ML20207H389

Safety Evaluation Supporting Amend 116 to License NPF-3

ML20207H389
Person / Time
Site:	Davis Besse
Issue date:	08/19/1988
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20207H386	List: ML20207H381 ML20207H389 ML20207H394
References
TAC-66699, NUDOCS 8808260362
Download: ML20207H389 (6)

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S SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.

116 TO FACILITY OPERAT!NG LICENSE NO. NPF-3 TOLEDO EDISON COMPANY, AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 DOCKET NO. 50-346

1.0 INTRODUCTION

By letter dated March 30, 1988 Toledo Edison Company requested approval to revise section 3/4.4.9 and bases section 3/4.4.9 of the Davis-Besse Technical Specifications; both relate to the pressure-temperature limits of the Reactor Coolant System and surveillance capsule withdrawal schedule.

The primary purpose of this request is to revise the existing pressure temperature limits and extend the operation period of the limits up to 10 effective full power years (EFPY). The revised pressure-temperature limits are based on the Babcock and Wilcox analyses (BAW-1882 and BAW-2011) of irradiation data taken from removed surveillance capsule TEl-A. The limits provide pennissible pressure and temperature for three operating conditions -

heatup and core criticality (Fig m 3.4-2), cooldown (Figure 3.4-3), and inservice leak and hydrostatic tests (Figure 3.4-4).

The licensee also requested approval to revise the existing surveillance capsule withdrawal schedule.

2.0 DISCUSSION Part of the NRC's effort to ensure integrity of the reactor vessel is to periodically evaluate the reduction in fracture toughness of the vessel material caused by neutron irradiation embrittlement. The effort consists of three steps.

First, the licensee is required to establish a survei' lance program in accord-ance with Appendix H of 10 CFR Part 50, which requiras periodic withdrawal of surveillance capsules from the reactor vessel. The capsules are instelled in the vessel prior to startup and they should contain test specimens that l

were made from the plate, weld, and heat-affected-zone materials of the reactor oeltline.

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. Secondly, the licensee is required to perform Charpy impact tests, tensile tests, and neutron fluence measurements on the specimens. These tests define the condition of vessel embrittlement at the time of capsule withdrawal in terms of the increase of the reference temperature, RT and upper shelf energy. ThelicenseemustalsopredictthefuturevesNI,embrittlementby calculating the adjusted RT and upper shelf energy at a specific EFPY. The licensee may use either RevNIon 1 or draft Revision 2 of Regulatory Guide 1.99 to calculate the adjusted RT The upper shelf energy is the average energy value for all specimens whosUDIe.st temperature is above the upper end of the transition temperature region. The licensee is required by 10 CFR Part 50 Appendix G, to provide assurance that the adjusted RT will not exceed 200'F and that the upper shelf energy will not be below 50 bib at the end of plant life.

Thirdly, the licensee is required to develop a set of pressure-temper.ture curves based on the adjusted RT of the limiting vessel materials. The curvesmustsatisfytherecommebdmethodsandrequirementsdescribedin 10 CFR Part 50, Appendix G, and Standard Review Plan 5.3.2.

3.0 EVALUATION Surveillance Program The surveillance program in Davis-Besse Unit 1 is part of the Integrated Reactor Vessel Material Surveillance Program sponsored by Bab Jck and Wilcox (Report No.

BAW-1543A), which the staff approved in 1985.

B&W 1ritiated the Integrated Pro-gram as a result of flow-induced vibration that cause cracking of surveillance capsule holder tubes in some B&W reactor vessels.

T! t capsules were ramoved from these nuclear plants, redesigned, and installed in other B&W reactors.

Davis-Besse, Unit I was selected as a host reictor.

The surveillance program in Davis-Besse consists of six surveillance capsules.

Capseles TEl-F, and TEl-B, and TEl-A have been removed at the end of fuel cycles 1, 3, and 4, respectively.

