ML20154F788

Forwards Safety Evaluation Re C Reactor Coolant Pump (RCP) Stud Integrity.Evaluation Supports Continued Pump Operation Until Nov 1988.Util to Submit Action Plan Addressing C RCP Leakage & Stud Integrity,W/Acceptance Criteria,By 880615

ML20154F788
Person / Time
Site:	Rancho Seco
Issue date:	05/17/1988
From:	Andognini G SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	Crutchfield D Office of Nuclear Reactor Regulation
References
GCA-88-335, NUDOCS 8805240079
Download: ML20154F788 (6)

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SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 S Street, P.O. Box 15830, Sacramento CA 95852-1830,1916) 452 3211 l AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA l@Y 17 1939' GCA 88-335 U. S. Nuclear Regulatory Commission Attn: Dennis Crutchfield Director, Division of Reactor Projects III, IV, and V and Special Projects 11555 Rockville Pike Rockville, MD 20852 Docket 50-312 Rancho Seco Nuclear Generating Station License No. DPR-54 REACTOR COOLANT PUMP STUD INTEGRITY

Dear Mr. Crutchfield:

Attached is the Safety Evaluation addressing Reactor Coolant Pump (RCP) stud integrity. The District performed this Safety Evaluation to support the decision to continue operation of the "C" RCP.

The Safety Evaluation concludes that the existing RCP studs are acceptable until November 1988 and continued operation does not introduce an Unreviewed Safety Question. The Safety Evaluation is supported by information provided from the pump manufacturer (Bingham-Willamette Company) in a letter dated May 9, 1984, and a report from the B&W Owners Group Materials Committee (BAW-1892P, Steam Generator Manway and Reactor Coolant Pump Bolted Closure Evaluation). The B&W Owners Group report assesses the integrity of RCPs with several degraded bolts. Specifically, the report uses Rancho Seco's RCPs for modeling purposes.

In performing the Safety Evaluation, Engineering determined that boric acid corrosion will not degrade the affected RCP studs to the minimum acceptable diameter until approximately November 1988. However, Rancho Seco will shut down for inspections and maintenance on or before July 15, 1988. During this planned out-age, Engineering will inspect "C" RCP and evaluate the physical condition of the affected bolts. Furthermore, the District will submit, by June 15, 1988, an action plan addressing the "C" RCP leakage and RCP stud integrity. This plan will contain accep-tance criteria for determining the continued use of the subject "C" RCP studs.

8805240079 880517 PDR i ADOCK 05000312 A001 P ppg p l

t RANCHO SECO NUCLEAR GENERATING STATION O 14440 Twin Cities Road, Herald, CA 95638 9799;(209) 333 2935

NAY 17 1988 Dennis Crutchfield GCA 88-335 Please contact me if you have any questions. Members of your staff requiring additional information or clarification may con-tact Steve Crunk at (209) 333-2935, extension 4913.

Sincerely, G. Carl Andogn di Chief Executiv Officer, Nuclear Attachment cc w/atch:

A. D'Angelo, NRC, Rancho Seco G. Kalman, NRC, Rockville (2)

J. B. Martin, NRC, Walnut Creek l

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SAFETY EVALUATION OF REACTOR COOLANT PUMP STUD INTEGRITY PDQ NO. 88-1007 PAGE 1 OF 4 DESCRIPTION:

The interim disposition for PDQ No. 88-1007 is to "accept-as-is" (i.e., not replace) studs number 9 and 10 on the "C" Reactor Coolant Pump. These studs have shown evidence of corrosion re-sulting from a borated water leak from the "C" Reactor Coolant Pump. The interim disposition is valid until the next scheduled outage (approximately July 15, 1988), at which time an additional inspection and/or measurement shall be accomplished and a final disposition made by revision to the subject PDQ.

REASON FOR CHANGES:

Purpose Gasket leakage from the "C" Reactor Coolant Pump (RCP) is causing a build-up of boric acid crystals around pump studs number 9 and 10, which could potentially cause corrosion of the studs to less than the acceptable minimum diameter.

