ML20235Q805
| ML20235Q805 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 09/30/1987 |
| From: | Cockfield D PORTLAND GENERAL ELECTRIC CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| TAC-65726, TAC-66589, NUDOCS 8710070705 | |
| Download: ML20235Q805 (28) | |
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M M W General BectricCompt 13r David W. Cockfield Vico President, Nuclear September 30, 1987 Trojan Nuclear Plant Pocket 50-344 Licenso NPF-1 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555
Dear Sir:
Pipo Support Design Verification - Lonn-Term Plan On August 18, 1987, Portland General Electric Company (PGE) pubmitted the results of the short-term Pipo Support Design Verification Program. This program consisted of verifying the design of 493 cupports which included all of the safety-related supports designed by the Civil Engineering Group of the Architect-Engineer ( A-E) for the Trojan Nuclear Plant. A long-term program has been developed to verify all other safety-related-large-bore pipo supports. A description of this long-term program is provided in
- Attachment A.
During the Nuclear Regulatory Commission (NRC) review of the short-term pipe Support Design Verification Program, 19 concerns woro raised.. Nine of those concerns were resolved by our letters of July 27, 1987; July 31, ! 1987; and August 18, 1987. Responses to the remaining concerns aro , provided in Attachment B. ! One of the concerns identified in the short-term program was the adequacy of rock bolto used in pipo support anchorages. The rock bolts in pipe supports woro demonstrated to be adequato and a commitment was made to demonstrate the adoquacy of rock bolto used in pipe whip restraints. The ) I program to conduct this verification is described in Attachment C. Sinco the original deficiency leading to the performance of the short-term Pipo Support Design Verification Program was attributed to the A-E's Civil Engineering Group, PGE intends to evaluate other design activities per- ! formed by this group to determine if additional design verification is 1 necessary. PGE's plan for performing this evaluation is provided in Attachment D. j 8710070705 , DR ADOCK PD ! 1 M{. : 1215.W S&non Stmet Port:anci Oregon 97204 i i
N , e Portland General BechicCornpany Document Control Desk September 30, 1987 Page 2 i As part of the short-term program, a quality assurance / technical audit of the A-E was conducted by an independent PGE consultant. During this audit, i several concerns were identified. In Attachment D2 of our August 18, 1987 letter, four of these concerns were resolved. The'A-E's response to the remaining findings and recommendations and PGE's assessment of the A-E's response is provided in Attachment E. PGE has selected another inde- ! pendent consultant and intends to conduct a two-week quality assurance / technical audit of the A-E's overall engineering activities. This audit will be conducted from October 12-23, 1987. The scope of the audit is described in Attachment F. Sincerely, Attachments c: Mr. John B. Martin Regional Administrator, Region V l U.S. Nuclear Regulatory Commission Mr. David Kish, Director State of Oregon ,
.I Department of Energy Mr. R. C. Barr NRC Resident Inspector Trojan Nuclear Plant i
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l f t , Trojan Nuclear plant Document Control Desk Docket 50-344 September 30, 1987 Licenso NpF-1 Attachment A j l l LONG-TERM pipe SUPPORT DESIGN VERIF7 CATION PROGRAM SCOPE i l l The long-torm pipe Support Design Verification program will verify the design of all large boro (1 -1/2 2 inches nominal pipe size) pipe supports on safety-related piping systems. This program will include pipo sup-ports for non-safety-related large-bore piping and all small-bov9 piping ) which are included in the solomic analysis for the safety-relathd large-bore piping. I The supports to be verified will be identified th.cm:gb a, review of cur-rent pipe stress analysos and drawings. The safety-reldted piping iso- { 1 metrics will be identified from the piping classification recordn. The pipe stress analyses for these isometrics will be reviewed to identify 1 the safety-related supports. All supports, snubbers, restraints, anchors, i und gravity supports will be identified. The support designs will be verified using the support loads generated by the current pipo stroca analysis and the support configuration identified in the current revision.sf the pipe support drawings. Where possible, support loads and other design basic information will be verified with the vendor responsible for this input during original construction. Utilizing this information, support celeulations will be reviewed. If ' the calculations are determined to be insufficient or incomplete, now calculations / evaluations will be perforned such that, at the completion of the program, all safety-related large-bore pipe supports will have complete design documentation. The review will extend from the attach-ment to the pipe to the Anchorage. The load rated components, structural steel, wolds, bolts, baseplates, and anchor bolts will be verified. A procedure will be developed for the verification program and will include the steps of the verification process, as well as acceptance cri-teria. A support design will be considered to be verified if.it meets the code allowables for the applicable load combinations from the Trojan Final Safety Analysis Report. 'lqu dngineers performing the verification will receivo training on the procedures and quality assurance requirements. SAB/kal 1716p.987 i l
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3 Trojan Uucicar plant Docunent Control Desk Docket 50-344 Snptember 30, 1987 License NPF-1 Attachment B
/ page 1 of 11 ;
'I 1 RESPONSES TO NRC CONCERNS pjoncern l' ;
l In calculation RC-61)1R-2-1-H-26, dated 4/22/83 by 7, Dang (Prcosurizer Blowdown System RC-01'EK1), the support calculation shows hanger H-26 and fi snubber SS-16 suppot ted by a beam. Huwever, the sopport drawing shows ) hanger H-26 and hunger H-27 (not SS-16) supported by a singic beam. j Comment on the apparent discrepancy.
