the tension / compression couple resulting from the applied moment be defined as either the tension resultant located at the centroid of the bolt group in tension or as the compression resultant located under the compressive edge of the member which is attached to and delivers the external moment to the plat The specified allowable loads in tension and in shear were found to have factors of safety in accordance with Bulletin No. 79-02; requirements and based on corresponding ultimate load capacities determined from static load test The design requirements for cyclic loading on baseplates are based on equivalent static pipe support design loads obtained from the dynamic analyses of piping systems. The dynamic load capacities were

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previously determined by testing programs to be approximately the same as their corresponding static load capacities. The minimum'

factor of safety utilized is therefore considered adequate for, expansion anchor bolts that may be sLojected to dynamic load A static testing program was conducted to confirm the manufacturer's published ultimate capacity values or to establish new ultimate capacity values. A technical reference document was generated to

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consider the various requirements for the design, installation, testing, and inspection of expansion anchor bolt The results of the NRC inspector's review inificated that the response to Bulletin No. 79-02 was technically correct and addressed the concerns listed in the bulleti This matter is considered close . Review of Design and Construction Deficiencies An inspection was conducted of selected deficiencies found in design and

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construction, which were reported pursuant to the requirements of 10'CFR Part 50.55(e), in order to determine whether corrective actions were taken as stated in submitted reports, and whether engineering evaluations were adequate and met regulatory requirements and commitment The following items were reviewed during this inspection: Design Loadings for the Isolation Valve Cubicle Walls (IRC 130]

On September 1, 1982, HL&P notified the NRC of a potential deficiency concerning the design pressure loadings for the Isolation Valve Cubicle (IVC). The IVC structural walls are required to provide protection and separation of the Auxiliary Feedwater System (AFWS)

trains as well as the main steam and feedwater isolation valves from both internal and external environmental conditions such as mass and energy releases, flooding and missiles. Brown & Root Interoffice Memoranda GM-33417, " Preliminary Feedwater Results for 4" Steamline Break in Aux Cubicles" and GM-19053, " Thermal Hydraulic Transient Resulting from Main Steam Line Rupture in the IVC," provide postulated pressure values and durations in the IVC due to main steam and feedwater pipe ruptures. This data was utilized to generate the design basis accident pressure loads (Pa) which in turn resulted in the design drawings for the 82-by 62-foot reinforced concrete structur A calculation performed by NUS Corporation (NUS) for a a'

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break in one of the 8" FW bypass lines within an AFW pump cubicle resulted in a Pa of 35.24 psid at 1.85 seconds. The original design of the wall utilized a Pa of 19.1 psid from an applied load of 10.6 psi. These values are contained in calculation No. SC068-3 The higher pressure calculated by NUS indicated a potential failure of the AFW pump compartment walls which couk, in turn, lead to ,'

failure of AFW pumps in adjacent cubicles. A review by Bechtel of NUS calculation No. A170XC602ANU disclosed that the assumptions used in the original analysis were overly conservative in that the

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operational alignment of the valves in the feedwater bypass line and the operating temperatures that were assumed in the analysis did not accurately reflect those that were predicted. Bechtel performed a reanalysis in order to incorporate more realistic assumptions and evaluate their effect on the calculated pressur .

The NRC inspector reviewed Bechtel calculation No. 7012, Revision 3,

, "J/C PT Analysis (COPDA Results) MSL Break." This calculation

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presents the pressure-temperature analysis due to a main steam line

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break in the IVC. All-thermodynamic and fluid flow assumptions and-

_ theory are: outlined in BN-TOP-4, Revision 1, "Subcompartment P/TL

Analysis." Bechtel calculation No. CC6251,-Revision 0, " IVC

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Reanalysis" was also reviewed by the.NRC inspector. This calculation presents'the final" design verification of the: IVC. A finite element analysis of the structure was made using Bechtel's BSAP computer-

cod In addition to the correct accident pressure' load, the walls were also; checked against tornado generated missile, impact ~and for strength of steel embed plates of the pipe whip restraints. . The NRC inspector verified the correct physical properties input, boundary conditions, and the selection of the critical loading combination It was found that all structural components were adequate >in-t strength, stiffness and reinforcement provtsions for.the design basis loads. Based on the results of the NRC inspectors review, this item is considered close Stress Analysis Performed for Anchor-Darling Valves (IRC 122)

