ML20029E632
| ML20029E632 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 05/13/1994 |
| From: | Pulsifer R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20029E633 | List: |
| References | |
| NUDOCS 9405190221 | |
| Download: ML20029E632 (28) | |
Text
,
bf/
[
h UNITED STATES
((!kfj j NUCLEAR REGULATORY COMMISSION
<; t{Qg:
,f WASHINGTON, D.C. 20555-0001 IES 9TILITIES INC.
CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET NO. 50-331 DUANE ARN0LD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 198 License No. DPR-49 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by IES Utilities Inc., et al.,
dated January 21, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that tht activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, Facility Operating License No. DPR-49 is amended to reflect -
the operating company name change in the title, l. A, l.E, 2, 2. A, 2.B.(1), 2.B.(2), 2.B.(3), 2.B.(4), 2.B.(5), 2.C.(1), 2.C.(3) and correct a typographical error in 2.B.(4).
9405190221 940513 PDR ADOCK 05000331 P
' Revise the title to read:
IES UTILITIES INC.
CENTRAL 10WA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET 50-331 DUANE ARNOLD ENERGY CENTER FACILITY OPERATING LICENSE Revise paragraph 1.A to read:
The application for license filed by IES Utilities Inc.,
Central Iowa Power Cooperative and Corn Belt Power Cooperative (the licensees) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I and all required notifications to other agencies or bodies have been duly made; Revise paragraph 1.E to read:
IES Utilities Inc. is technically qualified and the licensees are financially qualified to engage in the activities authorized by this operating license in accordance with the rules and regulations of the Commission; Revise paragraph 2 to read:
Facility Operating License No. DPR-49 is hereby issued to IES Utilities Inc. (IES), Central Iowa Power Cooperative (CIPC0) and Corn Belt Power Cooperative (Corn Belt) to read as follows:
Revise the first sentence of paragraph 2.A to read:
This license applies to the Duane Arnold Energy Center, a boiling water reactor and associated equipment (the facility), owned by the licensees and operated by IES Utilities Inc.
' Revise paragraph 2.B.(1) to read:
IES Utilities Inc., pursuant to Section 104b of the Act and 10 CFR Part 50, " Licensing of Production and Utilization Facilities," to possess, use, and operate the facility; and CIPC0 and Corn Belt to possess the facility at the designated location in Linn County, Iowa, in accordance with the procedures and limitations set forth in this license; Revise paragraph 2.B.(2) to read:
IES Utilities Inc., pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended as of June 1992 and as supplemented by letter dated March 26, 1993; Revise paragraph 2.B.(3) to read:
IES Utilities Inc., pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; Revise paragraph 2.B.(4) to read:
IES Utilities Inc., pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated radioactive apparatus components; 1.
1
Revise paragraph 2.B.(5) to read:
IES Utilities Inc., pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not to separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
Revise paragraph 2.C.(1) to read:
IES Utilities Inc. is authorized to operate the Duane Arnold Energy Center at steady state reactor core power levels not in excess of 1658 megawatts (thermal).
Revise the first paragraph of 2.C.(3) to read:
(3)
Fire Protection IES Utilities Inc. shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the Duane Arnold Energy Center and as approved in the SER dated June 1, 1978, and Supplement dated February 10, 1981, subject tu the following provision:
(2)
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 198, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendment is effective as of the date of issuance and shall be implemented within 60 days of the date of issuance.
FOR T E NUCLEAR R LATORY COMMISSION fm/
t' dl~
/
' Robert M. Pu sifer[ Project Manager _
Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: May 13, 1994
ATTACHMENT TO LICENSE AMENDMENT'NO.198 FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 P
Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.
The revised areas are indicated by marginal lines.
Remove Insert Appendix A Cover Sheet Appendix A Cover Sheet v
v 1.0-9 1.0-9 1.0-10 1.0-10 1.1-11 (Repositioned) 1.1-19 1.1-19 1.1-20 1.1-20 Figure 4.1-1 3.1-18 Figure 4.2-2 3.2-51 3.3-16 3.3-16 3.4-8 3.4-8 3.4-9 3.4-9 3.7-31 3.7-31 3.8-7 3.8-7 5.1-1 5.1-1 6.1-1 6.1-1 6.2-1 6.2-1 6.2-2 6.2 6.2-4 6,5-2 6.5-2 (Repositioned) 6.5-3 6.5-3 (Repositioned) 6.5-4 6.5-4 6.6-1 6.6-1 6.7-1 6.7-1 as a
~
APPENDIX A TO OPERATING LICENSE DRP-49 TECHNICAL SPECIFICATIONS AND BASES i
FOR DUANE ARNOLD ENERGY CENTER IES UTILITIES INC.
CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET NO. 50-331 FEBRUARY 1974 AMENDMENT NO.198
DAEC-1 TECHNICAL SPECIFICATIONS LIST OF TABLES Table Number Title Paae 1.0-1 Operating Modes 1.1-11 3.1-1 Reactor Protection System Instrumentation 3.1-3 3.1-2 Protective Instrumentation Response Times 3.1-7 4.1-1 Reactor Protection System Instrumentation 3.1-8 Surveillance Requirements 4.1-2 Deleted 3.2-A Isolation Actuation Instrumentation 3.2-3 4.2-A Isolation Actuation Instrumentation Surveillance 3.2-8 Requirements 3.2-B Core and Containment Cooling Systems 3.2-12 initiation / Control Instrumentation 4.2-B Core and Containment Cooling Systems 3.2-17 Initiation / Control Surveillance Requirements 3.2-C Control Rod Block Instrumentation 3.2-21 4.2-C Control Rod Block Instrumentation Surveillance 3.2-23 Requirements 3.2-D Radiation Monitoring Instrumentation 3.2-26 4.2-D Radiation Monitoring instrumentation Surveillance 3.2-27 Requirements 3.2-E Drywell Leak Detection Instrumentation 3.2-29 4.2-E Orywell Leak Detection Instrumentation Surveillance 3.2-30 Requirements 3.2-F Surveillance Instrumentation 3.2-32 4.2-F Surveillance Instrumentation Surveillance 3.2-33 Requirements 3.2-G (ATWS) RPT/ARI and E0C-RPT Instrumentation 3.2-35 (ATWS' RPT 3.2-36 Surve 11anc/ARI and E0C-RPT Instrumentatinn 4.2-G e Requirements 3.2-H Accident Monitoring Instrumentation 3.2-38 ~
4.2-H Accident Monitoring Instrumentation Surveillance 3.2-41 Requirements AMENDMENT NO. }')f, }7)', }77, }W.198 v
DAEC-1 34.
VENTING VENTING 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 not provided or required during the process.
Vent, used in system names, does not imply a VENTING process.
i 35.
The PROCESS CONTROL PROGRAM shall contain the current formulas, sampling, analysis, tests, and determinations to be made to ensure that.
processing and packaging 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 CFR Parts 20, 61, 71, state regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.
36.
MEMBER (S) 0F THE PUBLIC MEMBER (S) 0F THE PUBLIC are persons who are not occupationally associated with IES Utilities Inc. 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 IES Utilities Inc.
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.
l 38.
ANNUAL Occurring every 12 months.
For the purpose of designating surveillance test frequencies, ANNUAL l
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.
AMENDMENT No. J09,1f3,JE7, 1.0-9 JE0,JE A, 198
DAEC-1 40.
SHUTDOWN MARGIN SHUTDOWN MARGIN is the amount of reactivity by which the reactor is l
subcritical or would be subtritical 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.
l AMENDMENT No. A93,198 1.0-10
DAEC-1 TABLE 1.0-1 OPERATING MODES REACTOR MODE SWITCH AVERAGE REACTOR COOLANT OPERATING MODE POSITION TEMPERATURE 1.
RUN/ POWER OPERATION Run NA 2.
STARTUP Startup/ Hot Standby or NA Refuel (*)
3.
HOT SHUTDOWN (*)
Shutdown"'""
> 212*F 4.
COLD SHUTDOWN (*)
Shutdown")""')
s 212*F 5.
REFUELING"')
Shutdown or Refuel")")
NA (a) fuel in the reactor vessel with the reactor vessel head closure bolts fully tensioned.
(b) Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
(c) The reactor mode switch may be placed in the Run, Startup/ Hot Standby or Refuel position to test the switch interlock functions and related instrumentation provided that the control rods are verified to. remain fully inserted by a second licensed operator.
(d) The reactor mode switch may be placed in the Refuel position while a single control rod is being recoupled or withdrawn provided that the one-rod-out interlock is OPERABLE.
(e) The reactor mode switch may be placed in the Refuel position while a single control rod drive is being removed from the reactor pressure vessel per Specification 3.9.A.
(f) The reactor mode switch may be placed in the Startup position for i
demonstration of shutdown margin per Specification 4.3. A.I.
l 4
AMENDMENT N0.198 1.0-11
m._
I DAEC-1 12 e'
/
= ATie ro-i 1
4/
APRM SCR AM e 0.new + 52 RATID FLQw e og wigm,
. M.... _. 0_.