The proposed pressure-temperature limits were determined based on the data taken from Capsule TEl-A, which was positioned inside the reactor vessel between the thermal shield and the vessel wall.

This capsule contained Charpy V-notch test specimens fabricated from two base metals, a weld metal, two heat-affected-zone metals, and a correlation metal.

All test specimens were machined from the 1/4-thickness of the forging material and can be traced to the vessel material.

The licensee proposed to delete the existing capsule withdrawal schedule in section 3/4.4.9 of the Technical Specifications and replace it with the withdrawal schedule recomended for Davis-Besse in the Integrated Program in Bases section 3/4.4.9. The remaining three capsules in Davis-Besse will be removed at the end of the sixth, eighth and eleventh fuel cycles. The staff concludes that this withdrawal schedule for the remaining capsules satisfies the requirements of Appendix H to 10 CFR Part 50 and ASTM E 185-82.

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. UPPER SHELF ENERGY The licensee used methods in Regulatory Guide 1.99, Revision 2, and BAW-1803 (a B&W Owner Group method) to evaluate the reactor vessel end-of-life upper shelf energy (USE). The weld material in the Davis-Besse reactor used a submerged arc weld process with Linde 80 flux. This weld material has high copper content and low unirradiated Charpy USE; therefore, it is susceptible to neutron irradiation damage. The method given in Regulatory Guide 1.99, Rev. 2 estimated a USE of 47 ft-lbs whereas the BAW-1803 method estimated a 67 ft-lbs USE at end-of-life for weld WF-182-1. The BAW-1803 method uses a statistical analysis of all the existing surveillance Linde 80 weld metal; therefore, it estimates a higher USE than Reg. Guide 1.99.

The staff considers that the statistical analysis in BAW-1803 is an acceptable alternative to the R.G. 1.99 Rev. 2 method and that 67 ft-lbs is acceptabN. The licensee showed that none of the vessel raterial will have a USE below 50 ft-lbs at end-of-life at the T/4 wall location. The staff concludes that the USE's of Davis nesse reactor materials satisfy Appendix G of 10 CFR Part 50.

REFERENCE TEMPERATURE The observed herease in reference temperature (RTspecimens was detemined based o in the surveillance at the time of withdrawal and the initial RT The RT isthetesttemhSfaturethat correspondstothe30ft-lbimpaggT.absorptioMnergyintheCharpyv-notch impact tests.

From the Charpy test, the licensee determined that the maximum observed increase in the RT was 175'F and occurred in the middle circumference seamweld,WF-182-1.BasedBRTthe maximum RT shift, this weld metal is selected to be ghe limiting material.

CapsugTTEl-A received a neutron fluence of 1.29E19 n/gem and the inner wall fluance at 10 EFPY was calculated to be 5.63E18 n/cm. The ratio of the neutro i flux density at the capsule location to the maximum calculated neutron flux censity at the inside surface of the vessel wall (lead factor) was about 2.0, Since the pressure-temperature limits are usually needed for the future EFPY's, the licensee needs to predict the adjusted RT dt a could be used in detemining the hNssure future EFPY so that a valid RT temperature limits. The staff ush0Tthe Revision 2 method to evaluate the licensee's predictions of the adjusted RT The predicted RT depends on the neutron fluence at the specific vein 1 wall location, sp$SIfic EFPY, and copper content and nickel content of the vessel mater!al. The neutron fluence at T/4 vessel wall location (T=vesselthickness)and106FPYwasextrapolatedfromtheogservedneutron fluence of capsule TEl-A and was calculated to be 3.46E18 n/cm. The maximum 1

predicted adjusted RT was determined to be 146*F for the limiting weld NDT material,

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, The observed and predicted RT for base metal and heat-aifected-zone metal TheNedictedRT for the base metal exceeds the N

- are compared in Table 1.

observed RT while the predicted RT f5ETthe weld metal is less than the NDT observed RT Thelicenseeusedth5%redictedRT of 147'F to determine the pressurN[e.mperature limits instead of the obseNd value of 175'F. The requirements in subsection 2.1 of Rep. Guide 1.99, Rev. 2 allow the use of the lower of the two RT when the licensee uses previous knowr, surveillance data in the prediction. Nkbe staff finds that the use of 147'F for the RT in the proposed pressure-temperature limits is more reasonable than 175'F bEluse the RT of 175'F corresponds to a neutron fluence at 22 EFPY.