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On June 29, 1982, HL&P notified the NRC of a potential deficiency

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concerning thirteen valves supplied by Anchor-Darling Company. The-deficiency resulted from a situation identified at the Duane Arnold Nuclear Plant'wherein the seismic analysis performed by the "

Anchor-Darling Valve Company utilized weights of "ac" motor operators

'while the actual. operators supplied were "dc." The actual weights >

were greater than'those specified on the drawings. The NRC inspector reviewed the technical; adequacy of the licensee's evaluation of the- .

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deficiency. A review of the 13 stress analy;es involved indicated that.3 valves required reanalysis since the stress margin available-was potentially insufficient to compensate for the increase in the valve' operator weigh The~three valves involved were. located in the i Component-Cooling Water System (valve No. CC-392), Main _ Steam Line (valve No. MS-143), and the Liquid Waste Processing System (valve ~

No. WL-312). The respective increases in weight by percentage were calculated and incorporated in new stress calculations. The NRC inspector reviewed stress analysis calculations RC-1234 and RC-5028, which were generated using linear elastic analysis techniques. The l

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reanalyses showed that an overstress condition would not develop in

' the component cooling water and liquid waste processing system ,

Reanalysis of the main steam line valve was not performed due to a major' char.ge in piping. layout in'the main steam. isolation valve '

. cubicle. As a result of this piping modification, a new stress r

analysis was' generated which included the actual weight of_the 4 Anchor-Darling valve on the main steam line. Based on the results of the NRC inspector's review, this item is considered close ~

c. ~ Inadequately Designed Trapeze Supports (IRC 215)

J On 0ctober 22,'1984, .HL&P' notified the NRC of a deficiency identified ~

in the design of trapeze type support assemblies with mechanical snubbers used to support main feedwater line No. 2C369-P-FW-1014

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inside the Reactor Containment- Building. The design was such that if i

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left uncorrected, the trapeze assemblies could slide or rotate around the pipe causing these supports not to function as designed and consequently overstress the pipe. In addition, these assemblies utilized u-straps or u-bolts around the pipe for which there was no documentation showing that these items were qualified for the side loads induced by the dead weight of the support assembly itself. The applicable construction specifications and design drawings did not specify a required torque value for the u-bolts necessary to hold the support tight to the pip The NRC inspector reviewed the supporting documentation and technical adequacy of the licensee's corrective actions. The deficiency was investigated on a generic basis to assess the extent of supports in this category. Seventy-four trapeze type supports were identified on ASME Class 2 and 3 pipin Eleven supports were identified on ASME Class 1 piping. All Class 2 and 3 supports were modified by replacing u-bolts or u-straps with welded pipe (integral)

attachments. Supports for ASME Class 1 lines were redesigned to eliminate the unstable configurations and replace them with nonintegral structural frames. Training sessions were held with design engineers to prevent similar recurrence with unstatie configurations. Certified design reports were generated tc, quantify allowable normal and lateral loads for u-bolt assemblies. An interaction equation was also specified for combined normal and lateral loads. Bechtel calculation No. 5L349JC9944, "U-Bolt Tightening Torque to Prevent Slippage," was generated to provide a revision to the installation specification. Based on the results of the NRC inspector's review, this item is considered close d. Maximum / Minimum Soil Density Tests On Essential Cooling Water System Piping Backfill (IRC 54)

On June 20, 1980, HL&P notified the NRC of a potential deficiency concerning the maximum / minimum soil density testing of the essential cooling water (ECW) pipeline backfill. The determination of these densities, which was required for every fourth in place density test, was.not conducted in accordance with specification requirement This matter was identified during the Brown & Root reassessment of the Category I backfiT1 and compaction testing program conducted in response to Item 2, Appendix A of the NRC Order to Show Cause issued April 30, 1980. Portions of the ECW piping were subsequently excavated in order to provide access in response to ASME Piping Re-examination, Repair, and Restart Program. Backfill in these areas was replaced and tested in accordance with specification requirement For the limited sections of backfill which did not require excavation, a combination of backfill documentation surveys and a review by Bechtel of the previous Show Cause Expert Committee Report were used to provide a technical judgement as to the qualification of the backfill materia Based on the replacement of the ECW pipeline backfill and on the NRC inspectors review of the

