ML20004D972
| ML20004D972 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 06/03/1981 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20004D973 | List: |
| References | |
| NUDOCS 8106100424 | |
| Download: ML20004D972 (12) | |
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8 UNITED STATES
[ Q.,% 7 i NUCLEAR REGULATORY COMMISSION C
WASHING TON, O. C. 20555
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g, IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET NO. 50-331 AMENDf1ENT TO FACILITY OPERATING LICENSE Amendment No. 69 License No. DPR-49 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The applications for amendment by Iowa Electric Light and Power Company, Central Iowa Power Cooperative, and Corn Belt Power Cooperative (the licensee) dated April 23,1976, and October 26, 1977 comply with the standards etnd 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 applications, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the publip, 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, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No.
DPR-49 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.69
, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
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. 3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
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Thomas A. Jppolito, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: June 3,1981.
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ATTACHMENT TO LICENSE AMENDMENT NO. 69 FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331
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Remove the following pages and insert identically numbered pages into the Appendix A. Technical Specifications.
iii 1.0-5 3.12-1 3.12-2 3.12-3
-3.12-8 6.2-1 6.2-3 Remove the following page and insert an identically numbcrec page into the Appendix B Technical Specifications.
'2.3-8 i
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O-DAEC-1 PAGE NO.
SURVEILLANCE LIMITING CONDITION FOR OPERATION REQUIREMENTS 3.7 Containment Systems 4.7 3.7-1 A.
3.7-1 P
B.
Standby Gas Treatment B
3.7-15 C.
3.7-17
(
D.
Primary Containment Power D
3.7-18 Operated Isolation Valves 3.8 Auxiliary Electrical Systems 4.8 3.8-}
A.
Auxiliary Electrical Equipment A
3.8-1 B.
Operation with. Inoperable B
3.8-3 Components
.C.
Emergency Service Water System C
3.8-6 l
3.9 Core Alterations 4.9 3.9-1
~A.
Refuciing Int.erlocks A
3.9-1
'B.
Core Monitoring B
3.9-4 C.
Spent Fuel-Pool Water Level
... C 3.9-4 3.10 Additional Safety Related Plant 4.10 3.10-1 Capabilities A.
Main Control Room Ventilation A
3.10-1 B.
Emergency Shutdown Control B
3.10-2 Panel 3.11 River Level Specification 4.11 3.11-1 3.12 Core Thermal Limits 4.12 3.12-1 A.
Maximum Average Planar Linear A
3.12-1 He'at Generation Rate B.
Linear Heat Generation Rate B
3.12-2 C.
Mintmum Critical Power Ratio C
3.12-3 l
Amendment No. 69-iii
/
DAEC-1
- 19. ALTERATION'0F THE REACTOR CORE (CORE ALTERATION) l The addition,- removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor pressure ves' el with the s
vessel head removed and fuel in the vessel. Suspension of CORE ALTERATICNS shall not preclude completion of the movement of a component to a safe conservative position.
20.
REACTOR VESSEL PRESSURE Unless otherwise indicatad, reactor vessel pressures listed in the Tech-nical Specifications are those measured by the reactor vessel steam space detectors.
- 21. T'riERMAL PARAMETERS a.
Minimum Cirtical Power Ratio (MCPR) - The value of critical power ratio (CPR).for that fuel bundle having the lowest CPR.
b.
Critical Pcwer Ratio (CPR) - The ratio of that fuel bundle power which would produce boiling transition to the actual fuel bundle power.
c.
Tran:ition Boiling - Transition boiling means the boiling regime between nucleate and film boiling. Transition boiling is the regime in which both nucleate and film boiling occur intermittently with neither type being completely stable.
d.
Deleted e.
Linear Heat Generation Rate - the heat output per unit length of fuel pin.
f.
Fraction of Limiting Power Density (FLPD) - The fraction of limiting power density is the ratto of the linear heat generation rate (LHGR) existing at a given location to the desion LHGR for that bundle type.
g.
Maximum Fraction of. Limiting Power Density (MFL?D) - The maximum fraction of limiting power density is the highest value existing in the core of the fraction of limiting power density (FLPD).
h.
Fraction of Rated Power (FRP) - The fraction of rated power is the ratio of core thermal power to, rated thermal power of 1593 MWth.
Amendment No. 59 1.0-5 w
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DAEC-1 LIMITING CON'DITION FOR OPERATION SURVEILLANCE REQUIREMENT i
3.12 CORE THERMAL LIMITS 4.12 CORE THERMAL LIMITS Acolicability Aoplicability, The Limiting Conditions for The Surveillance Rei.p irements Operation associated with the apply to the parameters which fuel rods apply to those monitor the fuel rod operating parameters which monitor the conditions.
fuel rod operating conditions.
Ob jec'ti ve Objective The Objective of the Limiting The Objective of the Surveil-Conditions for Operation is to lance Requirements is to assure the performance of the specify the type and frequency fuel rods.
of surveillance to be applied to the fuel rods.
)
Soecifications Soecifications A.
Maximum Average Planar Linear A.
Maximum Average Planar Linear Heat Generation Rate (MAPLPGR)
Heat Generation Rate (MAPLHGR)
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During reactor power operation, The MAPLHGR f.or each type of the actual MAPLHGR for each type fuel as a function of average of fuel as a function of average planar exposure shall be planar exposure shall not exceed determined daily during the limiting value shown in Figs.
reactor operation at > 25%
3.12-2,
-3,
-4,
-5,
-6, and 7.
rated thermal power and If at any time during reactor following any change in power operation it is determined power level or distribution by normal surveillance that the that would cause operation with lic.iting value for MAPLHGR a limiting control rod pattern (LAPLMGR) is being exceeded, as described in the bases for action shall then be initiated Specification 3.3.2.
