ML20235K924

Safety Evaluation Re Reactor Pressure Vessel Flaw.Flaw Conditionally Acceptable Per Subarticle IWB-3123 of Section XI of ASME Code & Therefore Requires Augmented Inservice Insps Based on 10CFR50.55(g)(4)

ML20235K924
Person / Time
Site:	Point Beach
Issue date:	07/09/1987
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20235K884	List: ML20235K880 ML20235K924
References
NUDOCS 8707160490
Download: ML20235K924 (5)

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St.FETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT, UNIT NO. 1 l DOCKET NO. 50-266 l FLAW EVALUATION IN 1HE POINT BEACH NUCLEAR PLANT UNIT 1 REACTO INTRODUCTION For nuclear power facilities whose construction permit was issued prior to January 1, 1971, 10 CFR 50.55a(g)(4) requires that inservice inspection be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code. The ASME Code requirements used for inspection of the Point Beach Nuclear Plant Unit 1 (PBNP-1) reactor pressure vessel was the 1977 Edition through In addition, the licensee performed the Summer 1979 Addenda of Section XI.

examination based on Regulatory Guide 1.150. Article IWB-3600 of the Code establishes criteria for evaluating flaw indications discovered during inservice Article examination tnat exceed the allowable size criteria in Article IWB-3500.

IWB-3610 states that the evaluation procedures shall be the responsibility of the owner and shall be subject to approval by the regulatory authority having jurisdiction at the plant site. In a letter to the U.S. Nuclear Regulatory Commission dated June 2, 1987, the Wisconsin Electric Power Company (the licensee) submitted for staff review their evaluation of a flaw in the safety injection nozzle-to-shell weld of the PBNP-1 reactor pressure vessel.

DISCUSSION A. Ultrasonic Flaw Evaluation During the April 1987 inservice inspection, the PBNP-1 safety injection nozzle-to-shell weld at a vessel azimuth of 288.5 degrees was scanned from the nozzle bore using two separate search units. Based upon size and configuration of the nozzle, the optimum transducer arrangement was determined to be a 0-degree, 2.25 MHZ longitudinal wave (O L) and a 10 degree, 2.25 MHZ refracted longitudinal wave (10 RL). Prior to examination, the ultrasonic instrument was calibrated by measuring ultrasonic responses from known reference reflectors machined into the basic calibration block and constructing a distance amplitude correction The DAC curve was the primary reference level for recording (DAC) curve. Flaw indications were detected by threshold response signals reflectors.

required to be recorded by criteria given in Section XI using a scanning sensitivity of twice the primary reference level, and by interpreting the characteristics of the signals.

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After detection, flaw indications that exceed the recording criteria established by the ASME Code Section XI are evaluated to determine their size, location and characteristics. One flaw indication in the PBNP-1 safety injection nozzle-to-shell weld exceeded the allowable limits in Section XI, Article IWB-3500. The licensee performed beam spread correction, re-evaluated past ultrasonic and radiograph examination results, and reviewed the weld fabrication procedure to locate, size, and characterize the flaw.

The safety injection nozzle-to-shell weld configuration was a double "U" joint with the depth of the weld root from the inside surface being 5.2 inches and a 10 degree weld preparation angle. The weld was fabricated using the manual stick electrode process. The licensee indicates that an above-average amount of inclusions and voids have been observed in heavy vessel welds of this configuration.

The licensee indicates that the ultrasonic signal amplitude had several peaks, which suggested the indication was actually a series of small flaws. The flaw was located at the root of the inside "U" joint and, in accordance with the criteria in Article IWA-3300 of the ASME Code, it was characterized as a single embedded flaw. The flaw was sized without beam spread correction and with beam spread correction.

j B. Fracture Mechanics Evaluation  !

The licensee has provided a flaw evaluation chart for embedded f circumferential and longitudinal flaws in the PBNP-1 safety injection l nozzle-to-vessel weld. This chart was constructed using fracture mechanics analyses. The method and criteria used in the fracture mechanics-analyses are documented in Reference 1. The portions of this document that were relative to the PBNP-1 flaw evaluation were documented in Appendix B of the licensee's June 2, 1987 submittal. The fracture mechanics analyses that were performed to develop the flaw evaluation chart were in accordance with the methodology and criteria specified in Article IWB-3600 and Appendix A of the ASME Code Section XI except that stresses were not linearized and stress intensity factors were not calculated in accordance with the recommendations in Appendix A. In lieu of linearizing the stress, the method used represented the actual stress profile by a third order polynomial. Stress intensity factors l were calculated using the expressions of Reference 2. These stress intensity factor expressions have been shown to be applicable to vessels in Reference 3. These stress profiles and stress intensity factor expressions are believed to provide a more accurate determination of the critical flaw size, and are particularly important during the evaluation of emergency and faulted conditions where the stress profile is generally nonlinear and of ten very steep.

Important parameters in a fracture mechanics analysis are the materials' brittle fracture resistance and the projected flaw growth rate during operation of the component. The standard measurement of brittle fracture resistance for the vessel materials in the PBNP-1 reector vessel are their 1

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e crack initiation and arrest fracture toughness. These values of fracture toughness are used to determine a critical flaw size. Westinghouse indicates that the critical flaw size calculat#on used the crack initiation and arrest fracture toughness for vessel materials that are recommended in Appendix A of the ASME Code Section XI. The critical flaw size for the safety injection nozzle-to-vessel weld location was of 60 F and an upper determined using a reference temperature, RTThese values NI, acceptable for this shelf toughness of 200ksilin.

location in the reactor vessel, because the materials in this location are not subject to significant amounts of neutron irradiation.