Response
The pipe support calculation for RC-601R-2-1-H-26 that wt/2 generated for ) Revision 0 of calculation RC-01 for the pressurizer blowdown system, also j included the effects of SS-16 and H-27 connected to the same beam. The I hanger drawing RC-601R-2-1-SS-16 was subsequently "Voidod" as it was no j longer required in the piping stress analysis. When the loads from i Revision 1 to calculation RC-01 were reviewed, the portion of the support calculation tW L included SS-16 was not revised to deleto the offects of l the snubber loada, because the calculation was conservative with the ; l cr.tra load. The pipe suppest, calculation has been revisM to note this ; ) condition. CpJrjecen 2: Describo problems caused by revision of Computer program ME101, " Linear Elastic Analysis of Piping Systems", during the period of March 1984 to March 1987. Verify whether such probicms affect piping or support design in Trojan plant. Indicato correctivo actions if problems exist. i Response: f
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i During the period from March 1984 to March 1987, an error in the calculation <>f stresses may have occurred in ME101 if all of the following options were c;ecified:
- 1. CODE = SC3W75 (stress calculations in uccordance with ASME III 1974 s
through Winter 1975 Addenda)
- 2. CLASS = 1 (ASME III-NB design'ralen)
- 3. SOLVER = EAL (ME101 cubprogmm)
- 4. TEE = XXX (A tee section where branch pipe to run pipe diameter ratio is greater than 0.7) e 1
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Trojan Nuclear Plant Document Control Desk Docket'50-344 September 30, 1987 License NPF-1 Attachment B
'Page 2 of 1,1 y
Thic error has been corrected. I 1 d l J The error did not. affect any calculations performed for the Trojan , Nucicar Plant since no Trojan calculation used this combination of ME101-options.
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Concern 3: i j Verify extent'of missing original calculations of piping and' support l design in Trojan plant. Indicate corrective actions.to ensure design. lj adequacy if the problem exists beyond the scope of design performed byL the A-E Civil group. , s i
Response
This NRC concern was raised prior.to the development of the long-term ' Pipe Support Design Verification Program. -As a result of this program, all large-bore safety-related pipe supports will have design calculations on file demonstrating the adequacy of the support, as described in Attachment A. The long-term program will identify where original calculations for large-bore pipe supports do not exist or are inadequate, develop calculations as required, and perform support modifications if' 'l found to be necessary. l Concern 4: Identify procedures for controlling recalculations whenever there is a ! change of loads, including documentation requirements.
Response
Performance of calculations, including documentation, is governed by A-E Engineering, Department Procedures 4.37, " Design Calculations", and.4.25, "Desit,n Interface Control". Specific requirements regarding pipe support calculaM on updates are contained within Project Engineering Instruction (PEI) No. 11, and will be contained in the procedure for the long-term Pipe Support Design Verification Program. PEI-11 requires that pipe rupport load changes initiate a design load comparison, an evaluation of increased design loads, and a' revision of the support calculation
- l. to docuent the. acceptability of the support. Similar procedures for.
maintenance of pipe stress analyses and pipe support design.for work performed by.other than the A-E will be. formalized by January 1, 1988. f. l
Trojan Nuclear Plant Document Control Desk Docket 50-344 September 30, 1987 License NPF-1 Attachment B page 3 of 11 i Concern 5: In calculation J-307, for support SS-1047, sheet 44, it is stated that i reactions for SS-1047 are less than reaction used in Type II case 3 i generic support calculation. This is not true for bolt shear loads ; (SS-1047 V=20 kips while for Type II case 3. V=10 kips). Provide basis l to show higher shear load is acceptable. I
Response
The support for SS-1047 consists of a horizontal member and a knee brace, each having a baso plate as shown on Figure 1. By virtue of the location
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l of the applied loads, the 20 kips load is distributed to both base plates. The shear reaction to each base plate is approximately 10 kips. The critical base plate which is a'ctached to the knee brace has 2 bolts. For the Type II Case 3 support, which also has two bolts in the base l l plate, the bolts were evaluated for a 10 kips shear reaction and found to l l be adequate. The base plate attached to the horizontal member is acceptable by l comparicon, sinco it has 4 bolts rather than 2. Therefore, l support SS-1047 is acceptable. Concern 8: The original pipe anchor calculation #1 references design loads of M 594.2 kips-ft. and F=224 kips given by Stress group. What is the source of these loads and what do they represent? I
Response
These loads (Moment = 594.2 kips-ft, Force = 224 kips) were based upon pipe rupture jet forces and ultimate bending moment capacity of the pipe, ; and represented a conservative approximation of the load transmitted to the anchor. These loads were utilized since actual piping stress analysis had not been performed at the time of the anchor design. In the recent Pipe Support Design Verification Program, the pipe anchors were evaluated using current piping stress analysis-based loads. Concern 9: What is the basis for the Dywidag anchor bolt allowables? (Ref. original pipe anchor calculation #1). Why was tension-shear interaction not considered? How will they be treated in the new calculation?
