ML20059G651

Proposed Tech Specs Consisting of Editorial Changes to Correct Typos & Administrative Errors

ML20059G651
Person / Time
Site:	Duane Arnold
Issue date:	12/22/1993
From:	ILLINOIS POWER CO.
To:
Shared Package
ML20059G642	List: ML20059G651 NG-93-4345, Application for Amend to License DPR-49,requesting Rev to Ts,Per Proposed Change RTS-263 Which Consists of Editorial Changes to Correct Typos & Administrative Errors
References
NUDOCS 9401250016
Download: ML20059G651 (16)

Text

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t DAEC-1

27. FPEOUENCY NOTATION NOTATION TREOLTNCY _

S At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.  !

D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. ,

W At least once per 7 days.  ;

, M At least once per 31 days. i Q At least once per 92 days.-  !

SA At least once per 184 days. I A At.least once per year.

R At least once per 18 months. i S/U Prior to each reactor startup. ,

P Prior to each release.  !

'_ NA Not ap WIR2--69P7TJ.SSIO[Wplicable.

s- v --- ~ W ATER GYSTEMS- 28. DE'I E TED +

w N. --

s A fire suppression water system-shall consis_ta f-a-water source, pumps,.and f

control or isolation valves. Such distribution valves include piping with asso.ciateddet'IEalizing_ifalve-ahead

_ yard-hydra ~nt curb valves, the first of the water flow alar h sprinkler, hose standpipe or deluge system rise 7. ,j 2h. RI b b R Y SYTh$M EbSPONS ~ S b -

Reactor trip system response time is the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor untiil  !

deenergization of the scram pilot valve solenoids.

30. REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section. ,

50.73 to 10 CFR Part 50.  ;

, 31. OFFSITE DOSE ASSESSMENT MANUAL 4

The offsite Dose Assessment Manual (ODAM) contains the methodology and parameters to -[

be used in the calculation of offsite doses:resulting from radioactive gaseous and' i liquid effluents, in the calculation of' gaseous and liquid effluent monitoring alarm / trip setpoints, and in theiconduct of the Radiological Environmental Monitoring Program. The CDAM shall also contain (1) the Radioactive Effluent Controls and~

Radiological Environmental Monitoring program required by Section 6.9.4-and'(2) ,

descriptions of the information that should be included in the Semiannual Radioactive

• Material Release Report and Annual Radiological Environmental Report required by the-Technical Specification 6.11.1.

32. Deleted .

33. PURGE - PURGING ,

PURGE or PURGlHG is the controlled process of discharging air or gas from.a confinement to maintain temperature, pressure, humidity, concentration or-other ,

operating condition, in such a manner.that replacement air or gas is required to '

+

purify the confinement. -

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(- &1250016933ggy p p ADOCK 05000331 t!

pog e

-Amendment No. / DF s 164 1.0-8 RTS-263 I2AD -

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l DAEC-1 )

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34. VENTTNG I

VENTING is the controlleo precash of discharging air er gas from a confinesant i te maintain temperature, pressure, humidity, concentration or other operating j canoition, in such a manner that raplacement air or gas is net provided or '

required during the process. Vent, used in system names, does not imply a l VENTING process,  ;

35. l>_ROCESS CONTROL PROGPAM f PCP.1.

The PROCESS CONTROL PROGRAM shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and pac)sging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to ensure compliance with 10 CTR Parts 20, 61, 71, state regulations, burial ground requirements, and other requiremsnts governing the disposal of solid radioactive waste.

36. MEMBERfS) OT THE PUBLIC MEMBER (S) OT THE PUBLIC are persons who are not occupationally associated with Iowa Electric Light and Power Company and who do not normally frequent the DAEC site. The category does not include contractors, contractor employees, vendors or persons who enter the site to make deliveries or to service equipment.

