ML20205P803

Proposed Tech Specs Re Relaxation of Excess Flow Check Valve Surveillance Testing

ML20205P803
Person / Time
Site:	Duane Arnold
Issue date:	04/12/1999
From:	IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT
To:
Shared Package
ML20205P800	List: ML20205P803 ML20205P809 NG-99-0308, Application for Amend to License DPR-49,relaxing SR Frequency by Allowing Representative Sample of Excess Flow Check Valves to Be Tested Every 24 Months,Such That Each Will Be Tested at Least Once Every 10 Yrs
References
NUDOCS 9904200377
Download: ML20205P803 (8)

Text

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't SCR-010 to NG-99-0308 PageIot~l PROPOSED CilANGE TSCR-010 TO Tile Dl!ANE ARNOI.D ENERGY CENTER TECIINICAl, SPECIFICATIONS The holders oflicense DPR-49 for the Duane Arnold Energy Center propose to amend the Technical Specifications by deleting the referenced page and replacing it with the enclosed new page.

St ININ1 ARY OF CII ANGES:

Pace Description of Chances 3.6-14 Revises the description of the SURVEILLANCE for SR 3.6.1.3.7 3

to state, " Verify a representative sample of reactor instrunientation line EFCVs actuate on a sunulated instrument line break to restrict flow.'

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l PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.5 Verify the isolation time of each MSIV is In accordance

> 3 seconds and < 5 seconds.

with the Inservice Testing Program SR 3.6.1.3.6

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NOTE------------------

For the MSIVs. this SR may be met by any series of sequential, overlapping, or total system steps, such that proper operation is verified.

Verify each automatic PCIV actuates to 24 months the isolation position on an actual or simulated isolation signal.

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Verify eaW@ reactor instrumentation lir.e 24 months SR 3.6.1.3.7 actuat on a simulated instrument ine break to restrict flow.

SR 3.6.1.3.8 Remove and test the explosive squib from In accordance each shear isolation valve of the TIP with the System.

Inservice Testing Program (continued)

DAEC 3.6 14 Amendment 223

PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1 3.6 REQUIREMENTS (continued)

Automatic PCIVs close on a primary containment isolation signal to prevent leakage of radioactive material from primary containment following a DBA.

This SR ensures that each automatic PCIV will actuate to its isolation position on a primary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.1. " Primary Containment Isolation Instrumentation." overlaps this SR to provide complete testing of the safety function. A Note has been added for the MSIVs. that allows this SR to be met by any series of sequential, overlapping, or total steps so that proper operation of the MSIVs on receipt of an actual or simulated isolation signal is verified. The 24 month Frequency was developed considering it is prudent that this Surveillance be performed only during a unit outage since isolation of 3enetrations would eliminate cooling water flow and disrupt t1e normal operation of many critical components.

Operating experience has shown that these components usually pass this Surveillance when performed at the 24 month Frequency. Therefore. the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.6.1 3.7 ggg

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This SR requires a demonstration that reactor gC instrumentation line Excess F1 heck Valv3/TEFCV/T OPERABLE by verifying that th alvecausqf)amarked decrease in flow rate on a simulated instrument line break.

This SR provides assurance that the instrumentation line EFCVs will perform so that predicted rectological consequences will not be exceeded during the postulated instrument line break event evaluated in Reference 5.

15eN I v0perating experience has show" that these compenents-ttstteMy pass this Survei' lance ' hen performcd at the 2d =onth frequency.

Therefore, the Frequency was concluded to be acceptabic from a reliability 5tendpw ni..

(continued)

DAEC B 3.6-27 Amendment 223

9 Insert I to BASES for SR 3.6.1.3.7 The representative sample consists of an approximately equal number of EFCVs, such that each EFCV is tested at least once every 10 years (nominal). The nominal 10 year interval is based on other pertbrmance-based testing programs, such as Inservice Testing (snubbers) and Option B to 10 CFR 50, Appendix J. EFCV test fitilures will be evaluated to determine if additional testing in that test interval is warranted to ensure overall reliability is maintained. Operating experience has demonstrated that these components are highly reliable and that failures to isolate are very infrequent. Therefore, testing of a representative sample was concluded to be acceptable from a reliability standpoint (Reference 7).

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1 PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.8 REQUIREMENTS (continued)

The TIP shear isolation valves are actuated by explosive charges. An in place functional test is not possible with this design.

The explosive squib is removed and tested to provide assurance that the valves will actuate when required.

The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. Other administrative controls, such as those that limit the shelf life of the explosive charges, must also be followed. The Frequency of this SR is in accordance with the requirements of the Inservice Testing Program.

SR 3.6.1.3.9 The analysis in Reference 8 is based on leakage that is less than the specified leakage rate.

