ML20127P298
| ML20127P298 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 06/20/1985 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Corn Belt Power Cooperative, Central Iowa Power Cooperative, Iowa Electric Light & Power Co |
| Shared Package | |
| ML20127P303 | List: |
| References | |
| DPR-49-A-124 NUDOCS 8507020374 | |
| Download: ML20127P298 (17) | |
Text
,(jo nag Jg UNITED STATES y
p, NUCLEAR REGULATORY COMMISSION 5
.j WASHINGTON, D. C. 20555
...../
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. 124 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 June 14, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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 Comission; 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 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 Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by revising the effecti've date of Amendment No. 121 from May 28, 1985 to July 31, 1985.
8507020374 850620 PDR ADOCK 05000331 P
3.
This 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 Date of Issuance:
June 20, 1985 I
a
ATTACHMENT TO LICENSE AMENDMENT NO. 124
~ 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.
LIST OF AFFECTED PAGES 3.6-1*
3.6-1**
3.6-2*
3.6-2**
3.6-16*
3.6-16**
3.6-17*
3.6-17**
3.6-18*
3.6-18**
3.6-40*
3.6-40**
-s 3.6-41*
3.6-41**
Remove these pages from the Technical Specifications after July 31, 1985.
Effective on July 31, 1985.
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RECUIREMENT 3.6 PRIMARY SYSTEM BOUNDARY 4.6 PRIMARY SYSTEM BOUNDARY Acolicability:
Acolicability:
Applies to the operating Applies to the periodic status of the reactor examination and testing.
coolant system.
requirements for the reactor cooling system.
Objective:
Obiective:
To assure the integrity To detemine the condition and safe operation of the of the reactor coolant reactor coolant system.
system and the operation of the safety devices related to it.
Specification:
Specification:
A.
Themal and Pressurization A.
Themal and Pressurization L1m1tations Limitations 1.
The average rate of reactor 1.
During heatups and cooldowns, coolant temperature change the following temoeratures during nomal heatup or shall be logged at least every cooldown shall not exceed 15 minutes until 3 consecutive 0
100 F/hr when averaged readings at each given location over a one-hour period.
are within 5 F.
0
~
2.
The reactor vessel shall be ai.
Reactor vessel shell adjacent to vented and power operation shall shell flange.
not be conducted unless the reactor vessel temperature is b.
Reactor vessel bottom drain.
equal to or greater than that shown in Curve C of Figure 3.6.1. c.
Recirculation locos A and B.
Operation for hydrostatic or leakage tests, during heatup d.
Reactor vessel bottom head temperature.
or cooldown, and with the core critical shall be conducted 2.
Reactor vessel metal temoerature only wnen vessel temperature at the outside surface of the bottom is equal to or above that shown head in the vicinity of the control rod I
in the appropriate curve of drive housing and reactor vessel shell I
Fig. 3.6.1.
Figure 3.6.1 is adjacent to shell flange shall be effective througn 6 effective recorded at least every 15 minutes full power years. At least six during inservice hydrostatic or leak months prior to 6 effective testing when the vessel pressure is l
full power years new curves will
>312 psig.
be suomitted.
l 1
i Amendment No;
,56' 124 3.5-1 REMOVE FROM THE TECHNICAL SPECIFICATIONS ON JULY 31, 1985.
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 9
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 j
cooling system.
Objective:
Objective:
To assure the integrity and To determine the condition of safe operation of the reactor the reactor coolant system and coolant system.
the operation of the safety devices related to it.
Specification:
Specification:
A.
Thermal and Pressurization A.
Themal 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 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 l
Figure 3.6-1.
Operation for flange shall be recorded at i
hydrostatic or leakage tests, least every 15 minutes during during heatup or cooldown, and inservice hydrostatic or leak l
with the core critical shall testing when the vessel l
be conducted ortly.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 l
l effective through 12 effective l
full power years. At least l
six months prior to 12 effective full power years new curves will be submitted.
Amendment No. J6",J g 124 3
O gFFECTIVE ON JilLY 31, 1985.
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILtANCE REOUIREMENT 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' subjected to temperature of the vessel head the highest fluenct of greater flange and the head is greater than 1 MeV neutrons; shall be than 100*F.
installed in the-re" actor vessel adjacent to the vessel wall at 4
The pump in an idle the core midplane level.
The recirculation loop shall not specimens and sample program be started unless the shall conform to ASTM E 185-66 or temperatures of the coolant ASTM E-185-70 to the degree within the idle and operating discussed in Section 5.3.1.6 of recirculation loops are within the Updated FSAR.
50*F of each.Other.
Samples shall be withdrawn at 5.
