ML20117K400
| ML20117K400 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 06/05/1996 |
| From: | Kelly G NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20117K402 | List: |
| References | |
| NUDOCS 9606110196 | |
| Download: ML20117K400 (13) | |
Text
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UNITED STATES g
j NUCLEAR REGULATORY COMMISSION 1
WASHINGTON, D.C. 30seH001
- %...../
4 2
IES UTILITIES INC.
I CENTRAL IOWA POWER COOPERATIVE i
CORN BELT POWER COOPERATIVE l
DOCKET NO. 50-331 1
DUANE ARNOLD ENERGY CENTER j
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 214 License No. DPR-49 i
j 1.
The Nuclear Regulatory Commission (the Commission) has found that:
i A.
The application for amendment by IES Utilities Inc. et al., dated July 21, 1995, as supplemented August 8, 1995, and December 15 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's e
rules and regulations set forth in 10 CFR Chapter I; i
i B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; i
C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health i
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; 4
l ad i
E.
The issuance of this amendment is in accordance with 10 CFR Part i
51 of the Commission's regulations and all applicable requirements have been satisfied.
}
I 2.
Accortlingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and j
paragraph 2.C.(2) of Facility Operating License No. DPR-49 is hereby i
amended to read as follows:
i 0
i i
9606110196 960605 PDR ADOCK 05000331 P
i (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.214, 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 30 days of the date of issuance.
FOR THE NUCLEAR REGULATORY C0fMISSION lf
'6hfk
' GlennB.Ke'lly,Proje$tManager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance:
June 5, 1996 I
l i
ATTACHMENT TO LICENSE AMENDMENT NO.214 FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 Replace the following pages of the license and Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by vertical lines.
Remove
- Inser1, 4
4 3.5-10 3.5-10 3.5-23 3.5-23 3.8-4 3.8-4 3.8-6 3.8-6 5.5-1 5.5-1 5.5-la 6.5-3 6.5-3 6.8-1 6.8-1 6.8-2 6.8-2 1
i i
l
4 (3)
Fire Protection IES Utilities Inc. shall implement and maintain in effect all provisions of the approved fire protection program as described in 4
1 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 to the following provision:
j The licensee may make changes to the approved fire l
protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a-
- fire, d
(4)
The licensee is authorized to operate the Duane Arnold Energy Center following installation of modified safe-ends on the eight i
primary recirculation system inlet lines which are described in the licensee letter dated July 31, 1978, and supplemented by letter dated December 8,1978.
j (5)
Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans, including amendments made pursuant to the authority of 10 CFR 1
50.54(p).
The approved plans, which contain Safeguards i
Information as described in 10 CFR 73.21, ate collectively entitled:
j "Duane Arnold Energy Center Security Plan" dated December 1, 1978, January 19, March 9 and March 21, 1978, as revised i
throuah revisions dated January 1984 (transmittal letter dated January 12,1984), as revised by revision dated February 1984 (transmittal letter dated February 27,1984),
as revised by revision dated September 1984 (transmittal letter dated September 26, 1984); "Duane Arnold Energy Center Safeguards Contingency Plan," dated April 1980, as revised throuah revision dated January 1984 (transmittal i
letter dated January 12, 1984); "Duane Arnold Energy Center l
Guard Training and Qualification Plan" dated January 29, 1982, as revised April 1, 1982, as revised throuah revisions i
dated January 1984 (transmittal letter dated January 12, l
1984), as revised by updated revisions (transmittal letter dated July 30,1984), as revised by revision dated September
]
1984 (transmittal letter dated September 26, 1984) as i
revised by revision dated October 1984 (transmittal letter l
dated October 26,1984).
l Amendment No. 47,50,00,74,112,
DAEC-1 Im1 TING CONDITIONS FOR OPERATION SURVEHJ ANCE REOUIREMENTS a,
us i-
-Io w r;
- a t nni:.. w g,
n'
- Law r;, enni;.. and Dianal Generator Avadabahty D'-- ' C --
L - Ava:1-MI +v 1.
Dunng any penod when one diesel 1.
With one diesel generasor anoperable, generator is inoperable, <=*=uant reactor deterene that the OPERABLE diesel operation is penniesible only during the gener dor is not inoperable due to comunon
==-: ^; seven days unless such diesel cause failure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and perform j
generaser is sooner inade OPERABG, Surveillance Requirement 4.8.A.2.a.1.a 1
provided that the remnanang diesel within the Erst 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and every 72 generasar and all low pressure core and hours therenAer. In addition, all low
<=aminment coohng subsystems supported pressure core coohng and <=amiania=*
by the OPERABLE diesel generator are cooling subsystems supported by the OPERABLE. If this requirement cannot OPERABLE diesel shall be verified to be be meet, an orderly SHUTDOWN shall be OPERABLE.
initiated and the reactor shall 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 />.
