IR 05000454/1987007
| ML20235H208 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 07/08/1987 |
| From: | Danielson D, James Gavula, Liu W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20235H163 | List: |
| References | |
| 50-454-87-07, 50-454-87-7, 50-455-87-24, NUDOCS 8707150056 | |
| Download: ML20235H208 (70) | |
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{{#Wiki_filter:- _ _ - _ - _ _ - _ . _ - _ _ . U.S. NUCLEAR' REGULATORY COMMISSION REGION III 1 Reports No. 50-454/87007(DRS); 50-455/87024(DRS) l Docket Nos. 50-454; 50-455 Licenses No. NPF-37; NPF-60 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL' 60690 Facility Name: Byron Station, Units 1 and 2 Inspection At: Byron Site, Byron, Illinois l Nutech Engineers, Chicago, Illinois Inspection Conducted: February 9-12, 17, March 9-10, April 28, 30 and , May 4-5, 12, 20-21, 1987 at Byron ; February 18-19, 24-27, March 3-5, April 6-7 and May 13 and 18, 1987 at Nutech Inspectors- W. C. Liu 7/P[#J Date m A Ga Ie.y ' T, O f 7 ' Date
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Md D. H. Danielson, Chief Approved By: 7[//'d'/ Materials and Processes Section Date Inspection Summary Inspection'from February 9 through May 21, 1987 (Reports No. 50-454/87007(DRS); No. 50-455/87024(DRS)) . Areas Inspected: Special safety inspection of facility modifications associated with the-snubber reduction program (37701), snubber surveillance and functional testing (70370), training (41400), allegations (99014), onsite followup of events at operating reactors (93702) and followup on regional requests (92705).
Results: Two apparent violations were identified (inadequate design control -' Paragraph 3. f and- inadequate instructions for implementing the snubber reduction program - . Paragraph 3.d).
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i l l l DETAILS I Persons Contacted Commonwealth Edison Company (CECO) L. DelGeorge, Assistant Vice President, Nuclear Engineering
*R. Querio, Station Manager *R. Ward, Services Superintendent * Burkamper, Quality . Assurance Superintendent *T. Tramm, Project Engineer, SNED *T. Joyce, Assistant Superintendent, Technical Services * Schwartz,' Assistant Superintendent, Maintenance *F. Hornbeak, Technical Staff Supervisor *D. Flowers, Inservice Inspection Coordinator *S. Bakhtiari, Engineer, SNED * Kennedy, Technical Staff Engineer * Pirnat, Regulatory Assurance Staff C. Moerke, Engineer, SNED K. Ainger, Byron, N. L. B. Shelton, Projects Engineering Manager P. Donavin, Field Engineering Coordinator NutechEngineers(Nutechl *J. Cummings, Project Manager i B. Whiteway, General Manager J. Arterburn, Project Engineer T. Wenner, General Manager S. Mondrowski, Site Supervisor Sargent and Lundy Engineers (S&L)
8. Cleff, Project Director K. Green, Project Manager C. Lim, Mechanical Project Engineer R. Gerke, Project Engineer U.S. Nuclear Regulatory Commission
*P. Brochman, Resident Inspector *D. Danielson, Chief, Materials & Processes Section *J. Harrison, Chief, Engineering Branch M. Hartzman, Mechanical Engineering Branch, NRR D. Muller, Division of Reactor Projects, NRR L. Olsnan, Byron Project Manager, NRR * Denotes those attending the final exit interview on May 21, 1987.
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l 2. Telephone Conferences and Meetings i l During the course of this inspection, several telephone conferences and ! meetings were hold between the NRC staff and licensee representatives to discuss matters which included the review of Nutech's OPTPIPE computer program, benchmark calculations, rattle points evaluations, 3-inch cle6rance walkdown criterion, seismic impact analysis, seismic zero ; period accelerations (ZPA), seismic response spectrum used in design l calculations, field walkdown inspections, seismic anchor movements and diesel generator missile evaluation * Meetings were held on February 20, April 28, and May 13, 198 * Telephone confere.nces were held cn February 13, 17, and May 7, 198 . Snubber Reduction Program I Introduction The overall goal of the licensee's program was to remove or replace as many of the snubbers as possible, while maintaining the stress margins for the piping and associated support structure i Nutech was responsible for the engineering analysis portion of the ! snubber reduction program. The analyses were performed for individual i piping subsystems by making a series of OPTPIPE computer runs with l various combinations of snubbers either deleted or replaced by strut The project included an assessment option where new loads remained within support design loads and a configuration evaluation option where stresses remained within support capacities, Snubber Prioritization Snubbers were prioritized by CECO according to the following categories:
* Priority 1 Every Effort shall be made to remove this snubber including changes to existing support hardware on the piping system if require * Priority 2 It is very desirable to remove this snubbe * Priority 3 Removal of this snubber is desirable without'any other subsystem hardware changes or this snubber can be abandoned in place due to lack of acces _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _]
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* Priority 4 Snubber removal without hardware changes onl Nutech ~
should not perform additional calculations on any subsystem for' removal of Priority 4 snubber Number of Snubbers and Subsystems There are 67 subsystems containing Priority 1'and'2 snubbers. The total number of snubbers included in these subsystems is 496. Based on'information available on May 13, 1987, Nutech had completed'the
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analyses for all of the above' subsystems. .Results of the analysis' revealed that 302 snubbers can be deleted and 116 snubbers can be replaced with rigid struts, and only 78 snubbers will: remain in place. On'this basis approximately 84 percent of the. snubbers are to be deleted or replaced as a: result of the snubber reduction program.- Document and Procedure Review The NRC inspectors reviewed the relevant portions of the followin documents and procedures pertaining to the snubber reduction program to determine whether appropriate procedures.had been established-and whether they complied with NRC requirements and the licensee commitment * NUTECH, " Project Plant for Byron, Unit 1, Snubber Reduction," Revision 2, January 12, 198 * NUTECH, OPT-87-001, "0PTPIPE, Piping Analysis Computer Program Verification," Volumes I, II and III, Revision 0, February 15, 198 * NUTECH, OPT-86-001, "0PTPIPE Version 3.5, Computer Program ; User's Manual," Revision 2, January 22, 198 * NUTECH, COM-PI-BYR3, " Snubber Reduction Evaluations - Options ~1 , and 2," Revision 1, February 6, 1987, and Revision 2,-April 1, c 198 .j
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* NUTECH, COMP-PI-BYR5, " Pipe Support Calculation Revision,"
Revision 1, February 6, 198 ,
* NUTECH, COM-PI-BYR15, " Resolution of Rattle Point Interference Problems," Revision 0, December 15, 1986, and Revision 1, April 1, 198 * NUTECH, COM-PI-BYR19, " Thermal Monitoring / Seismic Interaction Walkdown Post-Installation Testing Requirements," Revision.2,.
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* NUTECH, COM-PI-BYR21, " Hot Functional Testing and Limited i Clearance Investigation,". Revision 0, April 16, 198 * NUTECH, COM-PI-BYR21, " Post Installation Limited Clearance and Thermal Monitoring Walkdown _ Procedure," Revision 3, May'12, 198 * Sargent & Lundy, " Statistical Study for Seismic Interaction for Piping Subsystems," Revision 0,. November 16, 198 The relevant portion of the.ab'ove procedures / instructions were' I reviewed with respect to the NRC requirements and the. licensee commitments relative to design analysis and field walkdowns for the safety-related piping systems. . The NRC inspectors,noted that- l certain procedures / instructions did not include quantitative or qualitative acceptance criteria for determining that important 3 activities had been satisfactorily accomplishe l 1' (1) Instructions did not address seismic effects.for all ,
portions of the piping subsystems: Paragraph 4.1.1 of Nutech Instruction BYR19, Revision 2 states that seismic. interaction walkdown shall be specified for portions of subsystems where seismic displacements exceed the original design seismic radial displacements by more than one inc Attachment E to Nutech Instruction BYR21, Revision 0, divided piping subsystems into
" piping within seismic boundary and piping outside seismic boundary" for seismic interaction walkdowns. The'use of " piping outside seismic boundary" was a misinterpretation of requirements for the safey-related piping subsystem All portions of the lines should be evaluated for seismic effects. Attachment A of the same instruction indicated that-only portions of the subsystems required seismic walkdown This is an example of a violation of 10 CFR 50, Appendix B, Criterion V (454/87007-01A).
(2) Instructions did not consider seismic anchor movement (SAM) i effects during seismic walkdowns: 'Nutech Instruction BYR21, Revision 1, was implemented at~the Byron site to evaluate the adequacy of spacing between piping systems and other plant components to accommodate thermal-and seismic-movemen . However, the effects of seismic anchor movements, which should ' be included in field walkdowns, were not addressed. This is an example of a violation of 10 CFR 50, Appendix B, Criterion V (454/87007-01B). This concern was resolved by issuing Revision 2 to the instructio (3) Instructions were insufficient in the OPTPIP' E User's' Manual: The NRC inspectors reviewed Nutech's OPTPIPE Computer Users' Manual,' Version 3.5, Revision 2, January 22, 1987. It wa .i noted that seismic zero period acceleration (ZPA) was listed on- < Pages 3.6, 4.2, 4.61, 4.65 and 4.66. However,~no instructions were given as to when and how the ZPA was.to be utilized in the
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computer application. The ZPA effects were not considered in the analysis of the piping subsystems for the snubber reduction program prior to the NRC inspection. As a result, the computer output showed zero accelerations for certain valves and supports during a seismic event. This is an example of a violation of 10 CFR 50, Appendix B, Criterion V (454/87007-01C).