Trojan Nuclear Plant Document Control Desk Docket 50-344 September 30, 1987 License NPF-1 Attachment B Page 4 of 11 - FIGURE 1 i 10K 10X
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7 Trojan Nuclear Plant Document Control Desk Docket 50-344 September 30, 1987 License NPF-1 Attachment B Page 5 of 11
Response
Dywidag anchor bolts are used to transfer pipe anchor loads to the !' concrete walls and are through bolts. The ultimate strength of the bolt material (ASTM A722) is 160 kips per square inch (ksi). The bolts are post-tensioned to obtain a final effective prostress force of 60 percent of the ultimate strength as permitted by ACI 318-63, Section 2606. The 60 percent of the ultimate strength (equal to 96 ksi) corresponds to an i effective prestress force of 120 kips and is used as the allowable tension load for the accident / safe shutdown earthquake (SSE) condition. For the allowable tension load during normal operation, the allowable ; tension load is reduced by a factor of 1.5 as required by i Section 3.8.4.3.3.2 of_the Final Safety Analysis Report. Shear forces are considered to be transferred through friction between the base plate and concrete wall. A coefficient of friction of 0.55 (steel on concrete) was used to determine the allowable shear loads. (Reference ACI 349-85, Section B.6.2.2.2.) The following table summarizes the allowable loads: Allowable Loads (pounds) (1-1/4 inch 0 bolt) Loading Tension Shear
- a. Normal Operating Condition 80 kips 43 kips
- b. Normal Operating condition 100 kips 53 kips ,
I with operating basis earthquako [=1.25x(a))
- c. Accident /SSE Condition 120 kips 65 kips
(=1.5x(a)) Shear-tension interaction was not documented in the old calculation since the shear force demand in general was low and, therefore, shear-tension interaction was not considered critical. Shear-tension interaction is conservatively assumed as linear in the new calculations, and the shear-tension interaction ratio was demonstrated to be less than 1.
.__-__-____-_____a
Trojan Nuclear plant Document Control Desk Docket 50-344 September 30, 1987 Licenso NPF-1 Attachment B Page 6 of 11 Concern 11: For the supports listed below, the eff6ct of welded attachment loads on the piping system was not considered.
- 1. EBB-3-2-SS-10 EBB-3-2-SS-11 EBB-3-2-SS-13 EBB-3-2-SS-14
Response
The local stcess effects due to welded attachment loads on the piping system were not documented in the piping stress analysis. The welded attachments for these 4 snubbers have design loads that are within the-vendor generic allowatla loads for these attachments. The documentation of the acceptability of these welded attachments on the piping system will be included in the pipe stress analysis by November 1, 1987. Concern 14: Support SR-1155 does not appear to fit into tho typical support categories referenced in the calculat.lon (Type I and V). Furthermore, drawing C-386 in conjunction with SI-2501R-2-3-SR-1155 does not give sufficient detail to define the actual support configuration without significant guesswork on the part of the analyst. How can we be assured that the au-built configuration matches the configuration assumed by the , analyst? This concern may affect other supports in the 265 group.
Response
Support SR-1155 does fit into the typical support categories referenced in the calculation, namely Type I generic support (see Figure 2). The braced cantilever shown on Drawing SI-2501R-2-3-SR-1155, is actually attached to the civil-designed portion of the pipe support. The civil-designed portion of the pipe support is a cantilevered column i section, similar to the Type I generic support. Therefore, the analyst for the civil-designed pipe support did use the correct configuration and i there was no need to question the as-built condition of the. design. l
- Trojan Nuclear Plant Document Control Desk Docket 50-344' September 30,'1987 Licence NPF-1 Attachment B Page.7 of 11 FIGURE 2 l ,
E I y i' Shown on l SI-250lR-2-3-SR-1155 t .(CesignedbyBergen-Paterson).
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Trojan Nuclear Plant Document Control Desk Docket 50-344 September 30, 1987 License NPF-1 Attachment B Page 8 of 11 Concern 17: 79-14 Program calculation review indicates: A.1. new support calculations are skimpy
- 2. pipe support detail drawings were not always updated for revised loads B.2. significant seismic load reductions were noted
Response
17.A.I. The long-term pipe support design verification program will review all large-bore safety-related pipe supports. The review will result in complete and documented pipe support calculations. I 17.A.2. The Architect-Engineer will identify all supports that require load updating during the long-term program and forward updated information to Portland General Electric, i I 17.B.2. Prior to NRC Inspection and Enforcement (IE) Bulletin 79-14 I
" Seismic Analysis for As-Built Safety-Related Piping Systems",
the stress analysis of the charging pump discharge piping l utilized the 61 foot elevation spectra that enveloped all l supports on the piping system. The IE Bulletin 79-14 review I identified differences between the as-built configuration and the original analysis. The following approach was used to i reanalyze the piping: I
- 1. The piping from the charging pungs up to the hanger on the 3 inch piping noted on Figure 3 and the remainder of the j piping was analyzed using the ground spectra which enveloped the spectra for that portion of the piping. This calcula-tion was reviewed by the NRC.
- 2. The 3 inch branch from the anchor at 59 feet 6 inches down to the charging pumps was reanalyzed as part of the Control Building modifications using the envelope of the 61 foot elevation and the ground spectra.
The supports for the piping system were evaluated for the design loads from both of the calculations.