37. SITE BOUNDARY The Site Boundary is that line beyond which the land is neither owned, nor leased, nor otherwise controlled by IELP. UFSAR Figure 1.2-1 identifies the DAEC Site Boundary. For the purpose of implementing radiological effluent controls, the Unrestricted Area is that land (offsite) beyond the Site Boundary

38. ANNUAL Occurring every 12 months.

For the purpose of designating surveillance test frequencies, annual surveillance tests are to be conducted at least once per 12 months.

39. CORI OPERATING LIMITS REPORT The Core Operating Limits Report is the DAEC-specific document that provides ,

cycle-specific operating limits for the current operating reload cycle. These cycle-specific operating limits shall be dstermined for each reload cycle in accordance with TS 6.11.2. Plant operation within these limits is addressed in individual technical specifications.

40. Shutdown Marcin Shutdown margin is the amount of reactivity by which the reactor is suberitical or would be suberitical assuming all control rods are inserted, except for the analytically stroagest worth control rod, which-is fullyswithd with the core in its most reactive state during the OPERATINGhCys4e4 C YC LE. .

i Amendment No. 109.743,767,780, 184 1.0-9 975-253 JUL 2 2 gy W

DAEC-1

8. Core Themsl Power Limit (Reactor Pressure 1735 psig or Core Flow 110%

of Rated)

At pressures below 7ES psig, the core e+aNation : essure droo (0 power, O flow) is greater than a.56 psi. At low power and 111 flows, this pressure differential is maintained in the bypass region of tne core. Sint.e the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and all flows will always be greater than 4.56 psi.

Analyses show that with a flow of 28 x 101bs/hr 3

bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus,

__ the bundle flow with a 4.56 psi driving head will be greater than 2B x 103 lbs/hr irrespective of total core flow and independent of bundle power for the range of bundle powers of concern. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors this corresponds to a core thermal power of more than 50%. Thus, a core thermal power limit of 25% for reac' tor pressures below 800 psia or core flow less than 10% is conservative. -

C. Power Transient" Plant safety analyses have shown that the scrams caused by exceeding any ..

safety setting will assure that the Safety Limit of Specification 1.1.A or 1.1.8 will not be exceeded. Scram times are checked periodically to a$sure the insertion times are adequate. The themal power transient resulting when a scram is accomp edather than by the expected scram signal (e.g., scram from neutron flux folowing close of the main turbine stop valves) does.not necessarily cause ue damage. However, for this specification a Safety Limit violation will be asstrned when a scram is only accomplished by means of a backup feature of the plant design. The concept of not approaching a Safety Limit provided scram signals are operable is supported by the extensive plant safety analysis.

Amendment No. 119 1.1 6 l RTS - 2 /o3

/2/93

DAEC-1

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2. APRM Hich Flux Scram (Refuel or Startuo & Hot Standby Model For operation in these modes the APRM scram setting of 15 tercent -

of rated power and the IRM High Flux Scram provide adequate thermal margin between the setpoint and the Safety Limit, 25 percent of rated. The margin is adequate to accommodate anticipated maneuvers associated with power plant startup. Effects of increasing '

. pressure at zero or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained to be uniform b ting procedures backed up by , cad the Rod ,

e A Scquence Contr:1 System. Worths of individual rods are very low in ,

a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most ~ probable cause of significant power rise.

Because the flux distribution associated with uniform rod

- withdrawals does not involve high local peaks, and because several  !

rods must be moved to change power by a significant-pen:entage of .

rated power, the rate of power rise is very slow. Generally, the heat flux is near equilibrium with the fission rate. In an assumed -

uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5 percent of rated power per m.inute, and the APRM system would be more than adequate to assure a scram I

before the power could exceed the Safety L'imit. The 15 percent Amendment No.119 1,1 13 g43 263 l n/Y: ,

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DAEC-1 4

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l 3- .LRM,  ;

The IRM system consists of 6 chambers, 3 in each of the reacter protection system logic channels. The IRM is a 5-decade instrument which covers the range of power level between that covered by the SRM and the APRM. The 5 decades are covered by the IRM by means of a