The aforementioned concerns were discussed in detail with licensee representatives. The licensee's corrective actions in resolving these concerns have been completed or are in progres e. OPTPIPE Computer Program Verification ,
The NRC inspectors noted that this was the first time that Nutech's i OPTPIPE computer program was used for any snubber reduction projec J The computer program had not been reviewed and accepted by the NR The licensee's approval for using of OPTPIPE on the project was 1 based on an independent technical review performed by S. Levy, Inc., ) in November 198 ! l The NRC inspectors reviewed the relevant portions of Nutech's OPTPIPE computer program verification (Volumes I, II and III) to determine whether the verification was adequately performed to assure conformance with NRC requirements and the licensee's commitments. The NRC inspectors noted that Nutech's verification was composed of two parts. The first part consisted of a comparison of results obtained from OPTPIPE to the results contained in the NRC benchmark problems contained in NUREG/CR-167 The second part consisted of a comparison of results for a series of additional test problems to solutions obtained from other recognized programs which have been accepted by the NR The NRC inspectors verified results from OPTPIPE to the results of the corresponding NRC benchmark problem Of the 11 benchmark . l problems examined, it was found that the results for all problems . were consistent with the exception of the two problems identified below: l
* Benchmark Problem No. 1, OPTPIPE Table H-5 Element Stress l Resultants (Solution 1A) - Forces J l 1 i NUREG OPTPIPE Axial Force Y-Axis Shear Z-Axis Shear q Element Element End-1 End 1 End 1 i Number Number Deviation Deviation Deviation ]
,l 2 31 15.43% 12.45% 24.33 l 6 4 8.29% 12.39% 15.84% )
13 11 25.22% 2.42% 22.52% i ,
* Benchmark Problem No. 1, OPTPIPE Table H-6 Element Stress Resultants (Solution 1A) - Moments
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NUREG OPTPIPE Torsional Y-Axis Z-Axial Element Element Momen . Moment Moment Number Number Deviatio Deviation Deviation 2 31 25.57%- 25.05% 12.53% 13 11 2.18% 22.80%. 4.46%
* Benchmark. Problem No. 4, OPTPIPE Table K-6 Support' Reactions I (Solution 4A) - Forces - 1 NUREG OPTPIPE Deviation Support . Support with Node N Node N NUREG j l
165-x 166-x 15.15% 165 y 166 y 15.10% 165-z 166-z- 14.86% ll The above deviations were noted during~a comparison of the OPTPIPE-computer runs and.the NRC benchmark solution In~accordance with y-NRC standard review plan (SRP) Section 3.9.1, an acceptable deviations- -) is 15%. It can be seen from the above tables that.some elements in-Problem No. 1 have deviations greater than 25% and some support . reactions in Problem No. 4 have deviations greater than_18%. Nutec I representatives agreed to perform further evaluation for Benchmark Problem No. 4. However, they felt that the solution for Problem fi No. I was misprinted, since it was identical to the solution l contained in NUREG/CR-1677, Volume 2,. Problem 1 The NRC inspectors reviewed CECO's response dated May 22, 1987, regarding the latest OPTPIPE verification for Problem No. 4 contained in NUREG/CR-1677, Volume 2. They noted that Nutech had input 533.884 for the "X" coordinate at Node 162 of the piping 1 model. The actual "X" coordinate used by NUREG/CR-1677 was 553.88 This resulted in a 20 inch difference in the pipe segment lengt Nutech's reanalysis using the correct length showed that the output of OPTPIPE was consistent with that of NUREG/CR-1677, Volume Furthermore as was stated by Nutech only NUREG/CR-1677, Volume _1,
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was utilized to qualify OPTPIPE for use in the snubber reduction program.
A memorandum dated April 29, 1987, from Mr. J. Richardson of Division of Engineering and Systems Technology, NRR. to-Mr. L. 01shan of Division of Reactor Projects, NRR, addressing ? the acceptance of OPTPIPE used for the Byron,_ Unit 1, snubber- ] reduction program, concluded that OPTPIPE was acceptable only for.- i the uniform (envelope) support motion response method of piping _ dynamic analysis, j Based on the reviews conducted by NRR and the Regional NRC inspectors, it was concluded that the use of OPTPIPE for the Byron, Unit 1, snubber reduction program was acceptable with the condition described abov l q
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-. f. Review of Design Analysis for Piping Subsystems The NRC inspectors randomly selected the following subsystems for review to determine whether the design analyses were adequately performed and evaluated in accordance with licensee commitments and NRC requirement Subsystem Number Piping System ! 1CV50 Chemical and Volume Control 1RY05 Rea'ctor Coolant Pressurizer 1FW05 Feedwater ICC35 Component Cooling The relevant portions of the above subsystem analyses were reviewed with respect to OPTPIPE computer applications, seismic effects, thermal considerations, support locations, rattle point evaluations, and code / standard implementatio (1) Effects of Zero Period Acceleration (ZPA) During the review of subsystem ICV 50, the NRC inspectors noted that the calculation for Valve 1CV8141A showed zero accelerations in three directions during the operating basis earthquake (0BE) and the safe shutdown earthquake (SSE). The use of these zero accelerations in the calculations resulted in incorrect values being compared to the prequalified allowable accelerations for the valve. In this case, the ZPA from the response spectra is 0.3g for OBE and 0.6g for SS The NRC inspectors also found other instances where the ZPA effects had not been considered in the evaluation of the piping subsystems. Subsequent to these findings, Nutech decided to run computer analyses considering the ZPA effects for the above four subsystems. Results of the analyses revealed that in several cases, accelerations and support loads were under predicated in the response spectrum analysis. Consequently, Nutech committed to include ZPA effects when qualifying all valves and supports for the piping subsystems. Furthermore, Project Instruction BYR3, Revision 1, was revised to include ZPA in the evaluation of all piping subsystems associated with the snubber reduction progra Based on Nutech's letter dated May 13, 1987 (CEC-78-003) the results of their evaluation of the ZPA effects were as follows: Supports
* Number reviewed: 1712 * Number requiring requalification: 106 l * Number requiring modification: 1 i
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* Number reviewed: 211 * Number requiring requalification: 9 ; * Number requiring modification: 0 Failure to include ZPA effects in the original calculations-is an example of a violation of 10 CFR 50, Appendix B, Criterion III (454/87007-02A).
(2) Justification of 3 inch Criterion for Seismic Interaction Walkdown Attachment C of Nutech's Instruction BYR21, Revision 3, states that seismic interaction walkdown clearance requirements will-be' verified at highest stress locations within the boundaries indicated:
* 3 inches pipe-to pipe * 3 inches pipe-to-component or building structures * rattle point dimension.if specified on Hunter Isometric or dimension specified on the PECN.-
Attachment A (Page 5) of Nutech's Instruction BYR15, Revision 1, states that. locations where the clearance between a piping system and other objects is less than 3 inches are referred to as rattle points. The analyst must review whether'these clearances,are sufficient so that the thermal and seismic pipe displacement do not exceed these clearances. If the clearances are exceeded b thermal plus seismic, the effect must be evaluate During review of the above instructions, the NRC inspectors noted that the 3 inch clearance criterion was used prior to the snubber reduction program. The 3 inch critorion was determined to be _ accxptable as a result of a statistical study for seismic ic ?;ction rurformed by S&L, dated November 16, 198 A total of"331 piping subsystems were included in the study. 'With: respect to the snubber reduction program, Nutech used'the same 3 inches criterion for the seismic interaction walkdowns. .An evaluation had not been performed to document the acceptance of the use of this criterion considering the increased pipin displacements that could result from the snubber reduction-progra In addition, portions of the subsystems were not
. included in the seismic interaction walkdowns, consequently, the seismic effects were not evaluated'for all portions of--
the piping subsystems. This is an example of a violation of 10 CFR 50, Appendix B, Criterion III (454/87007-028).
(3) 90% Allowable Stress and Insulation Criteria Attachment A (Page 6) of Nutech's-Instruction BYR15, Revision 0, states that seismic interferences are neglected for rattle point evaluation if one of the following condition is satisfied:
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* Insulation exists at the rattle poin * The seismic stresses are less than 90% of the allowables within the rattle point area Attachment A (Page 4) in the same instruction states that all methods described are based on engineering judgment I
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On June 19, 1985, CECO issued a letter to NRR committing j that where pipe supports are modified, moved, or deleted 1 as a result of the snubber reduction program,,the existing piping displacement limits will be maintained. In cases-where piping movements exceed the displacement limits, adjacent structures, components, and equipment will be reviewed for adverse effect ) The NRC inspectors reviewed the above instructions and the licensee commitments pertaining to the evaluation of seismic interferences associated with the snubber reduction progra It was noted that the insulation and the 90% seismic stress criteria used for the evaluation of rattle points were based'on engineering judgments rather than through technical justification In addition, the NRC inspectors noted during review of design analyses that, in several cases, where the seismic displacements were much greater than the original Westinghouse analysis, evaluations were not performed as a result of the aforementioned insulation criterio Assurance _can not be provided that existing piping displacements are maintained if seismic interferences are neglected for rattle point evaluation based on the insulation and 90% stress criteria. This is an example of a violation of 10 CFR 50, Appendix B, Criterion III (454/87007-02C).
