ML20237B414

Safety Evaluation Supporting Amend 92 to License DPR-54

ML20237B414
Person / Time
Site:	Rancho Seco
Issue date:	12/07/1987
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20237B399	List: ML20237B396 ML20237B405 ML20237B414
References
NUDOCS 8712160244
Download: ML20237B414 (6)

Text

___ _

~

,inc

)

o,,

UNITED STATES NUCLEAR REGULATORY COMMISSION n

p WASHINGTON, D. C. 20555 7,;

\\...../

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUFFORTING AMEMUMLN1 NO. 92 TO FACILITY OPERATING LICENSE NO. DPR 54 5ACRAMENTO MUNICIPAL UTILITY DISTRICT PANCHO SECO NUCLEAR GENERATING STATION DOCKET NO. 50-312 I

1.0 INTRODUCTION

By letter dated September 22, 1906 (Reference 1), Sacramento Municipal Utility District (SMUD or license) reouested an arrendment to Fecility Operating (RSNGS). License No. DPR-54 for the Rancho Seco Nuclear Generating Station This prcposed amendment would extend the surveillance period of inspection and operability testing of Reactor Vessel Internals Vent Valves (RVVV) to coincioe with the next reactor head removal during the Cycle 8 refueling outage currently scheduled for about May 1989. The proposed change involves Technical Specifications 4.1.2 and I.9.

Based on a review of Reference 1, the staff required SMUD to submit addi-tional information regarding this license amendment request.

Pursuant to this discussion the licensee submitted additional information (Reference

2) regarding mechanical design, materials of construction, and chemistry 1

of the reactor coolant system. A clarification involving cumulative testing intervals (Specification 1.9) was added to Technical Specification page 1-2b by submittal dated November 25, 1987 (Reference 3). The Noverrber l

25, 1987 submittal did not change the application as noticed in the Federal Register on August 26, 1987. On the basis of the information provided by the licensee, the staff perforned an assessment of the safety issues of the subject request.

Pegulatory and Plant Requirements The RSNGS Technical Specifications, Sections 4.1.2 and 1.9 require that the RVVV be demonstrated operable at least once per 18 months with a provision that an extension of 25% (= 4.5 months) may be granted for the 18 month period. Additionally, the total maximum combined interval time for any three consecutive tests may not exceed 3.25 times the 18 month surveillance interval.

In order to meet this Technical Specification inspection interval requirement, the valve operability at RSNGS should have been demonstrated by October 1986.

Licensee's Justification The licensee originally submitted Reference 1 as justification for deferring the testing of the RVVVs. The licensee states therein that the RVVVs were last inspected in April 1985 and by present Technical Specification requirements are again due to be inspected before restart 9712160244 871207 ADOCKOSOOg2 PDR P

l

' l 1

of the plant. The licensee states that after only two months of plant operation the inspection of the RVVVs is ur. warranted. The licensee also

)

states that the deferral of this surveillance would result in some reduction of personnel radiation exposure and that past testing has not

)

revealed any significant abnormalities.

The license also submitted Reference 2 which supplements the information in Reference 1.

This additional submittal includes details on the materia's and dimensions of the RVVV and an assessment of whether the valves sustained any significant corrosion since the last surveillance.

The RVVVs for Rancho Seco are of the same design and manufacture as the Davis-Besse RVVVs including the materials of construction and internal hinge clearences. The licensee has submitted the following information regarding Reactor Coclant System (RCS) chemistry and RVVV material corrosion:

1.

The oxygen levels in the RCS during the extended shutdown period have varied from low values to high values. The high values resulted from partially draining the RCS during which the RVVVs were above the water level.

2.

For most of the current fuel cycle, the PH,0, cl, and F concentrations in the RCS have been within tf,ie specifications except for brief periods when cl and F specifications were exceeded.

3.

Analyses have shown that there have been periods when sulfur levels were exceeded, but the sulfur existed rrostly as sulfate.

4 The general corrosion rate of the RVVV materials is in the range of 0.05 mils per year. The minimum internal RVVV hinge clearance is 3 mils such that the clearances will not close for this rate of corrosion.

5.

Excessive levels of chloride, fluoride, sulfur, and oxygen in the RCS have not affected the operability of the RVVVs. The valve materials are not very susceptible to stress corrosion caused by the presence of chloride and fluoride. The presence of oxygen was conducive to causing the sulfur to exist as sulfate rather than reduced sulfur.

2.0 DISCUSSION AND EVALUATION The internals vent valves are installed in the core support shield to prevent a pressure unbalance which might interfere with core cooling following a postulated inlet pipe rupture. The arrangement consists of vent valve assemblies installed in the cylindrical wall of the internal core support shield.

I

)

Internals vent valves are included in reactor internals to provide a l

direct path to the break for steam venting after a loss-of-coolant accident resulting from a postulated cold-leg rupture. The vent valves 1

.~

L___.__.__._

-3 are required because the arrangen'ent of the reactor coolant system can possibly inhibit the venting of steam generated in the core after the system is depressurized, if significant quantities of coolant remain in l

the reactor coolant pump suction piping at the end of the blowdown period.

