ML20071Q621

Proposed Tech Specs Revising Isi/Ist Program Requirements

ML20071Q621
Person / Time
Site:	Duane Arnold
Issue date:	07/29/1994
From:	IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT
To:
Shared Package
ML20071Q619	List: ML20071Q621 NG-94-2256, Application for Amend to License DPR-49,consisting of Request for TS Change (RTS-270),revising Isi/Ist Program Requirements
References
NUDOCS 9408110300
Download: ML20071Q621 (5)

Text

'

RTS-270 to

~

ViG-94-2256 Page 1 of 1 PROPOSED CIIANGE RT&270 TO TILE DUANE ARNOI.D ENERGY CENTER TECIINICAL SPECIFICATIONS The holders oflicense DPR-49 for the Duane Arnold Energy Center propose to amend Appendix A (Technical Specifications) to said license by deleting certain pages and replacing them with the attached new pages. The affected pages and a description of changes are given below.

AFFECTED PAGES 3.6-11*#

3.6-28*

SUMMARY

OF Cli ANGES:

PAGli DIISCRIPTION OF CilANGES 3.6-11 The reference in SR 4.6.G. I to written relief being granted by the NRC, which requires prior NRC approval, is being deleted. SR 4.6.G.3 is being added to make clear that surveillance frequency maximum time intervals are applicable to ISI and IST activities..

3.6-28 A statement is added to the Bases to state that the ASME Code requirements do not supersede the TS.

i l

• The proposed markup is based on a pending amendment which incorporates RTS-197A and RTS-261.

1. This page is also affected by RTS-249.

9408110300 940729 PDR ADDCK 05000331 P

PDR

DAEC-1 l

l LIMI' TING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS F.

Jet Pumo Flow Mismatch F.

Jet Pumo Flow Mismatch 1.

With core power greater than or 1.

Recirculation pump speed mismatch equal to 80% RATED POWER with shall be verified at least once both recirculation pumps at per day.

steady state operation, the speed of the faster pump may not exceed 122% of the speed of the slower pump.

2.

With core power less than 80%

RATED POWER with both recirculation pumps at steady state operation, the speed of the faster pump may not exceed 135%

of the speed of the slower pump.

3.

With the recirculation pump speeds different by more than the specified limits:

a.

restore the recirculation pump speeds to within the specified limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or b.

one recirculation pump 2.

See Surveillance Requirement shall be tripped.

See 4.3.F.4 for SLO requirements.

Specification 3.3.F.4 for SLO requirements.

G.

Structural Intecrity G.

Structural Intecrity 1.

At all times, the structural 1.

Inservice inspection of ASME integrity of the ASME Section XI Section XI Code Class 1, Class 2, code Class 1, 2, and 3 components and Class 3 components and shall be maintained in accordance inservice testing of ASME Section with Surveillance Requirement XI Codo Class 1, Class 2, and 4.6.G.I.

Class 3 pumps and valves.shall be performed in accordance with 2.

With the structural integrity of Section XI of the ASME Boiler and any ASHE Section XI Code Class 1 Pressure Vessel Code and or Class 2 component (s) not applicable Addenda as recuired conforming to the above 10CFR50 Se t on 50 55a e t requirements, restore the ipwr d tt af 4

structural integrity of the (6'e - 95 t'# b t

,purp apt affeeted component (a) to within 6 t6)DCFJBC',,A!ie of its limit or isolate the affected W. S K( g}{$ ) ti').

component (s) prior to increasing the Reactor Coolant System 2.

The augmented inspection program temperature above 212*F.

for piping identified in NRC Generic Letter 88-01 shall be 3.

With the structural integrity of performed in accordance with the any ASME Section XI code Class 3 staff positions on schedule, component (s) not conforming to methods, personnel, and sample the above requirements, restore expansion included in this Generic the structural integrity of the

Letter, affected component (s) to within

• ' ' ' ' 7'N its limit or isolate the affected C# Definb ow 3.

NC f ro V 83 I8 A5 component (s) from service-

[SMR V&lLL ANCE FO:GWGNCY) Art aglicdle * +be Geg uencies &

perhemig mservice'imf+eeHon as k,n sen;ee +esh ac iviks.

AMENDMENT NO.

3.6-11 RTS-2 70

DAEC-1 l

3.6.G & 4.6.G BASES:

I Structural Integrity i

A pre-service inspection of Nuclear Class I Components was conducted to assure freedom from defects greater than code allowance; in addition, this served as a reference base fer future inspections.

Prior to operation, the Reactor coolant System as described in Article IS-120 of Section XI of the ASME Boiler and Pressure Vessel Code was inspected to provide assurance that the system was free of gross defects.

In addition, the facility was designed such that grosc defects should not occur throughout plant life.

The pre-service inspection program was based on the 1970 Section XI of the ASME Code for in-service inspection. This inspction plan was designed to reveal problem areas (should they occur) before a leak in the coolant system could develop. The program was established to provide reasonable assurance that no LOCA would occur at the DAEC as a result of leakage or breach of pressure-containing components and piping cf the Reactor Coole.nt Sistem, portions of the ECCS, and l

portions of the reactor coolant associated auxlliary systems.

A pre-service inspection was not performed on Nuclear Class II Components because it was not required at that stage of DAEC construction when it would have been used.

For these components, shop and in plant examination records of components and welds well be used as a basis for comparison with in-service inspection data.