.l range switch and the 5 decades are broken down into 10 ranges, each being one-half of a decade in size. The IRM scram trip setting of i

120 divisions is active in each range of the IRM. For example, if the instrtment were on range 5, the scram would be 120 divisions on that range. Thus, as the IRM is ranged up to accommodate the increase in power level, the scram trip setting is also ranged up. The most I significant sources of reactivity change during the power increase - l are due to control rod withdrawal. For insequence control rod withdrawal, the rate of change of power is slow enough due to the --  !

physical limitation of withdrawing control rods that.the heat flux is

~

in equilibrium with the neutron flux, and the IRM scram would result i l

1 in a reactor shutdown well before any Safuty Limit is exceeded.

In order to ensure that the IRM provides adequate protection against-the single rod withdrawal error, a range of rod withdrawal accidents has been analyzed. This alysis included starting the accident at 1

i e  ;

various power levels; Thi.ostseverecaseinvolvesaninitial l condition in which the reactor is just subcritical and the.!RM system

)

is not yet on scale. This condition exists at quarter rod density.  !

I Additional conservatism was taken in this analysis by asstning that l

RTS-182 1.1-15 RTS-Zb3 01/85

-Amendment No. 120 1

1

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DAEC-1 SAFETY LIMIT LDiITING SAFETY SYSTEM SETTING , i C. Relief Valve Settings - Low-Low Set Function Valve Grouc Ooen Close Low (1 valve) 1020 psig 900 osio High (1 valve) 1025 psig 905 psig.

All settings are : 25 psi.

D. Safety Valve settings l  :

1240 psig)~+ 12 psi (2 valves

2. The shutdown cooling isolation valves snall be closed whenever the reactor vessel dome pressure is 2,135 psig.

Ibe;reatster vessel dome

2. href2re shall not exceed 135 psi a any time when operating the Residual Heat Removal pump in the shutdown cooling mode.

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ATS - 2 63 G/93 l

1.2-2 Amendment No.102.. l

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l DAEC-I l LIrilTING CONDITION FOR OPERATEN ' SURVEILLANCE REQUIREMENT

d. Each control rod shall be couoled d. When a control rod is withdrawn the to its drive.' If a centrei 4d becomes uncoupled.

first time after refueling, after CRD maintenance or when_ required by I (i) recouple the controi -

od within Specification 3.3.A.2.d(ii),

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and coupling integrity shall be (ii) verify couplino by perfctming verified by observing that the drive  ;

surveillance 4.3.A.2.d. does not go to the. overtravel ps n en e rod is My I (iii) If the control rod is not withdrawn, recoupled, declare the centroi rod inoperable. The actions stated in "

Specification 3.3.A.2.e shall be I i taken. >

e. A control rod that has been declared e. (not used) l inoperable for reasons other than being stuck shall:

1 (i) be fully inserted,** and (ii) disarm the associated directional control valves electrically. The control valves may be re-armed to permit testing associated with returning the control rod to OPERABLE status.

-(iii) Whenever the reactor is less than 20% power, verify all inoperable control rods not in >

compliance with BPV5 are separated by 2 or more OPERABLE control rods in any direction, including the diagonal.

(iv) Verify that no more than 8 inoperable control rods exist.

(v) If the requirements of Specification 3.3.A.2.e (i)-(iv) cannot be met, be in COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• This requirement does not apply in the refuel con 41ti n M Specifications 3.9.A. 4 or control rod req (3.9. A. s duri Jirerr6h refueling.

"The RVM may be bypassed, if required, I to allow insertion of inoperable .

control rods and continued operation.

RTS -2/o3 '

12 / 9 3 Amendment No. 101.120,JA2,_ 3.3-2 JM,180

• NAR 111992 -

. ]

i DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT i lC. Residual Heat Removal (RHR1 C. Surveillance of the AHR Service ( l Service Water System Water System i  !