I
+
/
/
/
(100.87)
=
/
/
^
I B
{
l E
l c
5 ME TdW AL powgR l
SH Actc Ang A iNeicA;gg AEGICN EXPANDED SY 4c LIMif WEN REACTOR a
PRES'iURE 15 < 785 J
EXTENDED LOAD LINI WIT ANALY313
~
PSIC OR CORE rLCW f
< 1!8 OF RATED i
~
j 25 -..J-.
/
mATLR AL CIRCLJLATION x
/
/
/
/
/
/
?
9 9
0 0
20 A0 60 80 10C CCAE PLOW ts C# RATI:1 3
DUANE ARNOLD ENERGY CENTER IES-UTILITIES INC.
TECHNICAL SPECIFICATIONS-APRM FLOW BIAS SCRAM ~ RELATIONSHIP TO NORMAL OPERATING CONDITIONS FIGURE 1.1-1 Amendment No. 720,198 1.1-19 i
.,.m v-
~v.---
. ~.
~-
DREC-1 4
/
NCauaL
- APmu senau TatPg ssNcLe Loop
- Atmu semau TmiP a[j
/
~
90 r
f/
/
/
MAL
- APmu m00 stoCE Tal*
stNC Loop
- APau m00 B CCE TmiP n.
50 -
b 50 -
0:
w 3
40 r
y 30 CU 20 10 0
0.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 80.0 90.0 100.0110.0120.0 W = RECIROULATION LOOP FLOW (:". RATED)
DUANE ARNOLD ENERGY CENTER g
IES UTILITIES INC.
l TECHNICAL SPECIFICATIONS CORE POWER Vs RECIRC LOOP FLOW FIGURE 2.1-1 Amendment No. J20,198 1.1-20
~ -
-. ~..
DAEC-l'
'o.
6 p
Uj i
C; I
i e
I 6
\\
6 2
L C'
r s
t Z
E e
Oe c 0 o<
k 2
e s.
Z 2 /
C 2
!. L 2E c-s 4.
i L <
02 I.-
E :'
6
- 2 i
- F 1
2 I
- 5 1
5*
i' i
c I
)
6
- 2c" j
j l
l l
c*
1 1
I I
C E
8 N
O
,c g.
2 S 3 en"11v4 3JvsNO JO b38WAN DUANE ARNOLD ENERGY CENTER IES UTILITIES INC.
TECHNICAL SPECIFICATIONS
. FAILURE HISTORY
-l FIGURE.4.1-1 Amendment NO.I98 3.1-18
l DAEC-I 10
-2 10 CURVE 1
[
~3 10
/
C 5
yURVE 3 h
~#
10 K
7 L
t-t/
4
_e t
10 -
O h
CURVE 2 E
10 CIr I
L CURVE 4S N CURVE 5 10~
I 2
d 5
10 10 10 10 10 to
^
TEST INTERVAL - (5) HOURS DUANE ARNOLD ENERGY CENTER IES UTILITIES INC.
TECHNICAL SPECIFICATIONS CHANNEL AVAILABILITY FIGURE 4.2-2 RMENOMEN! No.198 3.2-51 l
/
90 F
8005 LSAD
/
LDE r
g 80 A
///ssss
. 80 O
yo 50 ac nAD 50
.0 M4 4
_J
'/
l 395 g
I 30 Flow C
3 W
5 G
20 I
I W
l C
$0 C
/
NATURAL I
f!A::tLAT!DN O
10 20 30 40 50 60 70 80 90 100 %
45S TOTAL CORE FLOW
(% of RATED)
Regien I Two Loop Surv. Region, SLO Prohibited Region repires APM/LPRM noise monitoring Region 2:
Two Loop & SLO Sury. Region requires APRM/LPRM noise monitoring Region 3:
SLO Surv. Region
- requires APRM/LPRM & Core Plate D/P noise monitoring Region 4:
Extended SLO sury. Region requires Core Plate D/P noise monitoring Region 5:
Unrestricted Two Loop & SLO Region DUANE ARNOLD ENERGY CENTER l
IES UTILITIES INC.
