ML20126F528
| ML20126F528 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 05/28/1985 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Corn Belt Power Cooperative, Central Iowa Power Cooperative, Iowa Electric Light & Power Co |
| Shared Package | |
| ML20126F531 | List: |
| References | |
| DPR-49-A-121 NUDOCS 8506170518 | |
| Download: ML20126F528 (10) | |
Text
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8 4,1 UNITED STATES NUCLEAR REGULATORY COMMISSION h
j WASHINGTON, D. C. 20555
[
f IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET NO. 50-331 DUANE ARN0LD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 121 License No. DPR-49 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Iowa Electric Light and Power Company, et al, dated January 11, 1985, 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; 8.
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 the 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 Comission's regulaticns; 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 J
51 of the Comission's regulations and all applicable requirements j
have been satisfied.
2.
Accordingly, the license is amended by changes to the Technica.1 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
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amended to read as follows:
8506170518 850528 PDR ADOCM 05000331 P
.p.
(2) Technical Specifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No.121, 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.
FOR THE NUCLEAR REGULATORY COMMISSION Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing Attachmera:
Changes to the Technical Specifications Date of Issuance:
May 28, 1985 9
9
ATTACHMENT TO LICENSE AMENDMENT NO. 121 FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 Revise the Appendix "A" Technical Specifications by removing the current pages and inserting the revised pages listed below. The revised areas are identified by vertical lines.
AFFECTED PAGES 3.6-1 3.6-2 3.6-16 3.6-17 3.6-18 3.6-40 3.6-41 4
e
l DAEC-1 I
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT i
3.6 PRIMARY SYSTEM BOUNDARY 4.6 PRIMARY SYSTEM BOUNDARY Applicability:
Applicability:
Applies to the operating Applies to the periodic status of the reactor coolant examination and testing
- system, requirements for the reactor cooling system.
Objective:
Objective:
To assure the integrity and To determine the condition of safe operation of the reactor the reactor coolant syste1 and coolant system.
the operation of the safety devices related to it.
Specification:
Specification:
A.
Thermal and Pressurization A.
Thermal and Pressurization Limitations Limitations 1.
The average rate of reactor 1.
During heatups and cooldowns, coolant temperature change the following temperatures during normal heatup or shall be logged at least every cooldown shall not exceed 15 minutes until 3 consecutive 100*F/hr when averaged over a readings at each given location one-hour period, are within 5*F.
a.
Reactor vessel shell adjacent to shell flange.
b.
Reactor vessel bottom drain.
c.
Recirculation loops A and B.
d.
Reactor vessel bottom head temperature.
2.
The reactor vessel shall be 2.
Reactor vessel metal i
vented and power operation temperature at the outside shall not be conducted unless surface of the bottom head in the reactor vessel temperature the vicinity of the control rod is equal to or greater than drive housing and reactor that shown in Curve C of vessel shell adjacent to shell Fi ure 3.6-1.
Operation for flange shall be recorded at d
h rostatic or leakge tests, least every 15 minutes during d ring heatup or cooldown, and inservice hydrostatic or leak with the core critical shall testing when the vessel be conducted only.when vessel pressure is >312 psig.
temperature is equal to or above that shown in the appropriate curve of Figure 3.6-1.
Figure 3.6-1 is
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effective through 12 effective full power years. At least l
six months prior to 12 effective full power Wars new 3
curves will be submitted.
Amendment No.,56', 121 3.6-1 I
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DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.
The reactor vessel head Test specimens of the reactor bolting studs shall not be vessel base, weld and heat
)
under tension unless the affected zone metal subjecte.d temperature of the vessel head to the highest fluence of flange and the head is greater greater than 1 MeV neutrons than 74*F.
were installed in the reactor vessel adjacent to the vessel 4.
The pump in an idle wall at the core midplane recirculation loop shall not level at the start of be started unless the operation. The specimens and temperatures of the coolant sample program shall conform to within the idle and operating ASTM E 185-66. to the degree recirculation loops are within discussed in the FSAR.
50*F of each other.
Samples shall be withdrawn at 5 5.
The reactor recirculation and 15 effective full power ptmps shall not be started years in accordance with 10 CFR unless the coolant 50, Appendix H.
Neutron flux temperatures between the dome wires were installed in the and the bottom head drain are reactor vessel adjacent to the within 145'F.
reactor vessel wall at the core midplane level. The wires were removed and tested during the second refueling outage to experimentally verify the calculated values of neutron fluence at one-fourth of the beltline shell thickness that are used to determine the NDTT shift. Results of the flux wire test and the effects of copper and phosphorus on the beltline are reflected in Figure 3.6-1.
3.
When the reactor vessel head bolting studs are tensioned and the reactor is in a Cold Condition, the reactor vessel shell temperature immediately below the head flang shall be permanently recorded.
4.
Prior to and during startup of an idle recirculaticn loop, the temperature of the reactor coolant in the operating and idle loops shall be permanently
- logged, 5.
Prior to starting a recirculation pump, the reactor l
coolant temperatures in the i
dome and in the bottom head drain shall be compared and E* **"*"
99 Amendment No. J d, 121 3.6-2
DAEC-1 3.6.A and 4.6.A BASES:
Thermal and Pressurization Limitations The thermal limitations for the reactor vessel meet the requirements of 10 CFR 50, Appendix G, revised May 1983. (2) l The allowable rate of heatup and cooldown for the reactor vessel contained fluid is 100*F per hour averaged over a period of one hour. This rate has been chosen based on past experience with operating power plants. The associated time period for heatup and cooldown cycles when the 100*F per hour rate is limiting provides for efficient, but safe, plant operation.
