ML20095H237

Proposed Tech Specs,Incorporating EDG Conditional Surveillance & Editorial Clarifications

ML20095H237
Person / Time
Site:	Duane Arnold
Issue date:	12/15/1995
From:	IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT
To:
Shared Package
ML20095H233	List: ML20095H237 NG-95-3413, Amend to 950721 Request for TS Change (RTS) 285 to License DPR-49,consisting of RTS-285A,incorporating EDG Conditional Surveillance & Editorial Clarifications,Per NUREG-1433 & GLs 93-05 & 93-07
References
NUDOCS 9512220228
Download: ML20095H237 (15)

Text

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. RTS-285A Attachtrent 2 to NG-95-3413 Page1of2 ,

PROPOSED CHANGE RTS-285A TO THE DUANE ARNOLD ENERGY CENTER l TECHNICAL SPECIFICATIONS The holders oflicense DPR-49 for the Duane Arnold Energy Center propose to amend Appendix A (Technical Specifications) to said lic:nse as indicated on the attached marked-up pages. The List of Affected Pages is given below.

LIST OF AFFECTED PAGES Operating License Page 4 3.5-10 3.5-23 3.8-4 3.8-6*

5.5-1 6.5-3

• t 6.8-l

• 6.8-2 *

• Previously submitted as RTS-285, not affected by RTS-285A.

SUMMARY

OF CHANGES:

The following list of proposed changes is in the order that the changes appear in the Technical Specifications.

Page Descrintion of G1anges Operating Revise paragraph 2.C(4) to correct wording consistent with Amendment

! License 47.

page 4 3.5-10 Revise Surveillance Requirement to require a determination that the OPERABLE EDG is not inoperable due to a common cause within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and retain the requirement to perform the OPERABILITY test each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

3.5-23' Revise Bases to reflect above changes 3.8-4 Revise reference in TS Section 3.8.B.2.c from 3.7.D to 3.7.B.

951222O228 951215 PDR ADOCK 05000331 p PM t

. RTS-285A Attachment 2 to NG-95 3413 l Page 2 of 2 Eage Descrintion of Changes 3.8-6 Revise Su; :illance Requirement for one ESW pump or loop inoperable

. to delete the reference to Surveillance Requirement 4.5.G.1 and reiterate the requirement to verify all low pressure core cooling and containment cooling subsystems and the diesel generator associated with the OPERABLE ESW are also OPERABLE. .

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5.5-1 Reformat Section 5.5 to be consistent in content and format with NUREG 1433, Improved Standard TS. RTS-285A removes the i previously proposed limit on enrichment of new fuel.

6.5-3 Delete "and implementing procedures" from items i and j.

)

a r, ? 1 Replace " Procedures required by the Emergency Plan" with the word

" Deleted."

i 6.8-2 Replace " Procedures required by the plant Security Plan" with the word

" Deleted."

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S

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(3) Fire Protection Revision to CL Amenament #190 IES Utilities Inc. shall implement and maintain in effes..

01/93 all provistons of the approvea fire protection program as Revision to CL described in the Final Safety Analysts Report for the Amenneent #198 Duane Arnold Energy Center and as approved in the SER 05/12/94 dated June 1, 1978 and Supplement dated February 10, 1981, sun 3ect to the following provtston:

The licensee may make enacges to the approved fire protection program witnous prior approval of the Comunsaston only if those enanges would not adversely affect the ability to acnieve and matntain safe shutdown in the event of a fire.

Added to QL (4) The licensee is authorized to operate the Duane Arnold A==a - at #47 Enstrgy Center following installation of modified safe-ends 1/08/79 _ W ines on the eight ha & ? ? YM primary recarculatimp p$$e+Fe J u-ted f f y 3sg9 'IB and"sgkh+ed -by se tte r- d a+ <d Dece A*

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• % ' 416' (5) Physical ProtectiWfr ^

Added to OL The licensee shall fully implement and maAntain in effect Amenament #50 all provisions of the coussission arr. . l. physical 4/19/79 security, guard, training and qualification, and safeguards cow.ingency plans, including ===a**===ts made pursuant to Revision to CL the authority of 10 CFR 50.54(p). The approved plans.