1. 1. Surveillance of the RHR service Except as specifiedhn 3 5.C.2' wa e sys em s all be as follows:

l 3.5.C.3,3.5C.4/~5.5.C.5,'and i 3.5.G.3below,boIh'RRRsevice RHR Service Water Subsystem water subsystem-loops shall be Testing-operable whenever irradiated fuel Itjgg Frecuency is in the reactor vessel and reactor coolant temperature is a. Pump and Motor Once/3 months greater than 2121. operated valve operability,

b. Flow Rate after major Test-Each pump RHR service maintenance water pump and every 3 shall deliver months '

at least 2040 gpm at a TDH cf 610 ft. or more.

2. With one RHRSV pump inoperable, provided the remaining active components of both RHRSW subsystems are verified to be OPERABLE, restore the inoperable pump to OPERABLE status within 30 days or be in at least HOT SHUT 00VN 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 />.

3. With one RMRSW pump in each subsystem inoperable, provided the remaining active components of both RHRSW subsystems and the diesel generators are verified to be OPERABLE, restore at least one inoperable pump to OPERABLE status within 7 days or 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 />.

. RTS -2(o3 db'93 '

3.5-5 Amendment No. J08,J39,J70,174 AUG 121991

DAEC-1 LIMITING CONDITION FOR OPERATION l SURVEILLANCE REQUIREMENT 4 With one RHR5W subsystem inoperable, provided the remaining i RHR5W subsystem and its associated I '

diesel generator are verified to be OPERABLE, restore the inoperable system to OPERABLE status with at least one OPERABLE pump within 7 days or 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 />.

D. HPCI Subsystem D. HpCI Subsystem ,

1. The HPCI Subsystem shall be 1. HPCI Subsystem testing shall be OPERABLE whenever there is P' 'd ** ""

irradiated fuel in the reacter Item Frecuency  :

vessel, reactor pressure is '

greater than 150 psig, and prior a. Simulated Annual i to reactor startup from a COLD A CONDITIO 'fied in A @nTest 3.5.D.scd [ W.5.0.3 3 b !cw. b. Pump Operability Once/3 Months l

c. Motor Operated Once/3 Months ,

Valve Operability  !

d. At rated reactor -Once/3 months pressure demonstrate ability to deliver rated i flow at a discharge l pressure. greater ,

than or equal to that pressure required to accomplish vessel injection if vessel pressure were as high as 1040 psig. ,

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RTS -213 o

11/93 3.5-6 Amendment No. JJ5,J/3.Jf0,174 AUG 12 soo,

T. AEC-1 3.6.F & 4.6.F BASES:

Jet Pump Flow Mismatch The LPC1Aop_aelection logic has been previously described in the Updated FSAR

. 7. 3. 2. .', 2. 4 Section 1.3.1.1.24. jFor some limited 2 cw probability accidents with the recirculation loop operating with large speed differences, it is possible for the 4 logic to select the wrong loop for injection. For these limited conditions the core spray itself is adequate to prevent fuel temperatures from exceeding allowable limits. However, to limit the probability even further, a procedural limitation has been placed on the allowable variation in speed between the recirculation pumps.

The licensee's analyses indicate that above 80% power the loop select logic could be '

expected to function at a speed differential up to 14% of their average speed.  !

Below 80% power the loop select logic would be expected to function at a speed '

differential up to 20% cf their average speed. This specification provides margin  ;

because the limits are set at 110** and $15% of the average speed for the above and below 80% power cases, respectively. If the reactor is operating on one recirculation pump, the loop select logic trips that pump before making the loop selection.