Pump studs 9 and 10 were measured to evaluate potential corrosion loss as a result of crystal build-up. The latest measurerent of the studs (May 9, 1988) revealed the following diameters:

Stud # 9 -- 3.58 inches Stud # 10 -- 3.70 inches The nominal diameter of the RCP studs is 3.8 inches.

EVALUATION AND BASIS FOR SAFETY ANALYSIS:

Systems. Subsystems, Components Affected l

l The RCP studs in question are associated with the "C" RCP, which l is a part of the Reactor Coolant System (RCS). I Safety Functions of Affected Systems / Components The purpose of the Reactor Coolant System is to contain and cir-culate the reactor coolant at pressures and flow rates necessary to transfer the heat generated in the reactor core to the secon-dary fluid in the once-through steam generators. The RCS also acts as a barrier to prevent the release to the atmosphere of any fission products that could possibly escape through the fuel as-sembly cladding into the reactor coolant. (See USAR section 4.2 l

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SAFETY EVALUATION OF REACTOR COOLANT PUMP STUD INTEGRITY PDQ NO. 88-1007 PAGE 2 OF 4

- System Description and Operation.) The Reactor Coolant System is operated in accordance with Technical Specification 3.1 -

Reactor Coolant System.

The reactor coolant pumps provide the motive force for the forced circulation of reactor coolant within the RCS. Each pump has a capacity of 92,400 gpm at the design pressure and temperature of 2500 psig and 650 F, respectively. (See USAR section 4.2.2 Reactor Coolant Pumps.) Reactor coolant pump operation is per-formed as directed in Technical Specification 3.1.1.1 - Reactor Coolant Pumps.

The RCS Leak Detection System has the safety function of preven-ting Loss of Coolant Accidents by early detection and warning.

RCS leak detection is based on three separate and diverse methods: A) Containment sump level, B) Containment radioactiv-ity, and C) RCS Inventory calculation (USAR section 4.2.3.7). As amplified by the Bases of Technical Specification 3.1.6, "Every reasonable effort will be made to reduce reactor coolant leakage including evaporative losses (which may be on the order of 0.5 gpm) to the lowest possible rate and at least below 1 gpm in or-der to prevent a large leak from masking the presence of a smal-ler leak." The failure of the RCS Leak Detection System to per-form its safety function when needed (i.e., a LOCA which could have been prevented by early warning and action) is mitigated by the proper functioning of the Emergency Core Cooling System (see USAR section 14.2.2.5).

Effects on Safety Functions / Analysis of Effects on Safety Functions Previous measurements of the RCP studs were taken on 6/5/87. At that time, "C" RCP studs 9 and 10 diameters were measured as:

Stud # 9 -- 3.709 inches Stud # 10 -- 3.941 inches.

When the 1987 measurements are compared to the latest measure-ments, the corresponding differences (0.129 inches and 0.141 inches, respectively) indicate that stud corrosion due to the ef-fects of boric acid is less than 200 mils. This degradation is attributed to occurring during the past two months only, since seal leakage on the "C" RCP was first identified approximately March 7, 1988.

The RCP manufacturer (Bingham-Willamette Company) has stated that the minimum acceptable stud diameter is 3.20 inches (letter dated May 9, 1984). The letter states that the manufacturer's calcula-tions performed for Duke Power's Oconee Nuclear Station conclude that "With 19 studs at a diameter of 3.43 inches, the minimum ac-ceptable diameter of the 20th stud would be 3.20 inches..." Only

SAFETY EVALUATION OF REACTOR COOLANT PUMP STUD INTEGRITY PDQ NO. 88-1007 PAGE 3 OF 4 one other stud on "C" RCP has a diameter of less than 3.43 inches (stud # 3 - 3.415 inches); the remaining stud measurements are all greater than 3.8 inches (with the exception of the two studs in question). This indicates that the continued use of studs # 9 and 10 is acceptable, based on the manufacturer's recommendation of minimum stud diameter.

