IR 05000341/1987033

Insp Rept 50-341/87-33 on 870803-07.Violations Noted.Major Areas Inspected:Licensee Action on Previous Insp Findings, Onsite Followup of Events at Operating Reactors,Onsite Followup of Written Repts & Training

ML20237G855
Person / Time
Site:	Fermi
Issue date:	08/19/1987
From:	Danielson D, James Gavula, Liu W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20237G843	List: IR 05000341/1987033 ML20237G839 ML20237G847
References
50-341-87-33, NUDOCS 8708240269
Download: ML20237G855 (10)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION III

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I- Report No. 50-341/87033(DRS)

Docket No. 50-341 License No. NPF-43 Licensee: Oetroit Edison Company 2000 Second Avenue Detroit, MI 48224 Facility Name: Enrico Fermi Nuclear Power Plant, Unit 2 Inspection At: Enrico Fermi 2 Site, Monroe, Michigan Inspection Conducted: August 3-7, 1987 l

Inspectors: . C. Liu t//f/#7 Date I (h h.gt

.'A. Gavula P- /9- P 7 -

Date Approved By: D. H. Danielson, Chief (f/f//7 Materials and Processes Section Date Inspection Summary Inspection on August- 3-7, 1987 (Report No. 50-341/87033(DRS))

Areas Inspected: Special, announced safety inspection of licensee action on previous inspection findings (92702), onsite followup of events at operating reactors (93702), onsite followup of written reports (92700) and training (41400).

Results: Two apparent violations were identified (failure to. follow procedures - Paragraph 2.a and inadequate design control - Paragraphs and 4.b).

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DETAILS 1. Persons Contacted Detroit Edison Company (DECO)

• B. Sylvia, Group Vice President

• F. Agosti, Vice President, Nuclear Engineering and Services

• G. Overbeck, Director, Operator Training

• S. Latone, Director, Nuclear Training - General F

• G. Trahey, Director, Nuclear Quality Assurance L

• L. Schuerman, General Supervisor, Nuclear Systems

• S. Catola, Chairman, Nuclear Safety Review Group

• L. Fron, Supervisor, Projects and Plant Engineering

• R. McLeod, Supervisor, Nuclear Training - General

• S. Cashell, Licensing Engineer A. Colandrea, Lead Engineer, Projects and Plant Engineering S. Hassoun, Senior Seismic Engineer R. Bryer, System Engineer, Nuclear Engineering Stone & Webster Engineering Co. (S&W)

S. Leary, Lead Engineer, Structural / Engineering Mechanics C. Ng, Engineer, Structural / Engineering Mechanics Hopper and Associates D. Hopper, Engineer U. S. Nuclear Regulatory Commission

• W. Rogers, Senior Resident Inspector

• H. Miller, Director, Division of Reactor Safety

• J. Stefano, Project Manager, NRR

• Denotes those attending the final exit interview on August 7, 198 . Licensee Action on Previous Inspection Findings (Closed) Violation (341/85011-01): Inadequate corrective action was taken on spacing violations for concrete anchor bolts. This item related to DECO's March 25, 1981, 50.55(e) Item No. 42. Corrective actions implemented as a result of this violation are delineated in DECO's response from W. H. Jens to J. G. Keppler, April 18, 1985, and from F. E. Agosti to the NRC, May 5, 198 The following documentation was reviewed to verify compliance with NRC requirements and licensee commitment * Calculation DC No. 3200, Revision A, December 17, 1985,

" Severe Wedge Anchor Spacing Calculation".

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• Specification No. 3071-226, Revision G, July 15, 1985,

" Purchase and Installation of Concrete Anchors", Appendix C,

" Guide to Design of Expansion Anchors".

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• Calculation DC No. 668, Volume III, Revision 0, July 17, 1987, " Pipe Support E41-3163-G20".

Based on the review by the NRC inspector, the corrective actions taken by DECO were considered adequate and this item is considered close However, it was noted during the review of DC No. 688 that S&W had not utilized Specification No. 3071-226, Appendix C, Design Methodology in their calculation. This standard methodology was developed for plant wide use as corrective action to prevent further noncompliance for the above violation. During discussions ._

with S&W personnel, it was stated that although they were aware of the Appendix C methodology, S&W did not routinely utilize the referenced approach. Instead, S&W used their own internal design methodology for anchor bolt spacing violations. It was noted by T the NRC inspector that the existing approach used by S&W is at least as conservative as the methodology developed in Specification No. 3071-226, Appendix On this basis the safety significance of this issue is considered minimal. The failure of S&W to follow procedures, however, is an example of a violation of 10 CFR 50, Appendix B, Criterion V. (341/87033-01) (Closed) Violation (341/85011-02): Inadequate design control was used for concrete expansion anchor work. Corrective actions implemented as a result of this violation are delineated in DECO's response from W. H. Jens to J. G. Keppler, April 18, 1985, and from F. E. Agosti to the NRC, May 5, 198 The following documentation was reviewed to verify compliance with NRC requirements and licensee commitment * Calculation DC No. 974, Revision C, February 22, 1985,

" Installation Torque for Wedge Anchors".