" Expert Committee's Final Report Concerning Show Cause Item 2 -

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Structural Backfill Investigation,' February 27, 1981," this item is considered close American Bridge Structural Steel (IRC-84)

On January 8,1981, HL&P notified the NRC of a potential deficiency concerning nonconformances of welds in vendor-fabricated Category I structural steel. The welded connections of concern were performed at the vendor's facility and were not associated with the onsite welding program. A reinspection program was initiated which resulted in the identification of weld conditions which deviated from design drawings, specifications, and welding code (AWS D1.1) requirement This matter was identified by the NRC as Unresolved Item 50-498/8130-01 and has been discussed previously in NRC Inspection Reports 50-498/81-31, 50-498/81-34, and 50-498/82-08. The unresolved item was closed in NRC Inspection Report 50-498/82-15 based on visual and dimensional examinations performed by NRC inspectors. As a result of these previous reviews, this matter is considered close . Followup on Allegation Concerning Stress Analysis for Containment Liner Penetrations (4-86-A-076)

This allegation was received by NRC Region IV from a reporter who called and stated a source had advised him that the stress analysis done during the Brown & Root era was incorrect. The reporter related that his source had identified the affected areas as main steam and auxiliary feedwater line penetrations. The reporter stated that the specific concern related to a difference between the penetration sleeve thickness as installed and the thickness called for on the drawings and that which was analyzed in the design calculation In response to tuis allegation, the NRC inspector reviewed the current design specification, analyzed the design stress analyses, reviewed the vendor supplied certified material test reports, and the results of independent ultrasonic thickness measurements witnessed during this inspection. As a result of this review, it was determined that the containment penetrations involved satisfied the design requirements for class MC components of the ASME B&PV Code,Section III as required by STP FSAR Section 3.8.2.1.4. Brown & Root Specification No. 20099NS054, Revision E provided the original requirements for the design, fabrication, and delivery of the containment mechanical penetration assemblies. As a result of Bechtel's engineering review of all Brown & Root design calculations and specifications, Brown & Root Specification No. 20099NS054 was superseded by Bechtel Specification No. 2C090RS1000, Revision 3,

" Containment Mechanical Penetrations." Bechtel subsequently reanalyzed all penetration assemblies and generated the following calculations which were independently evaluated by the NRC inspector: Calculation No. 2L469RC9910, January 3, 1984, " Stress Report for Penetration Nos. M94/M95."

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' ; Calculation No. 2L469RC9513, March 5,1986, "Re-evaluation o '

Penetrations M94 and M95'."

l Calculation No. 2L469RC9911, March 12 1984, " Containment Mechanical

' Penetrations M83 and M84."

. Calculation No. CC-5707, August 15, 1985, " Mechanical Penetration

. Sleeve Thickness Qualification-RCB."

The listed calculations consist of different load combinations for stress

' analyses using mechanical loads for static stress, results of thermal

. analyses, transient analyses, and fatigue evaluations performed'in accordance with ASME B&PV Code Section III. Appendix XIV. These '

reevalatuions of the stress intensity values'at critical sections were necessitated by changes in design loads resulting from the Bechtel reanalysis of piping system stresses. The following calculations provide the new design inputs and were also reviewed by the NRC inspector: Calculation No. 2C159RC5049, August 4,1986, " Auxiliary Feedwater Piping From Penetration M94 to Steam Generator 1A."

. Calculation No. 2C159RC5050, March 27, 1986, " Auxiliary Feedwater

Piping From Penetration M95 to Steam Generator IB."

~ _ Calculation No. SM159RC6529, April 29, ~1985, " Auxiliary Feedwater ' -

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Piping From Penetration M-94 to Electrical Auxiliary Feedwater Pump

No. 11."