RTS -2 63

/2/93 Amendment No. JJf,183 3,g_33 JUN 2 41992

7 .,

DAEC-1 The pressure suppression pool water provides the heat sink for the reactor i

primary system energy release following a postulated rupture of the system. '

The pressure. suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blowdown from 1040 psig. Since all of the gases in the drywell are purged into the ]

o pressure suppression chamber air space during a loss-of cglant accident, the_ '

I pressure resulting from isothermal compression plus the vapor pressure of the licuid must not exceed 62 psig, the suppression chamber maximum

• allowable pressure. The design volume of the suppression chamber (water and 1

air) was obtained by considering that the total volume of reactor coolant to

, be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

i Using the minimum or maximum water volumes given in the specification, 4

I i

I containment pressure during the design basis accident is approximately 43 l ,

psig which is below the design pressure of 56 psig. The minimum volume of l

58,900 ft' results in a submergence of approximately 3 feet. Based on  ;

i Humooldt Bay, Bodega Bay, and Marviken test f acility data as' utilized in l General Electric Company document number NEDE-21885-P and data presented in_

1 l

Nutech document, Iowa Electric document number 7884-M325-002, the following  ;

j technical assessment results were arrived at:

i

1. Condensation effectiveness of the suppression pool can be maintained for both short and long term phases of the Design Basis Accident (DBA), Intermediate Break Accident (ISA), ~and Small Break i

Accident (SBA) cases with three feet submergence.

R1;S -263 3 7-31 Amendrent No.115 .

11s ' ? 3 w m

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DAEC-1

8. Procedures required by the plant Security Plan.

9. Operation of radioactive waste systems. '

10. Fire Protection Program implementation. ,

11. A preventive maintenance and periodic visual examination program to reduce leakage from systems outside contairnent 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 perfomance of periodic systems leak tests of each system once per operating cycle.

12. Program to ensure the capability to accurately detemine the airborne iodine concentration in vital areas under accident conditions, including training 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. j

15. Process Control

16. Quality Control Program for effluents. j l

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

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6.8.3 Temporary minor changes to procedures described in 6.8.1 above which i do not change the intent of the original procedure may be made with j 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. Subsequent incorporation, if' necessary, as a permanent change, shall be in accord with 6.8.2 above.

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RTS -Zb 3 Amendment NkPRTYM' 12/93 APR 181989

DAEC-1 i

',, (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.

6.11.2 CONE OPERATING LIMITS REPORT

a. Core cycle-dependent limits shall be established prior to each reload cycle, or prior to any remaining part of a reload cycle, for the followings  ;

1) Haximum Average Planar Linear Heat.

Generation Rate (MAPLHGR) -

Specification 3.12.A.

2) Linear Heat Generation Rate (LHGR) -

Specification 3.12.B.

3) Hinimum Critical Power Ratio (MCPR) -

Specification 3.12.C.

4) MAPTAct and MAPTAC, Pacte,rs which multiply the MAPLHGR limits -

Specification 3.3.F.4.a.

These limits shall be documented in the CORZ OPERATING LIMITS REPORT.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A, (CESTAR II).*  !

c. The core operating limits shall be determined such that all applicable limits (gsgt fuel thermal-mechanical limits, core thermal hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limit s ) of the safety analysis are met. I

d. The CORE OPERATING LIMITS RIPORT, including any  !

mid-cycle revisions or aupplements shall be  !

provided upon issuance, for each reload cycle, to ,

the NRC Document Control Desk with copies to the {

Regional Administrator and Resident Inspector.

6.11.3 UNIOUE REPORTING REOUIREMENTS Special reports shall be submitted to the Director of Inspection and Enforcement Regional Office within the time period specified for each report. These reports shall be submitted civering the activities identified below pursuant to the recuireaants of the applicable reference i specification.

a. Reactor vessel base, weld and heat affected zone metal test specimens (Specification 4.6.A.2).

b. deleted 1

c. Inservice inspection (Specification 4.6.G. ) .

a

d. Reactor C,antainment Integrated Leakage Rate Test (Specification 4.7.A.2(f)  ;

3 '

" Approved revision number at time reload fuel analyses are performed.