Trojan Nuclear Plant Document Control Desk Docket 50-344 September 30, 1987 License NPF-1 Attachment B Page 9 of 11 In addition to Calculation 15-22, seventy-two other IE Bulletin 79-14 calculations have been reviewed. Including Calculation 15-22, only five stress calculations were found to have similar load reductions (>10 percent) on supports. The 73 calculations include piping in the following systems: Chemical Volume and Control System Component Cooling Water System
- Containment Vent Monitoring Diesel Generator Fuci Oil System
- Spent Fuel Pool Cooling System Main Steam System i Miscellaneous Small Pipe (<2 inches in diameter) l Reactor Coolant Pump Seal Injection System l Residual Heat Removal System
- j Safety Injection System
- Service Water System
- The systems identified with an asterisk (*) are those in which the five calculations were found to have similar load reduc-tions. The load reductions for each calculation were evaluated and determined to be justifiable and reasonabic. Reasons for the load reductions are as follows:
- 1. Grouted penetrations have been included in the piping analysis which has the not offect of adding more restraints to the original analysis.
- 2. Actual As-Built geometry was included in the analysis instead of the original design geometry.
l
- 3. The reanalysis used a response spectra that more closely represented the spectra for the piping system than the original conservative response spectra. In addition to the spectra changes in Calculation 15-22, calculations in the diesel rooms use spectra generated for those specific. areas instead of the original design spectra.
Concern 18: The NRC IE Bulletin 79-02, " pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts", calculation for CS-151R-7-51-H-44 contained calculations for anchor bolts and allowable load interaction. The package did not contain information listing what type of anchor bolt was used. The calculation used allowables for 3/8 inch Phillips bolts in block wall. Later the allowables were changed to concrete allowables with no explanation as to why. Was the correct allowable used? What l
l i { Document Control Desk j Trojan Nuclear Plant September 30, 1987 ) Docket 50-344 Attachment B License NPF-1 Page 10 of 11 - i FIGURE 3 (QA M pJf POMP DISCH. D DOW < 9reses Cat c. 15-22 1 i I 4 9 e !
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Trojan Nuclear Plant Document Control Desk Docket 50-344 September 30,.1987 License NpF-1 Attachment B
'Page 11 of 11 controls were in place to assure that the correct allowables were used?
I Also,' base plate flexibility was not considered.
Response
The calculation for the base plate for Support CS-151R-7-51-H-44 (H-44), was inadvertently provided to the NRC as part of the calculation record for IE Bulletin 79-02. Support H-44 is located on a small-bore (2 inch) line. Small-bore piping was not part of the 79-02 program because its supports were not based upon detailed. computer analysis, but.on a simpli-fled (span chart) methodology. The calculation for.H-44 was initiated but was not finalized since the small-bore pipe supports were not part of the 79-02 program requiring analysis.of.the base plates. However, the support calculations in question have been evaluated. It has been verified that the support H-44 is attached to a masonry. wall. , I The change to concrete wall allowables was not correct. A detailed. piping analysis was performed. Using the calculated loads and load combinations in the FSAR, H-44 was evaluated for design adequacy. The pipe support met the appropriate allowables for masonry walls. ! Bechtel's calculation procedures in effect during the 79-02 program required that the calculations be subject to independent checks which would ensure that proper allowables were used, and that base plate flexibility was considered. The long-term design verification program will review all large-bore safety-related pipe supports and ensure that base plate. flexibility has -been considered and that appropriate allowables are utilized. SAB/kal 1712p.987 1
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l Trojan Nuclear Plant Document Control Desk Docket 50-344 September 30, 1987 License NPF-1 Attachment C l PIPE WHIP RESTRAINT ROCK BOLT VERIFICATION PROGRAM SCOPE As part of the short-term Pipe Support Design Verification Program, the i rock bolts used in support anchorages were demonstrated to be adequate l for the design loads. It was identified during that verificat'on that .] rock bolts are also utilized to anchor many of the pipe whip restraints. ) It is the intent of this program to demonstrate the adequacy of the rock ) bolts in these pipe whip restraints. j The verification will be performed by first establishing rock bolt acceptance criteria from analyses and measurements of the embedmont depth of the pipe whip restraint rock bolts. The loads on the rock bolts will then be determined based on the whip restraint resistance mechanism and compared to the acceptance criteria. Additionally, pipe thermal move-ments will be determined and, along with the pipe break thrust loads, will be used to verify the impact loads for evaluation of the whip restraints. These analytical methods are described in FSAR Section 3.6.1. The calculations for the whip restraints, including the rock bolts, will be verified for the required loads. ; i SAB/kal 1716P.987