Specific analyses were made based on a heating and cooling rate of 100*F/ hour applied continuously over a temperature range of 100*F to 546*F.
Calculated stresses were within ASME Boiler and Pressure Vessel Code Section III stress intensity and f atigue limits even at the flange area where
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maxistan ' stress occurs.
Chicago Bridge and Iron Conpany performed detailed stress analysis as shown in the Updated FSAR Appendix SA, " Site Assembly of the Reactor Vessel."
This analysis includes more severe thermal conditions than those which would i
be encountered during normal heating and cooling operations.
t The permissible flange to adjacent shell tenverature differential of 145*F is the maximum calculated for 100*F hour heating and cooling rate applied continuously over a 100*F to Amendment No.% 121 3.6-16
DAEC-1 i
550*F range. The differential is due to the sluggish tegerature response to the flange metal and its value decreases for any lower heating rate or the same rate applied over a narrower range.
The coolant in the bottom of the vessel is at a lower tegerature than that in the upper regions of the vessel when there is no recirculation flow. This colder water is forced up when recirculation pinps are started. This will not result in stresses which exceed ASME Boiler and Pressure Vessel Code,Section III limits when the temperature differential is not greater than 145'F.
The reactor coolant system is a primary barrier against the release of fission products to the environs. In order to provide assurance that this barrier is maintained at a high degree of integrity, restrictions f
have been placed on the operating conditions to which it can be subjected.
I
(
The enerating limits in Figure 3.6-1 are derived in accordance with 10 CFR 50 Appendix G, May 1983 and Appendix G of the ASME Code.
Conditions in three regions influence the curves: the closure flange region, the non-beltline region which ir.cludes most nozzles and discontinuities, and the beltline region which is irradiated with fluence above 1017 n/cm2 during the vessel operating life.
Irradiation causes an increase in the nil-ductility teraperature (RTNDT) of t,he beltline materials, possibly to the point where the beltline region impacts the pressure-teenperature limits for the vessel. However, for Figure 3.6-1, effective to 12 EFPY, the beltline which has an AmendmentNo.)f,121 3.6-17
DAEC-1 RTNDT of 40*F is less limiting than the non-beltline regions which generally experience higher stresses at nozzles and discontinuities.
The limiting RTNDT of 58'F for the Standby Liquid Control Nozzle (M10) is the highest RTNOT of any component in the non-beltline region.
The closure flange region, with RTNDT = 14*F, has a bolt preload l
and minimum operating temperature of 74*F.
This exceeds original requirements of the ASME Code (Winter 1967 Addendum) and provides extra margin relative to current ASME Code requirements.
1 Neutron flux wires and suples of vessel material are installed in the I
reactor vessel adjacent to the vessel wall at the core midplane level. The wires and samples will be removed and tested according to 10 CFR 50 Appendix H.
Results of these analyses will be used to adjust Figure 3.6-1 as appropriate.
As described in paragraph 4.2.5 of the Safety Analysis report, detailed stress analyses have been made on the reactor vessel for both steady state and transient conditions with respect to material f atigue. The results of these transients are compared to allowable stress limits. Requiring the coolant temperature in an idle recirculation loop to be within 50*F of the operating loop temperature before a recirculation pep is started assures that the changes in coolant temperature at tne reactor vessel nozzles and bottom head region are acceptable.
Amendment No.,,K,121 3.6-18
DAEC-1 associated installation and maintenance records (newly installed snutoer, seal replaced, spring replaced, in high radiation area, in high temperature area,etc...).
The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of age and operating conditions.
Due to implementation of the snubber service life monitoring program after several years of plant operation, the historical records to cate may be incomplete.
The records will be developed from engineering data available.
If actual installation data is not available, the service life will be assumed to commence with the initial criticality of the plant.
These records will provide statistical bases for future consideration of snubber service life.
The requirements for the maintenance of records and the snubber service life review are not intended to affect plant operation.
3.6 and 4.6 References S*
- 1) General Electric Company, Low-Low Set Relief Loaic System and Lower MSIV e
Water Level Trip for the Duane Arnold Energy Center, NEDE-30021-P, January, 1983.
- 2) General Electric Company, Duane Arnold Energy Center Reactor Pressure Vessel Fracture Toughness Analysis to 10 CFR 50 Appendix G, May 1983, NEDC-30839, Decenber,1984.
3.6-40 Amendment No. M 121
1000 l
I I
l l
CURVE A - HYDROSTATIC PRESSURE TESTS CURVE B - NON-CRITICAL HEATING AND COOLDOWN CURVE C - CORE CRITICAL OPERATION 1400 l
I i
f I
LIMITING RTNOT*
E 1000
/
3 3
O CURVE A.
.B- :=
/
/
/
f CURVE 8, 000
>/
f f
SOLT PR E LOAD 400 TEMPERATURE = 74'F I
FLANGE REGION MTNOT = 140F g
j CURVE C
/
/
s, 0
0 100 200 300 MINIMUM VESSEL METAL TEMPERATURE (*F)
Figure 3.6-1.
Pressure versus Minimum Temperature Valia to Twelve Full Power Years, per Appendix C of 10CFR50 Amendment No. M 121 3.6-41
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