ML20078F127

Advises That Util Elected to Perform Once Per Cycle Requirement Delineated in SR 4.6.D.1 for Valve MS-RV-70ARV During Next Refueling Outage Scheduled to Commence Oct 1995

ML20078F127
Person / Time
Site:	Cooper
Issue date:	01/26/1995
From:	Mueller J NEBRASKA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NLS950029, NUDOCS 9502010327
Download: ML20078F127 (5)

Text

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,,i' COOPER NUCLEAR ST ATION Ndbraska Public Power District

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a-1 NLS950029 January 26,1995 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555

Subject:

Cycle Extension Cooper Nuclear Station NRC Docket No. 50-298, License No. DPR-46

Reference:

Letter (No. NLS940133) to the U.S. NRC Document Control Desk from Nebraska Public Power District dated December 27,1994;

Subject:

Exemption Request - 10 CFR 50 Appendix J, Paragraph III.D.2(a) t Gentlemen:

Cooper Nuclear Station (CNS) has been in a forced outage since May 25,1994. The length of this outage, in combination with the longer than expected 1993 refueling outage and other unplanned outages, has extended the length of the current operating cycle (Cycle 16) for CNS by approximately 12 months. Currently, the Nebraska Public Power District (District) plans to commence the next refueling outage during the month of October 1995.

In support of the extended cycle, the District has reviewed those CNS Technical Specification requirements which require surveillance testing at least once per operating cycle, once every 18 months, refueling cycle, or refueling outage; otherwise categorized as " cycle-related technical specifications." The CNS Technical Specifications define " operating cycle" as the

" interval between the end of one refueling outage and the end of the next subsequent refueling cutage". Many surveillances are performed on a frequency of at least once per operating cycle.

Though not required by CNS Technical Specifications, the District has recently defined " cycle" as 18 months (plus/mmus a maximum of 25% interval based on the provisions specified m CNS Technical Specification Definition 1.0.Y) for tracking and scheduling purposes.

In accordance with this internal definition, the District has reviewed both its cycle-related Tecimical Specification surveillance requirements and non-Technical Specification surveillance requirements for determination as to which of these surveillance requirements should be completed during the current forced outage. Out of approximately 175 surveillance procedures (both Technical Specification and non-Technical Specification) reviewed, the District has elected to perform all but one of the cycle-related Technical Specification surveillances. The i

District has also elected to perform all but 18 of the cy:le-related non-Technical Specification procedures. Written justifications supporting the perfo1 ming of the 18 surveillance procedures during the 1995 refueling outage have been prepared and are available for NRC inspection.

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9502010327 950126

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'U'S. Nuclear Regulatory Commission January 26,1995 Page 2 of.2 The one remaining cycle-related Technical Specification is Surveillance Requirement (SR) 4.6.D.1, which requires, in part, that "approximately half of the Main Steam Safety Valves (SVs) be checked or replaced with bench checked valves once per operating cycle." The specific SV which would, under the internal definition of " cycle", require bench checking during the current outage is MS-RV-70ARV. Ilowever, the District has performed a 10 CFR 50.59 evaluation and has determined that adequate justification exists to perform SR 4.6.D.1, as it pertains to this valve, during the 1995 refueling outage. Per discussions with the Nuclear Regulatory Commission (NRC), the District is providing a summary of thisjustification in the attachment to this letter.

Portions of SR 4.7.A.2.f, " Local Leak Rate Tests" also uses terminology similar with the District's internal definition of the term " cycle" (at least once per ope.ating cycle, once every 18 months, refueling cycle, or refueling outage). Ilowever, the schedule requirements for this SR is driven by 10 CFR 50, Appendix J, and requires testing in all cases within two years.

For this reason, the District has already requested (see reference) a schedular exemption to implement SR 4.7.A.2.f, as it pertains to the Type L1 local leak rate test requirements for the Drywell llead and Manport (Penetration X-4), during the 1995 refueling outage. The District will submit a one-time Technical Speciiication change to support the exemption. Because an exemption request has already been submitted (see reference), it has not been included within the scope of this cycle extension letter.

In keeping with the philosophy that the surveillance interval for cycle-related Technical Specifications is 18 months (+/- 25%), these surveillances will be performed on an 18 month interval unless adequate justification can be otherwise demonstrated. The District has taken a conservative approach in providing your office infbrmation regarding its plans to perform SR 4.6.D.1, as it pertains to Main Steam Safety Valve MS-RV-70ARV, during the 1995 refueling outage. The District has determined that no Technical Specification change is required or warranted for this situation.

If you have any questions or need additional information, please contact me.

Sincerely, f_)J l}n..[0 John 11. Mueller Site Manager

/dnm Attachment ec:

Regional Administrator USNRC Region IV NRC Resident inspector Cooper Nuclear Station NPG Distribution

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'A'ttachment to NLS950029 Page 1 of,2 The'following information provides a brief description of the justification supporting the District's decision to perform CNS Technical Specification Surveillance Requirement (SR) 4.6.D.1 during the 1995 refueling outage for Main Steam Safety Valve MS-RV-70ARV. A 10 CFR 50.59 safety evaluation justifying this test schedule and concluding that no unreviewed

. safety question exists has been approved by the CNS Station Operations Review Committee (SORC).

