ML20063B946

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Proposed Tech Specs Changing Ielp to Ies Utilities,Inc & Title Manager-Nuclear Div to Vice President,Nuclear
ML20063B946
Person / Time
Site: Duane Arnold NextEra Energy icon.png
Issue date: 01/21/1994
From:
IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT
To:
Shared Package
ML20063B943 List:
References
NUDOCS 9402030241
Download: ML20063B946 (20)
v • d • e


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34.

VENTING VENTING is the controllec procesh of discharging air er gas from a confinemar.:

te maintain temperature, pressure, humidity, concentration or other operating concition, in such a manner that replacement air or gas is not provided or required during the process. Vent, used in systaa names, does not imply a VENTING process.

35.

P_R_OCESS CONTROL PROGRAM fPCP)

The PROCESS CONTROL PROGRAM shall contain the current formulas, samplir.g, analyses, tests, and determinations to be made to ensure that processing and pack 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 regulat, ions, burial ground requiraments, and other requirements governing the disposal of solid radioactive waste.

36.

BXMBERfS1 OF THE PUBLIC MBER(S) OF THE PUBLIC are p_erson o are not occupationally associated with g514bTMt1LCdE G E h Ot M PerrrW ~=" nd who do not normally frequent the DAEC

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@danrt contractors, contractor employees, vendors, or persons who enter the site to make deliveries or to service equipment.

37.

SITE BOUNDARY ICS Me hW8J I" The Site Boundary is that line beyo which the land is neither owned, nor leased, nor otherwise controlled by

_c) UFSAR Figure 1.2-1 identifies the DAIC 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.

CORE 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 determined 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 strongest worth control rod, which is fully withdrawn, with the core in its most reactive state during the OPERATING Cycle.

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Sodium Pentaborate Solution Volume Concentration Requirements FIGURE 3.4-1 RT.S-245 0/ / i'l -

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3.4-9 Amendment No.151 01/94 JUL.i B6E 4

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

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

The pressure suppression chamber water volume must absorb the associated 2

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decay and structural sensible heat released during primary system blowdown l

from 1040 psig. Since all of the gases in the drywell are purged into the l

i pressure suppression chamber air space during a loss-of-colant accident, the

.i pressure resulting from isothermal compression plus the vapor pressure of

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the liquid must not exceed 62 psig, the suppression chamber maximum allowable pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to

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be condensed is discharged to the suppression chamber and that the drywell i

volume is purged to the suppression chamber.

Using the minfmum or maximum water volumes given in the specification.

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containment pressure during the design basis accident is approximately 43 l

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psig which is below the design pressure of 56 psig. The minimtsn volume of 4

5 58,900 ft' results in a submergence of approximately -3 feet. Based on Humboldt Bay, Bodega Bay, and Marviken test f acility data as utilized in j

General Electric Costany do.cument-p=her NEDE-21885-P and data presented in Nutech document, towa-E_-leettfe-poetmei)tt number 7884-M3

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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 (IBA), and Small Break Accident (SBA) cases with three feet submergence.

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Amendment Nie.10tter dated 8/4/933,s.6a f.orrection RT5-2ta5 Of /94

i DAEC-l' 5.0 DESIGN FEATURES R

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5.1 SITE The Duane Arnold Energy Center site is located on the western side of a north-south reach of the Cedar River, approximately 2-1/2 miles north-j i

northeast of the Village of Palo, Iowa.ys.J:9MI f'

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approximate 1L500 acresgedJyfth Ne Ostric Lig.:

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(C e plan of the site is shown on Figures 1.2-1 and 1.2-2 of the Updated FSAR. The minimum distance to the boundary of the i

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.. exclusion area as defined in 10 CFR 100.3 is approximately 1000 feet.

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6.0 ADMINISTRATIVE CONTROLS 6.1 MANAGEMcNT - t,UTHORITY ann RESPONSIBILITY

' 6.1.1' The Plant Superintendent-Nuclear has prima s

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e operation of the DAEC, and reports to r.;r.;;;; " -'n-Divu Vlce PraSidenf hfulleM.

p fire erotection % gfirTEr DAEC is-7-Wr14-ThrsveT1r1T~re.es4bt13@" ;.;;r DiiFi;;;r,? The DAEC Plant f 1/t assignedtoth7"eee.

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pWrtprendent-Nuclear is nap)RITBTE Tift directing the' operating piar.t v

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fire protection program.

6.1.3 The Manager, Corporate Quality Assurance is responsible for i i implementation of the Quality Assurance Program at the DAEC.

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Amendment No. pp,U S,166 6.1-1 RT.5-2 lob JUI.6 1990 01/94 1

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6.2 ORGANIZATION I'

ti. 2.1 ONSITE AND OFF5ITE ORGANIZATIONS I

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Onsite and offsite organizations shall be established for unit oper: tion

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and corporate management, respectively.

The onsite and offsite organizations shall include positions for activities affecting tho';:.tety of the nuclear power plant.

Lines of authority, responsibility, and communication shall be a.

establisheo and defined from the highest management levels through

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intermediate levels to and including all operating organization positions.

These relationships shall be documented and updated, as appropriate in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and Job descriptions for key personnel positions, or in equivalent-forms of documentation.

These requirements shall be documented in the Duane.

Arnold Energy Center Updated Final Safety Analysis Report and updated in accordance with 10 CFR 50.71(e).

b.

The Plant Superintendent-Nuclear shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

V:ce Peesided, duelear c.

The.ansger-k hall have corporate responsibility for overa M nt nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

i d.