2.
Any combination of inoperable cornponents in the core and contaanment cooling systems shall not defeat the capability of the remaining OPERABLE components to fulfill the cooling functions.
3.
When irradiated fuel is in the reactor vessel and the reactor is in the COLD SHUTDOWN Condition or REFUEL Mode:
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 b.
If work is being performed which has the potential for draining the reactor vessel, at least two of any combanation 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 Amendment No. !!9,!97, 214 3.5-10
\\
DAEC-I i
4.5 BASES Core and Containment Cooling Systems Surveillance Frequencies The testing interval for the core and containment cooling systees 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.
The Diesel Generators are critical to operation of all core and containment 8
cooling systems.
Therefore, it is imperative that they be maintained in a standby readiness condition.
In the event that one Diesel Generator is made or found to be inoperable, the remaining Diesel Generator must be shown to not j
be susceptible to the same condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This evaluation may be performed by analysis or inspection or by demonstration of OPERABILITY. The OPERABLE Diesel Generator must also be demonstrated to continue to be OPERABLE t
each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during the period that the other Diesel Generator is inoperable.
l The RHR valve power bus is not instrumented.
For this reason surveillance requirements require once per shift observation and verification of lights and instrumentation operability.
Amendment No. 47,143,150,174,210, 3.5-23 214
-. ~.
DAEC-1 LIMITINO CONDITIONS FOR OPERATION SURVEH I ANCE REOUIREMENTS chargese for the 24 Volt Systems, two of the voltage shall be measured and recorded.
three batsery chargers for the 125 Volt Systems, and one of the two battery chargers b.
Each three ma=tha the essential battenes' for the 250 Volt System shall be OPERABLE.
voltage of each cell to the nearest 0.01 Volt, specific gravity of each cell, and temperature of every fiRh cell shall be measured and j
recorded.
J c.
Once each OPERATING CYCLE, the essential battenes shall be subjected to a Service Discharge Test (load profile). The specific gravity and voltage of each cell shall be deterinined aner the diarharge and recorded d.
Once every five years, the essential batteries shall be subjected to a Performance Discharge 4
Test (capacity). This test will be performed in i
lieu of the Service Test requirement of j
j 4.8.B.I.c above.
4 l
2.
Operation with Inoperable Components.
{
2.
Surveillance Requirements with Inoperable Components.
c.
With normal battery room ventilation j
unavailable, portable ventilation equipment With the battery room ventilation unavailable, a.
shall be provided.
j samples of the battery room atmosphere shall be taken daily for hydrogen concentration b.
With one of the two 125 Volt DC Systems determmation, j
moperable, verify that Specification 3.5.G is
)
met, and within 3 days either:
1)
Restore the inoperable 125 Volt DC System to 4
OPERABLE status, or l
2)
Be in at least HOT SHUTDOWN within the i
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 l
within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
I c.
With the 250 Volt DC System inoperable, the HPCI System and other affected primary rmamm- =t isolation valves shall be considered inoperable and the requirements of
]
Specifications 3.5.D and 3.7.B respectively shall be snet.
J d.
With one of the 24 Volt DC Systems anoperable, the requirements associated with the affected instruments of Specificassons 3.1 and 3.2 shall be met.
Amendment No. 153,107,- 214 3 8-4
1 DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS E.
Emergsacy Service Water System E.
F-nency Service Water Sv=*> =
1.
Except as roquared in Specification 3.8.E.2 1.
Emergency Service Water System surveillance below, both F-ay Service Water System shall be as follows loops shall be OPERABLE whenever irradiated fuel is in the reactor vessel and reactor coolant a.
Simulated auto-once/
temperature is greater than 212*F.
matic actuation OPERATING test.
CYCLE l
i b.
Pump and motor As specified in i
operated valve the IST Program j
I c.
Flow Rate Test Each Emergency After major pump Service Water maintenance and pump shall once per 3 months, deliver at except weekly least that flow during periods of deternuned 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 2.