The amount of projected flaw growth was detennined to be negligible. The calculation was performed for the reactor vessel design transients that are listed in Table 4-1 of the June 2, 1987 submittal. The rate of fatigue growth was calculated using the ASME reference curve for air environment. Since the flaw under evaluation is embedded, this method of calculating the flaw growth rate is acceptable.

The flaw evaluation chart was constructed from the analysis of all reactor vessel design transients that are listed in Table 4-1. The limiting transient for the safety injection nozzle location was determined to be the large break loss-of-coolant accident. The flaw evaluation chart indicates that the flaw sized with or without beam spread correction, meets the fracture mechanics criteria in Article IWB-3600 for the 40 years of service life of the plant.

In addition to the reactor vessel design transients, which are listed in Table 4-1, the licensee performed a fracture mechanics analysis of a postulated low temperature overpressure (LTOP) event, which was not mitigated by the LTOP protection system. The licensee's analysis of the postulated LTOP event indicates that the flaw satisfies the fracture The LTOP mechanics criteria in Article IWB-3600 for faulted conditions.

event assumed that there were multiple failures of the LTOP protection The system and the pressure rose to 1500 psi, at a temperature of 150*F.

event was classified as a faulted condition because multiple failures were assumed.

The NRC required licensees to install LTOP protection sy' tens in 1979. ,

The staff's evaluation of the PBNP-1 LTOP protection system is contained i in Reference 4. Since the industry installed LTOP protection systems, there has been only one event in which the LTOP system did not mitigate the event. This event occurred on November 28, 1981 at Turkey Point Unit 4 (Reference 5). In this event, the pressure rose to 1100 psi, at a temperature of 110 F. Pressurized Water Reactors (PWR's) have accumulated approximately 400 years of plant operation since installation of LTOP protection systems. Because only one event has occurred in 400 years of accumulated PWR plant operation, the event would not be expected to occur during the 40-year life of a PWR nuclear power plant. Hence, according to Appendix A,10 CFR Part 50, the event is not an anticipated operational occurrence and may be considered a faulted condition.

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1 LTOP protection systems are installed to prevent overpressurization of l

the reactor vessel when its materials are least resistant to fracture. '

As a result of PBNP-1 operating procedure, the PBNP-1 reactor vessel is most susceptible to overpressurization between 100 F and 370 F during heat up and 160 F and 350 F during cooldown. The fracture resistance of the P8NP-1 reactor vessel materials decreases with decreasing temperature.

Hence, the PBNP-1 reactor vessel would be most susceptible to fracture from overpressurization when the vessel temperature is 100 F. The licensee's evaluation was performed at a temperature greater than 100 F; therefore, it does not represent a bounding LTOP event for the PBNP-1 ,

l reactor vessel.

To conservatively bound LTOP events for the PBNP-1 reactor vessel, the staff has performed a fracture mechanics analysis for the PBNP-1 reactor vessel in which the postulated event occurred at 100 F and pressurized the vessel to 1500 psi. The analysis was performed using the methodology described in Appendix A of ASME Code Section XI. The staff's evaluation indicates that for a postulated LTOP event, the PBNP-1 flaw satisfies the fracture mechanics criteria in Article IWB-3600 for faulted conciitions.

CONCLUSIONS (1) The evaluation performed by the licensee is acceptable except that the LTOP event analysis did not provide sufficient information to justify considering it a f aulted condition and did not conservatively bound low temperature operation of the PBNP-1 reactor vessel.

(2) Based on the licensee's evaluation and the staff's evaluation of a postulated LTOP event, the flaw satisfies the analytical evaluation criteria in Article IWB-3600. Based on these analyses, the flaw in the safety injection nozzle-to-vessel weld will not grow during the life of the plant to a size that will affect the integrity of the reactor vessel. The reactor vessel is acceptable for the 40 years of service life of the plant.

However, this flaw is conditionally acceptable in accordance with subarticle IWB-3123 of Section XI of the ASME Code and, therefore, requires augmented inservice inspections based on 10 CFR 50.55a(g)(4). The staff will require that the best available techniques and instrumentation be used during the next inservice inspection to dimension and characterize the flaw. Future inspections should be performed with transducers with optimum characteristics at the flaw depth. The staff will require that the licensee provide a report at least 6 months before the scheduled outage describing detailed plans for this inspection. The staff intends to perform a detailed review of the examination results to determine whether the flaw has been accurately sized and growth has not occurred. The staff will consider beam spread correction procedures that have been demonstrated to be technically valid on reflectors of different dimensions and types.

1 Principal Contributor: B. Elliot Date:

hL o 91997

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REFERENCES

1. Bamford, W. H., et al., " Background and Technical Basis for the Handbook on Flaw Evaluation for the Point Beach Units 1 and 2 Reactor Vessels,"

Westinghouse Electric, WCAP-11498, April 1987.

2. Shah, R. C and Kobayashi, A.S., " Stress Intensity Factor for an Elliptical Crack Under Arbitrary Loading," Engineering Fracture Mechanics, ,

Vol. 3, 1981, pp. 71-96. 1

3. Lee, Y. S. and Bamford, W. H., " Stress Intensity Factor Solutions for a Longitudinal Buried Elliptical Flaw in a Cylinder Under Arbitrary Loads,"  ;

presented at ASME Pressure Vessel and Piping Conference, Portland, I Oregon, June 1983. Paper 83-PVP-92.

4. Letter dated May 20, 1980 to S. Burnstein, Wisconsin Electric Power i

Company, from R. A. Clark, NRC.

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5. W. D. Lanning, " Low Temperature Overpressure Event at Turkey Point Unit 4," Case Study Report by Office for Analysis and Evaluation of I Operational Data, NRC, March 1984.

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