, Calculation No. SM159RC6530 May 30, 1985, " Auxiliary Feedwater e Piping From Penetration M95 to Electrical Auxiliary Feedwater Pump 1' No. 12."

A minor discrepancy was identified by the NRC inspector during the review of Bechtel Calculation No. 5707. It was determined that an incorrect thickness was used in the calculation for the sleeves of penetrations M94

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and M95. Design drawing No. 9-5-1009 listed schedule 40 material while the calculation utilized the thickness for schedule 80 material. The NRC- -

( inspector performed a reanalysis utilizing the correct schedule and

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determined that no detrimental affect resulted which invalidated the ability of the penetration sleeves to satisfy code requirements.

- In order to verify the as-ouilt condition and the validity of _ the vendor

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' supplied documentation, the NRC inspector requested and witnessed

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ultrasonic thickness measurements on either side of the penetration sleeve

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welds for a' 1 penetrations involved. It was determined that all minimum

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wall thicknesses required to satisfy the ASME Code stress requirements

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As a result of this inspection, the allegation is considered

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unsubstantiated. A minor discrepancy was found but it is not considered s

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! ,to be related to the allegation and it was found to be a minor error by.

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. , 1 l 5.. -Licensee-Actions on Previously Identified Inspection Findings L

An inspection was conducted of the licensee's response to the following items identified during the NRC Construction Appraisal Team (CAT) c l

. inspection (NRC Inspection Report 50-498/85-21; 50-499/85-19): SeismicSeparatidnBekweenReactorContainmentBuilding(RCB)andthe Mechanical / Electrical Auxiliary Building (MEAB) - (CAL 85-21-35)

During a general walkdown, the NRC CAT l inspector identified a crack

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in the Unit 2 azimuth 304 tendon access wall at elevation (-)13 ft 3 inches which was subsequently ~ chipped ou It was identified that the 3-inch seismic joint material had not been installed between the Reactor: Containment Building mat and the tendon access wall as

. required by drawing 3M01-9-C-4320 Revision Bechtel Engineering (BEC) issued Nonconformance Report (NCR) HCO3170 and determined that the disposition.o~f this NCR would be to "use as is." The: basis of the disposition was that all settlements had taken place, the bearing surface area was small, and the vertical seismic movements would be.small at this locatio The NRC CAT. inspector expressed concern that BEC's disposition did not adequately address the seismic movement of the mat and the forces on the tendon access wall that would result from the predicted relative. heaves of the two adjoining buildings ~once the dewatering system is discontinue The' basis for the disposition of NCR HCO3170 was not adequately explained at the time of the CAT inspection and in subsequent meetings. With regard to the' settlement considerations used as part of the basis for the NCR disposition, the correct statement should have been that the observed settlements between the RCB and the MEAB near the wall in question did not indicate any tendency for the MEAB wall to be progressively bearing on the RCB basemat. The NRC-inspector verified this behavior by reviewing three time history-differential displacement curves covering a period from October 1976 to October 1986. These curves are generated by monitoring the'

relative movement of a pair of opposing structural benchmarks. A-p comparison of the actual relative movement of these six benchmarks 4 indicated that the RCB, which was the deeper and heavier of the two -

i structures, was settling at a slightly higher rate than the MEA This was consistent with the predictions of analytical studies. The heave resulting from a return of the water table to its normal position will not reverse the observed differential settlement rates since the mechanism causing this heave will have a uniform effect on both structures. With regard to the seismic movements, the NRC inspector verified that the maximum relative vertical displacement

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calculated between the MEAB snd the RCB during the SSE was

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0.23. inches. !A displacement.~of this low magnitude did not represent

/ ;a potential for structural damage to the wall in question. Based on the results of the NRC inspector's review, this matter is considered

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. LowTohqueonSmallPortionofSampledStructuralBolts(CAL 85-21-36)

AttherequEstoftheNRCCatinspector,6487/8-inchdiameterd325 ' '

high strength bolts.were checked for proper. installed torque for friction type connections. The installed torque values'of 23

't(approximately.4 percent) of the 648 bolts inspected were found to be

' below the torque requirement of 450 ft-lbs. Nonconfonnance reports CC03132 and CC03134 were written to repair the improperly installed