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• Engineering Package EDP-2356, Revision 0, May 11, 1985, l " Revision of Project Specification 3071-266, Revision F and N Drawing SC721-2002, Revision H".

• Drawing SC721-2002, Revision 1, August 9, 1985, and Revision H, May 13, 1983, " Wedge Anchor Standard Details".

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• Specification No. 3071-266, Revision G, July 15, 1985,

" Purchase and Installation of Concrete Anchors".

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Based on the review by the NRC inspector the corrective actions

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taken by DECO were considered adequate and this item is considered close ,

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However, during the course of the review of the above documentation the following deficiencies were identifie (1) In DC No. 974 the torque requirements for 1 inch diameter wedge anchors was calculated using the shear capacity of the bolt instead of the tensile capacity. This approximately 6% nonconservative error potentially affects the required installation torque of this size bol (341/87033-02A)

(2) On Drawing SC721-2002 the minimum edge distance for 1 1/4 inch diameter wedge anchors was given as 6 inches whereas the previous revision and all other documentation gives 6 1/4 inch for this dimension. This error occurred during the current revision to incorporate EDP-2356 when the drawing I was reproduced on the computer aided drafting system.

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l l (3) In Specification No. 3071-226, Appendix A, " Anchorage in l Concrete Masonry Wall," the definition for " Manufacturer or Supplier" was incorrectly transcribed from E0P-2356. The definition for " Seller or Distributor" was completely omitte (341/87033-02C)

The above three instances are examples of a violation of 10 CFR 50, Appendix B, Criterion III in that the design bases were incorrectly translated into the design document . Review of Licensee's DER 86-191 for the Steam Tunnel Cooler Platform Background Information On November 21, 1986, the licensee initiated a deviation event report (DER)86-191 which noted that some structural steel assemblies for the steam tunnel cooler platform were not installed in accordance with the detailed drawings. Furthermore, the licensee's review of l 00-0576 calculations revealed the following:

(1) The calculations did not consider lateral loads in the upper end connection design.

i (2) The calculation did not consider the effects of eccentricity I between the longitudinal bracing and the center of the mass of the uni (3) No frequency calculations were made other than in the vertical direction of the platfor Seismic Evaluation The seismic evaluation of the steam tunnel cooler platform was performed in December 1986 by the licensee's subcontractor, Hopper and Associates. The evaluation demonstrated that the platform would deform and yield during a seismic event. However, with the

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consideration of ductility effects of the steel platform, it L' -would not collapse and therefore not affect any safety-related l' . equipment.

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'Since both the coolerLand the steel platform are not safety-related'

items, the evaluation ~of the platform for seismic effects appeared to be acceptabl Corrective Action The steam tunnel cooler platform has been evaluated _with respect to the consideration of eccentricity of the structural assemblies and seismic affects. Design modifications were documented in DC No. 576, Revision 4 and structural modifications were specified in-Engineering Design Package (EDP) 6720. The modifications included supports t the ceiling, a brace to the adjacent wall, and the addition of. steel members for. internal bracings. All modification work activities including verification of as-built configuration were completed on March 17, 198 Based on the above review and discussions with licensee representatives regarding.the. aforementioned concerns, it was determined that the issues contained in DER 86-191 were resolved and this item was considered close Within the' areas inspected, no vioiations or deviations were identifie . ' Review of Licensee's DER 86-167 for the Masonry Wall Evaluations j Background Information  !