Amendment #65 which contain Safeguards Information as described in 10 3/03/81 CFR 73.21, are collectsvoly entitled:

Revision to CL "Duane Arnold Energy Center Security Plan" dated Amenament #74 Decommer 1, 1978, Januncy 19, March 9 and March 21 6/09/82 1978, as revised through revisions dated January 1984 (transmittal letter dated January 12, 1984), as Revision to CL revised by revision dated February 1984 (transmittal Asenament #112 letter dated February 27, 1984), as revised by 2/26/85 revision dated Septemmer 1984 (transmittal letter dated Septemmer 26, 1984); "Duane Arncid Energy Center Safeguards Contingency Plan," dated April 1980, as revised throuan revision dated January 1984 (transmittal letter cated January 12, 1984); "Duane Arnold Energy Center Guard Training and Qualification Plan" dated January 29, 1982, as revised April 1, 1982, as revised throuch revisions dated January 1984 (transmittal letter named January 12, 1984), as revised by undated revisions (transmittal letter dated July 30, 1984), as revised by revision dated Septemper 1984 (transmittal letter dated Septoneer 26, 1984) as revised by revision dated Octomer 1984 (transmittal letter dated October 26, 1984).

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, ,, DAEC-1 LIMITING COND2TIONS FOR OPERATION SURVEILLANCE REOUIREMENTS G. Minimum Low Pressure Coolina and G. Minimum Low Pressure Coolina and Diesel Generator Availability Diesel Generator Availability w

1. During any period when one diesel 1. "h _ it is-detes.aii. 4- thet-one-generator is inoperable, --di;;;l ve..-. ter-is-inoperable,--

continued reactor operation is -the . - rir.inv di==wl v e..; :ter permissible only during the ch:11- h; ' - ;;treted Le 6 -

succeeding seven days unless such 07:72"i5 in eeee 4;;;; -lih-diesel generator is sooner made OPERABLE, provided that the

- 0;; ifi;; tie.

th; fi.et 24 hwu. and -

4. G. A.2.e. l.- -lu.in-i

-.y remaining diesel generator and all low pressure core and Q :ddition,:T In a  ;..tall'y low he se ih- --Ge..

pressure core containment cooling subsystems cooling and containment cooling supported by the OPERABLE diesel subsystems supported by the generator are OPERABLE. If this OPERABLE diesel shall bt verified requirement cannot be met, an to be OPERABLE.

orderly SHUTDCAfN shall be initiated and the reactor shall h-C be in at least HOT SHUTDOWN ine c, U,,

within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in 6 m Al de de fe rm ine- M ne rde rOF6~%E'(C

+b" COLD SHUTDOWN within the

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gree / ener4for B not ), p edle following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4y

2. Any combination of inoperable -

wiON 24 h o u.ns ctn/ fe rv, m components in the core and containment cooling systems shall g"C "" b ", C " "p "'

not defeat the capability of the $

( 4 g 4 L g, ( 4 - 47n g remaining OPERABLE components to \

fulfill the cooling functions. [res i 72 h ou r s <tod ave ry

3. When irradiated fuel is in the ~/ 2 ha es fhereet Oc r. #

reactor vessel and the reactor is in the COLD SHUTDOWN Condition or REFUEL Modes -s n jm/

g[ l l

a. If no work is being performed which has the potential for draining the reactor vessel, both core spray and RHR systems may be inoperable; or i

b. If work is being performed I which has the potential for draining the reactor vessel, at least two of any combination of core spray and/or RHR (LPCI or shutdown cooling mode) pumps shall be OPERABLE (including the capability to inject water into the reactor vessel with suction from the suppression pool) except as 3.5-10 Amendment No.70$el97 N TS - 2 L> d

'Y S

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DAEC-1 4.5 BASES

" Core and Containment Cooling Systems Surveillance Frequencies The testing interval for the core and containment cooling systems is based on industry practice, quantitative reliability analysis, judgement and practicality.

