ML20247R467

Application for Amend to License NPF-3,revising Tech Spec 3/4.4.10 Re 18-month Surveillance Interval for Insp & Operability Testing of Reactor Vessel Internals Vent Valves

ML20247R467
Person / Time
Site:	Davis Besse
Issue date:	08/01/1989
From:	Shelton D TOLEDO EDISON CO.
To:
Shared Package
ML20247R463	List: ML20247R460 ML20247R467 ML20247R470
References
1674, NUDOCS 8908080013
Download: ML20247R467 (8)

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Docket number.50-346'

,, Licensa' Furabar!NPF-3.:

s. .

', Szrial Nurbari16?4 Enclosure v Page 1 1

APPLICATION */OR AMENDMENT-TO FACILITY OPERATING LICENSE NUMBER NPF 3.

FOR ---

DAVIS-BESSE NUCLEAR POWER. STATION l

UNIT NUMBER 1 l

i Attached are requested changes to the' Davis-Besse Nuclear. Power Station, Unit' Number 1 Facility Operating' License Number NPF-3. ~Also included are the Technical Description and Significant Hazards' Consideration.

The proposed changes'(submitted under cover letter Serial Number 1674) concern j Technical Specification Surveillance Requirement 4.4.10.1.b.

l-By:

D. C.*Shelton, Vice PresideM , Nuclear

\

Sworn and subscribed before me this 1st day of August, 1989.

et- b ,k ~

Notary Public, State of Ohio t.AURIE A. HINKLE N:2v Public. State of Ohio '

My Ommission Expires May 15.1991 8908080013 890601 PDR ADOCK 05000346 P- PDC .

. . , Docket Number 50-346

, License Number NPF-3 Serial Number 1674 Enclosure Page 2 The following information is provided to support issuance of the requested changes to the Davis-Besse Nuclear Power Station, Unit Number 1 Operating License Number NPF-3, Appendix A, Technical Specifications.

A. Time Required to Implement: This change vill be implemented within 45 days following NRC issuance of the License Amendment.

B. Reason for Change: (LAR Number 88-0007/DCR Number 89-0063) Revise the 18-month surveillance interval for inspection and operability testing of Reactor Vessel Internals Vent Valves to eliminate unplanned, mid-cycle shutdowns.

C. Technical

Description:

See attached Technical Description (Attachment Number 1). l D. Significant Hazards Consideration: See attached Significant Hazards Consideration (Attachment !! umber 2).

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. . , Docket Number 50-346 Licensa Number NPF-3 Serial Number 1674 Attachment 1 Page 1 TECHNICAL DESCRIPTION TITLE Revise the 18-month Surveillance Interval for Inspection and Operability Testing of Reactor Vessel Internals Vent Valves DESCRIPTION 0F PROPOSED CHANGE In 1975, the NRC revised 10 CFR 50.55a to require an " Inservice Inspection" of various safety related components, including pumps and valves, to be performed in accordance with the ASME Boiler and Pressure. Vessel Code,Section XI, "to the extent practical vithin the limitations of design,~ geometry, and materials of construction." The L' avis-Besse Technical Specification Section 4.4.10.1.b' requires that the Reactor Vessel Internals Vent Valves (RVVVs) be demonstrated ,

operable at least once per 18 months with a provision that an extension of 25%1 i (4.5 months; may be granted for the 18-month period. '

The purpose of this Technical Description is to address proposed changes to '

the Technical Specification Surveillance Requirement 4.4.10.1.b 18-month .  !

surveillance interval described in Appendix A, Technical Specifications, of l the Davis-Besse Nuclear Power Station (DBNPS) Operating License. This {

proposed amendment vill change the surveillance period for inspection and i operability testing of the RVVVs from "at least once per 18 months, during shutdown" to "during shutdova for refueling". This vill allow the performance of the surveillance testinS, which requires removal of the reactor vessel head, to correspond to si.aeduled refueling outages. This change vill also i

delete the applicability of Technical Specification 4.0.2 from Surveillance  !

Requirement 4.4.10.1.b and delete an outdated, Cycle Five-specific footnote. I SYSTEM AND COMPONENTS AFFECTED Reactor Vessel Internals Vent Valves (RVVVs)

SAFETY FUNCTION OF AFFECTED SYSTEMS AND COMEONFHTS Technical Specification 4.0.2 provides allowable tolerances for performing surveillance activities while ensuring that'the reliability-associated with the surveillance activity is adequate when deviations'from the nominal specified interval occur. Specification 4.0.2 provides additional margin necessary for operational flexibility because of scheduling and performance considerations. Specification 4.0.2.a provides a maximum allowable extension of 25 percent for each inspection interval. This provides a maxinum 4.5 calendar month extension for 18-month Surveillance Requirements, resulting in a maximum single interval length of 22.5 calendar months. - Specification 4.0.2.b requires that the total combined interval for'any three' consecutive tests not exceed 3.25 times the specified surveillance interval. This allows a maximum of 58.5 calendar months for three consecutive test intervals.