Trojan Nuclear Plant Document Control Desk ) Docket 50-344 September 30, 1987 License NPF-1 Attachment D DESCRIPTION OF PLAN FOR EVALUATING OTHER ARCHITECT-ENGINEER CIVIL DESIGN ACTIVITIES I In view of the identification of several original design deficiencies caused by the Civil Engineering Group of the Architect-Engineer for the 1 Trojan Nuclear Plant, it has been determined an evaluation of other Civil design activities is prudent. This evaluation will be performed for the ! purpose of identifying any additional areas where potential design l deficiencies may exist. The major Civil-designed structures at Trojan are those required to resist design basic seismic loads (ie, Intake Structure, Containment, Containment internals, Turbine Building, Fuel Building, Auxiliary Building, Control Building, Diesel Generator Enclosure ' Main Steam Support Structure, and Auxiliary Feedwater Pump Enclosures). In addition, the Civil group was involved in the design I of cable tray supports, equipment supports, and field-erected tank I foundations. A technical audit of selected design documentation for each of the Civil designs will be performed before July 1, 1988. The technical audit will 4 l consist of a review of the design documentation for consistency with the I Final Safety Analysis Report design bases. A representative sample of j the calculations will also be checked. This audit will be conducted by a J team consisting of experienced civil engineces. 1 Based on the results of these technical audits, a more detailed design verification may be performed. The scope and schedule for the design l verification will be determined as part of the evaluation. i i l l SAB/kal 1716P.987
i Document' Control Desk
, . September _ 30, 1937 Attachment E BechteI~ Western Po4Fdoddany-Engineers - Constructors Fifty Beale Street San Francisco, California Mall Address: PO Box 3965. San Francisco, CA 94119 In' reply please reference:
Doc. Control No. T029428 BP- 13109 SEP 3 01987 Mr. A. N. Roller Manager, Nuclear Plant Engineering
'ortland General Electric Company 1:1 S. W. Salmon Street rortland, Oregon 97204
Subject:
Portland General Electric Company-Trojan Nuclear Plant - Job 11760 Resolution of Remaining. Quality Assurance Audit Findings and Recommendations
Reference:
BP-13044 dated August 11, 1987
Dear' Mr. Roller:
Portland General Electric contracted with Impell to conduct an audit of Bechtel of the outage " Pipe Support Verification Program" in our offices on August 4 and 5,1987. The reference letter contained responses to 3 findings and 1 recommendation which have previously been transmitted to the NRC by PGE. This letter transmits the responses to the remaining findings (1,2 and 6) and recommendations for your use. If you have any questions on this material, please do not hesitate to call us. Very truly yours, R.-W. Fosse Written Response Reg'd: No RWF/eng
Attachment:
As stated TL17/42 12649 A Unit of Bechtel Power Corporation
Document Control Des D September 30 1987 , - Attachment E l Page 2 of 7 I RESOLUTION OF REMAlNING QUALITY ASSURANCE AUDIT FINDINGS AND RECOMMENDATIONS i Findings
- 1. Quality Assurance (QA) Document Control q Deficiency:
i It was observed that Controlled Copy 6 of the Project Engineering Procedure I Manual (PEPM), Revision 42, for the Trojan Nuclear Plant (Trojan) Job 11760 ; which was assigned to Karl K. Gross, Civil Group Supervisor, contained the following discrepancies:
- a. Engineerin9 Department Project Instruction (EDPI)-4.27.1, Revision 1, issued April 9,1981 was missing. (
Reference:
Engineering Department l Procedure (EDP)-5.10, Revision 2, Paragraph 3.2.) -
- b. Twenty-nine documents (procedures, etc.) were filed in the PEPM i that were not listed in the Revision 42 Table of Contents. '
Corrective Action: Revision 1 of EDPI-4.27.1 was inserted in PEPM Copy 6. Other documents were removed from the cited PEPM to be in agreement with the Revision 42 Table of Contents. During the investigation, it was found that one other copy of the PEPM lacked the cited EDPI and was corrected. A verification has been completed of all other PEPMs to ensure they are current, and training has been conducted to reemphasize the need to promptly update the j PEPM after receipt of revisions.
- 2. QA Corrective Action Deficiency:
It was observed that Trojan Project Audit 3.2.12 dated May 20-22 and May 25-29,1987 and issued June 5,1987 required a written response by July 17,1987. As of August 4,1987, neither a written response requesting an extension ar documented evidence of completed corrective action had been received by the Project Quality Assurance Engineer (PQAE). Corrective Action: Interoffice memoranda were issued by the Project Engineer that requested extensions until October 18, 1987 to respond to Audit 3.2.12. TPF41/9-1
_ Document Control Desk
, ,' - September 30-, 1987 Attachment E'
- 6. Data Transmittal Traceability Deficiency:
Support loads were transmitted to the A-E.through telecopies (e.g. civil calculation No. J-307, Revision 2). Calculation filed have been approved I as " final" without formal transmittals. Corrective Action: The calculations will be revised to include the formal transmittals of the support loads. The calculations will be reviewed to verify that the formal loads have been noted in the support' evaluation. The calculations should have been approved as " Committed" status and-upgraded to " Final" upon receipt of formal load transmittals. Other calculations were found to have informally transmitted 19 formation f rom . Portland General Electric (PGE), for which formal transmittals have been requested. These calculations will also be revised to include the formally transmitted information. The need for formal information in " final" status l calculations will be reinforced in project training. l l Recommendations
- 1. Recommendation:
In Hangers Guidance Tables of Calculation No. S-1-300 for. Main Steam Fact < Valve Closure Analysis, Revision 0, dated June 17, 1987, support loads ) due to operating basis earthquake (0BE) and safe shutdown earthquakes j (SSE) are from superseded calculations. (Revision 2 of Calculations 1-29 -{ l and 1-30 were used instead of current Revision 3.) No explanation is ' given in S-1-300. Project personnel indicated that Revision 3 considered l additional impact loads of main steam piping interfering with rupture restraints during seismic events. Recent information from field walkdowns. l shows that this interference would not happen. Therefore, Revision 2 I support loads are still valid.