Component Id:

MS-RV-70ARV Manufacturer:

Dresser Type 377QA-RT22 SR Requirement:

SR 4.6.D.1 requires, in part, that half of the Main Steam Safety Valves be checked or replaced with bench checked valves once per operating cycle. All valves will be tested every two cycles.

Date last performed:

November 5,1991 District Plan:

The District will perform the once per cycle requirement, delineated in SR 4.6.D.1, for Valve MS-RV-70ARV to the 1995 refueling outage, currently scheduled to commence October 1995.

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Justification for the District Plan:

Technical Specification 4.6.D.1 requires that approximately half of the SVs shall be checked or replaced with bench checked valves "once per operating cycle". Technical Specification 4.6.D.1 also requires that all SVs be tested "every two cycles". In addition, no work has been performed which would negatively affect Safety Valve performance during the current shutdown. Therefore, no Technical Specification or procedure change is required to support performing this SR during the 1995 refueling outage.

1 The subject Main Steam Safety Valve (SV) is of Dresser Maxiflow design.

The valve components evaluated for this assessment are those subjected to the potentially highly humid Main Steam line internal environment such as internal mechanisms, discs, and pressure boundary components.

i The SV utilizes beh carbon steel and stainless steel structural materials. Carbon steel is of primary concern for low temperature corrosion influences. Ilowever, the critical components of the valves whose function may be affected by corrosion (e.g., discs and seats) are manufactured from corrosion resistant stainless steel. This material is chosen for its overall resistance to corrosion in all potential environments including both cold and operating i

temperature conditions. These components will not be impacted by the shutdown. For these corrosion resistant components. corrosion virtually stops during the shutdown and will increase upon return to high temperature operating conditions. Therefore, performing maintenance is not necessary as a resA of the extended outage.

The length of time a vaive may be left installed in a plant prior to its removal is based on IIngineering Judgement and experience. The Main Steam Safety Valves are resistant to on-line storage conditions and will be unaffected by the shutdown. They do not have tight clearances subject to the relatively humid steam line environment. The components subject to the inlet environment are made of corrosion resistant materials (i.e., not carbon steel). The SVs do not

' Attachment to NLS950029 Page 2 of,2 requ' ire actuation upon return to service and will therefore not be affected by upstream steam line corrosion. Also, the SVs have been tested and will continue to be tested in accordance with ASME Section XI requirements (and within the ASME bounded 5 year frequency requirement).

Surveillance history also supports performing SR 4.6.D.1 for MS-RV-70ARV during the 1995 refueling outage. The most recent surveillance, performed in 1991, test data indicates that lift occurred at 1205 psig which was outside the 1240 psig i 1% Technical Specification tolerance (NUREG 1433, Standard Technical Specifications utilizes a 3% tolerance). The reduced lift setpoint is an operational concern rather than a safety concern because the drift is in a consesative direction. Substantial margin remain between the normal plant operating pressure of 1000 psig and the as-found setpoint of 1205 psig. A review of the 1989 surveillance data indicated that the valve had minor seat leakage but the as-found setpoint was 1234 psig which was within the i 1% Technical Specification tolerance.

The District has concluded, as documented in the SORC approved 50.59 analysis, that performing SR 4.6.D.1 during the 1995 refueling outage does not result in an increased probability of an occurrence or increased consequences of an accident previously evaluated in the USAR. The District has also concluded that this plan does not create the possibility of an accident of a different type than that evaluated in the USAR, nor is the margin of safety impacted.

None of the events described in Chapter XIV of the CNS USAR categorized as " accidents" are initiated by the operatir,n, or...isoperation, of the SVs. Furthermore, continued use of the SVs (in particular.'1S-RV-70ARV) Bllowing an extended outage; 1) will not degrade the quality of the Reactor Coolant Pressure Boundary component; 2) does not change the required maintenance or insersice inspection intervals; 3) will not create or require any new operator procedures or training; 4) does not add new equipment to a safety-related system or create a new failure mode; 5) does not change an instrument accuracy or response time; 6) does not operate the system outside of its design envelope; 7) does not present a new radioactive leakage path; 8) does not increase n USAR Chapter XIV accident dose calculation result to be greater than its License Acceptance Limit; 9) does not increase onsite or offsite radiation doses; and 10) does not increase personnel hazards. For these reasons, along with additional reasons identified in the SORC approved 50.59 analysis, the District has concluded that performing SR 4.6.D.1 during the 1995 refueling outage will not result in the creation of a safety conectn.

l LIST OF NRC COMMITMENTS l ATTACHMENT 3 l Corlespondence No:

NLS950029 The following table identifies those actions committed to by the District in this document. Any other actions discussed in the submittal represent intended or planned actions by the District. They are described to the NRC for the NRC's information and are not regulatory commitments.

Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.

COMMITTED DATE COMMITMENT OR OUTAGE During the 1995 ment SR 4.6.D.1 for Main Steam Safety Valve (MS-RV-cr y

uled to commence October 1995 To support NRC approval of the Technical The District will submit a one-time Technical Specification prior Specification change to support the schedular exemption to July 17, 1995 for the Drywell Head and Manport.

(Current due date of the subject Type B LLRT)

PROCEDURE NUMBER 0.42 l

REVISION NUMBER 0 l

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