(4) Design Verification (a) Impact Analysis Impact analysis for piping clearance report (PCR) i No. 15I01-8 was reviewed by the NRC inspectors. The ! inspectors noted that improper calculations were performed i to calculate the impact velocity. The delta used by the i engineer was a gap of 0.913" at operating condition. In accordance with Nutech Instruction BYR21, Revision 3, Attachment D, delta is defined as the interference at impact point due to seismic excitation (not the gap). The analysis package was signed and checked on May 16, 198 (b) Seismic Response Spectrum Section 3.7.3.8.4 of the Byron FSAR requires that the most severe floor response spectrum corresponding to support locations is to be used in the evaluation of piping systems supported at different elevations. The NRC inspectors
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_ - reviewed the calculation package for subsystem 1FW0 The inspectors noted that support 1FW05007X was attached to the building structure at elevation 421 f A review of the seismic response spectra revealed that the highest floor response spectrum used in the analysis was at elevation 412 ft. Since the floor response spectrum at higher elevation would generally produce greater acceleration than that of at lower elevation, the use of the envelop response spectra up to elevation 412 f appeared to be unconservative and inconsistent with the FSAR commitment j
! (c) Missing Node Points j The NRC inspectors reviewed calculation package for ;
subsystem ICV 50 to determine whether the calculations were- 1 performed in accordance with licensee commitments and NRC requirements. The inspectors noted that two node points
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were missing on Sheet 21 of 32. The predicted dicolacements at rattle points were tabulated for each node point so that further evaluations could be performed with respect to the thermal and seismic effects. Although the design documents were signed and checked, the missing node points i provided evidence that the design review might not have
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been adequately performed in accordance with the instruction The above concerns were discussed in detail with licensee representatives. The licensee's corrective actions in resolving these concerns have either been completed or in progress. The above instances are an example of a violation of 10 CFR 50, Appendix B, Criterion III (454/87007-020).
(5) Friction Consideration in Support Design Friction forces due to thermal movement previously were l not considered by Nutech in the support analyses. This has since changed as a result of an independent design review performed by S&L personnel in January 1987. The NRC inspectors reviewed Nutech's Instruction BYR5, Pipe Support Calculation, Revision 1, February 6, 1987. The inspectors found that neither the formula for determining ~1 the friction force was provided nor was the coefficient l of friction specified in the instructio Subsequent to i the discussion with licensee representatives regarding the above concern, Nutech revised its instruction to include I the calculation of friction force in the support analysi This matter is resolve g. Review of Field Walkdown Documents The NRC inspectors reviewed the documents associated with the field walkdown activities pertaining to the snubber reduction progra ______
(1) Justification of 3 inch envelop for potential interferences: !
Source Target (Piping Supports) Piping 251 19 I Pipe Supports 69 6 i Miscellaneous Supports 31 1 l Conduit 53 4 HVAC 2 0 Cable Tray . 19 0 Miscellaneous Structural Steel 243 9 Building Structure 47 4 Equipment 24 2 Totals 739 45 Resolution of the above interferences: Number Percent Displacement check 617 79% Monitoring 28 4% Interaction Analysis 16 2% Insulation Coping 114 14% Interference Removal 9 1%
(2) Justification of 1 inch delta walkdown for potential interferences (for the entire subsystem)
Source ! Target (Piping Supports) Piping 43 2 Pipe Supports 15 1 1 Miscellaneous Supports 4 0 l Conduit l HVAC 1 0 Cable Tray 1 0 Miscellaneous Structural Steel 49 1 Building Structure 8 1 Equipment 1 0 Total 134 5 Resolution of the above interferences: Number Percent Displacement check 107 78% Monitoring 2 1% Interaction Analysis 3 2% Insulation coping 25 18% Interference Removal 2 1%
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f I' l l At the time of this inspection, the licensee's evaluation with l regard to the above interferences was still in progres j h. Training and Qualification (1) Training Records at Nutech's Engineering Office I The NRC inspectors reviewed the training records pertaining to l the snubbers reduction program to determine whether Nutech had adequately conducted the training sessions for those personnel who were actively participating the design analysis for the piping subsystems. The NRC inspectors noted that several entries on the training records contained missing and/or incorrect dates. In some cases the training sessions conducted for the snubber reduction program covered too many instructions in one session and appeared to be too generic. Furthermore, this was the first time that Nutech's OPTPIPE computer program was used for any snubber reduction analysis. There were no records to show that personnel using the program had received l formal training for its proper us Nutech later updated the training records as a result of discussions with NRC personne Nutech's position with regard to the training activities is as follows: The training records document only the formal training that was conducted on the project. Other training was provided to the team leaders by the project engineer and by the team leaders to members of their team in one-on-one sessions and small meeting Nutech believes that this type of training is much more effective than would be derived from larger, more formal training session However, in response to NRC's concern Nutech initiated formal I training sessions on OPTPIP (2) Interview of Nutech's Personnel of Byron Site The NRC inspector interviewed six randomly selected individuals i from the site engineering and walkdown group. The discussions ' focused on the use of applicable procedures, instructions and field walkdown activities associated with the snubber reduction program. The interviews of the three field engineers and three field walkdown personnel disclosed the following:
* They are familiar with the procedures and instructions used for the field walkdown activitie * They understand the contents and requirements in those procedures and instruction * They feel that the walkdown activities are much more extensive than the previous walkdowns (prior to snubber reduction program).
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* They feel' comfortable with respect to the program and'th work they performe * They -have a total of 36 years experience' with the Byron project, or an average of six years experience-for each individua On the basis of the above interviews, the'NRC inspector determined that the site personnel-appeared to be qualifi_ed for the use,of the procedures and.the performance of walkdown activities associated with the-snubber reduction progra '
Except 'as noted, no violations or deviations were identifie : 4. Allegation Followup " (Closed) Allegation (RIII-87-A-0056): On April 24,.1987, an anonymous individual contacted Region III:with concerns'that
' dealt with the stress analyses performed.for the Byron snubber ;
reduction program. Subsequently,'on April'29, 1987, the alleger ! met with Region III staff and expressed the following concerns-associated with the rattle point evaluations:
(1) " Classification of seismic portion versus non-seismic I portion on seismically qualified line (2) Use of only thermal movement in non-seismic parts in a seismically qualified pip (3) Basis for determining 1/4 inch hits on non-seismic portion (4) Consideration of seismic anchor movement (5) Consideration of 1/2 iach insulation deformation regardless of insulation thicknes (6) The impact of piping on other components and the impact of other components on pipin (7) Adequacy of procedures and instruction (8) Personnel qualifications for using these procedures'and instructions."
In response to the above concerns, the NRC inspectors conducted inspections on May 4-21, 1987, at the Byron Site and at the_Nutech Chicago'0ffice to investigate if any potential deficiencies existe Two violations were identified pertaining to,the' alleger's 1 concern !