Response
Stress Calculation S-1-300, Revision 1 and the calculation ' log have been updated to provide traceable documentation of pipe support load inputs. Calculation 5-1-300, Revision I superseded Calculations 1-29, Revision 3 and 1-30, Revision 3.
- 2. Recommendation:
Review of the support qualification file (J-301) indicated that the latest loads have been used for the support qualification. However, all support drawings reviewed have not been updated to reflect the latest design load conditions. Example: The Main Steam Piping Support Hanger EBB-1 SS-80. TPF41/9-2 l
Document Control Desk September 30, 198) L} Attachment E
. Page 4 of 7 Document OBE Loads (lb) SSE Loads (lb)
D'rawing EBB-1 SS-80 3,434 5,735 Revision 3 (11/26/86) Calculation 1-29 9,032 14,232 Revision 1 (7/25/85) Calculation 1-29 7,475 9,113 Revision 3 (1/2/87) A-E project personnel indicated that Revision 1 loads of 1-29 were never transmitted to PGE and Revision 3 loads of 1-29 were transmitted to PGE on May 1987. It is recommended that PGE and the A-E update all support hanger drawings to reflect current loads during the long-term program.
Response
During the course of the long-term design verification program, pipe support drawing loads that do not reflect the latest piping stress analysis hanger guidance will be identified to PGE, along with the current loads to enable PGE to update the drawings.
- 3. Recommendation:
Review of the main steam piping analysis indicated that the evaluation is correct. However, the main steam piping analysis is documented in various 4 calculations. i Main steam piping analysis calculation files involved different organiza-tions and different time frames as follows: Gravity and hydro analysis: Subvendor calculation MS-C8 & C9 (Circa 1973) Thermal and seismic anchor movement analysis: A-E Calculations 1-29 and 1-30 (Revision 3,1/87) Seismic inertia analysis: A-E Calculation (Circa 1975) Turbine trip analysis: A-E Calculation S-1-300 (Revision 0, 6/87) There is no discussion of analysis history or calculation file cross-referencing in recently completed work. Example 1: Supports Guidance Tables in Calculations 1-29 and 1-30, j Revision 3 (1/87) have not been superseded, although in Calculation S-1-300, ) Revision 0 (June 1987), revised Supports Guidance Tables have been developed j and used in this Support Verification Program. i TPF41/9-3
Document Control Desk September 30, 1987
. Attachment E Page 5 of 7 Example 2: The OBE and SSE support loads in Hangers Guidance Tables of Calculation S-1-300, Revision 0 are the square root sum of the squares' values of the seismic inertia and seismic anchor movement as taken from 1-29 and 1-30, Revision 2. A definition of OBE and SSE support loads in Hangers Guidance Tables of BPC S-1-300 are recommended. (Note: The OBE and SSE support loads in Hanger Guidance Tables of 1-29 and 1-30, Revision 3, include seismic inertia, seismic anchor movement, and seismic loads.)
This documentation issue does not impact the support verification effort, however, it is recommended to add cross referencing or consolidate all I main steam calculations in one document during the long-term program. 1 Rerponse: All main steam piping inside containment was evaluated in Calculation l S-1-300, Revision 1 which consolidated the individual load case computer ! calculations. I \ l l 4. Recommendation: This recommendation is related to documentation of calculations and has no I impact on support qualification. Insufficient details exist for the l qualification of certain support components of the following supports: SR-82 The gravity loads taken from the subcontractor load summary SR-85: generated in 1972 are not traceable to the isometric l available for the reviewer to verify the load application. SR-82 As-Built sketch indicates missing bolt, but calculation SR-85: considered all bolts. Even though the side of the support with the missing bolt would be in compression, the missing i bolt should be documented and evaluated accordingly. SR-82 Through-bolts used 8th edition of American Institute of SR-85: Steel Construction Code (AISC), although design criteria PC-ll760-P-003 indicates 7th edition. Proper reference should be used. l SS-87 Judgement was used to assess the adequacy of the support l members. Member W8x17 was judged to be noncritical and was not evaluated in the calculation. Documentation should be l provided for this member. SS-1052: The snubber orientation in the calculation is shown as 17 . The system analysis model (Drawing 7108-S1-2501 R-19, Revision 15) shows the snubber orientation to be 8.5 . 1 The stresses are quite low such that this discrepancy in the orientation will not fail the support based on engineering judgement. Piping analysis review is recommended, however, to evaluate the impact of the discrepancy on adjacent supports. The A-E should document the interface with the stress group concerning the discrepancy between the support drawing and piping layout drawing. TPF41/9-4
.keh$N NE$hb AttachmenttE-Page 6.of'7 '!
'l SS-1052: The baseplate. and bolt stresses were verified by comparison. l The support is similar to a Type IV Generic' Support H-19, j
' Revision 1, Sheets 125-127 Job 6478. This'should be documented R in the calculation.