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Subsequently, it was determined by an engineering review of the structural. design that the connections identified in the above NCRs are bearing-type connections and need not be considered as friction-type connections. Bearing-type connections, even though normally tightened the same as friction-type connections, actually do not require tightening to a specific torque in order to develop their shear load capacity. Therefore, the finding of some not-fully torqued or even loose ~ bolts is inconsequential -for the structural-integrity of the identified bearing-type connections.' Accordingly, no corrective action other than retightening all bolts for good workmanship was required for those connections. Although all connections were originally installed as friction type connections, only about three hundred (including Units l' and 2) must be friction connections to satisfy design requirements. Of these three hundred, over one hundred were reexamined during the review of Brown & Root .

documentation (reference NCR GC-00152). -Closure of this NCR provided verification of..the adequacy of those connection ,

.To resolve the question of adequacy. of quality control inspections, a statistical survey was undertaken. The results of that' survey showed

! a 95 percent confidence that at least 95 percent of all connections were satisfactorily inspected (95/95 confidence).

The statistical approach used was the " Likelihood Density Function Method" (NUREG/CP-0063, Proceedings of the 1984 Statistical Symposium

.on National Energy. Issues, October 16-18,,1984; paper entitled

" Sampling Inspection-of Nuclear Power Plants" by Julius Goodman, pages 213-227). Utilizing this method, with an infinite-population, a sample size of 59 with no failures provides 95/95 confidence.' A

, computerized random number generator was used by HL&P QA to select 59 L connections from a population of 5,363 connections with 2, 3, or 4 bolts per connection. Two, three, or four bolt connections were chosen to allow a pass / failure criterion that: two bolts meeting

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minimum torque requirements, verify that the connection had been

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adequately inspected in accordance with American Institute of Steel Construction Code (AISC) requirement During the survey, 59 connections with a total of 194 bolts were examined (16 bolts were inaccessible). All connections passed the pass / fail criteria. (Only one bolt failed to achieve full torque, 280 vs. 430, in a three bolt connection.) Based on these results, a 95/95 confidence was achieve The NRC inspector reviewed Surveillance Reports SH-0761 and SH-0762 which provided documentation of the bolt reinspection In friction-type connections, proper tightening verified by inspecting torque is essential for their structural integrity. The adequacy of the STP friction-type connections was resolved by demonstrating the reliability of the STP Quality Control Inspection program for the installed connections. The inspection performed-in accordance with the AISC Specification provisions for an arbitration inspection is not 100 percent, but is limited to 10 percent of the bolts with a minimum of two bolts per connection. Accordingly, the isolated findings of under-torqued bolts discovered during a 100 percent inspection of certain connections are not necessarily indicative of a deficient QC inspection program, but nevertheless dictated additional consideration through a statistical survey of installed and inspected connections, as described abov The statistical survey sample taken provided the needed confidence in the past Quality Control Inspection Program and eliminated the need for a reinspection of the limited number of friction-type connections identified in the STP desig In bearing-type bolted connections, the shear force is transmitted by bolts in fitted (non-slotted) holes bearing against the connected parts. The shear load capacity of the connection does not depend on the amount of bolt pretension introduced into bolts tightened to specific torques. Therefore, the capacity of a partially tightened bolt will be,the same as a fully tightened bolt as long as the bolt remains in plac The nuts, even if not tightly torqued, assure that the bolts will remain in place and will not fall out due to vibration, if an It is recognized in the design process that occasional bolts may be installed with less than fully torqued values. This is acceptable for bearing type connections based on the design assumptions made and the inherent factors of safety provided in structural steel design by the AISC Code. Based on the NRC inspector's review of the analyses conducted by the licensee, it has been determined that no safety hazard exists from the conditions identified during the CAT inspection. This matter is considered close , --. - - - -. -. .. .- . , - _ --.