Leakage through each MSIV must be s 100 scfh when tested at 2 24 psig.

The combined maximum ?athway leakage rate for all four main steam lines must De 5 200 scfh when tested at 2 24 psig.

If the leakage rate through an individual MSIV exceeds 100 scfh. the leakage rate shall be restored to s 11.5 scfh.

This ensures that MSIV leakage is properly accounted for in determining the overall primary containment leakage rate.

The Frequency is required by the Primary Containment Leakage Rate Testing Program.

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REFERENCES 1.

UFSAR, Chapter 15.6.

2.

UFSAR Table 7.3-1.

3.

10 CFR 50. Appendix J. Option B.

4.

UFSAR. Section 7.3.1.1.1.7.

5.

UFSAR. Section 6.2.4.2.4; 6.

J. Franz (IELP) to T. Murley (NRC). " Revised Response to NRC Position on Operability of Safety-Related Dual Function Valves." NG-93-5124 December 7. 1993.

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(continued)

DAEC B 3.6-28 Amendment 223

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TSCR-010 Anachment 3 to NG-99-0308 Page1ofI SAFETY ASSESSMENT l

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Introduction:==

By letter dated April 12,1999, Alliant-lES Utilities Inc. submitted a request for revision of the Technical Specifications (TS) for the Duane Arnold Energy Center (DAEC). The proposed amendment would revise TS SR 3.6.1.3.7.

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Evaluation:

The amendment will relax the frequency of SR 3.6.1.3 7 from testing each EFCV every cycle (24 months) to testing a representative sample of EFCVs every cycle.

In general, EFCVs have low failure rates, with zero failures to isolate at the l

DAEC. This high reliability and the low significance associated with an EFCV I

failure are the primary bases for this change as documented in a BWROG report included as Attachment 5 to this submittal. The instrument lines at DAEC include a flow restricting orilice upstream of the EFCVs to limit t ; actor water leakage in the event of a rupture. Previous evaluation of such an instrument line rupture l

1 (DAEC UFSAR 1.8.11), which the EFCVs are designed to mitigate, do not credit the isolation of the line by the EFCVs. Thus, a failure of an EFCV, though not expected as a result of this change, is bounded by the previous evaluation of an instrument line rupture. The radiation dose consequences of such a break are not impacted by this change.

l The reduced testing associated with this change will result in potential dose savings during the outages in which the testing is performed. An increase in the l

availability ofinstrumentation during the outage and cost savings are also considered potential benetits from this change.

l Therefore, we conclude that the proposed revision to the DAEC TS is acceptable.

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TSCR-010 Anachment 4 to NG-99-0308 Page 1of1 ENVIRONMENTAL CONSIDERATION 10 CFR Section 51.22(c)(9) identities certain licensing and regulatory actions uhich are eligible ihr categorical exclusion from the requirement to pertbrm an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the f acility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration: (2) re.sult in a significant change in the types or significant increase in the amounts of any eftiuents that may be released offsite; and (3) result in a significant increase in individual or cumulative occupational radiation exposure. IES Utilities Inc. has reviewed this request and determined that the proposed amendment meets the eligibility criteria tbr categorical exclusion set ibrth in 10 CFR Section 51.22(c)(9). Pursuant to 10 CFR Section 51.22(b).

no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the amendment. The basis tbr this determination follows:

Basis The change meets the eligibility criteria for categorical exclusion set forth in 10 CFR Section 51.22(c)(9) tbr the following reasons:

1.

As demonstrated in Attachment I to this letter, the proposed amendment does not involve a significant hazards consideration.

2.

The proposed change reduces the number of Excess Flow Check Valves (EFCVs) that are tested on a cyclic basis. A representative sample (instead of every valve) of EFCVs will be tested each cycle pursuant to TS SR 3.6.1.3.7. Failures that occur within the sample population will be evaluated to determine if additional testing is warranted. Operating experience demonstrates a high reliability and a very low fhilure rate for these EFCVs. Therefore, there will be no significant change in the types or significant increases in the amounts of any eftluents that may be released offsite.

3.

The proposed change will not change the way the EFCVs or the systems they are part of are operated. The EFCVs installed at the DAEC connect to the Reactor Coolant Pressure Boundary (RCPB). There are no EFCVs at DAEC that connect to primary containment atmosphere. The function provided by isolating an RCPB instrument line in the event of an instrument line break is not impacted by this change. This change does not impact the radiation dose results of a previous evaluation of an instrument line rupture. Theretbre, there will be no significant increase in either individual or cumulative occupational radiation exposure.

1 1 to NG-99-0308 GE Nuclear Energy

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B21-00658-01

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