The rates of degradation for the studs have been calculated (based on a conservative "start time" of March 7, 1988) as fol-lows:

Stud # 9 -- 0.065 in/ month Stud # 10 -- 0.071 in/ month By extrapolating the calculated rate of degradation, stud #9 (the most eroded stud) would reach the minimum acceptable diameter of 3.20 inches in November 1988. This indicates that the continued use of studs # 9 and 10 until approximately July 15, 1988 is ac-ceptable, based on the manufacturer's recommendation of minimum stud diameter.

The Babcock & Wilcox (B&W) Owner's Group contracted B&W to anal-yze stud failure on the RCP main closure due to observed corro-sion wastage in pump closure studs in several plants. B&W chose the Rancho Seco coolant pumps for modeling purposes. The resul-ting analysis showed that a loss of seven studs would not lead to catastrophic failure of the pump closure. Total failure of five studs would result in a minimum calculated leakage of 4.8 gpm, well above the detectable leakage value, and above the Technical Specification acceptable value. The maximum leakage which would be encountered subsequent to the total failure of five studs is 100 gpm, within the capability of the makeup pump (rated at 300 gpm). (Reference BAW-1892P, May 1986, Steam Generator Manway and Reactor Coolant Pump Bolted Closure Evaluation.)

Summary The interim disposition for PDQ No. 88-1007 is to "accept-as-is" (i.e., not replace) studs number 9 and 10 on the "C" Reactor Coolant Pump. These studs have shown evidence of corrosion re-sulting from a borated water leak from the "C" Reactor Coolant Pump. Gasket leakage from the "C" Reactor Coolant Pump (RCP) is causing a build-up of boric acid crystals around the pump studs, which could potentially cause corrosion of the studs to lass than the minimum acceptable diameter.

The RCP studs in question are associated with the "C" RCP, which is a part of the Reactor Coolant System (RCS).

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i SAFETY EVALUATION OF REACTOR COOLANT PUMP STUD INTEGRITY <

PDQ NO. 88-1007 PAGE 4 OF 4 Manufacturer's recommendations limit the minimum acceptable stud J diameter to 3.2 inches. This minimum diameter is acceptable even if the remaining studs have a diameter of only 3.43 inches. The other studs on "C" RCP have a measured diameter of greater than 3.8 inches (with the exception of studs # 9, 10, and one other).

This indicates that the continued use of studs # 9 and 10 is con- '

servative and acceptable based on the manufacturer's recommenda-tion of minimum stud diameter. ,

By extrapolating the calculated rate of degradation, stud #9 would reach the minimum acceptable diameter of 3.20 inches in November 1988. This indicates that the continued use of studs #

9 and 10 until approximately July 15, 1988 is conservative and acceptable based on the manufacturer's recommendation of minimum stud diameter.

A B&W analysis performed for the B&W Owner's Group showed that a loss of seven studs would not result in catastrophic failure of the pump closure. Total failure of five studs would result in a maximum calculated leakage of 100 gpm, well within the capability of the makeup system. Only two studs on the "C" RCP have indica-tion of wastage due to boric acid corrosion. This indicates that the continued use of studs # 9 and 10 is conservative and accept-able.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously eval-uated in the safety analysis report will not be increased because no failure is calculated or anticipated (i.e., until approximate-ly July 15, 1988), based upon criteria supplied by the pump manu-facturer. In spite of the level of assurance of the foregoing, if a failure should occur, the resulting leakage is within the capability of the makeup system.

The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report will not be created because a total failure of five RCP studs .

would be the equivalent of a small-break LOCA, analyzed in USAR  !

sections 14.2.2.5.4.C., Small Leak Analysis, and 14.2.2.5.6, Sup-plemental Small Break Analysis.

The margin of safety as defined in the basis for any Technical Spocification is not reduced because any postulated leakage fol-loiting total failure of the studs in question will be within the j ca? ability of the makeup pumps, discussed in the Bases for Tech- I ni:al Specification 3.2 - High Pressure Injection, Chemical Addi-tion, and Low Temperature Overpressure Protection Systems.

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