NDT The RT at end-of-life and T/4 location was calculated to be 222'F. Davis-Besse NI a part of the pressurized thermal shock study that has an allowable RT of 300'F, as aescribed in section 50.61 of 10 CFR Part 50. The staff, thE[ fore, finds the end-of-life RT acceptable.

NDT PRESSURE-TEMPERATURE LIMITS The staff used the methods described in Standard Review Plan 5.3.2 to evaluate three proposed pressure-temperature limits for heatup and core criticality, cooldown, and hydrostatic leak test. The SRP 5.3.2 method uses the principle of linear elastic fracture mechanics to calculate the pressure-temperature limits. The basic parameter is the stress intensity factor, which is a function of pressure and thermal stresses and flaws in the vessel material. The stress intensity factor is related to the reactor coolant system temperature and prNIure,the vessel material by an exponential curve. For a given operating RT of a reference stress intensity factor is calculated and via the exponential curve, the system operating temperature could be detennined. Aside from the pressure and thennal stresses, the pressure-temperature limits should also consider the bolt preload stress at reactor closure flange regions as described in Appendix G to 10 CFR Part 50. The staff finds that the proposed pressure-temperature limits in Figures 3.4-2, -3, and -4 are in accord with SRP 5.3.2 and meet the requirements of Appendix G to 10 CFR Part 50, 4.0 FINDINGS l

The staf' finds that the proposed surveillance capsule withdrawal schedule is acceptab'.e based on the requirements of Appendix il to 10 CFR Part 50 and ASTM l

E 185-82.

The staff has concluded that the proposed pressure-temperature limits on the reactor coolant system for heatup, core criticality, cooldown, and hydrostatic test operating conditions are in conformance with requirements of Appendix G to 10 CFR Part 50. The proposed limits are acceptable up to 10 effective full power years.

The staff agrees that the proposed surveillance withdrawal schedule and pressure-temperature limits may be incorporated into the Davis-Besse Technical Specifications, i

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5.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environwntal assessment and finding of no significant impact have been prepared and published in the Federal Register (July 18,1988,53FR27095). Accordingly, based upon the environmental assessment, the Comission has determined that the issuance of this amendment will not have a significant effect on the quality of the human environment.

6.0 CONCLUSION

The staff has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance,that the health and safety of the public will not be endangered by operation in the )roposed manner, and (2) such activities will be conducted in compliance with tle Commission's regulations, and the issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public.

Principal Contributor:

J. Tsao Dated: August 19, 1988 i

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l TABLE 1 i

Adjusted RTun, 9 10 EFPY and T/4 watt' location Neutron Copper Nickel Fluence 9 Observed Staff Pre-Licensee Specimen (1)

Content Content InnerSur{ ace RT diction Prediction n/cm WET

'F

'F Base Metal BCC-241 0.02 0.81 5.63E18 28 78 78 Base Metal AKJ-233 0.04 0.77 5.63E18 2

56 56 Heat Affected 0.02 0.81 5.63E18 34 78 Not Zone BCC-241 Available Weld Metal WF-182-1 0.24 0.63 5.63E18 175 146(2) 147(2) i Notes (1) All surveillance specimens can be traced to the vessel material.

(2) Use capsule fluences and observed RT shifts from capsules TEl-A, 1

TEl-B and TEl-F to calculate the chhIstry factor per subsectioli 2.1 of Reg. Guide 1.99 Rev. 2.

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