The core cooling systems have not been designed to be fully testable during operation. For example, in the case of the HPCI, automatic initiation during power operation would result in pumping cold water into the reactor water vessel which is not desirable. Complete ADS testing during power operation causes an undesirable loss-of-coolant inventory. To increase the availability of the core and containment '

cooling systems, the components which make up the system, i.e., instrumentation, pumps, valves, etc., are tested frequently. The test intervals are based upon Section XI of the ASME Code. A simulated automatic actuation test once per year combined with frequent tests of the pumps and injection valves is deemed to be  !

adequate testing of these systems.

When components and subsystems are out-of-service, overall core and containment cooling reliability is maintained by evaluating the operability of the remaining equipment. The degree of evaluation depends on the nature of the reason for the out-of-service equipment. For routine out-of-service periods caused by preventative maintenance, etc., the evaluation may consist of verifying the redundant equipment is not known to be inoperable and applicable surveillance intervals have been satisfied.

However, if a failure due to a design deficiency caused the outage, then the evaluation of operability should be thorough enough to assure that a generic problem does not exist.

3 FThe RHR valve power bus is not instrumented. For this reason surveillance l requirements require once per shift observation and verification of lights and

/ instrumentation operability.

( L v- _- ,- y~f 'V~~w-w Diesel G e n e m+o r s ace er:freal to ope rx+ ion da sn ee a ed e ,n+ m er a epo ir,3 sys k s. ~rk r dorn 1+ is WP' *0i" \

4hd + hey he wa r n + .a n e <4 in a sfs <d b y r ead ?" e n condrfr om.

_D +he e ve nl -lhsh one b re s el dene edo e is mah oe 4vu J Jo be i n op e.ca b l e, & rnwy v re sd Ge<- cJo r-mcf he s how n ho no4 b .- s u re p+; ble ho +he ca. m-wd:4;en wNkin Pfl hov e s.  % e vA dre n he-pe rforne d by a na l ys t s or in s p a eJ7e n or by dwo n d rko n )

cf O PE R A i5ILI T Y, The OPER A BLE .p ra wl de m e rdo r m d s)w be de m s4rs le d co d t,w e lo O PER A& LE euk 72 Q +c. be

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hoves ducie3 ha pe r ;o d Od i k- o %e r D re w I G er A e r /

\ is nopeca W .

A_.

mm Amendment No. C7,14' ,100,174, 210 3.5-23 RTS - 2 8 5 A 12 / s

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, DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS l chargers for the 24 Volt Systems, voltage shall be measured and two of the three battery chargers recorded.

for the 125 Volt Systems, and one of the two battery chargers for b. Each three months the essential the 250 Volt System shall be batteries' voltage of each cell to OPERABLE. the nearest 0.01 Volt, specific gravity of each cell, and temperature of every fifth cell shall be measured and recorded.

c. Once each OPERATING CYCLE, the essential batteries shall be subjected to a S.?rvice Discharge Test (load profile). The specific gravity and voltage of each cell shall be datermined after the discharge and recorded.

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d. Once every five years, the essential battaries shall be l subjected to a Performance

' Discharge Test (capacity). This test will be performed in lieu of the service Test requirement of 4.8.B.1..c above.

2. Operation with Inoperable 2. surveillance Requirements with Components. Inoperable components.

a. With normal battery room a. With the battery room ventilation ventilation unavailable, portable unavailable, samples of the ventilation equipment shall be battery room atmosphere shall be provided. taken daily for hydrogen concentration determination.

b. With one of the two 125 Volt DC Systems inoperable, verify that Specification 3.5.G is met, and within 3 days either:

1) Restore the inoperable 125 Volt DC System to OPERABLE status, or

2) Be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c. With the 250 volt DC System inoperable, the HPCI System and other affected primary containment isolation valves shall be considered inoperable and the requirements of Specifications 3.5.D and 3. .J( 6 respectively shall be met.

d. With one of the 24 Volt DC Systems inoperable, the requirements associated with the affected instruments of Specifications 3.1 and 3.2 shall be met.