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+ OvertightStructural/~oltsiNSlidingConnections(CAL B 85-21-37,

PEA 6.f, Violation 86-12-I.A 6) l

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.Nine structural: steel slotted connections containing'68 7/8 inch

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diameter.A325 high strength' bolts were inspected by the NRC CAT

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> inspector for proper installation torque. Forty-three bolts were-

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found to be installed at torque values _ greater-than'the. inspection

. torque value required by the design drawings. 'Nonconformance

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-Report-(NCR) No. CC03190 was issued to document:the' identified 2 . -deficiency. The NRC CAT inspection, finding concerned the~1ack of

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specificity' of QC inspections for " snug-tight": bolting as evidenced by :the finding of over-tightened bolts. Although no specific torqui.ng requirements, other than snug-tight, were specified in the

' design drawings, the basic definition of a snug-tight connection, as invoked by the-AISC Specification, " Structural Joints Using ASTM A325 _

or A490 Bolts," zis described as the full force of a man on a spud -

wrench. The, torque valve ordinarily associated with this definition

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is 150 ft-lbs.

4 Structural steel connections inside the Reactor-Containment- .

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' Building (RCB) and the Isolation Valve Cubicle (IVC) are designed to

. provide free' movement'of the members to-relieve thermal stresses. -

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s' .resulting from accident condition thermal loads. Slotted connections with snug-tight bolts arelspecified for this purpose. As a result of

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_the over-tightening of the bolts that were specifled to-be~only-snug-tight, a reanalysis was performed to evaluate the effect of the accident condition ~ thermal loads on-the.as-built condition.- Bechtel

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Calculation No. CC-5805. documents this analysis and was evaluated by-

.the NRC inspector. When accident condition thermal movement of structural steel framing is inhibited, axial stresses induced in the

. members can be: relieved only if the connections at the supporting

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wall allow displacements to take' place. This movement can be accommodated by either the clip angle connection.on the embedded plate or by yielding of-the anchorage embedded in concrete'. All ,

structural steel members are attached with heavily stiffened clip angles that are primarily designed to resist lateral seismic. load Relief from thermal ~ loads can be expected from bolt hole installation

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diameters which, being larger than the bolts, allow a small amount of

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movement and by yielding of the embedded plate anchorages. Embedded F 1 plate anchorages consist of ASTM A-36 rod material and Nelson studs conforming'to ASTM A-108. The reviewed analysis showed that the embedded plate anchorages would yield to permit the calculated

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displacements of the structural steel members to relieve accident T condition thermal ' stresses without compromising their ability to withstand the design vertical and lateral loads. Calculated wall  :

displacements under postulated accident conditions were reviewed and-

found to be very small and able to be accommodated by the analyzed

, deformations of the connecting clip angles and embedded plate anchorages,-thus-preventing any significant load transfer. Since the

' load carrying capability of the over-tight connections is greater ,

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1 than the expected loads, no unacceptable failure which would impact on the safety-related function of the structures in question is expected to occur under the design loading condition In response to the lack of design drawing specificity which led to excessive torquing by construction and did not provide QC inspection with verifiable inspection acceptance criteria, revisions to all pertinent design drawings were issued to provide an acceptable range of torque values. Regardless of the analytical disposit. ion of the over-tight connections identified by the NRC CAT inspection, three nonconformance reports were issued for the purpose of disposition by reworking. The NCRs and documented dispositions were reviewed by the NRC inspector. Based on the results of this inspection, this matter is considered close Welding Across Beam Flanges (CAL 85-21-38)

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The NRC CAT inspector reviewed Bechtel project STP Specification [

No. 3A010550030, Revision 5, " Erection of Structural Steel and {

Miscellaneous Steel," and noted that welding across the flanges on (

fully loaded structural steel members was permitted. The NRC CAT inspector asked if an engineering evaluation had been performed (similar to that indicated in AWS D.1.1, Section 7.5.1) to determine whether or not a member is permitted to carry a live-load stress while welding on i The following provisions for transverse welding across flanges are

- defined in Subsection 7.2.2.9 of the STP Specification 3A010SS0030:

" Transverse welding across flanges is permissible in all instances if the weld size is equal to or less than 0.75 times the flange thickness. If the weld size is larger than 0.75 times the flange thickness, the weld may not be made in the middle one-third of the member's length, but is permitted elsewhere. If the member is supporting only its own dead weight (i.e., concrete slabs, or upper levels are not in place), the above limitations do not apply. The above limitations do not apply for welding onto auxiliary steel members and onto structural steel members which are part of cable tray, conduit or HVAC duct supports."