Without the venting of the steam, a pressure differential would exist between the core region and the reactor vessel internals inlet annulus region, where emergency core coolant is injected, which would prohibit flow into the core. To eliminate the problem, the vent valves are installed in the reactor internal to provide a flowpath from the region above the core directly to the pipe rupture location. The flowpath provides for pressure equalization and permits emergency coolant water to reflood the core rapidly.

The staff took into consideration the record of the past tests of similar valves, which represented about 420 RVVV inspections and exercises, at B&W facilities. This indicates that RVVVs have demonstrated a high degree of reliability and no failures were found.

Recent information has been submitted by Toledo Edison Company (Reference 4) which indicated that the typical span between RVVV inspection and exercise was IF - 18 months with a maximum test interval (with the exception of TMI-1) of about two years.

In the case of TMI-1, the interval would be about 49 rronths if the testing is deferred until the next refueling outage.

The staff also evaluated the information pertaining to RCS chemistry, the compatibility of the mating materials and their corrosion resistance, and the reactor coolant environment. The parts of interest which could be vulnerable to corrosion are the shaf t, bushing, and the body. These corrponents are respectively constructed of type 431 martensitic stainless steel, Stellite No. 6, and Type 304 austenitic stainless steel. The corrosion rates of these materials in the RCS hot operating conditions have been verified by the staff in professional literature (Reference 5) to be in the range of 0.05 mils / year or less. Since the thickness of the deposit is about three times the rate of corrosion, the expected thickness of the deposited meterial would be 0.15 mils per year. Since the minimum cold ciearance gap dimensions vary from 3 to 60 mils, the gap would not close for this rate of corrosion. For the cold RCS condition, the staff has determined that the general corrosion rate is even less than for hot conditions (varying exponentially with temperature).

The staff has also reviewed the effect of the high concentration of RCS contaminants. Any one or combination of the contaminants (i.e., cl.

F, 0,, and S) can have a significant effect on the corrosion rate, especially on stress corrosion. Particularly, the fact that the oxygen levels were so high at times is potentially most significant.

Even with the valves located above the water surface, the air space above the water surface would be saturated with water vapor and condensation would form on all valve surfaces including the sliding surfaces between the moving parts. However, even under these adverse conditions, the materials of valve construction are such that the corrosion rates would still be low relative to the amount of the clearance gaps.

In addition, the pres.ence

of oxygen in the RCS has contributed to cause most of the sulfur to exist as sulfate rather than reduced sulfur. While corrosion due to recuced sulfur can be significant even to these valve materials, corrosion due to sulfate is low. Therefore, the specific type and amounts of these RCS contaminants which have been experienced at RSNGS should not have produced a significant amcunt of corrosion of RVVV materials.

In sumary, the proposed change would extend the current surveillance period for testing the RVVVs until the next scheduled reactor head removal but not later than May 1989. The staff evaluation is based on review of test results of similar valves at other facilities, a detailed review of the valve construction, and analysis of Rancho Seco specific information concerning RCS chemistry and contaminants. The staff concludes that minimal amounts of corrosion of RVVV materials would occur over the extendeo surveillance period and would not affect the reliability of RVVVs.

Consequently, the one-time extension of the 18 month refueling interval beginning April 1985 to May 1987 (totalling approximately 49 months) is acceptable.

Additionally, the staff concurs with the licensee that the extended surveillance period should be excluded from the computation required by Specification 1.9, which states that the maximum interval for any three consecutive surveillance periods not exceed 3.25 times the 18 month surveillance interval.

In effect, the first interval for the computation required under Specification 1.9 will not commence until the next RVVV surveillance.

3.0 CONTACT WITH STATE OFFICIAL The NRC staff has advised the Chief of the Radiological Health Branch, State Department of 9eath Services, State of California, of the proposed determination of no c!gM ficant hazards consideration. ho comments were j

received.

]

4.0 ENVIRONMENTAL CONSIDERATION

This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.

We have determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no I

significant hazards consideration and there has been no public comment on I

such finding. Accordingly, this amendment meets the eligibility criteria u

for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to D

10 CFR 51.22(b), no environmental impact statement or envircanental d

assessment need be prepared in connection with issuance of the amendment.

1 1

l l

(r.;

.o. f 5

5.0- 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 proposed manner, (2). such activities will be conducted in compliance with the Comission's regulations, and (3) the issuance of the amendment will not be inimical f.:

to common defense and' security or to the health and safety of the public.

PRINCIPAL CONTRIBUTOR: Gary Hanmer Dated:

December 7,1987 1

m. '

q

References:

1.-

Letter from John E. Ward, SMUD, to Frank J. Miraglia, NRC, dated September 22, 1986.

.2.

Letter from John E. Ward, SMUD, to John F. Stolz, NRC, dated February 20,.

1987.

3.

Letter.from Joseph F. Fir 11t, SMUD.'to-Frank J. Miraglia, NRC, dated November 25, 1987.

1 4

. Letter from Joe Williams, Jr., Toledo Edison Company, to John F. Stolz, NRC, dated June 6, 1986.

'5.-

'Uhlig, Herbert, H., " Corrosion and Corrosion Control,* John Wiley and Sons

~

Inc., 2nd Edition, 1971.

_ _ _ _ _ _. - _ _ _ _ _ _ _. _ _ _ _ _ _ _. _ _