TECHNICAL SPECIFICATIONS THERMAL POWER Vs CORE FLOW LIMITS FOR THERMAL HYDRAULIC STABILITY SURVEILLANCE FIGURE 3.3-I u.EreMENT 30. 179,120,780,7g3,198 3.3-I6
DAEC I g
2$
r 1
i i i
r e
i i i i i i
~
Constant Boron Mass Cunne 20 (750 com Ecuivalent)
Overflow vol
~
Y g
w Region of Allowable 3
Vol Conc-A i
L w
j li --
r r
9.i
~
t Ei r
c
.g e0 25 2 g an Pumo Rate Qg
>=
j g4 p
n1a*.i
~
45
~
2a 10 1
i t-L
~
L j
P 1
L L
t I
S i
+
10CC 1$30 2000 2500' 3000 m
Net Tana Volume (980 DUANE ARNOLD ENERGY CENTER IES UTILITIES INC.
i TECHNICAL SPECIFICATIONS SODIUM PENTABORATE SOLUTION VOLUME CONCENTRATION REQUIREMENTS.
FIGURE'3.4-1 Amendment No. J57,198 3.4-8
f DAEC-1 i.
90 J
M 80 Repon of Allowacle Solution Temperature 3
Minimum Solution E*
E Temperature Curve 4
[
(Sat, Temp.. $')
5 I
w F
l e
i E --
g-___
155 E
- F l
r t
1 l
I P
50 -l l
i i
t-I i
f I
l I'
, m 8 8.1
+
I 4-10 12 14 16 18 S::+m Pentaterate Solution (Per:ent ey weig-ti DUANE ARNOLD ENERGY CENTER IES UTILITIES INC.
~
TECHNICAL SPECIFICATIONS
.-i MINIMUM TEMPERATURE OF j
SODIUM PENTABORATE SOLUTION FIGURE 3.4-2 i
Amendment No. J5J,198 3.4-9 1
DAEC-1 The pressure suppression pool water provides the heat sink for the reactor 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 pressure suppression chamber air space during a loss-of-coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of 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 be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.
Using the minimum or maximum water volumes given in the specification, containment pressure during the design basis accident is approximately 43 psig which is below the design pressure of 56 psig.
The minimum volume of 3
58,900 ft results in a submergence of approximately 3 feet.
Based on Humboldt Bay, Bodega Bay, and Marviken test facility data as utilized in General Electric Company document number NEDE-21885-P and data presented in Nutech document, IES Utilities Inc. document number 7884-M325-002, the following technical assessment results were arrived at:
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 (IBA), and Small Break Accident (SBA) cases with three feet submergence.
Amendment No. UE,198 3.7-31
DAEC-1 2200
/
x 1100 8
?1 Eg 1000 C
b 900 0
I b
m X
g 800
/
E, b
3 ao 700 600 bd 5
N 80 85 90 95 River Water Temperature - Degrees F DUA'.E' ARNOLD ENERGY CENTER IES UTILITIES INC.
l TECHNICAL SPECIFICATIONS DAEC EMERGENCY SERVICE WATER FLOW REQUIREMENT FIGURE 4.8.C-1 AMENDMENT I40. /0,198 l
DAEC-1 5.0 DESIGN FEATURES 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 nort-northeast.of the village of Palo, Iowa. The site consists of approximately 500 acres owned by IES Utilities Inc.
The plan of the site is shown on Figures 1-.2-1 and 1.2-2 of the Updated FSAR.
The minimum distance to the boundar/ of the exclusion area as defined in 10 CFR 100.3 is approximately 1000 feet.
Amendment No. JJ A,198 5.1-1
.- -~
DAEC-1 6.0 ADMINISTRATIVE CONTROLS 6.1 MANAGEMENT - AUTHORITY AND RESPONSIBILITY 6.1.1 The Plant Superintendent-Nuclear has primary responsibility for the safe operation of the DAEC, and reports to the Vice President, Nuclear.
6.1.2 The overall responsibility for the fire protection program for DAEC is assigned to the Vice President, Nuclear. The DAEC P.lant Superintendent-Nuclear is responsible for directing the operating plant fire protection program.
6.1.3 The Manager, Corporate Quality Assurance is responsible for implementation of the Quality Assurance Program at DAEC.
Amendment No. ED,736,766,198 6.1-1 l
DAEC-1 6.2 ORGANIZATION 6.2.1 ONSITE AND OFFSITE ORGANIZATION Onsite and offsite organizations shall be established for unit operation and corporate management, respectively.
The onsite and offsite organizations shall include positions for activities affecting the safety of the nuclear power plant.
a.
Lines of authority, responsibility, and communication shall be established and defined from the highest management levels through 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 descrf.tions 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
'e plant.