Visual examinations for leaks will be made periodically on ASME Section XI Class 1, 2 and 3 systems. The inspection program specified encompasses the major areas of the vessel and piping systems within the ASME Section XI boundaries.

The type of examinations planned for each component depnds on location, accessibility, and type of potential defect.

Direct visual examination is proposed wherever possible since it is fast and reliable. Surface examinations are planned where practical, and where added sensitivity is required. Ultrasonic examination or radiography shall be used where defects can occur in concealed surfaces.

Section 5.2.4 of the Updated FSAR provides details of the inservice inspection program.

Starting with the cycle 9/10 Refueling outage, an augmented inspection program was implemented to address concerns relating to Intergranular Stress corrosion cracking (IGSCC) in reactor coolant piping made of austenitic stainless steel.

The augmented inspection program conforms to the NRC staff's positions set forth in Generic Letter 88-01 and NUREG-0313, Revision 2 for inspection schedule, inspection methods and personnel, and inspection sample expansion.

The first 10-year interval for inservice testing of pumps and valves in accordance with the ASME Code,Section XI commenced on Februit; 1, 1975 and ended on January 31, 1985.

The second 10-year inservice testing interval commenced on February 1, 1985 and is scheduled to end on January 31, 1995.

The second 10-year testing program addresess the requirements of the ASME Codt, Saction XI, 1980 Edition with Addenda through Winter 1981, subject to the limitations and modifications of 10 CFR 50.55a.

Section 3.9.6 of the Updated FSAR describes the inservice testing program.

Nc4W.

in %c A S u%E bcMer & ('ren ar e Vc.ss e l f le s k 'l be.

W s4 mea 40 sqrseA N re drsmed 5 of cury Tb.

AMENDMENT NO.

3.6-28 R 3~5 - 2~7O

RTS-270 to NG-94-2256 Page1ofI SAFETY ASSESSMENT Intmduction By letter dated July 29,1994, IES Utilities Inc. submitted a request for revision of the Technical Specifications, Appendix A to Operating License No. DPR-49, for the Duane Arnold Energy Center (DAEC). The proposed change would delete the requirement to obtain prior NRC approval of relief requests and would provide the licensee the authority to implement relief requests pending NRC approval, provided that certain internal review processes are completed.

Additionally, provisions are added to make clear that the surveillance fiequency maximum time intervals are applicable to inservice inspection and inservice testing activities and that the ASME Code requirements do not supersede the TS requirements.

A_sgs_sment Drafl NUREG-1482, " Guidelines for Inservice Testing at Nuclear Power Plants," recommended that licensees revise their Technical Specifications (TS) to incorporate the revised Standard Technical Specifications (STS) wording for inservice testing (IST) programs. Upon determining that an ASME Code requirement is impractical because of prchibitive dose rates or limitations in the design, construction, or system configuration, the licensee can implement a Code relief without prior NRC approval, provided that the relief request had been reviewed and approved by the plant stafTin accordance with IST program administrative procedures, reviewed by the plant safety committee (DAEC Operations Committee), and processed under 10 CFR 50.59.

The change is consistent with the recommendations provided in draft NUREG-1482 and with the STS (NUREG-1433). Minor dilTerences exist between these documents and the DAEC TS due to the DAEC custom TS format. The STS reflects the position that the licensee must establish and implement the ISI and IST programs in accordance with 10 CFR 50.55a, but does not require that relief requests be granted by the NRC before they can be implemented. Rather,10 CFR 50.55a(f)(5)(iv) and 10 CFR 50.55a(g)(5)(iv) allow a licensee up to a full year after the beginning of the updated interval to inform the NRC of any new ASME Code requirements which cannot be met and to request relief. The regulations require the licensee to submit relief requests within 12 months of the interval start date, or during the interval if the licensee fmds specific need for relief.

i This change will enable a licensee to avoid situations where compliance with the TS cannot be achieved for the period between the time of preparation and submittal of a relief request and NRC granting of the relief. The NRC will still grant final approval and this, in conjunction with the required internal review and approval process of the licensee, will ensure that only the appropriate relief from the ASME Code is granted.

Based on the above evaluation, we conclude that the proposed Technical Specification change is

)

acceptable.

4 RTS-270 to NG-94-2256 PageiofI ENVIRONMENTAL CONSIDER ATION 10 CFR 51.22(c)(9) identifies certain licensing and regulatory actions eligibk for categorical j

exclusion from the requirement for an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation cf the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration; (2) result in a significant change in the types or significant increase in the amounts of a iy efiluents that may be released ofTsite; and (3) result in an increase in individual or cumulative occupational radiation exposure. IES Utilities Inc. has reviewed this request and determined that the proposed amendment meets the criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmentalimpact statement or j

environmental assessment needs to be prepared in connection with the amendment. The basis for this determination follows:

Ilasis The change meets the criteria for categorical exclusion set Ibrth in 10 CFR 51.22(c)(9) for the following reasons:

1.

As demonstrated in Attachment 1, the proposed amendment does not involve a significant hazards consideration.

2.

The proposed change to the ISI/IST programs will have no effect on the types or amounts of eilluents released offsite. The requirements of the ASME Code will still be complied with except where written reliefis required. The licensee may implement the relief request pending NRC approval, but fmal NRC approval must still be granted.

3.

The proposed change to the ISI/IST programs will have no elTect on individual or cumulative occupational radiation exposures.

i t

[

i

{

?

?

l i

i u

1 r-m

,wr

- +

fr'