ML20206F304

Safety Evaluation Accepting Util Review Re Cold Pressurization of Secondary Side of Steam Generator During Testing on 850906

ML20206F304
Person / Time
Site:	Davis Besse
Issue date:	05/27/1986
From:	Office of Nuclear Reactor Regulation
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Shared Package
ML20206F301	List: ML20206F296 ML20206F304
References
TAC-60062, NUDOCS 8606240415
Download: ML20206F304 (4)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l COLD PRESSURIZATION OF SECONDARY SIDE OF STEAM GENERATOR DURING TESTING i DAVIS-BESSE NUCLEAR POWER STATION DOCKET NO. 50-346

1.0 INTRODUCTION

By letter dated October 18, 1985 (Reference 1), Toledo Edison, the licensee for Davis-Besse, informed the NRC that on September 6, 1985, during special testing of the Davis-Besse Nuclear Power Station No.1  ;

auxiliary feedwater pump turbine (AFPT) 1-1, the secondary side of j the once-through steam generator (OTSG) 1-1 was subjected to a naxi-mum pressure of 1058.5 psig. This cold pressurization was discussed in Licensee Event Report 85-017, dated October 4, 1985.

Subsequent to this event, an engineering evaluation was performed by the licensee as required by Section 3.7.2.1.b of the Davis-Besse Technical Specifications. This evaluation included an analysis of l the potential growth of existing flaws in the secondary side shell and main steam nozzle regions by Babcock and Wilcox (B&W). In addition, Toledo Edison and Bechtel Power Corporation reviewed the existing analysis of structural loadings associated with water filled main steam lines.

Based upon the information on the transient, stress, and fracture mechanics analyses, Toledo Edison conducted a 10 CFR 50.59 review.

This review concludes that the September 6, 1985 pressurization of the OTSG 1-1 had no adverse structural effects and that the steam generator remains acceptable for continued operation at full power conditions. Toledo Edison has concluded that this is not an unreviewed safety question.

2.0 EVALUATION

1. Effects of secondary side overpressurization on the Once Through Steam Generator with respect to brittle fracture and crack growth.

The licensee has performed an analysis in accordance with the linear elastic fracture mechanics methods recommended in ASME Code Section XI, Appendix A (1980 Edition) to show that an assumed flaw one quarter times the shell thickness in depth and having an aspect ratio of 6:1 will not grow at the secondary side pressure of )

i 8606240415 860527 gDR ADOCK 05000346 PDR u . _ . _ _ _ _.

, 1080 psi and temperature of 100*F which occured during the over-pressurization event. Since the most highly stressed locations are evaluated, other locations which might contain indications are bounded by this analysis. The minimum allowable temperaturas to preclude crack growth in steam generator shell and steam outlet nozzle were determined to be 19 and 89*F, respectively. Since both these temperatures are lower than the actual temperature (100 F) during the overpressurization event, no crack growth initiation is expected. No crack growth initiation at 100*F is also predicted for the shell to tube sheet junction.

The technical specification limits of 237 psi and 110 F for the OTSG are to preclude failure of the secondary side. The values were arrived at in accordance with 10 CFR 50, Appendix G guide-lines (Reference 2). These guidelines state that protection against non-ductile failure is achieved by limiting the pressure to 20 percent of the preservice system hydrostatic test pressure, which for this plant is 1312.4 psi. Taking into account instru-ment errors of up to 25 psi results in the technical specifica-tion limit of 237 psi. This limit is conservatively set to ensure that the materials toughness is on the upper shelf. Other values are acceptable if the conservatisms and margins are met.

The pressure and temperature evaluated by the licensee is one of the many sets which meet the above criteria.

. Conclusion Staff has reviewed the calculations performed by the licensee to evaluate the effect of the secondary side pressurization of Steam Generator 1-1 of Davis-Besse. Analysis indicates no adverse effect on the steam generator. These results are not in conflict with the fact that the material toughness precludes brittle fracture. The technical specification limit of 237 psi provides an acceptable factor of safety in lieu of specific analysis to justify other acceptable pressure and tenperature conditions. The licensee has demonstrated that the conditions experienced during this incident represent one of the many sets which are acceptable.

2. Evaluation of the weight effects of water in steam outlet nozzle because of the cold pressurization event.

The pressure and temperature of 1080 psig and 100 F, respectively, achieved inadvertently in the process of testing the Davis-Besse AFW system are less severe than the Code hydrotest conditions, even  ;

though the steamline was filled with water. Because of the addi-tional weight effects, the licensee has performed calculations to )

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justify the effect of the conditions on the nozzle. 1 The steam outlet nozzles were originally analyzed as part of a Class 1 vessel. The applicable Code was ASME, 1968 Edition with Addenda thru Summer 1968.

r From Para. N-713.2, the applicable stress limits are:

P,< 0.9 Sy PL+PB < 1.35 Sy More recently, the 1980 Edition of the Code specifies stress requirements for pneumatic test to be:

P, < 0.9 S y PL+PB < 1.35 Sy when P, < 0.67 Sy

< 2.15 Sy - 1.2 Pm when 0.67 Sy < P, < 0.9 Sy The loadings on the nozzle due to the added weight of the water were determined to be Forces (lbs): 3,853 (axial), 17,013 (shear), and 4,660 (shear)

Moments (ft-lb): 11,651 (torsion), 31,555 (bending), 116,846 (bending)

From these loadings the maximum stress intensity, pressure, dead-weight and thermal stresses were determined. All these stresses were determined to be within the ASME Code allowable stresses.

Conclusion Based on a review of the analysis performed by the licensee to determine the structural loading imposed on the main steam nozzles -

as a result of water filled main steam lines, the staff concludes that the change in loadings as a result of the localized water load has a negligible impact on the previously calculated stresses as documented in the Code Stress Report for Davis-Besse.

NRR Personnel contributing to this evaluation: J. Rajan Dated: May 27,1986

References

1. Letter from J. Williams, Toledo Edison, to J. Stolz, NRC, dated ,.

October 18,1986, " Cold Pressurization of Secondary Side of the Once Through Steam Generator at Davis-Besse Unit 1."

2. Babcock and Wilcox Report BAW 10046P - Methods of Compliance With Fracture Toughness and Operational Requirements of 10 CFR 50, Appendix G, March 1976.

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