With one Emergency Service Water System System pumps or loops inoperable, REACTOR pump or loop inoperable, the OPERABLE POWER OPERATION must be limited to l
pump and loop shall be verified to be seven days unless OPERABILITY of that OPERABLE. In addition, all low pressure system is restored within this period. During care cooling and conamnnwat cooling subsystems and the diesel generator supported such seven days all active components of the other Emergency Service Water System shall by the OPERABLE ESW loop shall be verified be OPERABLE, provided the requirements of to be OPERABLE.
Specification 3.5.0 are met.
8 i
3.
If the requir==anas of Specification 3.8.E l
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 />.
Amendment No.W,32,130,143,lSO,197,210,214 3.8-6
DAEC-1 5.5 SPENT ANO NEW FUEL STORAGE 5.5.1 Criticality 5.5.1.1 The spent fuel storage racks are designed and shall be maintained with:
- a. Fuel assemblies having a maximus k, of 1.31 in the normal reactor core configuration at cold conditions and a maximum initial uniform average U-235 enrichment of 4.6 weight percent.
- b. k,,, s 0.95 flooded with unborated water.
5.5.1.i The new fuel storage racks are designed and shall be maintained with:
- a. Fuel assemblies having a maximus k, of 1.31 in the normal reactor core configuration at cold conditions.
- b. k,,, s 0.90 dry and s 0.95 flooded with unborated water.
5.5.2 Capacity 5.5.2.1 The spent fuel storage 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 storage vault is equipped with racks for storage of up to 110 fuel assemblies in a vertical orientation only.
Bases The basis for the k, limit is described in Reference 1 for the GE-designed new fuel storage racks. Compliance with this specification is demonstrated by comparing the beginning-of-life, uncontrolled k, values for the fuel type of interest to the 1.31 limit. For GE-supplied fuel, k, values can be found in Reference 2.
The k, values found in Reference 2 represent the maximum, exposure-dependent lattice reactivity and can be conservatively applied to the new fuel limit.
Calculations have been performed (Reference 3) to determine the bounding reactivity limits for bundles of GE-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 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 wt% 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.
Amendment No. 44ETM57 214 5.5-1
DAEC-1 References 1)
General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A.*
2)
General Electric Fuel Bundle Designs, NEDE-31152-P.*
3)
Licensing Report for Spent Fuel Storage Capacity Expansion, Duane Arnold Energy Center, Holtec Report HI-92889.
- Latest NRC-approved revision.
l 4
4 Amendment No. 214 5.5-la
DAEC-1 f.
Review of all Reportable Events.
g.
Review of facility operations to detect potential safety hazards.
h.
Performance of special reviews, investigations or analysis and reports thereon as requested by the Chairman of the Safety Committee.
4 i.
Review of the Plant Security Plan.
j.
Review of the Emergency Plan.
i k.
Review of every unplanned release of radioactivity to the environs for which a report to the NRC is required.
4 1.
Review of changes to the Offsite Dose Assessment Manual and changes to the Process Control Program.
a.
Review of the Fire Protection Program and implementing procedures.
6.5.1.7 Authority The Operations Committee shall:
3 a.
Recommend to the Plant Superintendent-Nuclear written approval or disapproval 4
j of items considered under Specification 6.5.1.6 (a) through (d) above.
Amendment No. 100,100-214 6.5-3
DAEC-1 6.8 PLANT OPERATING PROCEDURES 6.8.1 Written procedures involving nuclear safety, including applicable check-off lists and instructions, covering areas listed below shall be prepared, and approved as specified in Subsection 6.8.2.
All procedures shall be implemented and maintained.
1.
Normal startup, operation, and shutdown of systems and components of the facility.
4 2.
Refueling operation.
3.
Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, including responses to alarms, suspected primary system leaks, and abnormal reactivity changes.
4.
Emergency and off-normal condition procedures.
5.
Preventive and corrective maintenance operations which could have an effect on the nuclear safety of the facility.
6.
Surveillance and testing requirements of equipment that could have an effect on the nuclear safety of the facility.
7.
Deleted.
Amendment No. -19h-214 6.8-1
1 i
)
DAEC-1 i
8.
Deleted 1
1 9.
Operation of radioactive waste systems.
l 10.
Fire Protection Program implementation.
11.
A preventive maintenance and period'.c visual examination program to i
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.
j 12.
Program to ensure the capability to accurately determine the airborne I
iodine concentration in vital areas under accident conditions, including training of personnel, procedures for monitoring and provisions for maintenance of sampling and analysis equipment.
i 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.
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 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.
Amendment No.1^^,125,128,10,12-6.8-2 199,202,214