SR-1155: Qualification of Support SR-il55 on Calculation J-307,
-Revision 2, is done by similarity to generic support. Type 1 Case 9. Insufficient documentation is ~ provided in the calculation to justify the ; dissimilarities between support SR-1155 which is a braced-cantilever frame using W4x13 members under a 3.31 K. load versus Type 1 Case '9 generic l support which is- a W12x53 cantilever member with a 0.5 K
~ load.
Response
- a. SR-82-85:
l The lack of cortplete documentation in the subcontractor weight calcu-I lation on the main steam piping will be resolved during the long-term program. The calculation will be revised to reflect that one. bolt is missing i and reference the 7th edition of the AISC. 1
- b. SS-87.
The W8x17 is a short compression member and was deemed to be adequate based on engineering' judgement. The calculation will be revised to address the adequacy of the W8x17 more thoroughly. In general, temperature effects, where judged to be critical, were addressed and documented in the support verification program. In the example in the recommendation, temperature had no effect.and in this case was not documented. Documentation regarding temperature effects will be addressed in the long-term program.
- c. SS-1052:
The use of 17 in the calculation was conservative, but will be corrected in the long-term program. The as-built support detail and isometric both snow the 8.6* angle, , The analysis and as-built reconciliation account for this angle; therefore, there will be no impact on adjacent supports. The calculation for SS-1052 does state that the support 'is similar to Type IV generic support. TPF41/9-5 j i l
Document Control Desk' ! September 30,.'1987 Attachment E d .- SR-1155 : Type 1, Case 9 generic support is a W12x53 cantilever member'with a 10 K load. The braced cantilever frame, using W4x13 members shown on Drawing SI-250lR-2-3-SR-ll55, is actually attached to the civil- ; designed portion of the pipe support. The civil-designed portion of 'i the pipe support is a cantilever, similar to a Type I generic support. Therefore, qualification of SR-il55 in Calculation' J-307 is correct ' (see response to Concern 14'in Attachment B).
- 5. Recommendation:
This recommendation is strictly related to the documentation of calcula-tions and has no impact on the support qualification. Assumptions and/or. references are not clearly stated in the support calculations of the.following supports: ; 1 Support Calculation Revision- ; 7 J-308 1 9 J-308 1 SS-1047 J-307 2 i SS-ll33 J-307- 2 j' SS-1259 J-307 2 SS-1052 J-307 2 SS-1053 J-307 2 SS-ll41 H-307 2 SS-1154 J-307 2-SS-1155 J-307 2 5S-1174 J-307 2 SS-1177 J-307 2 SS-1263 J-307 2 SS-1268 J-307 2 Turbine STR 1 J-302 0 I SS-1130 RC-01 0 SS-135 J-307 2
Response
The calculations will be reviewed for completeness during the long-term program and assumptions / references will be added as necessary. TPF41/9-6 i
1 Trojan Nucicar Plant Document Control-Desk Docket 50-344 September 30, 1987 License NPF-1 Attachment F Page 1 of 5 DESCRIPTION OF TECHNICAL AUDIT PLAN { l During the period of October 12-23, 1987, a technical audit will be conducted of overall engineering activities of the Architect-Engineer ) (A-E) for.the Trojan Nuclear Plant. The audit will focus on recent work j tasks performed by the A-E which include: J Mechanical Syste.ns
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t Task ( Description I 121 Feedwater Pump Bypass to Condenser j 136 Feedwater Pump Low Suction Pressure Trip l 137 Auxiliary Feedwater System Operating Procedure ] 165 MOV Switch Setting Program l 175 Design Review of Main Control Room (HVAC) 1 177 Refueling Water Storage Tank Level Study I 181 Emergency Diesel Cenerator Room Heatup Analysis j 191 Turbine Building Pipe Break Analysis 202 Steam Cenerator Wet Lay - Up Recirculation System 207 I&E Bulletin 85-03 210 Main Steam to Auxiliary Feedwater Pump Turbine Check Valve Change 220 CCW System Network Analysis ; 228 Moderate Euergy Line Break Design Analysis 232 CCW System Concerns Electrical 063 Prepare Electrical Load List 064 Failure and Effects Analysis for 125V DC and Instrument AC System 124 Auxiliary Feedwater System Reliability 201 Identify Hazards Area 223 Auxiliary Feedwater Valve EQ Concerns Mechanical Components 056 RHR Letdown Lines - Addition of 2 MOVs and Associated Pipe ' Support Changes 058 Seismic and Stress Analysis of Reactor Coolant Pump Seal Water Retutm Line 060 Hold-Up Tank Cover Gas Piping Seismic Upgrade Evaluation 077 Install Isolation Valves for Two Reactor Coolant System Sample Lines 094 Evaluate Feedwater Bypass Lines to Condenser Trains A and B 116 Reanalysis of Main Steam Piping for Steam Generator Movement 120 Evaluation of Failed Mechanical Snubbers
Trojan Nuclear Plant Document. Control Desk j Docket 50-344 September 30,.1987 ) License NPF-1 Attachment F Page 2 of ,5 Mechanical Components (cont) Task # Description
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142 Condenser Uplift i 146 Reanalysis of Chilled Water Isolation Valve Change 149 Auxiliary Feedwater Suction Piping Analysis . 161 Analysis of Service Water Discharge Piping i 196 Main Steam Isolation Valve Bypass Piping Analysis 203 MOV Weight Changes 211 CCW Flange Misalignment at CV-3287 230 Evaluation of Service Water Piping l l Civil / Structural 051 Feedwater Heater Retubing - Turbine Building Z Line, i Structural Evaluation and Modifications ! 073 Evaluate Non-Seismic 1 Cable Trays and Support for SSE l 085 Safety Evaluations for Pipe Supports and Anchors Having Loose Nuts l 087 Turbine Building South Wall Evaluation l 097 Design a Lateral Support for North Side Block Wall in Control l Building 098 Evaluate Turbine Building Wall to Withstand SSE With Bracing Removed for Condenser Tubing 101 Identification Numbers for Safety-Related Masonry Walls 112 Civil Engineering Support 117 Seismic Category I Boundary Verification 123 Core Exit Thermocouple Upgrade 131 Foodwater Heaters 5, 6, and 7 Replacement - Building Modifications 139 Control Room Ceiling Evaluation l 184 Condenser Retubing - Building Modifications 191 Turbine Building HELB Analyses Additional items may be added or items deleted' subsequent to the initial l review by considering additional scopo details if appropriate. A review of the original design bases criteria and requirements will also be made to establish commitments made for the design of the system or structure. Several of the task packages listed above involve extensive multidisci-pline review. Primary responsibility for the initial review has been . assigned to one discipline who will coordinate design review activities l between disciplines as required.