.! Amendment No. 709.728.767.770.164 6.11-4 RT~C-263 i

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n/92 -

JUL 2 21992  :

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DAEC-1

e. deleted l f. deleted

g. deleted

h. Radioactive Liquid or Gaseous Effluent - calculated dose ,

exceeding specified limit (oDAM Sections 6.1.3, 6.2.3 and 6.2.4).

i. Off-Gas System inoperable (ODAM Section 6.2.5).

l

j. Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level a values of ODAM Table 6.3-3 when avera any calendar  !

quarter sampling period (ODAM Section 7.3.2.2 .

(. 3 2.2 .,

k. Annual dose to a MEMBFN OF THE P 'LIC e m ned to exceed a 40 CFR Part 190 dose 12.mit (oDAM Section 6.3.1.1).

1. Radioactive liquid waste h a,ser",without treatment when activity concentratic exceed: (p QLp,cif4tQD,AQection c - ,

6 1 4 1)- ts equal to er greater than

m. Explosive Gas Monitoring ns N entation Anw u (Specification 3.2.I.1).

~

l n. Liquid Holdup Tank Instrumentation Inoperable (Specification 3.14.B.1).

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R TS-2 b 3 i (2/93 . -

i Amendment No. 190 6.11 5 l if0Y53is92

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.; ~ i l-H RTS-263 Attachment 3-to .

NG-93-4345 I Page 1 l SAFETY ASSEPSMENT )

Introduction  ;

By letter dated December 22, 1993, Iowa Electric Light and Power '

Company (IELP) submitted a request for revisionlif the Technical  !

Specifications (TS) for the Duane Arnold ~ Energy Center (DAEC).  !

The proposed changes are editorial in nature and will eliminate j discrepancies discovered in the recent comparison.of the IELP j controlled copy with the NRC authority copy of the DAEC TS. '

Assessment The proposed changes to the DAEC TS correct administrative and  :

typographical errors. None of the revisions would result in a l

change to the systems, structures or components in the plant or operation-thereof. .The revised TS will result in continued safe' '

operation of the DAEC and elimination of some administrative and typographical errors in the TS~. .

Based upon the above information, we have concluded that the proposed changes to the DAEC TS are acceptable. d l

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l RTS-263 Attachment 4'to i NG-93-4345  !

.Page 1 ,

l ENVIRONMENTAL CONSIDERATION  !

10 CFR 51.22(c)(9) identifies certain licensing and regulatory actions  !

- which are eligible for categorical exclusion from the requirement to r perform an environmental assessment. A proposed amendment to.an operating f

'llcense for a facility requires no environmental assessment if operation of j the facility in accordance with the proposed amendment would not: . (1)  ;

involve a significant hazards consideration; (2) result in a significantL -!

. 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 individuali  ;

or cumulative occupational radiation exposure. Iowa Electric Light and Power has reviewed this request and determined that the proposed. amendment

. meets the eligibility criteria for categorical exclusion set forth in 10 i

CFR'51.22(c)(9). Pursuant to 10 CFR 51.22(b),.no environmental impact- l

- statement or environmental assessment needs to De prepared in connection j with the issuance of the amendment. The basis for this determination  ;

follows: 'j Basis i;

'i The change meets the eligibility criteria for categorical exclusion set  ;

forth in 10 CFR 51.22(c)(9) for the following reasons: {

l

1. As demonstrated in Attachment 1 to this letter, the proposed Amendment-  ;

does not involve a significant hazards consideration.  !

i

2. The proposed Amendment makes editorial changes to correct  ;

administrative and typographical errors. No changes in.either design or operation of the DAEC will be made as a result of these changes; 'j thus, there will be no increase-in either the types or amounts of ,

effluents that may be released offsite.  :

l 1

3. The proposed Amendment makes editorial changes to correct  :

administrative and typographical errors. No changes in either design  !

or operation of the DAEC will be made as a result of these changes; thus, there will be no significant increase in either individual'or j cumulative occupational exposure.  !

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