The Vice President, Nuclear shall have corporate responsibility for c.
overall plant nuclear safety and shall take.,any measures needed to ensure acceptable performance of the staff in operating, Amendment No. 69,72E.JE6,198 6.2-1
~.
DAEC-1 l
maintaining, and providing technical support to the plant to ensure nuclear safety, 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 manger; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.
6.2.2 PLANT STAFF ORGANIZATION The following manning requirements shall be met:
1.
All CORE ALTERATIONS 'shall be directly supervised by either a Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.
2.
At all times when there is fuel in the reactor:
a.
A senior reactor operator shall be on the plant site.
b.
A reactor operator shall be in the control room.
Two reactor operators shall be in the control room during startup, c.
scheduled shutdown, and during recovery from trips caused by transients'or emergencies.
d.' Minimum operating shift crew compositions shall conform to those shown in Table 6.2-1.
At least one member of each operating shift crew shall be qualified e.
to implement radiation protection procedures.
Amendment fio. 190,198 6.2-2
DAEC-1 6.5.1.4 Meeting Frequency The Operations Committee meet at least once per calendar month and as convened by the Operations Committee Chairman or.Vice Chairman.
6.5.1.5 Quorum A quorum of the Operations Committee shall consist of the chairman or Vice Chairman and five members including alternates.
6.5.1.6 Responsibilities The Operations Committee shall be responsible for:
a.
Review of (1) all procedures required by Specification 6.8, Plant Operating Procedures, and changes thereto, (2) any other proposed procedures or changes thereto as determined by the plant Superintendent-Nuclear to affect nuclear safety.
b.
Review of all proposed tests and experiments that affect nuclear safety.
Review of all proposed changes to the Technical Specifications.
c.
d.
Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.
Investigation of all violations of the Technical Specifications e.
including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Vice President, Nuclear and to the Chairman of the Safety Committee.
Amendment No. E0,JE7,198 6.5-2 r--
v v
r-
= - - -.
+
DAEC-1 f.
Review of all Reportable Events.
g.
Review of facility operations to detect potential safety hazards, h.
Performance of special rev'ews, investigations or analyses and reports thereon as requested by the Chairman of the Safety Committee, i.
Review of the Plant Security Plan and implementing procedures.
j.
Review of the Emergency Plan and implementing procedures.
k.
Review of every unplanned release of radioactivity to the. environs for which a report to the NRC is required.
1.
Review of changes to the Offsite Dose Assessment Manual and changes to the Process Control Program.
l m.
Review of Se Fire Protection Program and implementing procedures.
1 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.
Amendment No.J90,198-6.5-3
4 DAEC-1 b.
Render determinations in writing with regard to whether or not each item considered under 6.5.1.6 (a) through (e) above constitutes an unreviewed safety question.
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 Vice President, Nuclear and the Safety Committee of disagreement between the Operations Committee 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 Record The Operations Committee shall maintain written minutes of each meeting and copies shall be provided to the Vice Present, Nuclear and the Chairman of the Safety Committee.
6.5.2 Safety Committee f
6.5.2.1 Function The Safety Committee shall function to provide independent review and audit of disignated activities in the areas of:
^
a.
Nuclear power plant operations.
b.
Nuclear Engineering.
i Amendment No.20,J36198 6.5-4
DAEC-1 6.6 REPORTABLE EVENT ACTION 6.6.1 The following. actions shall be taken for REPORTABLE EVENTS.
a.
Each REPORTABLE EVENT shall be reviewed by the Operations Committee, and a report shall be submitted to the Safety Committee and the Vice President, Nuclear and b.
The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50.
Amendment No. J05.73f;,198 6.6-1 w
-e-1e ied' 7
+
'=
N a
-*11 F
r" t'
DAEC-1 6.7 ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED 6.7.1 If a safety limit is exceeded, the reactor shall be shut down and reactor operation shall only be resumed when authorized by the NRC.
6.7.2 An immediate report shall be made to the Vice President, Nuclear and the Safety Committee.
The Vice President, Nuclear shall promptly report the circumstances to the NRC as specified in Subsection 6.11, Plant Reporting Requirements.
6.7.3 A complete analysis of the circumstances leading up to and resulting from the situation together with recommendations to prevent a recurrence shall be prepared by the Operations Committee.
This report shall be submitted to the Vice President, Nuclear and to the Safety Committee.
Appropriate analyses or reports 1.;.1 be submitted to the NRC by the Vice President, Nuclear as specified in Subsection 6.11, Plant Reporting Requirements.
Amendment No. JJ#,135,198 6.7-1
,