3.8-4 kI b ~ ?-'05A Amendment No.J33,197 IE / 7dI

. .- DAEC-1 LIMITING CONDITIONS FOR OPERATION

! SURVEILLANCE REQUIREMENTS i

E. Emeroency Service Water Svitg5 4

E. Emeroency Service Water System

1. Except as required in Specification 1. Emergency Service Water System 3.8.E.2 below, both Emergency Service Water System loops shall be surveillance shall be as follows:

OPERABLE whenever irradiated fuel a.

is in the reactor vessel and Simulated auto- once/

reactor coolant temperature is matic actuation OPERATING CYCLE greater than 212*F. test.-

b. Pump and motor Asspecifiedinf operated valve the IST Program j OPERABILITY.

c. Flow Rate Test Each Emergency After major pump Service Water maintenance and pump shall once per 3 months, deliver at except weekly 7-06 least that flow during periods of 8-01 etermined from time the river Figure 4.8.E-1 water temperature for the exceeds 80*F.

existing river

[ water temperature.

2. With one of the Emergency Service Water System pumps or loops

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/. . With one Emergency Service Water i i

inoperable, REACTOR POWER OPERATION System pump or loop inoperable, the OPERABLE pump and loop sh be 9'-02 must be limited to seven days 2

unless OPERABILITY of that system varifiQ be?;;;.;i.

Gfditi::,th: OPE ;B.;; t:-Ie- ef is restored within this period. '

During such seven days all active E;;;ffi::ti: 4 . 5. 0 .; ;t,;11 b; : : n components of the other Emergency I n a h +:e n , eli los p r<.s m co re c..li g a. s c< 4 - w e d c e = 't a s Service Water System shall be sdystus e,J M< (

OPERABLE, provided the requirements drese / i of Specification 3.5.G are met. geneed-r surparhd o rn a s t.s em I o r s h a ll 4

7 "e be /

1 g v e r (, a f. b ,. o rca A BL s.

3. If the requirements of Specification 3.8.E cannot be met,  %

be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

7Ts- zes Amendnent No.20,22,120,100,100,)97 210 7 3 . 8-6_. . _ _ _ _ _ _ _ -- --

p 5EIC~

DAEC-1 l pDN ANDy.UEL_We #

/ 1. The new fuel storage facilit t effective neutron-iEuttQlication factor (k ,,) yofshall the be such fuel, b that i

N floodel s less than 0.95. These k,g val ess than 0.90 and maximum in ' te lattice multiplicati a satisfied if the /

fuel bundles is 1.31. actor (k. f the individual

/

2. /

' The k,,,loftothe fuel in -

or equa 0 95 T k spent fuel storage pool shat 1 be les than /

ue is satisfied if .ttie maximum, i exposure-depende ' 1 initial unifo average ofenrichment be indh ualis fuel bundies is s 1.31 and the 4.6,.wt% U-235. (

3. Spen el shall only be stored in the, pent uel pool in a vertical or entation in approved storage racks'.

UN -

B.B.Lt1 The fuel storage basis ferracks. the k, limit is described in Reference I for the GE-designed new Compliance with this specification is demonstrated by comparing the interest to the 1.31 limit. For GE-supplied beginning-of-life, uncontrolled k, values for the fuel type of Reference 2. fuel, k, values can be found in The k, values found in Reference 2 represent the maximum, exposure-dependent new fuel limit. lattice reactivity and can be conservatively applied to the Calculations have been performed (Reference 3) to determine the bounding reactivity limits for bundles of CE-designed fuel, when stored in the spent fuel storage racks of an approved design. These analyses were performed conservatively assuming uniform average ~ initial enrichments in a parametric i evaluation .for fuel with enrichments up to 4.6 wt% U-235 initially. The bounding limit of an infinite multiplication factor of 1.31 for fuel of 4.6 wtX enrichment (or less) was evaluated at the maximum k, over burnup and includes a conservative allowance for possible differences between the rack design calculations and the fuel vendor calculations.

. References

1) General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A.* 1 l

2) General Electric Fuel Bundle Designs, NEDE-31152-P.*  !

3)  !

Licensing Report for Spent Fuel Storage Capacity Expansion, Duane Arnold j Energy Center, Holtec Report HI-92889.

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• Latest NRC-approved revision.

Amendment No. JJg, 195 5.5-1 I RTS-zesA

/2/95

fNSERT .___ '

.p 3 5.5.1 Criticality N- 1 5.5.1.1 The spent fuel storage racks are designed and shall be maintained with: l

a. Fuel assemblies having a maximum k., of 1.31 in the normal reactor core i configuration at cold conditions and a maximum initial uniform average U-235 /

enrichment of 4.6 weight percent.