The individuals who train the operating staff and those who carry out health physics and quality assurance-functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

i Amendment No. 69,D E,166 6.2-1 RT5 - 2 L,5 i

JUL 6 1990 i

OAEC-1 Investigation of all violations of the Technical Specifications e.

including the preparation and forwarding of repo noevaluation qd recommendations to prevent recurrence to the

_n: c:-N_ clear L,JaE and to the Chairman of the. Safety Commi e,,f, A,Ja,-

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evi w of all Reportable Events.

g.

Review of facility operations to detect potential safety hazards, b.

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

t, Review of the Plant Security Plan and implementing procedures.

Revtew of the Emergency Plan and implementing procedures.

k.

Review of every unplanned release of radioactivity to the environs for wnich a report to the NRC is required.

Review of changes to the Offsite Dose Assessment Manual and changes to the Process Control Program.

l m.

Review of the Fire Protection Program and implementing procedures.

6.5.1.7 Authority The Operations Committee shall:

a.

Recommend to the Plant Superintendent-Nuclear written approval or disapproval of items considered under Specification 6.5.1.6 (a) through (d) above.

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R T.5 - ?. 65 Amendment No.190 6.5 3 OI/94

DAEC-1 4,

Rencer determinations in writing with regara to whetner or not eacn item b.

considereo unoer 6.5.1.6 (a) througn (e) acove constitutes an unreviewea safety question.

Vies President Wuelear l

c.

Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Mar.cger 'Nci,ccr Divisica-and the Safety Comittee of disagreement between the Operations Comittee and the Plant Superintendent-Nuclear; -however, the Plant Superintendent-Nuclear shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1 above.

6.5.1.8 Recoro The Operations Comittee shall maintain written minutes of each meeting and copies 9

l snall be provided to the and the Chairman of. the Safety-Comitt ee.

6.5.2 Safety Comittee 6.5.2.1 Function The Safety Comittee shall function to provide independent review and audit of designated activities in the areas of:

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e a.

Nuclear power plant operations.

b.

Nuclear Engineering, 6.5-4 RTS - Zld Amendment No. S9, 136 OlN4

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6.6 -

REPORTABLE EVENT ACTION

- 6.6.1 The following actions shall be taken for REPORTABLE. EVENTS.

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a.

Each REPORTABLE EVENT shall be reviewed by the Operations Comittee, an{_eygshagsubmitted to the Safety Cassnittee and.the

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b.

The h be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50.

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Amewheent No. !$5,136 6.6 1

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6.7 ACTION TO 8E TAKEN IF A SAFETY 1.!MIT 15 EXCEEDED-6.7.1 If 'a safety limit is exceeded, the reactor. shall be' shut down and reactor operation shall only be resumed een authorized by the letc..

Vica Presiderst, Maelet.r 6.7.2 An immediate report shall be a.

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Vice Pr esidenh Wae. fare and the Safety Committee. The m _,,.. _ _.nr Oivi;ir, shall l

promptly report the circumstances to the Mtc as specified.in' Subsection 6.11, plant Reporting Requirements.

l 6.7.3 A complete analysis of the circunstances leading up to and resulting -

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i from the situation together with mecamendations to prevent a recurrence shall be prepared by MCnsittee. This '

Vice President, dadere report shall be submitted to

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t Safety cannittee. Vice Pr@esident, Haa. lear p

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Subsection 6.11. Plant Reporting Requirmeents.

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Amendment so. !!i.136 5.7 1

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RTS-265 to NG-94-0018 Page 1 SAFETY ASSESSMENT Introduction By letter dated January 21, 1994, IES Utilities Inc. submitted a request for revision of the Operating License and Technical-Specifications for the Duane Arnold Energy' Center (DAEC).

The proposed changes revice the name of the corporation authorized to own l'

a share of.and operate the DAEC from Iowa Electric Light and Power Company (IELP) to IES Utilities Inc. and riange the title of Manager-Nuclear Division to Vice President, Nuclear.

Assessment IES Industric e owned two electric ut j subsidiaries:

Iowa Electric Light and Po. r Company and Iowa Southern Utilities Company (Iowa Southern).

Effective January 1, 1994, Iowa Southern was merged.into IELP and the corporation was renamed IES Utilities Inc.

Under this corporation, the Nuclear Division will continue as before.

The.DAEC Operating License and Technical Specifications must be revised to reflect the new name of the corporation.

The change in position title results from elimination of a layer of management.

Formerly, the Nuclear Division was headed by the Manager-Nuclear Division, who reported to the Vice President, Production who in turn reported to the President of IELP.

Now the Nuclear Division is headed by the Vice President, Nuclear who reports directly to the President of the corporation, l

The changes will not result in any change in the operation of the DAEC.

The merger of the two utilities results in a stronger, more dynamic company which can respond readily to a rapidly changing

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industry while assuring the continued safe operation.of the DAEC.

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i RTS-265 'to NG-94-0018 Page 1 t

t ENVIRONMENTAL CONSIDERATION l

10 CFR 51.22(c)(9) identifies certain licensing and regulatory actions-i 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) q 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 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 for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 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 i

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

1.

As demonstrated in Attachment 1 to this letter, the' proposed Amendment does not involve a significant hazards consideration.

2.

The proposed Amendment includes changes to the operating company name and a title.

No changes in either plant design or i

operational strategies will be made as acresult of these changes; thus, there will be no increase in either the types or amounts of effluents that may be released'offsite.

1 3.

The proposed Amendment includes changes to'the operating company.

name and a title.

No changes in either plant' design or operational strategies will be made as a result of these changes; thus, there will be no significant increase in either individual or cumulative occupational exposure.

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