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Attachment E of Nutech Instruction BYR 21, Revision 0, April 16, 1987, classified piping subsystems into " piping j within seismic boundary and piping outside seismic boundary for I
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rattle point seismic interaction evaluations." Further, the NRC inspectors reviewed Nutech's field walkdown drawings and ! noted that these drawings were marked up with seismic portions ) and non-seismic portions. The use of " piping outside seismic ! boundary" was a misinterpretation of requirements for the safety-related piping subsystems. All portions of the lines should be evaluated for seismic effects. Attachment A of the same instruction indicated that only portions of the subsystems required seismic walkdown Conclusion The concern was substantiated by the fact that Nutech used the seismic boundary and non-seismic boundary breakdown of the safety-related piping subsystems for rattle point evaluation This was evidenced by NRC review of the field walkdown drawings and Nutech's Instruction BYR21, Revision 1, May 4, 1987. One violation delineated in Paragraph 3.d(1) was identified in conjunction with this allegatio (b) NRC Review for Item 4.a(2) In general, thermal and seismic effects are to be considered in the safety-related piping subsystems to determine for potential interferences due to thermal movements and seismic excitation The NRC inspectors reviewed Nutech's Instructions BYR15 and BYR21 pertaining to rattle point evaluations. It was found that Nutech's seismic interaction walkdowns were performed only at potential high stress locations or within the seismic boundary of the subsystems. For non-seismic portion of the subsystems, only thermal movements effects were considere This was due to the criteria contained in Instruction BYR15, Attachment A, which states that seismic effects are neglected I if the seismic stresses are less than 90% of the allowable I and if insulation exists at the rattle point Conclusion The concern was substantiated and one violation delineated in I Paragraph 3.f(3) of this inspection report was identified in l conjunction with this allegatio (c) NRC Review for Item 4.a(3) The NRC inspectors reviewed the applicable instructions and held discussions with licensee representatives regarding the 1/4 inch criterion used for the thermal walkdowns. It was
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, _ - _ _ - i noted that the intent of the thermal walkdowns was to identify q structures which could interfere with the pipe expansion during i heat-up. A 1/4 inch clearance was selected to facilitate j identification and resolution of potential interfaces prior to j heat-up. This 1/4 inch criterion was obtained through Nutech's l engineering judgmen l J Conclusion This allegation was substantiated in that the basis for determining 1/4 inch hits on non-seismic portions of the piping subsystems was to facilitate identification and resolution of potential interferences prior to heat-u The 1/4 inch clearance selected was based on Nutech's engineering judgment. Nutech's Instruction BYR21, Revision 3, added new requirements in which all gaps less than or equal to 1/4 inch are to be identified by , the hot functional walkdown team. The use of 1/4 inch clearance ' criterion for the thermal monitoring walkdown appeared to be acceptable in terms of identifying potential interferences for the piping subsystem J (d) NRC Review for Item 4.a(4) ! Discussions with Nutech representatives revealed that Nutech initially identified all rattle points where the Westinghouse > previously identified displacements and rattle point dimensions were exceede In maintaining consistency with the Westinghouse resolution method, seismic anchor movements (SAM) were not considered. In a subsequent evaluation, Nutech reviewed all SAMs and determined that with the exception of decoupled branch lines (SI-26, 27, 32 and 34) and lines off the RC pumps, the SAMs are insignificant. As a result, Nutech limited its assessment of rattle points relative to SAM at these locations
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and included these movements to determine the requirements for additional rattle point walkdowns. This was implemented on May 12, 1987, and documented in Nutech's Instruction BYR21, Revision 3, May 12, 198 Conclusion The allegction was substantiated by the fact that Nutech did I not consider SAM in rattle point evaluation prior to May 12, j 1987, when Instruction BYR21, Revision 3, was issued for implementatio The SAMs were included in the reevaluation of all rattle points that were conducted after implementation of this revision. This concern is a part of the violation identified in Paragraph 3.d(2) of this inspection repor (e) NRC Review for Item 4.a(5) The NRC inspectors held discussions with Nutech representatives i regarding the consideration of 1/2 inch insulation deformatio The response from Nutcch on May 20, 1987, was that the
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l I insulation was made of sheet metal, with an inner core of
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reflective metal foi The total thickness of the insulation is typically two to three inche The foil core is a composite ) of foil layers separated by spacers. This core has very little i structural capacit This type of insulation is easily crushed l locall On this basis, the allowable crush distance of 1/2 inch was established. The NRC inspectors noted that the 1/2 inch criterion was changed to 1/4 inch when Instruction BYR21, ) Revision 3, was issued on May 12, 1987, for implementation, j
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Conclusion
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This allegation was substantiated in that the 1/2 inch insulation { deformation criterion was used until May 12, 1987. The 1/4 inch ' criterion has been implemented since then.' However, use of these criterion based on the above description appears to be acceptable for the evaluation of seismic interference (f) NRC Review for Item 4.a(6) The NRC inspection associated with the snubber reduction l program started on February 9, 1987. During the course of this inspection, a number of discussions were held with licensee representatives to resolve NRC's concerns. The impact of piping on other components and the impact of other components i on piping was one of the NRC's concerns and was discussed in ' detail with Nutech representatives for proper resolutio Nutech's Instruction BYR21, Revision 0, April 16, 1987, provided the method for impact analysis for piping, equipment nozzles, and supports resulting from anticipated interference Revision 2, May 11, 1987, of the same instruction documented additional requirements. Paragraph 2.6 of this instruction 1 states that both the impacts to the targets and the effects I of the targets on the piping shall be considered. In instances where the target is outside the scope of-the snubber reduction project, Nutech shall interface with the responsible Architect / Engineer for further evaluation of piping and target. This was evidenced by the fact that Nutech transmitted R.C. pump support to Sargent & Lundy and Class D pipe line re-route to Westinghouse on April 28, 1987, for impact evaluation Conclusion i This allegation was partially substantiated in terms of evaluation of the impact of piping on other components on piping prior to April 28, 1987. In accordance with Instruction BYR21, Revision 3, and discussions held with Nutech's representatives, j the NRC inspectors noted that Nutech was responsible for the I evaluation of the impact of other components on piping and Sargent & Lundy/ Westinghouse were responsible for the evaluation of piping on other components when the potential interferences arise. All the subsystems associated with the snubber reduction j l i
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____ program were included in the impact evaluatio The stresses I resulting from the impact analysis are the additional stresses that have to be added in the combined stresses in comparison ; with the allowable stresses for determining the adequacy of the { piping subsystems and the affecting component I (g) NRC Review for Item 4.e(7) The NRC inspection activities associated with this allegation are described in Paragraph 3.d of this inspection repor Although Nutech established various procedures / instruction for implementing snubber reduction program, some of the procedures / instructions contained inadequate information in terms of performing engineering analysis and implementing field walkdown verification Instruction BYR19, Revision 2, did not address seismic effects for all portions of the piping subsystem Instruction BYR21, Revision 1, did not consider the effects of seismic anchor movements (SAM), and Nutech's OPTPIPE Computer Program User's Manual provided no instructions as to when and how the zero period acceleration (ZPA) was to be utilized in its computer application Conclusion This allegation was substantiated for some of the procedures / instructions identified in Paragraph 3.d of this inspection repor One violation in conjunction with this allegation was identified in the same paragraph. The above concerns were discussed in detail with licensee representative The licensee's corrective actions in resolving these concerns have either been completed or in the progres (h) NRC Review for Item 4. The NRC inspection activities associated with this allegation are described in Paragraph 3.h(2) of this inspection repor Conclusion This allegation was not substantiated based on interviews of three field walkdown personnel and three field ongineer Although some procedures / instructions contained inadequate safety considerations, personnel qualifications appeared to be acceptable in terms of performing work activities in accordance with applicable procedures and instruction b. (Closed) Allegation (RIII-86-A-0118): As-built piping drawings at Byron, Unit 1, have discrepancies and therefore violate the requirements of IE Bulletin 79-1 This allegation was originally addressed and closed in the NRC Inspection Report No. 50-454/87027.
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j Prior to the issuance of the above report, an attempt was made to meet with the alleger to confirm that the Region III investigation < had covered all of his related concerns. However, based on the l alleger's requests, this meeting was not held until January 26, l 1987, after the report was issue A copy of the presentation ; material used by the NRC during the meeting is contained in i Attachment I of this repor l Following the NRC presentation of the results and conclusions of their investigation, the alleger was given an opportunity to present his assessment of the situation. After a lengthy discussion between 3 the alleger and regional staff, it was concluded that there was a j technical difference of opinion concerning the significance of the !
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as-built drawing errors. The alleger ultimately did not concur with the conclusions presented by the regional technical staff in their J inspection repor ' Prior to the conclusion of the meeting, the alleger presented one ; additional example of an as-built drawing error. On Byron, Unit 2 Drawing: Isometric No. CC-21, Revision 8J, dated December 16, 1983, the as-built dimension of 35'-11/4" appeared to be discrepant by over seven fee NRC Review A subsequent walkdown of the above specific portion of pipe by the J NRC inspector indicated that there was in fact an erro Based on ! the remeasurements, it appeared that the 35'-11/4" dimension line was drawn to the wrong elbow on the as-built drawing. This observation - was later confirmed by reviewing the original as-built walkdown package. The marked up drawing from the field showed the accurate ] i dimension to the correct elbow. The discrepancy was introduced when l Revision 8 of the isometric was issued on March 11, 1982. Instead of drawing the arrowhead to the dimension line for the first 45 elbow, the draftsman drew the arrowhead to the second 45 elbo This cause an apparent 7'-7" overall discrepancy in the pipe lengt l The review of the analysis for this section of pipe, contained in S&L Calculation EMD-030667, revealed that the subject length of pipe was correctly modeled and analyzed at 35.093 feet. The review of pipe supports, 1CC03117R and 1CC03116R located on this section of i pipe, accurately located the first 45 elbow relating to the plant column lines. The locations of these supports were also accurately located based on the walkdown informatio The NRC inspector verified that the true as-built configuration of the pipe matched the as-analyzed configuration. Although a drafting
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error was made during transposition of as-built data to the final drawing, the pipe was ultimately reconciled to the actual configuratio _ _ _ _ _ _ . _
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j As a result of this inspection and the previous' inspection
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noted above, the NRC inspector reviewed the following drawings q and verified that the previously identified drafting errors had ! been correcte CC-21, Revision 10 CC-37, Revision 12M CC-46, Revision 7Q SI-01, Revision 118 SI-04, Revision 108 WO-40, Revision 5C Conclusions Although the original allegation specifically addressed Byron, Unit 1, subsequently statements by the alleger identified specific concerns at Byron, Unit-2. However', due.to the nature of the original allegation inspection, a detailed investigation reviewed the as-built dimensions for both Units 1 and 2. put of the 900 as-built , dimensions reverified during the original inspection, over 550 were i obtained from Unit 2. Using this information as a basis, no further review of Unit 2 as-built dimensions is. deemed necessary as a result of this latest concer Based on the lack of significance for the previous drafting errors and the lack of significance for the additional example provided by the alleger, the previous conclusions : given in the original inspection report are still valid and no ( further follow-up is warrante l l 5. Snubber Visual Inspection and Functional Testing Background The Byron Unit 1 snubber population had'approximately 1239 mechanicals ' and 16 hydraulics at the start of the outage. As specified in CECO's
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letter from K. A. Ainger to J. G. Keppler, dated January 15, 1987, the sample plans selected for the snubber functional tests were the 10% plan for hydraulics and the 55 snubber initial sample plan for , the mechanicals. These alternate sample plans are specified in I Technical Specification Section 4. . i Procedure and Documentation Review 'l The NRC inspector reviewed the following Ceco and Paul Munroe j Procedures: .q
! * " Visual Examination of Safety-Related Snubbers," No.1BVS j 7.8-1, Revision l * " Visual Examination Requirements for Safety-Related Snubbers,"
No. BVP 200-5, Revision l
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* " Functional Testing of Safety-Related Mechanical and Hydraulic Snubbers," No. 1BVS 7.8.2, Revision * " Functional Testing'. Requirements for Safety-Related Snubbers," I No. BVP 200-6, Revision 2 * " Visual Examination of Safety-Related Component Supports,"
No. 1BVS 4.10-6,' Revision '
* Paul Munroe Q.A.' Procedure'No. 055-QAP-11.1, " Functional !