\., )

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b. kms 0.95 flooded with unborated water.

5.5.1.2 The new fuct storage racks are designed and shall be maintained with:

a. Fuel assen:blies having a maximum kJf tsasto /

confi uratiott at cold conditiod5r.e;1 dr.e; . m.;.e! Jfu.... evere es11-235 M i d d " E ;'.: p ; A.d. f-y ,,7 m

~ n -

b. kg < 0.90 dry and s 0.95 flooded with unborated water. I l

5.5.2 Capcity 5.5.2.1 The spent fuci stcrage pool has been analyzed to allow storage of a maximum of 3152 fuel assemblies in a vertical orientation only.

5.5.2.2 The new fuel stomge vault is equipped with racks for storage of up to 110 fuel assemblies

/  !

in a vertical orientation only.

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RTS - 2 6 5'A IL /95

, DAEC-1

, fl Review of all Reportable Events,

g. Review of facility operations to detect potential safnty hazards.

4 h.

Performance of special reviews, investigations or analyses and reports thereon as requested by the Chairman of the Safety Committee.

1. Review of the Plant Security Plam,=d i.T+k...atir.; prretr::.

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j. Review of the Emergency Plan . ed-/ i.T+R.;..ti r.; precat. ;; .-

W /

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k. Review of every unplanned release of r Mioactivity to the environs for which a report to the NRC is required.

1.

Review of changes to the Offsite Dose Asses ment Manu;31 and changes to the Process Control Program.

l m.

Review of the Fire Protection Program and implementing procedures.

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6.5.1.7 Authority I

The Operations Committee shall:

l a.

Recommend to the Plant Superintendent-Nuclear written cpproval or i

disapproval of items considered under Specification 6.5.1.6 (a) through (d) above.

Amendment No. 198,198 6.5-3 07/95

j DAEC-1 j .

i 6.8 PLANT OPERATING PROCEDURES 6.8.1 Written procedures involving nuclear safety, including applicable I '

check-off lists and instructions, covering areas listed below shall be l prepared, and approved as specified in Subsection 6.8.2. All l .

t procedures shall be implemented and maintained.

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1. Normal startup, operation, and shutdown of system: and components  !

of the facility.

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f 2. Refueling operation.

3. Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, including responses to l alarms, suspected primary system leaks, and abnomal reactivity changes.

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4. Emergency and off-normal condition procedures.  !

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5. Preventive and corrective maintenance operations which could have l l

an effect on the nuclear safety of the facility. l

6. Surveillance and testing requirements of equipment that could have an effect on the nuclear safety of the facility.

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7. 4 ::: t :: -eT & cd 53 tr.: Scr;;;:y "hn. De.le.fe)

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, 6.8-1 Amendment No.109 RTS-285 C'/? C

DAEC-1

8. recedsg; r;;2ir:d by th: ?h t h::Mty ."hr..- Deleb

9. Operation of r aste systems.

10. Fire Protection Program implementation.

11. A preventive maintenance and periodic visual examination program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient to as low as practical levels. This program shall also include provisions for performance of periodic systems leak tests of each system once per OPERATING CYCLE.  !

12. Program to ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions, including trainir of personnel, procedures for monitoring and provisions for maintenance of sampling and analysis equipment.

13. Administrative procedures for shift overtime for Operations personnel to be consistent with the Commission's June 15, 1982 policy statement.

14. OFFSITE DOSE ASSESSMENT MANUAL.

15. PROCESS CONTROL PROGRAM.

16. Quality' Control Program for effluents.

6.8.2 Procedures described in 6.8.1 above, and changes thereto, shall be reviewed by the Operations Committee as indicated in Specification 6.5.1.6 and approved by the Plant Superintendent-Nuclear or designee prior to implementation, except as provided in 6.8.3 below.

6.8.3 Temporary minor changes to procedures described in 6.8.1 above I which do not change the intent of the original procedure may be made with the concurrence of two members of the plant management staff, at least one of whom shall hold a senior operator license.