The evaluation required by AWS D1.1 Code is intended as a precaution for welding operations performed on members carrying live-load stresses that presumably are relatively high and approach the allowable capacity of the members. The evaluation assigned to the Engineer by the code is fulfilled by the following considerations:

(1) Beams for Category 1 structures are characterized as simply-supported beams with maximum fiber stresses at midspa _ _ _ _ _ _ - _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _

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Therefore, the provision to exclude large welds from the middle third of the member's length prohibits large welds in the high stress area (2) At least 35 percent of the total governing flexural stress of the Catagory I structural beams originates from seismic stresses. Thus, members are stressed, at most, to 65 percent of-their allowable capacity, assuming the unlikely case that full loading existed when the welding was don (3) The primary concern with welding on an installed member is the reduction in yield strength that occurs at elevated temperature The primary means of minimizing the temperature increase in the member is to limit the size of the transverse weld. The provisions of the Bechtel specification address weld size limitation Based on these considerations and the provisions of the specification, welding across flanges will not degrade the structural integrity of the members in question. This matter is considered close e. Expansive Clay Under ECW Pipe (CAL 85-21-39)

The NRC CAT inspector identified a potential problem which they felt may not have been addressed by the license The concern specifically related to the potential af the expansive A2 clay' layer to swell upon return of ground water to normal levels and the potential effect of this swelling on the ECW pipin The NRC CAT report states in part, "If the moisture content of the clay layer during the dewatering period has been reduced significantly and then the dewatering system is discontinued, the clay layer when exposed to the returned ground water is expected to expand."

The NRC inspector reviewed Woodward-Clyde Consultants (WCC) Report No. D160XR035-WL and WCC Letter ST-XC-YB-0152, February 17, 198 I, was determined that the potentially expansive nature of the A2 clay layer was recognized and taken into consideration in the foundation design. WCC performed laboratory tests on samples from the A2 clay layer and concluded that the clay exhibited expansive behavio However, the A2 clay layer was not expected to experience significant moisture content changes except near exposed surfaces in excavation Construction procedures were implemented to minimize the amount of time the clay in excavations remained expose The moisture content of the clay was not changed significantly either from exposure during construction or from dewaterin During construction, the clay in the trench excavation was exposed for only short periods before it was insulated from the effects of sunlight and win r.:

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Dewatering does not dry out the soil, it only removes the free water in the voids that is not retained by capillary actio The degree of saturation of the A2 clay layer prior to construction was 100 percent; the degree of saturation during construction is conservatively estimated to be 90 to 95 percen A comparison of test results from the preconstruction period to the test results obtained during 1982-1983 ECW trench excavation (about 7 years after the start of dewatering) demonstrates that the average moisture content of the A2 clay layer changed very little (24.8 percent to 24.6 percent).

When the moisture content is above 20 percent and the degree of saturation is greater than 90 percent, there is very little potential for volume change. Based upon linear swell tests and the moisture conditions at.STP, WCC estimateo the upper bound of volume change in the A2 clay layer to be on the order of 0.25 percen In addition, the confining pressure of the trench backfill, which is approximately equal to the average swell pressure of 0.75 tons per square foot for the A2 clay layer, will virtually eliminate the small amount of swell otherwise anticipate Finally, design criteria provisions require that a certain amount of differential movement be considered in the analysis of piping. The ECW pipe is designed to withstand the effects of 1-1/2 inch differential _ settlement per 1,000 feet of pipe length. Any volume change due to rewatering will occur over a large area. Consequently, the potential for significant differential movement due to clay volume change is' negligible. The flexibility of the piping system would enable the pipe to accommodate any differential movement that might result from clay expansion. This further alleviates any concern about swelling of the clay. Based on these facts, this matter is considered close . Exit Interview The NRC inspector met with the licensee management and their staff at the conclusion of the inspection to inform them of the dispositian of the followup items. Findings were discussed with established management representatives during the course of the inspection.

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