Trojan Nuclear Plant Document Control Desk Docket 50-344 September 30, 1987 , License NPF-1 Attachment F Page 3 of 5 i The activities to be reviewed will be as follows: ) l' Mechanical Systems Review individual task packages-selected to determine the effect'of the work performed on the system or other support systems to meet established commitments and design requirements. Review discipline interface requirements to insure that any change has i been properly evaluated and documented in all affected disciplines. I Review calculations and associated documentation for each task to assure that recommendations are adequately substantiated and documented. Con-firm that evaluations are complete and verified in accordance with BPC's procedures which implement the requirements of ANSI N45.2.11. Review task packages, where applicable, to assure that provisions have been made to perform post modification testing of those changes which I affect the capability of the system to perform its safety function. Confirm that changes made which require modification of operating proco-duros, operator actions, technical specifications and response times are adequately identified and have been forwarded to PGE. Electrical Power Systems Review individual task packages to determine the effect of the work on the capability of the electrical system to perform its safety function (s) j and meet design requirements and commitments. l l Review related equipment qualification requirements and records to assure that the equipment is environmentally qualified to perform its safety function for the design basis specified. Review documentation substantiating tasks to confirm that the analyses are verified and completed in conformance with BPC's procedures which implement the requirements of ANSI N45.2.11. Review interdiscipline design requirements to assure that any change has been properly evaluated and documented by all affected disciplines. Mechanical Components Review individual task packages selected to determine if the recommenda-tior.s have been adequately addressed without reducing the original design margin or performed in a manner inconsistent with licensing commitments, design requirements or accepted industry practices (where appropriate).
Trojan Nucicar plant Document Control Desk Docket 50-344 September 30, 1987 License NPF-1 Attachment F page 4 of 5 Review discipline interface requirements to insure that any change has ! been properly evaluated and documented in all affected disciplines. { Review calculations and associated documentation for each task to assure that the recommendations are adequately substantiated and documented. Confirm that calculations are complete and verified in accordance with BPC's procedures which implement the requirements of ANSI N45.2.11. I Review the source of design inputs to assure that they are current and j consistent with the affects of the task on the component being evaluated. Review the.affects of the change on any vendor supplied design data and stress / test reports. In addition, make certain that substitution of equivalent components is completely substantiated. j Review the applicability of alternate analyses and design approaches used I to insure that they are equivalent or consistent with accepted industry i practice. ) Review the application of ANSI or other governing codes, correct use of l response spectra, FSAR commitments for load combinations and stress limits, piping nozzle loads, design temperatures and pressures, as well 3 as any transient dynamic affects. l l Civil / Structural i Review individual task packages selected to determine if the recommenda-tion has been adequately addressed without reducing the original design i margin or performed in a manner inconsistent with licensing commitments, l design requirements or accepted industry practices (where appropriate). l Review discipline interface requirements to insure that any change has been properly evaluated and documented in all affected disciplines. Review calculations and associated documentation for each task to assure that the changes are adequately substantiated and documented. Confirm that calculations are complete and verified in accordance with BPC's procedures which implement the requirements of ANSI N45.2.11. Review the source of design inputs to assure that they are current and consistent with the affects of the task on the structure or local area l being modified. 1 Review the affects of any recommendation on vendor supplied hardware such as expansion anchors or electrical strut hardware. In addition, make certain that any substitution of equivalent hardware is substantiated by test or analytical methodc.
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. l Trojan Nuclear Plant Document Control Desk Docket 50-344 September 30, 1987 ,
License NPF-1 Attachment F l Page 5 of 5 ( i Review suitability of base plates used, anchor bolts, welding, and high j strength bolts. Confirm that any special installation requirements (eg, I torquing of high strength bolts) are uniquely celineated and have been , fotvarded to PCE. l Review the application of AISC or ACI codes, FSAR commitments for load l combinations and stress limits as well appropriate consideration of l thermal or transient dynamic affects. l l SAB/crt 1725P l l l l l l l l l
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