In accordance with the Fermi FSAR requirements on masonry walls, the. licensee must insure that the masonry walls are qualified as seismic II/I structures. These walls are not to be used as-seismic I supports and must not collapse on any safety-related equipment during a seismic event. In October 1986, the licensee discovered that some masonry wall connections were not installed per detailed drawings. .The as-built conditions had not been  !

addressed in the original design calculations. As a result,  !

these walls may not provide the required strength and may result in damage to nearby safety-related items during a seismic even Accordingly, the licensee initiated DER 86-167 on October 27, 1986, for further evaluations of the masonry walls for seismic effects due to the nonconforming condition of these wall Review of Design Calculations for Seismic Effects The NRC inspector reviewed several letters pertaining to the evaluation of the masonry walls. Letter NE-87-0045, from L. Simpkin to R. Lenart, dated April 10, 1987, states that " Nuclear Engineering has completed a seismic analysis of the walls. The analysis is documented in DC 4479A. This calculation demonstrates that the

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i walls previo'u sly identified wi111not collapse during a seismic Levent." Letter NE-PJ-87-0276, from C. Gelletly.to S. Frost, dated )

L May 5, 1987, states that "The operability evaluation was performed 'l in DC 4479 and was to'show that walls with safety-related items in-close proximity would not collapse during a seismic event." Letter

NE-PJ-0368, from C. Gelletly to. S. Frost, dated June 8,1987, states

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that " Design Calculation 4479A,. entitled Evaluation of Masonry Walls for DER 86-167, demonstrates that the nonconforming masonry walls will not collapse:in a seismic event".

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,The NRC. inspector noted that the key document used for the evaluation of the masonry walls was Design Calculation DC 4479, Revision A, dated April 13, 1987. 'This design calculation had been verified and approved by the licensee's responsible personnel. The inspector'

reviewed the above design calculation to determine whether it meets ,

the licensee commitments and the NRC requirements. The inspector i noted that the following design errors were contained in the design

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calculation:

(1) Masonry Wall No. 219 on Page 7 of 15 (a) At the top of the wall the calculated maximum moment  ;

(equal to 0.045 qa2 ) was intended to be the horizontal momen This moment was actually in the: vertical direction and should be zero at top of the wall. The calculated moment should be distributed to the bottom of the wal (b): The maximum moment at top of the. wall in the horizontal direction was not performed as require (c) At the bottom of the wall,.the maximum moment (equal to 0.120 qa2 ) in the vertical direction should be increased to account for the moment carried over from top of the wal i

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(d) The calculated tensile stress at top of the wall was in the vertical direction and should be zero. The tensile stress in the horizontal direction was not performed as require (e) The calculated bending stress at the bottom of the wall in the vertical direction should be increased due to additional moment being carried.over from top of the wal (f) The design assumptions are acceptable, but additional explanations must be provided to justify the vertical boundary condition at top of the wal .- - _ - -__

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(g) Portions of the lateral seismic loads were. distributed to the door frame which was assumed to be a simply supported  !

l- structure. However, the door frame was not evaluated to  !

L determine whether it can withstand the calculated seismic load !

(h) The shear stress during a seismic event was not considered  !

in the design evaluatio (2) Masonry Wall No. 234 on Page E-5 (a) The calculated maximum moment, (equal to 0.0503 qa2 ) in the vertical direction should be the moment in the horizontal directio (b) The calculated maximum moment, (equal to 0.0694 qa2 ) in the horizontal direction should be the moment. in the i vertical directio (c) The calculated bending stress in the vertical direction should be the bending stress in the horizontal direction for both the OBE and SSE case (d) The calculated bending stress in the horizontal direction should be the bending stress in the vertical direction for i both the OBE and SSE case (e) The shear stress during a seismic event was not considered in the design evaluatio (3) Masonry Wall Nos. 216, 218 and 221 on Page E-7 (a) The ratio of the horizontal to the vertical dimensions  !

of the wall was not in accordance with the specified reference formula. As a result, the calculated natural i frequency for the wall was incorrec l (b) A factor of 1.5 for seismic loads was not included in the design calculations. This was inconsistent with the previous similar design evaluation (c) The shear stress during a seismic event was not considered in the design evaluation, i

(4) Masonry Wall No. 216 With Different Boundary on Page E-8 (a) The ratio of the horizontal to the vertical dimensions of the wall was not in accordance with the specified reference formula. As a result, the calculated natural frequency for the wall was incorrec _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _

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(b) A factor of 1.5 for seismic loads was not included in the design calculations. This was inconsistent with previous similar design evaluation (c) The shear stress was not considered in the design evaluatio (d) The use of the Zero Period Acceleration (ZPA) value, based on a calculated frequency of 12.9 HZ, was inappropriat The corresponding seismic response spectrum should be verified prior to using the ZPA value in the design evaluatio ;

(5) Section 6-6 on Design Drawing No. 6C721-2608, Revision H,

" Design-Reinforcement for Existing Block Walls", could not be found. Revision G of this drawing noted that Section 6-6 was deleted per DCN 10831. However, Section 6-6 was still shown on the drawing in Revision The above design discrepancies were identified during the course ;

of this inspection. These design discrepancies were discussed in l the detail with licensee representatives. No dissenting comments were received from the licensee. The discrepancies identified above are other examples of a violation of 10 CFR 50, Appendix B, Criterion II (341/87033-020) .