Testing of. Pacific Scientific Shock Absorbers," dated ' February 22, 198 i The NRC inspector had no adverse comments on.any of the above - procedures. Acceptance criteria were adequacy' delineated and j
. acceptable instructions:were specified for the functional test gj It was noted that due to the current snubber reduction effort, 1 the checkoff lists associated with BVP'200-5 and BVP 200-6 will eventually need to be revise The calibration records for the snubber test stands were also j reviewed. The. Functional Test Stand Position Measurement System, 1 Model STADAS 4120, Serial No.'653A, was scheduled for recalibration j on January 27, 1988. All calibration documentation and operator certifications were in orde ) Test Results "
The 10% sample of the hydraulic snubbers consisted of two snubber Both of these units were on the steam generators and met the specified acceptance criteria. No visual indications were noted , during the visual examinations by. CEC ] The functional tests for the mechanical snubbero resulted in a total of seven failures. This required a total sample of approximately 248 snubbers be tested. The nature of the' failures were as follows: Snubber Number Failure Description l 1CV25021S Locked up during test 1CV660295 High Drag Forc FWO7010S Locked up during' test 1MS93014S Locked up during test 1RH09044S Locked up during test 1RH09045S Received locked up 15I01018S Locked up during test The piping systems associated with the above failures were reanalyze by Nutech or Westinghouse. Where appropriate,'the snubbers were either modeled as rigid supports or completely removed. In each case, the subsystems were determined to be operable due to the additional loads imposed by the failed snubber >
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' Visual Inspections In accordance with Byron's Technical Specification 3/4.7.8.b,.
all safety-related snubbers were visually' inspected this outag Approximately 33 snubbers had " recordable indications" requiring an operability determination. Typical indications given on the visual inspection sheets were:
* Boric acid residue on outer cove ! * Loose bolt * ~ Bound u . * Spherical bearing. displace *' Paint on' extension-tub * Grout on inner support tub With the exception of one' snubber, all of the above indications-were determined to'be operabl On 15002029S, however, the snubber was declared inoperable as a result of incorrect staking,of th spherical bearing. Since all snubbers were visually; inspected this. outage, no further inspections were required to determine th extent of the problem. However, this inoperablo snubber will require 'the inspection interval for all inaccessible snubbers to be reduced to twelve month Training The training and personnel certification records were reviewed for-the following Paul-Muuoe technicians:
D. P. Graham VT-3, VT-4, Level II Certified January 17, 1987 E. Beverly VT-3, Level I Certified November 11, 1986' J. DeNicolo Technician Certified November 17, 1986-All certifications were current and appropriate for the functional examinations being performe No violations or deviations were identified. Main Steam Line Transient Event On May 19,-1987, a significant waterhammer occurred in'the non-safety related portion of the Unit 1 "B" main steam lin Although the precise initiation of the event is somewhat in question, the root.cause of the waterhammer was a pool of water trapped on the downstream side of the 1B main steam isolation valve (MSIV). The 8 main steam line-is the.'only on of the four main Steam lines that has a section of pipe slopping back toward'the MSIV. This approximately 40 foot long section of pipe has the potential of trapping about 94 cubic feet of water behind the closed MSI It should be noted that this identical pipe configuration was implicated in the waterhammer event at Braidwood, Unit 1, in March of 198 (See the closure of Open Item No. 454/87007-01 in NRC Inspection Report No. 50-456/86028.)-
l l l Event Scenario l During plant heat up, as part of the return to power from the first l refueling outage, the downstream side of the MSIVs sas being heated by opening the MSIV bypass valves. At this time the reactor coolant system was at 340 F and 375 psig. The secondary side of the steam generators was at approximately 96 psig. After operating approximately two hours in this configuration, a CECO engineer, sent to tour the MSIV room for other purposes, discovered significant damage on the "B" main steam line. No noise had been heard nor was there any other indication that an event had occurred. The area had been toured 31/2 hours earlier with no indication of any damag The postulated initiation of the event involves a differential thermal expansion due to thermal stratification in the 1B main steam lin The
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thermal stratification results from the relatively cold pool of water trapped behind the MSIV (200 F or less) and the hot steam above this pool on the upper portion of the pipe (estimated at 300 F to 330 F). This ! uneven temperature distribution caused the 32" diameter pipe to warp l downward and impose very large tensile loads on the first rod hanger in l the system. This excessive load caused the threaded rod to pull out of the weldless eye nut and eventually fail completely. When this hanger failed, the trapped pool of water surged forward as the line dropped such that a large slug of water formed inside the pipe. It is postulated that l this initiated a steam-driven-water-slug type of waterhammer.
l l Damage Description As a result of the waterhammer event, five other rod hangers (in addition to the first rod hanger that potentially initiated the event) failed completely. In four of these cases the " welded beam attachment" was torn away from the embedment plate at the attachment weld. In the fifth case, the entire embedment plate was pulled out of the ceiling when the welds to the rod supports embedded in the concrete failed. One additional rod hanger embedment plate was also deformed, due to an excessive vertical load. In addition, three snubbers were significantly overloaded resulting in obvious deformation or failure to properly strok Four other snubbers were also damaged when the 1B line dropped approximately 20 inches and impacted these lateral restraints for the 18 and 1C main steam line As part of the overall damage assessment, the entire length of the 1B and 1C main steam lines were walked down. Twenty-four total supports were visually inspected with three integral attachments examined by magnetic particle and eleven snubbers manually stroked. The walkdowns included all attachments to the 18 and 1C lines, all surrounding pipes, hangers, etc., individual hangers components, adjacent concrete and the pipes themselve In addition, a total of 21 pipe welds were inspected to assess any piping ! damag Thirteen welds on the IB line and eight welds on the IC line were I inspected using a wet fluorescent magnetic particle technique. Also the
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three welds with the highest calculated stress were ultrasonic 1y examine One weld had a double wall radiographic examination as well. No other damage was found as a result of the above inspections. Based on the analyses performed to simulate the event and the above inspection results, it was concluded that the main steam piping itself was not degraded and i is fully capable of supporting all plant operation I Corrective Actions All of the damaged pipe supports were reinstalled to their original configuration except for two rod hangers. - For these supports, th j embedment plates had been damaged and new baseplates were installed 3 adjacent to the existing locatio Otherwise the installations were I identica In order to eliminate the potential for water accumulation on the downstream side of the 18 MSIV, additional modifications were made to the ' existing drain piping. As a result of the previously mentioned Braidwood waterhammer, Post Fuel Load Engineering Change Notice (PECN) No. P-569 was implemented to add a 4" drip leg and a 1 1/2" drain line on the downstream side of the IB MSI This line had two manual ~ valves and one i check valve prior to teeing into the upstream drain line which runs back to the flash tank. Although this modification was completed during the ; current outage, the system had not been turned over to operations and i as a result could not be utilized during the pla'nt start-u j i Based on additional reviews conducted after the Byron waterhammer, an j additional drain line to the floor drain'was added to the downstream { drain line. This modification added additional assurance that any water on the downstream side of the MSIV would be removed if the drain valves were opene This eliminated the need for any pressure in the main steam line to force the water out and instead relied solely on gravity as the driving forc This modification was completed prior to Unit'l start-u As additional insurance against the build up of water in the main steam lines, changes to operating procedures were made with appropriate steps and cautions during heatu No specific procedure changes were delineated by the license On May 19, 1987, Ceco met with members of the Region III staff and representatives from the Division of Engineers and Systems Technology and the Division of Reactor Projects III, IV, V and Special Projects from Headquarter A copy of the presentation material is included in Attachment III. Based on the information presented during this meeting,. as well as the inspections and reviews conducted at the site by the NRC inspector, it is- felt that no additional followup is require . Emergency Diesel Generator Common Mode Failure Due To Missile Generation-As a result of the recent Cooper-Bessemer diesel engine failures at Zion and Palo Verde, CECO performed a reevaluation of the effects-of potential missiles generated for similar engines at Byron. The recent industry
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_ _ _ _ _ _ , failures indicated that, contrary to previous assumptions, the crankcases of the Cooper-Bessemer diesel engines would not contain all possible , internally generated missile l The basic concern addressed in CECO's reevaluation was the potential for a common mode failure due to the rupture of the "B" diesel's Essential Service Water Cooling piping by an "A" diesel generator engine missil This concern was originally identified by the NRC in Noncompliance 454/80-12-01; 455/80-11-01 and subsequently closed in NRC Inspection Reports No. 50-454/83-49; No. 50-455/83-35. The basis for the closure of this item was a reexamination of Byron's FSAR, Section 3.5, " Missile Protection." As stated: "There are no internally generated missiles postulated outside the containment, other than the main turbine missiles described in Subsection 3.5.1.3." Based on the recent Zion and Palo Verde experiences, this statement no longer appears to be true. As a result,. the licensee agreed that an amendment to the FSAR would be issued in the future to address this situation. Pending review of the proposed FSAR Amendment, this is considered an Open Item (454/87007-03; 455/87024-01).