Such changes shall be documented and promptly reviewed by the '

Operations Committee and by the Plant Superintendent-Nuclear or designee. Subsequent incorporation, if necessary, as a permanent change, shall be in accord with 6.8.2 above. '

6.8-2 AmendnyntNo. 109,126,128,1 0 ,157, JSp.20s . , g 5-O 7 /y5,

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RTS-285A Attachment 3 to

, NG-95-3413 l Page 1 of 2 SAFETY ASSESSMENT I INTRODl!CTION By letter dated December 15,1995, IES Utilities has proposed revisions to the Duane l Amold Energy Center (DAEC) Technical Specifications (TS) to provide administrative I improvements. These changes include correcting erroneous references in the Operating License (OL) and TS Section 3.8.B.2.c, reformatting Section 5.5 on Spent and New Fuel Storage and Sections 6.5 and 6.8 to remove the requirement for Operations Committee review of procedures in support of the Emergency and Security Plans. The current i Surveillance Requirement 4.8.E.2 inappropriately requires demonstration of Emergency l Diesel Generator (EDG) OPERABILITY when one Emergency Service Water (ESW) pump or loop is inoperable. The current Surveillance Requirement,4.5.G.1, requires demonstration of EDG OPERABILITY within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after having found the other inoperable. The proposed revision would require an evaluation of the OPERABLE EDG to verify that it is not inoperable due to a common cause within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and continue to require the demonstration of OPERABILITY every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

l ASSESSMENT The proposed revisions will provide administrative. enhancements to the OL and TS and l the process for certain procedure revisions. No changes will be made to the existing limits on spent or new fuel storage. The previously approved analytical limits on fuel I enrichment, design and quantity of spent fuel assembly storage will be incorporated. The l l proposed revisions are consistent with the Improved Standard TS, NUREG 1433. i Elimination of the requirement to review certain procedures will allow the Operations Committee to concentrate on other issues more pertinent to its function. The procedures implementing the Security and Emergency Plans will still be maintained and any changes will be reviewed by appropriate members of IES staff. This revision is consistent with the guidance provided in NRC GL 93-07.

The changes to Surveillance Requirements correct an inappropriate conditional surveillance and improve another. The revised requirements will still serve to assure OPERABILITY of the affected systems. The current conditional surveillance for ESW I requires demonstration of EDG OPERABILITY. The purpose of any conditional surveillance is to prove that whatever condition or event degraded one division of equipment is not common to the other. The link between the ESW conditional surveillance and the EDG conditional surveillance is erroneous. A condition which makes one division of ESW inoperable would not typically be suspected to make the s

opposite division EDG inoperable. There? ore this conditional Surveillance may be eliminated with no adverse impact. In both cases, for inoperable components, a review is performed of the degradation to determine the likelihood of a similar situation existing in the opposite division. The EDG conditional Surveillance requires that when one EDG becomes inoperable, the other must be tested within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This test is unnecessary

I RTS-235A Attachment 3 to

,. NG-95-3413

. Page 2 of 2 when the OPERABLE EDG can be shown to have not been affected by the condition making the other EDG inoperable. The surveillance would still require a demonstration of OPERABILITY every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

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E RTS-285A Attachment 4 to

, NG-95-3413 Page1ofI ENVIRONMENTAL CONSIDERATION 10 CFR Section 51.22(c)(9) identifies certain licensing and regulatory actions which are eligible for categorical exclusion from the requirement to perform an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of 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 any effluents that may be released offsite; 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 eligibility criteria for categorical exclusion set forth 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 for this determination follows:

Basis The change meets the eligibility criteria for categorical exclusion set forth in 10 CFR Section 51.22(c)(9) for 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 changes are administrative; no physical changes are made to the plant. The proposed changes do not alter any plant parameters, revise any safety limit setpoints or provide any new release pathways. Thus, there will be no change in the types or increase in the amounts of any effiuents that may be released offsite.

3. The proposed changes are administrative; no physical changes are made to the plant. The proposed changes do not alter any plant parameters, revise any safety limit setpoints or provide any new release pathways. Thus, there will be no increase in either individual or cumulative occupational radiation exposure.