. Docket Number 50-346 Licensa Numbar NPF-3 Serial Number 1674 Attachment 1 Page 2 The RVVVs are large, swing check valves mounted vertically between the inlet and outlet sides of the core support shield. The core support shield directs cold leg (inlet) flow devnvard into the annular space just inside the vessel

]1 and contains core outlet flow in the central portion, directing it upward to the hot leg nozzles. The vent valve assemblies are installed so they can-swing outward into the cold leg water space should pressure on the outlet side of the core exceed inlet pressure. Under normal operating conditions, the vent valves are closed, J

J The RVVVs are discussed in Section 4.2.2.2, Core Support Assemblies,-and l Section 4.2.2.3, Evaluation of Internals Vent Valve, of the Updated Safety Analysis Report (USAR).- The RVVV materials were selected on the' basis of their corrosion resistance, surface hardness, anti-galling characteristics, and compatibility with mating materials in the reactor caolant environment.

The valve disc, hinge shaf t, thaf t journals (bushings), disc journal ,

receptacles, and valve body journal receptacles have been designed to l vithstand, without failure, the internal and external differential pressure loadings resulting from a loss-of-coolant accident. These valve materials are non-destructively tested and accepted in accordance with the ASME Code III 1 requirements for Class A vessels as a reference quality level. The valve materials are listed in USAR Tables 4.2-3 and 4.2-4. Table 4.2-3 also I includes the vent valve shaft and bushing clearances.

l Design criteria for these valves included (1) functional integrity, (2) ,

I structural integrity, (3) individual part-capture capability, (4) functional reliability, (5) structural reliability, and (6) leak integrity through the design life. i The RVVV hinge assembly provides eight loose totational clearances and two end-clearances to minimize any possibility of impairment as disc-free motion in service. In the event that one rotational clearance s%6uld bind in service, seven loose rotational clearances would remain to allow-unhampered disc-tree motion. In the vorst case, at least three clearances must bind or seize solidly to avversely affect the valve disc-free motion. '

In addition, the valve dise hinge loose clearances permit disc.self-alignment so that the external differential pressure adjusts the disc seal face to the valve body seal face. This feature minimizes the possibility of increased leakage and pressure-induced deflection loadings on the hinge parts- ,

in-service.

The internals vent valves are installed in the core support shield to prevent a pressure unbalance which might delay or interfere with emergency core cooling following a postulated inlet pipe rupture. The arrangement consists of four 14-inch inside diameter vent valve assemblies installed in the cylindrical vall of the internal core support shield. The internals vent valves provide a direct path to vent steam in the upper plenum through the break following a postulated cold-leg rupture. The vent valves are required because the arrangement of the RCS could delay the venting of steam generated in the core after the system is depressurized, if significant quantities of '

coolant remain in the reactor inlet piping at the end of the blowdown period.'

Without venting of the steam, the pressure in and above the core region could be greater than the pressure in the reactor vessel inlet annulus where

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. Docket Numb 2r 50-346 Lictnsa Numbar NPF-3 Serial Number 1674 .j Attachment 1 I Page 3 L

emergency coolant is injected. This pressure differential could retard flov-into the core. The vent valves provide a flovpath.from the region above the core directly to the pipe rupture location. This flovpath allows the pressures to equalize and permits emergency coolant water to reflood the core rapidly. 3 EFFECTS ON SAFETY Surveillance testing of each of the four RVVVs at DBNPS has been successfully completed nine times, since 1977, resulting in a total of 36 exercises of the valves. The results of the last inspection after an extended surveillance interval of 42 calendar months demonstrate. satisfactory operability. The trend of this data is consistent with that of the other B&W operating reactors.

A survey was performed by B&W on the RVVVs and the surveillance tests performed on them at operating B&W 177 fuel assembly plants. The information gathered on the RVVV surveillance testing is summarized in the table below:

Te::h Spec. Number of Extended 1986 Test Times Number of Test ,

Plant Cycle Interval Tested Failures. Interval j i

Oconee 1 9 Every Refueling 8 0 <22 mo. )

Oconee 2 8 Every Refueling 7 0 <22 mo.  ;

Oconee 3 9 Every Refueling 8 0 <22 mo. )

ANO-1 7 Every Refueling 6 0 23 mo. j Rancho Seco 7 Every Refueling 6 0 25 mo.