Based on CECO's reevaluation of the potential effects of a diesel generator missile, only one case exists where missiles in one diesel room could potentially affect functions of the other diesel or any 3 i other systems in the redundant divisio This case is the previously ' identified Essential Service Water Cooling Lines for the "B" diesel passing through the "A" diesel roo i Using very conservative analytical assumptions, CECO concluded that for Unit 1 the piping in question was not in the define missile path. On
- this basis, there is no potential for damage to the cooling water line l due to geometric constraints and protective structures. For Unit 2, even though no credible missile was feasible, an even more conservative analysis indicated that even if an impact occurred, the piping in question would remain functiona l The above information was presented to the Region III staff on February 20, 1987, by licensee representatives. A copy of the presentation material is contained in Attachment II. Following this meeting, Ceco documented their evaluation in their letter from K. A. Ainger to H. R. Denton (NRC),
dated March 17, 198 This information was subsequently reviewed by NRR and the regional staf Based on verbal concurrence between the above organizations, the NRC concluded that this issue does not present a safety concern and is therefore close . Open Item Open Items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involves some action I on the part of the NRC or licensee or both. Open Items disclosed during l this inspection are discussed in Paragraph 7.
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. Exit Interview The Region III inspectors met with the licensee representatives (denoted in Paragraph 1) at the conclusion of the inspection on May 21, 1987. The inspectors summarized the purpose and findings of the
. inspection. The licensee representatives acknowledged this informatio .The inspector also discussed the likely informational content of the inspection report with regard to documents or processes reviewed during the inspection. The licensee representatives did not identify any such documents or processes as proprietar i I
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Anacaneur .'
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EETING WITH COMPUTERIZED J.hTJtFEREN.CE. ELIMINATI.O.N,..I.N.C (CJK)
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REGARDING ALLEGATIONS.0F AS-BUILT f DRAWING P_RQB_LES. AT .T!(E_BVtDN STAT).01 AGENDA
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INTRODUCTION J. G. KEPPLER-l
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GENERAL REPARKS C. J. PAPERIELLO -
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INSPECTION SCOPE, FINDINGS AND CONCLUSIONS J. J. HARRISON
J. A. GAVULA
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DISCUSSION ALL
/ ' * l CLOSING REFARKS !j J. G. KEPPLER l ' r';
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-1 LBERAL IWO.RMATION
PLANT NAE: BYRON STATION
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LOCATION: OGLE COUNTY, ILLIN0IS, 17 MILES S0lJTHWEST OF ROCKFORD,ILLINDIS LICENSEE: C0VfDNWEALTH EDISON COMPANY
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TYPE REACTOR: Th01130 VEGAWATT (ELECTRICAL) WESTINGHOUSE PRESSURIZED WATER REACTOR , ARCHITECT-ENGINEER: SARGENT AND LUNDY
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LICENSE ISSUED: ' UNIT I - FEBRUARY 14,1985, FULL POWER ; UNIT 2 - NOVEMBER 6, 1986, LOW POWER i
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BACK. GROUND . I
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ALLEGATION:
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ljYRE AS-BUILT DRAWINGS ARE DISCREPAhT
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IEE 79-14 REQUIRED ACCURACY WAS NOT KT-
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STRESS REPORTS F%Y LE AFFECTED 1
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SOURCE:
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C0FPUTERIZED lhTERFERENCE ELIMINATION, INC. (CIE) MR, SCHlNLER MITCHELL
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C______________.____________________._._________________ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ . . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , ______j
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KEY DATES: 1
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JULY 1986
- NRC NOTIFIED BY CIE (f0EROUS CONTACTS) l
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NRC DEVELOPED ALLEGATION REVIEW PLAN ,
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STATE OF ILLIN0IS CONTACTED BY CIE
* AUGUST 5,1986 - NRC INTERVIEWED CIE (TRANSCRIBED)
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AUGUST 18, 1986 - IDNS MEETING WITH CIE IDNS EETING WITH NRC-l
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SEPTDEER 12, 1986 . PROTOCOL AGPEEENT: NRC/IDNS
* SEPTEMBER 14,1986 - INSPECTION STARTED i * SEPTEMBER 19, 1986 - NOTIFIED OTHER NRC REGIONS (TRANSCRIPT)
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OCTOBER 14, 1986 - MEETING WITH IDNS TO DISCUSS STATUS
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NOBEER 13, 1986 - INSPECTION COMPLETED
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DECEMBER 17, 1986 - NRC MEETING WITH STATE OF ILLINDIS OFFICIALS
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JANUARY 26, 1987 - NRC FEETING WITH CIE
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. 1DEFINIJ10NS1
- DESIGN - IS TE PROCESS BY WHICH PIPING SYSTEM COPPONEhTS ARE SELECTED, i
LOCATED, ORIENTED, CONFIGURED, AND ANALYZED TO SATISFY CONDITIONS j SET FORTH IN THE PIPING DESIGN SPECIFICATION STRESS / DESIGN REPORT - CONSISTS OF THE DESIGN CALCULATIONS AND STRESS ! ANALYSIS PERFORED TO SHOW TE ADEQUACY OF DESIG l
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DESIGN SIGNIFICANT DIENS10NS - DIMENSIONS USED TO ANALYZE TE PIPING CONFIGURATIONS l
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AS-BUILT - ACTUAL INSTALLED CONDITION OF PIPING SYSTEFS j
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PECONClllATION - RESCLUTION OF SIGNIFICANT DIFFERENCES BETWEEN AS-EUILT AND DESIGN
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WALKDOWNS (INSPECTIONS) - PERFORED TO VERIFY AS-BUILT EETS TE DESIGN l AND TO IDEffflFY DIFFERENCES OR DISCREPANCIES BETWEEN DESIGN AND INSTALLATION FOR TEIR IMPACT ON PIPING SYSTEh STRESS ANALYSIS, EASURE OF ASSURANCE THAT THE DESIGN WAS ET,
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___ _ . 1 1 - - DIFFERENES:
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MINOR DIFFERENCES "USE-AS-IS"
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SIGNIFICANT DIFFERENTS
- ENGINEERING JUDGEENTS !
RE-ANALYZE
! - MODIFICATION /PDERK -
IE BULLETIN 79-114 i
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* VERIFICATION - DOES THE AS-BUILT CONDITION MEET THE DESIGN?
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TOLERANES
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INSTALLATION - ACCEPTABLE DEPARTURE FROM NOMINAL DIMENSIONS.WIDilN PRACTICAL LIMITATIONS
* EASUREMENT - DIFFERENE BETWEEN ACTUAL DIMENSIONS AND DRAWING DIMENSIONS *
RECONCILIATION - FAXIMUM ALLWABLE DEPARTURE BEWEEN AS-BUILT AND , AS ANALYZED DIfENSIONS i i i k
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'E!'HO.N. AS,.E!UlkT. LEA.SC.RElf#T TOERANCES ' LINEAR DIFE S10NS: 0 - 9.9' 1"'
10 - 19.9' 2" 20 - 29.9' t3" ET ANGULAR DIPD SIONS: 5* FOR EBOWS, BENDS AND RELATIVE LEG ORIENTATIONS
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IMON AS-BU.l.L.T. RIEL @_fiE_CO.NCJ.LJATJ.0N..DJ.FFERENCE ' (SIMPLIFIED)
)f_SLl_NJC OUSE -
SEGMEhT LENGTHS - LESS THAN 5 F "
- GREATER THAN 5 F %
BRANCH LOCATIONS - NOMINAL 0.D. DOWN TO' 3" NOMINAL 0.D. UP TO 18" SUPPORT LOCATIONS - N WINAL 0.D. DOWN TOl 3" NOMINAL'0.D. UP TO 18" SEGMEhT ORIEfffATION - 10* SARGEh18. .L.U}i SEGtfNT LENGTHS - 0-10 F " i AND BRANCH LOCATIONS
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11-15 F " 16-20 F " 21-25 F " l i SUFFORT LOCATIONS
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NOMINAL 0 D. DOWN TO 12" l SECfEhT OklENTATION - 10'
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SYSTEM NUMagg g
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SUMMARY OF RESULTS Note Pipe anale and lenoth chances are noted on the drawinos in parenthese SYSTEM 1: 8" SCH BOS Original valve weights: V21 = V24 = 700 l . Modified valve weights: V21 = 809 l V24 = 833 , 1 , ,
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Driginal maximum '9E' -Stress = 23817 psi G.C19B i Tolerance maximum '9E' Stress 5 21449 psi e C19B
% Change = 10%
SUPPORT DIRECT CHANGE ORIGINAL TOLERANCE PER CENT NAME (DEG) LOAD LOAD CHANGE R7 Y 4 1332 1155 -13 RB X 1 1005 '1052 5 Y 4 2408 2298 -5 R9 (SPR) Y 5 938 948 1 SYSTEM 23 8" SCH 80 ]
-
Original valve weights: 783 l Modified valve weights: 932 l Original maximum 'yE' Stress = 28215' psi e 62 Tolerance maximum 9E' Stress = 27916 psi G 62
% Change = 1%
SUPPORT DIRECT CHANGE ORIGINAL TOLERANCE PER CENT NAME (DEG) LOAD LOAD CHANGE H5 X 2 1252 1257 O H6 2 5 2628 2412 8 H4 2 3 1692 1780 -5
_ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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SYSTEM 3: 3" SCH 40
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Original valve weights: 90 l Modified valve weights: 104.l Original maximum '9E' Stress = 17268 psi @ A1 Tolerante maximum '9E' Stress = 16088 psi e A1-
.