Within the areas inspected, one violation was identifie . High Pressure Coolant Injection (HPCI) Piping Transient Event During startup testing of the HPCI System, the system failed to meet its j time to rated flow criteria as defined in the Technical Specification As a result, the system was declared inoperable on July 4, 1987. On July 5,1987, a quick start vessel injection was conducted to determine the cause of the observed deficiency. During this test, a HPCI overspeed ,

trip occurred and a subsequent overpressurization was indicated by the HPCI suction line instrumentation. As a result, the system was evaluated to determine the cause of the overpressurization transient and to i identify any damage to the HPCI suction pipin A detailed walkdown was conducted on the HPIC suction piping from the pump to the condensate storage tank, including all branch piping and !

tubing to the first closed isolation valv '

The scope of the walkdowns included all piping, pipe components, and fittings, and addressed obvious physical deformation, leakage, and piping ovality. Also, all pipe t supports, instrument tubing, instruments, valves and in-line components were visually inspected for any signs of damag In addition to the walkdown inspections, approximately 32 pipe welds were non-destructively examined (NDE) using ultrasonic, dye penetrant and radiographic techniques. Also a number of valves were stroked to verify proper operatio j i

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A stress analysis was also performed on the HPIC suction piping in order to quantify the support loads and pipe stresses that resulted from the indicated pressure wave. Three analytical models were required to evaluate the piping with each model isolated by anchors, The analyses extended from the HPCI pump to the HPCI torus penetration, back to the condensate storage tank, over to the Reactor Core Isolation Cooling (RCIC) pump and the RCIC torus penetratio A time history forcing function was utilized to drive each mode The forcing function characteristics were derived from the pressure data recorded in the GETARs system. Using conservative analytical assumptions, the amplitude, period and shape of the pressure wave were calculated for.each segment of the pipe in the models. This loading ,

information was applied at appropriate points on the models assuming '

conservative loading sequence Using the preliminary. piping analyses as a basis, the support loads, pipe stresses and pipe deflections were compared with the observed damage from the walkdown This review concluded that there may be a substantial degree of conservatism inherent in the piping analysis. This conservatism is most likely due to the assumptions made for the forcing functions since the time history analytical method is relatively precise. It was noted, however, that for at least one support, the transient load exceeded the support's ultimate load capacity. On this basis, the behavior of the piping system was nonlinear and as such could not be precisely predicted using linear techniques. The results would, however, indicate which portions of piping would be expected to experience the most damag Ultimately, the operability of the subject piping was based on the extensive post event inspections including the NDE data and walkdown ,

information. Only one support sustained any damage as a result of the event and this was corrected prior to declaring the system operable again. All of the other inspection data indicated that the system was not degraded and is capable of fulfilling its safety functio ,

The following documentation was reviewed during the course of the inspectio * OC 4771, Volume I, Revision 0, July 16, 1987, "Waterhammer Loads on HPCI Suction Pipe Due to HPCI Turbine Overspeed Trip System Test".

• DC 4692, Volume 2, Preliminary, Part A, B and C, " Fluid Transient Analysis HPCI Suction".

• DER 87-232, July 5, 1987, "HPCI Overspeed Trip During Startup Testing".

Within the areas inspected, no violations or deviations were identifie '

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' Training The qualification and training records for the following individuals were

, reviewed. .-These individuals were directly-involved in the technical work

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reviewed during the course of this inspection.

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A. F. Colandrea, Principal Engineer R. A. Bryer, Systems Engineer A. I. Hassoun, Senior Engineer J. P. Thorpe,-Engineer

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All qualification and training records were satisfactor It was noted that .one individual had not yet received the required QA Awareness trainin No violations or deviations were identifie ! Exit Interview The Region III inspectors met with the licensee representatives (denoted i in Paragraph 1) at the conclusion of the inspection on August 7, 198 .The inspectors summarized the purpose and findings of the inspectio The licensee-representatives acknowledged this information. The inspectors also' discussed the likely informational content of the inspection report with regard to documents or processes reviewed during the inspectio The licensee representatives did not identify any such i documents or processes as proprietar !

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