• Crystal  !

River 3 7 24 Months 7 0 <29 mo. l THI-l 5 Every Refueling 5 0 37 mo.  ;

• Davis-Besse 6 18 Months 9 0 42 mo.

• Updated data for currer cycle in 1988.

The data shown above reprasents approximately 400 RVVV inspections and exercises, without a failure, at eight operating plants over the past ten years. This data demonstrates that the RVVVs have exhibited a high degree l

of reliability with no observable degradation in valve operability with reactor age. The typical time between RVVV inspections and exercises is 12-18 calendar months, with a maximum test interval of approximately two years for all plants other than THI-l and Davis-Besse. As noted in the table above, the majority of the plants surveyed defined the test interval as described in i their Technical Specifications to be "Every Refueling", exclusive of a fixed interval length. The definition "Every Refueling" is consistent with Davis-Besse's proposal of "during shutdown for refueling." As summarized by the survey performed by B&V, no failures of the RVVVs have occurred. Problems associated with the RVVVs are discussed in the following paragraphs. However,  !

these problems did not render the RVVVs incapable of proper operation.  !

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- Dockat Number 50-346 1

, Licensa Number NPF-3  !

Serial Number 1674 Attachment 1 Page 4  ;

1 The only degradation of these valves was discovered in November 1978 at two j other B&W operating reactors. At that time, vent valve jackscrew locking - i mechanism wear was discovered in the RVVVs adjacent to the reactor vessel j outlet nozzle. This vear is not expected to occur at Davis-Besse because, ,

unlike the'other B&V 177 plants, Davis-Besse has only four reactor vessel vent I valves instead of eight and none of the four are located near.the outlet nozzles. The only problem encountered at Davis-Besse occurred in 1973, prior to operation, and involved seizing of one jackscrew, as described in Davis-Besse's USAR Section 4.2.2.3, " Evaluation of Internals Vent Valve."

This was attributable to an excessive thickness of " Electrolyze" which spalled off the screw threads. This problem was corrected and no further jackscrev j problems have occurred or are anticipated on the basis that the surfaces are i now separated by a lov friction " Electrolyze". RCS chemistry, material- l compatibility and corrosion revistance, and the reactor coolant environment are also important in ensuring acceptable performance of the RVVVs. These aspects are discussed below. The chemistry of the RCS vater is controlled in accordance with Technical Specification requirements to minimize corrosion and  ;

material activations and to assure the reliability of reactor and steam 1 generator equipment. Reliability of these valves is expected to be' l maintained. l The RVVV parts vulnerable to corrosion are the shaft, bushing, and the body. I These components are constructed of Type 431 martensitic stainless steel, i stellite Number 6, and Type 304 austentic stainless steel, respectively. l Available date for the RCS hot operating conditions indicate that the general corrosion rates of these materials, are in the range of 0.05 mils / year or less. This information was also verified independently by the NRC staff in scientific literature (Uhlig, Herbert H., " Corrosion and Corrosion Control",

John Viley and Sons Inc., 2nd Edition, 1971). Since the accumulation of the corrosion deposit is about three times the corrosion rate, the expected ,

thickness of the deposited materials would be 0.15 mils per year. The minimum  !

clearance gap dimensions could vary from 3 to 60 mils, therefore the gap would not close and hinder the operation of the valve during the period of time until the next test of the RVVVs.

There is information available from both TMI-1 and Davis-Besse to indicate ,

that the RVVVs will remain operable even for unusually long intervals between '

tests. TMI-1 experienced one test interval of 37 months and Davis-Besse experienced an interval of 42 months between tests. In both cases, the valves remained operable. Additionally, using the corrosion rates and clearance gap  :

measurements discussed in the previous paragraph, corrosion would not close the gap and hinder valve operation even during such extended intervals.

Additional support for this change is provided in the NRC Safety Evaluation Report (dated Nosember 19, 1975, from A. Schwencer, NRC, to K. E. Suhrke, B&V), which relieved B&V plants of a five percent flow penalty on operational

and accident transient analyses. The conclusion of this Safety Evaluation Report states "the NRC staff vill require testing to be conducted each refueling outage to confirm that no vent valve is stuck in an open position and that each valve continues to exhibit complete freedom of movement." It is evident from this statement that the surveillance requirement was intended to  !