'% Change = 7%
SUPPORT DIRECT CHANGE ORIGINAL TOLERANCE PER CENT NAME (DEG) LOAD LOAD CHANGE RH1 Y 2 315 334 6
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RH2 Y 2 371 398 7 l l RH2 LATERAL 4 233 261 12 i SYSTEM 4: 3/4" SCH BO Original maximum '9E' Stress = 18073 psi e S3 (15847 e 1000) Tolerance maximum '9E' Stress = 19800 psi G 1000 (18179 e S3)
% Change = 10% in maximum stress SUPPORT DIRECT CHANGE ORIGINAL TOLERANCE PER CENT NAME (DEG) LOAD LOAD CHANGE S1 LATERAL O 12 10 -20 S3 LATERAL 4 24 30 25 Y O 45 46 2 56 LATERA , 11 10 Y O 29 28 -3 2 l
.
_ . _ _ _ _ _ _ . _ _ _ _ . _ . - _ _ . _ _ __..d
- _ _ _ - _ _ _ _ - _ - _ _ _ _ _ - _ _ _ _ _ _ _ _
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l lI.E._IiDENT.ifl.ED DISdiEPMCIES:
- ON 60 DRAWINGS LOCATED IN 1 QUADRAffT OF UNIT 1 CONTAINENT )
- CHAIN-DifENSIONS - 4
- ANGULAR ORIENTATIONS -3
-
l- - REFERENCE DIMENSIONS - i
- TOTAL OF 15 PROBLEMS RELATED TO 26 DIMENSIONS
!
!
- _ _ - - _ _ _ _ _ - - - _ _ _ _ _ _ . - - - - _ -l
. _
liR.c. ALLEGATION.RIEKPLAN:
- PERF0W. WALKDOWNS - OBSERVE ACTUAL MEASUREFEATS i - REVIEW RECONCILIATION PROCESS; AS-BUILT TO DESIGN ,
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iINSPECT10fLSyfARY:
- SAF' PLED RANDOM DIFENSIONS I * UNIT 1 CONTAINFEhT 80 * UNIT 1 AUXILIARY BUILDING 279 * UNIT 2 CONTAlffENT A10 AUXILIARY EUILDING 553 TOTAL 912
- DIMENSIONS SELECTED hERE CRITICAL-TO-DESIGN (SIGNIFICANT) ,
1 R
- 8 REFERENCE DIFENSIONS WERE INACCESSIBLE; WERE EVALUATED, NOT USED IN THE ANALYSES l
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_ _ - _ - _ _ _ _ - _ _ _ _ - _
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"8YRON AS-BUILT SYSTEM REVIEWS !
(LNI,T; N CONTAINMENT I
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SAFETY INJECTION COMPONENT COOLING CHILLED WATER ESSENTIAL SERVICE WATER FEEDWATER REACTOR COOLANT PRESSURIZER
,
U_ NIT 1 AUX. BUILDING CHEMICAL AND VOLUME CONTROL l SAFETY INJECTION l PRIMARY WATER REACTOR COOLANT PRESSURIZER . FEEDWATER ! AUXILIARY FEE 0 WATER CONDENSATE MAKEUP ESSENTIAL SERVICE WATER RESIDUAL HEAT REMOVAL COMPONENT COOLING UNIT 2 CONTAINMENT & AUX _. B_UILDING_ ESSENTIAL SERVICE WATER STEAM GENERATOR BLOWDOWN REACTOR COOLANT REACTOR COOLANT PRESSURIZER ( CONTAINMENT EQUIPMENT DRAINS i FUEL POOL COOLING " AUXILIARY FEEDWATER CHEMICAL AND VOLUME CONTROL l
;
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DESCRIPTION OF MITCHELL' DIMENSIONAL ERRORS SI-04 ANGULAR TOLERANCE EXCEEDED ON DRAWING
,
CC-46 ANGULAR TOLERANCE EXCEEDED ON DRAWING .i
!
CC-42 AS-BUILT DRAWING UNCLEAR-CC-37 CONFLICTING AS-BUILT DIMENSION ON DRAWING- j WO-1 WITHIN SEGMENT TOLERANCE SX-36 WITHIN SEGMENT TOLERANCE SI-12/SI-15 CONFLICTING AS-BUILT DIMENSION ON DRAWING FW-13 REFERENCE DIMENSION INCORRECT CC-37 REFERENCE DIMENSION INCORRECT
! !
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_ _ _ _ _ N i RESULTS: - LINEAR DIENSIONS
*
1 DIEhSION EXCEEDED RECONCILIATION TOLERANES (NRC)
* ERROR WAS CONSERVATIVE
- ANGULAR DIENS10NS
. * SOE WERE NOT ANNOTATED ON TE DRAWINGS *
I 1 WAS OUTSIDE RECONCILIATION TOLERANCES ),
- WAS PESOLVED THROUGH E-ANALYSIS AND - ET ASE CODE STRESS ALLOWABLES d
- NRC ISSUED A PROCEDURAL VIOLATION ON THIS ANGULAR PROBLEM :
l - ECO's RESPONSE TO NRC's FINDING
* PEPFORED DESIGN DOCUENT REVIEks * AS-BUILT REVIEkS * REVERIFICATION - 105 DIENSIONS * ALL ANGLES WERE WITHIN RECONCILIATION TOLERANES
- NRC REVIEWED EC0's ACTION - ACCEPTABLE -l
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- CIE IIAD IDENTIFIED 5 ORIGINAL AND 10 ADDITIONAL POTENTIAL PROBLEMS l *
1 WAS OUTSIDE THE RECONCILIATION TOLERANCES j i
*
RE-ANALYSIS FOUND THIS 1 TO KET ASME CODE STRESS ALLOWABLES l l l
1 i i l l l t ' t .l- '. [ ------ ---- - - -
_ CONC.LUSION: - DRAWING DISCREPANCIES DID EXIST
* DEGkEE OF ACCURACY '
- CEC 0'S OVERALL AS-BUILT PROGRAM WAS EFFECTIVELY IMPLEEhTED - NkC IEB 79-14 - OVERALL CLJECTIVE WAS ET
* AS-BUILT PIPING SYSTEM ET THE DESIGN
- ALL PIPING STRESSES ET THE ASE CODE
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- BYJO!N/BRAIDWOOD
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* OTHER ROUTINE NRC INSPECTIONS .;
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* CECO REVIEWS / INSPECTIONS l - OTHER REGION I INSPECTION AT SEABROOK (CIE ALLEGATION) SAE BASIC CONCLUSION ;
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ATTACHMENT II
' Meeting Attendees ,
Region III Byron Diesel Concerns 2/20/87 Name Organization Title J. A. Gavula RIII Reactor Insp - Mec , K. J. Green S&L Project Manager l Brent Shelton CECO Projects Eng, Mg ! , Ninu Kaushal CECO Pro. Engr._ Byron /Braidwood j l Duane Danielson NRC RIII Section Chief 1 J. J. Harrison NRC RIII Engineering Branch Chief '
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DIESEL GENERATOR MISSILES
*
DESIGN BASIS
*
POTENTIAL MISSILE HAZARDS
*
POTENTIAL FOR DAMAGE TO SX PIPING l
*
l CONCLUSION i l i l l l
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1
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DESIGN BASIS
*
MISSILES CONTAINED IN ENGINE
-
ORIGINAL SPECIFICATION REQUIRED IDENTIFICATION ' OF MISSILES i
-
BASED ON CRANKCASE STRENGTH; NO MISSILES WERE l l
'
l IDENTIFIED i
*
i SUPPORT SYSTEMS SEPARATED l l \
-
D!vlSIONAL INDEPENDENCE OF DIESEL SUPPORT ! SYSTEMS l '
-
DIESEL SELF-CONTAINED (IO EXTENT POSSIBLE)
-
FIRE WALLS BETWEEN REDUNDANT DIESELS
-
ISOLATED FROM NON-DIESEL HAZARDS -
,
l t--_-_---__---__---_---_--___-_--_----_---_--- - . - - - - - -
. .