During within 15 minutes to restore operation with a limiting control operation to within the pre-rod pattern, the MAPLHCR shall scribed limits.
If the MAPLHGR be determined at least once (LAPLHGR) is not returned to per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
within the prescribed limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, reduce reactor power to < 25% of Rated Thern ? '
l Power within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
i Surveillance and corresponding action shall continue until the prescribed limits are again being met.
Tonendment No. yf, 69
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENT B.
Lineae Heat Generation Rate B.
Linear Heat Generation Rate (LNGR)
(LNGR)
The LHGR as a function of core 1.
During reactor power opera-tion, the linear heat genera-height shall be checked daily tion rate (LHGR) of any rod during reactor operation at in any 7x7 fuel assembly at
> 25% rated thermal power i
ahdfollowing any change in any axial location shall not exceed the maximum allowable power level or distribution that would cause operation LHGR as calculated by the with a limiting control following equation:
rod pattern as described in 38 S r Pecification LHGR
< LHGR
- ((a max (L] max-d 3.3.2. During operation with a limiting control rod pattern d.= Design LHGR = 18.5 KW/ft the LHGR shall be determined at LHGR (7x7 array) least once per 12 hours. (aP/P,,, = Maximum power spiking penalty = 0.025 LT = Total care length - 12 feet L = Axial position above bottom of core. I 2. .During reactor power operation the linear heat generation rate (LHGR) of any rod in any 8x8 fuel assembly shall not exceed 13.4 KW/f t. If at any time during reactor power operation it is deter-mined by normal surveillance that the limiting value for LHGR is being exceeded, action shall then be ini-tiated within 15 minutes to restore operation to within the prescribed limits. If the LHGR is not returned ta within the prescribed 1_
- s within 2 hours, reduce r-Mr power to 1 25% of Rated Thermal Power within the next 4 hours. Surveillance and corresponding action shall continue until the prescribed i
limits are again being met. I
- 3. 12-2
/ Ame.dment No. 69 1 ( I 1
DAEC-1 LIMITING CONDITI"0NS FOR OPI; RATION SURVEILTANCE REOUTREMENTS C. Minimum Critical Power Ratio C. Minimum Critical Power Ratio (MCPR) (MCPR) During reactor power operstions, MCPR shall be determined daily MCPR shall bc E values as during reactor power operation indicated in Table 3.12-2 at h 257. rated thermal power - at rated power and flow. If at .and following any change in any time during reactor power power level or distribution operation it is'dctermined by that would cause operation with normal surveillance that the a limiting control rod pattern limiting value for MCPR is be-as described in the bases for ing exceeded, action shall then Specification 3.3.2. During be initiated within 15 minutes operation with a limiting to restore operation to within control rod pattern, the MCPR the prescribed limits. If the shall be determined at least i operating MCPR is not returned once per 12 hours. to within the prescribed limits within 2 hours, reduce reactor power to < 25% of Rated Thermal Power within the next 4 hours. Surveillance and corresponding action shall continue until the presecibed limits are again being met. For core flows other than rated the MCPR shall be E values as indicated in table 3.12-2 times K,, where K, is as shown in 0 Figure 3.12 1. t i I I l 1 Amendment No. 69 3.12-3 y
.DAEC-1 For coeration in the automatic flow control mode, the same procadure was em-ployed except the initial power distribution was established such that the r -MCPR was equal to the operating limit MCPR at rated power and flow. r r The'K factora shown in Figure 3.12-1 are conservative for Duane Arnold opera- [ g tir ' ecause the operating limit MCPR of values as indicated in Table 3.12-2 a is greater than the original 1.20 operating limit MCPR used for the generic f derivation of K. g I t i h l F I 4 Amendment No. 69 3.12-8
y DAEC-1 6.2 . PLANT STAFF ORGANIZATION. 6.2.1 The plant staf f organization shall confom to that shown in Figure 6.2-1. 6.2.2 The following manning requirements shall be met: 1. All CORE ALTERATIONS shall be directly supervised by either a Senior Licensed Operator or Senior Licensed 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 licensed operator shall be on the plant site. b. A licensed cperator shall be in the contrel room. Two licensec' ope rators shall be in. the control room c. during startup, scheduled shutdown,.and during recoverf from trips caused by transients or emergencies. d. Minimum operating shift crew compositions shall conform to those shown in Table 6.2-1. .N Amend:nent No. 69 6.2-1 e a
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= e 2.3-8. TABLE 3.3-1 ~ RADIDACTLVII LIQUID WASTE SAMPLING AND ANALYSIS Sample (5) Sampic Type Samply Frequency Sample Analysis Detectable Limit (2) Waste Tar.k to be released Each Batch Camma Scan (3) 5 x 10~7 uC1/mi Preportional Composite of Tritium 1 x 10-5 uCi/ml Batches Monthly Gross alpha 5 x 10~7 uCi/m1(4) Proportional Qua r te rly Sr00, Sr89 5 x 10-8 pci/m$ Coinposite of
- Batches 0se Batch Mduthly Dissolved noble
.1 x 10-5 uC1/mi gases Notes: 1. Certain mixtures of radionuelldes may cause interference in the measure- ~ ment of individual radionuclides at their detectable limit especially if other radionuelides are at much higher concentrations. Under these circumstances use of known ratios of radionuclides will be appropriate to calculate the levels of such radionuclides. 1 2. The above sample detectable limits are applicable to grab samples used to determine liquid waste release levels. Reported data shall re flect any improvement in detectable limits as such improvements are achieved. 3. S ignifican t radlunuclides are to be identified and where possible, quantitative values obtained. 4. Self absorption will result in a higher detectable limit for alpha counting. 5. Sample detectable limits are subject to revision. The values listed are believed to be attainable. Amendment No. 69 ..}}