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Docket Number 50-346 l License Number NPF-3 i Serial Number 1674 At'achment 1 ,

Page 5 be performed during rer;ularly scheduled refueling outages. Further, past operating experience r.t Davis-Besse and other B&V plants has shown the RVVVs are highly reliable and that increasing the surveillance interval to

" refueling" rather than 18 months is justified.

The proposed amendment identifies that the provisions of Specification 4.0.2 vould not apply to this surveillance. There is no documented or expressed basis for either the 18-month surveillance interval or for the tolerance values allowed by Specification 4.0.2, and experience has shown that surveillance is not required as frequently as specifled by these conditions in order to reliably ensure RVVV operability. Therefore, with this change, Specification 4.0.2 applicability is no longer necessary.

This proposed change to the surveillance interval vill eliminate the necessity i for mid-cycle shutdown to perform this surveillance. This change would also reduce the need.for unnecessary handling of Reactor Vessel internals.

Unnecessary handling of Reactor Vessel internals increases the probability of their damage. Furthermore, handling of Reactor Vessel internals solely for the performance of this surveillance vill be contrary to the ALARA philosophy for radiation exposure.

In addition, this license amendmenc application proposes the deletion of an outdated footnote. The footnote allowed a one time extension of the RVVVs' Surveillance Requirement which was to be performed no later than the fifth refueling outage. Since this requirement has been complied with and the time frame has expired, the footnote is no longer applicable. This is an administrative change.

UNREVIEVED SAFETY OUESTION CONCLUSIONS The following conclusions are provided as a result of Toledo Edison's safety review and evaluation of the proposed changes to the Davis-Besse Nuclear Power Station, Unit Number 1 Operating License, Appendix A, Technical Specifications.

The proposed action would not increase the probability of an accident previously evaluated in the USAR because the RVVVs are not related to the probability of any accident analyzed in the Accident Analysis (USAR Chapter 15). Additionally, the criterion of Technical Specification 4.0.2 was not considered in the evaluation of the probability of events analyzed in the USAR. Therefore, no increase in the probability of an accident vill occur (10CFR50.59(a)(2)(i)).

The proposed action veuld not increase the consequences of an accident previously evaluated in the USAR because tha proposed inspection and operability test interval for the RVVVs would still be frequent enough to ensure proper operation, as supported by past performance and industry experience. Operating experience with these valves in the industry and at Davis-Besse has shown that they will remain operable for periods far greater than 18 months. Therefore, no accident consequences would be affected, and i the accident response vill be as assumed in the USAR (10CFR50.59(a)(2)(i)).

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Page 6 i The proposed action would not increase the probability of a malfunction of ,

equipment important to safety. Operating experience with these valves in the j industry and at Davis-Besse has shown that they will remain operable for a periods far greater than 18 calendar months. Testing performed at " refueling" )

would be conducted frequently enough to ensure there is no increase in the 1 probability of a malfunction of these valves, as supported by past performance L and industry experience, even for unusually long periods between tests. No other equipment would be affected (10CFR50.59(a)(2)(i)).

i The proposed action vould not increase the consequences of a malfunction of ]

equipment important to safety because the proposed amendment does not involve  !

a modification to any system or a change in operation of the existing I system (s). Furthermore, it vill not prevent any system from functioning as assumed in the USAR (10CFR50.59(a)(2)(i)). 1 The proposed action would not create a possibility for an accident of a different type than any evaluated previously in the USAR because this proposed amendment does not add or modify any existing equipment. The inspection and i operability test interval for the RVVVs vill still be frequent enough to  !

ensure proper operation, such that they would not become an accident icitiator. This is supported by past performance and industry experience. No new accident would be created by this Technical Specification change (10CFR50.59(a)(2)(ii)).

The proposed action vould not create a possibility for a malfunction of equipment of a different typa than any evaluated previously in the USAR because no equipuent is being added or modified. All equipment will continue to function as r.ssumed in the USAR (10CFR50.59(a)(2)(ii)).

The proposed action would not reduce any margin of safety as defined in the basis for any Technical Specification because the inspection and operability test interval for the RVVVs will be frequent enough to ensure the Limiting Condition for Operation is met. The Technical Specification Basis is not affected, as supported by past performance and industry experience. Since equipment reliability will be maintained, the margin of safety vill also be maintained (10CFR50.59(a)(2)(iii)).

Based on the above, it is concluded that the proposed Technical Specification changes do not constitute an unreviewed safety question.

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