POTENTIAL COMMON MODE FAILURES
-
SINCE THE ZION AND PALO VERDE COOPER BESSEMER DIESEL ENGINE FAILURES, THE POTENTIAL FOR MISSILES TO PENETRATE THE CRANKCASE AFTER AN INTERNAL FAILURE HAS BEEN REVIEWED
-
SX SUPPLY / RETURN TO "B" TRAIN ROOMS WHICH IS COMMON TO BOTH DIESEL DIESEL (IS 10gNLY SYSTEM SCHED I 40 PIPE) THIS PIPING PASSES THROUGH DIESEL ROOM "A"
-
REASON FOR ROUTING g4
' -
AVOID PENETRATION OF FLOORS C AM '
-
AVOID SAFETY RELATED FLUID SYSTEM (SX) ROUTING IN UNNECESSARY AREAS
-
ORIGINAL COMMON FAILURE MODE ANALYSIS
-
DIESEL ROOM PROTECTED FROM EXTERNAL MISSILES,
'
HELB
-
FIRE PROTECTION ANALYSIS SHOWS FIRE IN DG RM A WILL NOT ADVERSELY AFFECT SX PIPE TO RM B INTERNAL MISSILES NOT BELIEVED PROBABLE
-
- _ _ - _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ .
_-_ _
. .
POTENTIAL FOR MISSILE GENERATION
, -
POTENTIAL CAUSES OF INTERNAL FAILURE
-
OVERSPEED PROTECTION FAILURE
-
SAFETY RELATED MECHANICAL OvERSPEED PROTECTION l l - NON-SAFETY ELECTRICAL OVERSPEED PROTECTION l - FUEL AND AIR SHUT 0FF CAPABILITY
-
IMPROPER COMPONENT ASSEMBLY
-
DIESEL CLASSIFIED AS SAFETY RELATED COMPONENT
--
MAINTAINED UNDER QA PROGRAM j
-
DIESEL IS REGULARLY IESTED, MONITORED, AND
!
MAINTAINED .!
-
INTERNAL PARTS FAILURE
-
IDENTIFIED SUSPECT COMPONENTS (RODS) REPLACED i
-
QA PROGRAM APPLIED TO REPLACEMENT PARTS L l l
--------------.-.-----._--_.-..-_a- - - - _ - - - - - - - - - - , - - _ - - - - - - - - -
_ - - - - - - - - _ - -
, - - - - - - .
MISSILE GENERATION CRITERIA
1
-
DESCRIPTION OF MISSILE
-
200 LB MAXIMUM WEIGHT # jf>M
-
SHARP EDGED
-
INITIAL VELOCITY 69 FPS (660 RPM AT 1 FT CRANK RADIUS)
;
,
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AIMED AT EXPLOSION DOORS (ONLY-POINT WEAK EN0 UGH FOR j PENETRATION)
-
MISSILE EJECTION MECHANISM - BOLT FAILURE IN COMBINATION WITH LOCAL CRANKCASE FAILURE -
-
DIRECTION OF MISSILE I i
-
INITIAL DIRECTION PERPENDICULAR TO CRANKSHAFT CHANGE IN DIRECTION POSSIBLE DUE TO IMPACT
- -
POTENTIAL MISSILES +/- 250 FROM PERPENDICULAR f
(CONSISTENT WITH R.G. 1.115)
l - TRAJECTORY IS HORIZONTAL OR LOWER (HIGH TRAJECTORIES WILL NOT STRIKE EXPLOSION DOORS)
-
RESULTS OF CRANKCASE FAILURE ANALYSIS l - MINIMUM OF 36% OF INITIAL ENERGY ABSORBED IN EXPLOSION l D0OR AND BLOCK FAILURE
-
EXIT VELOCITY PREDICTED TO BE MAXIMUM OF I48 FPS
,
I I
_ _ _ _ - _ -
. .
POTENTIAL FOR DAMAGE TO SX
*
UNIT 1 :
-
PIPING IS NOT IN DEFINED MISSILE PATH
-
NO POTENTIAL FOR DAMAGE TO PIPING DUE TO GE0 METRY AND PROTECTIVE STRUCTURE
*
UNIT 2 CONSERVATIVE ANALYSIS
-
ASSUMPTIONS FOR ANALYSIS i
-
MAXIMUM EXIT VELOCITY (48 FPS) !
-
HORIZONTAL INITIAL TRAJECTORY
-
HORIZONTAL ANGLE WITHIN +/- 250 0F PERPENDICULAR s
-
ENERGY LOSS IN DEFLECTION IGNORED )
-
RESU TS OF ANALYSIS-
-
6 BELOW SX PIPING AT NQ6 REST POINT j
- -
NEAREST SX PIPE BEYOND 4b0 (ACTUAL 320) . MISSILE DOES NOT STRIKE SX P1 PING q
-
HORIZONTAL MISSILE REQUIRED /2 FPS TO STRIKE PIPING LIMITING ANALYSIS
-
ASSUMPTIONS
-
SAME AS ABOVE EXCEPT ALL ANGLES PERMITTED (N0 i RESTRICTIONS ON EITHER HORIZONTAL OR VERTICAL ANGLES) l
-
RESULTS OF ANALYSIS I AT CERTAIN ANGLES SX PIPE COULD BE IMPACTED
- - l SOME SUPPORTS COULD BE DAM $GED ' -
PIPING RUN BENT LESS THAT 3.50 AT ANCHOR
-
PIPING WILL REMAIN FUNCTIONAL ZION COMPARISON l
-
AT ZION MISSILES WERE APPROX 150 LBS AS OPPOSED TO 200 LBS ASSUMED IN BYRON ASSESSMENT ,
-
BYRON PREDICTION IS 30 FT TOTAL TRAVEL OF MISSILE
'
FROM ENGINE AS OPPOSED TO MAXIMUM OF 15 FT AT ZION
- -_ -- - - - )
, ..
CONCLUSIONS ,
;
,
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ANALYSIS METHOD USED CORRESPONDS WELL WITH EVENT AT ZION
-
MISSILE IMPACT ON REDUNDANT DIESEL SX SUPPLY
-
PIPING WILL REMAIN FUNCT!0NAL AFTER ARBITRARY POSTULATED MAXIMUM IMPACT
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. DESCRIPTION P NDE FAILURE MODE ANALYSIS t
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WELD INSPECTIONS l i ' B ! l l
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. .
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DAMAGE ASSESSMENT VT'S
! -PERFORMED BY SARGENT'& LUNDY NUTECH HUNTER ,
l-ALL ATTACHMENTS TO B & C LINES '
-ALL SURROUNDING PIPE, HANGERS, ETC.! ; -CONCRETE ,
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-AREAS INSPECTED l 13 WELDS -
B l , 8 WELDS - C 1
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-
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-DOUBLE WALL RT-1 WELD l
-WALL THICKNESS AT WELDS AND BOTH SIDES OF WELDS
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- ___________-_____
- __- , -i
./ ; FAILURE MODE ANALYSIS -
, , .
SINGLE HANGER FAILURE
! ' -
LOCKED SNUBBER WATERHAMMER
. .
f i
. ' ' .
_
..
j ..+e e g ,g
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l SINGLE HANGER FAILURE '
. -NORMAL DEADWEIGHT LOAD 16-23 KIPS i-HYDRO LOAD 30-39 KIPS ' ! -DESIGN CAPACITY 40 KIPS-DESIGN LOAD (WT+TH+SSE) g 33 KIPS .
e s
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.
D L R E E W E P_ P T I .
- N Z .
E M D R H - s L E C U G A O N T C A T H A E R E E U V V L I I I T T A C C F E E F F E E E L D D B U F F S O O O E D C N N . N S O O E U .I I U O T T Q E A A E N C C S A I I N T D D O L N N C U I I M O I O O N S N N
- - - -
(
s
\
1 a a f LOCKED SNUBBERS J
'1l N
'
-ALL LATERAL & LONGITUDINAL il . :) ' -HOT LOCKUP !
i-
-COLD LOCKUP l
, . &
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-- _ _ _ _ _ _ _
_
- - % ,
WATERHAMMER
.
,
-DAMAGE VERY SIMILAR TO BRAIDWOOD-
-SHEAR FAILURE AT ALL ROD HANGERS i-ALSO DUCTILE TEARING AT WEST END
.
-INTERNAL DAMAGE TO 1228 SNUBBER-SINE WAVE AT 70' KIPS I i
J l y j -PIN DESTRUCTION ESTIMATED AT l 100 KIPS OR GREATER
. I i-VERTICAL PIPE MO EMENT AT EAST END "
-FATIGUE EVALUATION -
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_ _ _ _ _ _ _ _ _ _ _ _
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.1 CONCLUSION
'
:
-SCENARIOS EXAMINED .
-
-WATERHAMMER 9-PIPE UNDAMAGED i -REINSTALLED TO ORIGINAL CONFIGURATION I
I , 1 -
,
e t
' - - . _ - _ _ - _ _ - - - - - . _ _ _ '____ a-
~,- ;
,
N
,
SUMMARY - i
'
i i-APPRO ' PIPE AFFECTED-LIMITED LATERAL MOTION, HIGH LOADS E
.: -HANGERS FAILED VERTICALLY IN SHEAR-NO NOISE REPORTED-PIPE IS INTACT AND UNDAMAGED \' - - - - - . . _ _ _ _ _ . - - .=- ____ J
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