ML20235J384

Safety Evaluation Concluding That Continued Operation of Facility Not Affected by Steam Generator Tube Failures Experienced by Advanced gas-cooled Reactors

ML20235J384
Person / Time
Site:	Fort Saint Vrain
Issue date:	02/13/1989
From:	Office of Nuclear Reactor Regulation
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ML20235J374	List: ML20235J372 ML20235J384
References
TAC-65991, NUDOCS 8902240260
Download: ML20235J384 (4)

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. ,f l'o UNITED STATES a

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SAFETY EVALUATION BY THE_0FFICE OF NUCLEAR REACTOR _ REGULATION RELATING T0_ FORT ST..VRAlftSTEAM_G_E_NERIC BIMETALLIC WELDS FORT ST. VRAIN NUCLEAR GENERATING STATION PUBLIC SERVICE COMPANY OF COLORADO I

DOCKET NO. E0-267 1.0 _

INTRODUCTION Or March 17, 1986, a steam generator (btiler) tube failed at one cf the l Advanced Ges-Cooled Reactors (AGR's) at Hartlepool in the United Kingcom. i This event was of imn.9diate interest to the staff because of the sirnilarity of the hartlepool AGR steam genetutcr design to that at Fort St. Vrain (FSV).

The FSV steam gErieratorf each Consist cf 6 modules. The gas coolar.t flows down thrcugh tha modules, first over the reheater tubes, then over the main steam tubes. The tube bundles are helically wound. The main steam l secticn tubes are febricated frcn n.ultiple rnaterials, each tailored for the specific operating terrperature rar.ge. The stearr lines exit through penetrations in the bottom of the reactor vessel.

The AGR steam generators tre sir..ilar in several major respects. The pas coolant also flows down through the steam generator, flowirg first over a separate reheater tubc tundle. The tube bundles are also belically wound.

The nain steam secticr. tubes a e fabricated from multiple materials. One majcr difference is that the stearn lines are connected externally through the top of the reactor vessel.

Table 1 (attached) summarizes the temperatures and pressures of the reactor (primary) and secondary coolant. Exce severe conditions in the FSV reactor (primary)pt for the coolant, slightlycoolant secondary more ccrditions are very similar.

1 Subsequently, proprietary information was received from the OECD Nuclear Energy Agency, Incident Reporting System (IRS). The specific IRS reports are numbers 07.92.10 and 07.92.20. These reports concerned superheater '

transition joint cracking, found both at Hartlepool and Heysham 1. The staff's evaluation is based on this information, and knowledge about the construction of the FSV steam generators.

8902240260 890213 PDR ADOCK 05000267 p PDC

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2.0 EVALUATION 2.1 Mete 11urgy The staff's evaluation focused on the welding techniques used to icin dissimilar tubing materials in the AGR and FSV steam generators.

The staff has reviewed the welding procedures for making the bimetallic transition welds in the steam generators at the Hartlepool and Heysham AGRs and at the FSV reactor. Inspection of the affected area in the AGRs showed that cracks were formed in the biretellic transition weld in a brittle martensitic zone between the 2.25% Cr-1% Mo and Inconel materials.

The procedure used to fabricate the Hartlepool and Heysham bimetallic welds involved butterinc the 2.25% Cr-1% Mo pipe with Inconel 182 and stress relieving (700 : 15*C. for 3 h.) the buttered welds in the shop. After pressure testing, the buttered pipe was transferred to the site and welded to the Type 316 stainless steel steam generator nozzle, using Inconel 182 filler metal. Neither a preheat nor a stress relief heat treatment was carried out on the welds on site. This procedure would account for the brittle martensitic zone observed at the jurction of the 2.25% Cr-1% Mo and Inccnel materials.

The procedure used to fabricate the FSV steam generator bimetallic welds was in compliance to Sterns-Roger Welding Procedure Specifications WS-5, Specification No. 205. Welder qualification tests were made and documented in accordance with the requirements of Sections III and IX of the AShE Eoiler and Pressure Vessel Code for Class A pressure vessels and certified by the Hartford Steam Boiler Inspection and Insurance Company.

The Sterns-Roger specifications required that the weld be slowly cooled from the welding temperature and that the 2.25% Cr-1% Mo (P-5) material be preheated prior to welding to a minimum of 300'F. The filler material was to ccnform to ASME material specification SB-304, F-43 (ER NiCr-3).

Preheat was not specified for the P-45 (33%Ni 21%Cr 43%Fe) material. This procedurc ensured that a brittle martensitic zone would not form on the 2.251 Cr-If Mo surface during welding of the steam generators at the Fort St. Vrain reactor. Theanticipatedlifeofthegimetallicweldsprepared to the Sterns-Roger specifications exceed I x 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

2.2 System Safety The staff also examined the metallurgical findings relative to the actual Fort St. Vrain operational experience. As of December 31, 1988, Fort $t.

Vrain had accumulated 37,245 hours0.00284 days <br />0.0681 hours <br />4.050926e-4 weeks <br />9.32225e-5 months <br /> of critical reactor operation and 25,073 hours8.449074e-4 days <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br /> of generator on-line operation. The latter figure more closely represents actual hours of operation of the steam generators at

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, rated temperature and pressure conditions. At the present time, the steam generators are operating well within their design lifetime. A more extensive consideration of this issue would be required if anticipated-plant operation actually approached the estimated bimetallic weld lifetime given above. .  ;

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3.0 C0NCLUSIONS __

The staff finds that the steam generatur tube failures experienced by the AGR's. in the United Kingden. do not affect near term operation of Fort St.

Vrain. The estimated life of Fort St. Vrain's bimetallic welds greatly exceeds actual service life. Therefore, the staff concludes that continued operaticr, of Fort St. Vrain is r.ot affected by the steam generator tute failures experienced by the AGR's.

4.0 REFERENCES

1. The mechanical design and validation of the helical tube boilers for Hartlepool and Heysham AGR stations. No. 32, " Gas-Cooled reactors today" ENES, Londor, 1983.

2. AGR boiler materials selection considerations Nc. 109, " Gas-cooled reactors today," BNES, London, 1983.

Date: February 13, 1989  ;

Principal Contributors: F. Litton, EMTB K. Peitner, PDIV

Attachment:

Table 1 i

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TABLE 1 -1 C0tlPARIS0N OF TEMPERATURES AND PRESSURES FORT ST. VRAIN STEAM GENERATORS VS HARTLEPOOL AGR BOILERS .

AGR Gas - Reactor (Primary) Coolant FSV (Hartlepool)

Maximum Temperature, 'F 1427 1189 Maximum Pressure, psia 700 573 Steam - Secondary Coolant Main Steam Maximum Temperature, 'F 1000 1009 Maximum Pressure, psia 2400 2484 Reheat System Maximum Temperature, 'F 1000 2002 Maximum-Pressure, psia' 600 602 m - - - - - _ - - . _ _ . _ - _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ - - - . - - . - - - - - - - - - - - - - -.