The reactor recirculation one-fourth and three-fourths pumps shall not be started service life in accordance with 4
unless the coolant 10 CFR 50, Appendix H.
temperatures between the dome A
and the bottom head drain are within 145*F.
~
A.*
J
~
3.
When the reactor vessel head bolting studs are ten:ioned and '
the reactor is in a Cold Condition, the reactor vessel shell temperature immediately below the head flange shall be permanently recorded.
4 Prior to and during startup of an idle recirculation 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 coolant temperatures in the dome and in the bottom head drain shall'be i
compared and permanently logged.
- 0~2
'g Amendment No. p( 124
, _ REMOVE FR_01 THE TECHNICAL SPECIFICATI0NS ON JULY 31, 1985.
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 4
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 sbbjected 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 puinp 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.
1 Samples shall be withdrawn at 6 5.
The reactor recirculation and 15 effective full power i
pisnps 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 i
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 innediately below the head flang shall be permanently recorded.
4.
Prior to and during startup of an idle recirculation loop, the temperature of the reactor coolant in the operatTng 'id idle loops shall be permanently logged.
5.
Prior to starting a recirculation pisnp. the reactor coolant temperatures in the dome and in the bottom head drain shall be compared and Amendment No.g. J21' 124 permanently logged.
3.6-2 0
EFFECTIVE ON JilLY 31, 1985.
_ -. _ _.~..
DAEC-1 3.6.A and 4.6.A BASES:
4 Thermal and Pressurization Limitations
- i The thermal limitations for the reactor vessel meet the requirements of 10 CFR 50, Aopendix G.
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 heatuo and cooldown cycles when the 100*F per
. tour rate is limiting provides for efficient, but safe, plant operation. '
A..*
J 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 maximum stress occurs.
Chicago Bridge and Iron Company perfonned detailed stress analysis as shown
]
in the Updated FSAR Aopendix 5A, " Site Assembly of the Reactor Vessel."
i This analysis includes more severe thermal conditions than those which would be encountered during nonnal heating and cooling operations.
The pennissible flange to adjacent shell temocrature differential of 145*F is the maximum calculated for 100*F hour heating and cocling rate applied continuously over a 100*F to 3.6-16 Amendment No. Jg 124.
REMOVE FROM THE TECHNICAL SPECIFICATIONS ON JULY 31, 1985.
_ 1. ^_ _ _
DAEC-1 3.6.A and 4.6.A BASES:
Thermal and Pressurization Limitations 3
The thermal limitations for the reactor vessel meet the requirements of 10 CFR 50, Appendix G, revised May 1983. (3) 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 a
associated time period for heatup and cooldown cycles when the 100*F per hour rate is limiting provides for efficient, but safe, plant operation.
m 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 fatigue limits even at the flange area where maximum stress occurs.
Chicago Bridge and Iron Company performed detailed stress analysis as shown in the Updated FSAR Appendix 5A, " Site Assembly of the Reactor Vessel."
This analysis includes more severe thermal conditions than those which would be encountered during normal heating and cooling operations.
The permissible flange to adjacent shell temperature differential of l
145*F is the maximum calculated for 100*F hour heating and cooling rate applied continuously over a 100*F to Amendment No. g 124 3.6-16 0
gFFECTIVE ON JilLY 31, 1985.
DAEC-1 550*F range.
The differential is due to the sluggish temperature response to the flange metal and its value decreases for any lower heating rate or the same rate cpplied over a narrower range.
The c:olant in the bott:m of the vessel is at a lower temperature than that in
- ne up er regions of the vessel when there is no recir:ulation flow.
This calcer water is for:ed uo when recirculation pumps are started.
This will not result in stresses whi:h exceed ASME Soiler and Pressure Vessel Code,Section III limits wnen the temoerature differential is not greater than 145'F.
The reacg:r :colant system is a primary barrier ag' inst the release cf fission a
- r:due:s to the envir:ns.
n ceder to provice assuranc.e tnat t.lis barr.i.er is maintained at a hign cegree of integrity, restrictions have been olaced en the
- rating ::nci ions to *nich it can be subjected.
l l
=
The nil-ductility transitien (NDT) temperature is cefined as the temperature be'.ow whicn ferritic steel breaks in a brittle rather : nan ductile manner.
l Radia:icn exposure from f ast neutrons ( > 1 mev) acove aceu 1017 nyt may snif t l
- ne NOT temperature cf the vessel base metal above the initial value.
Extensive tests have establisnct the magnituce of changes as a function of the integrated neutren exposure.
~
Neutren flux wires and samples of vessel material are installed in the reacter vessel adjacent to the vessel wall at the core midplane level.
The wires and samoles will be Amendment No. 124 3.6-17 I
REMOVE FROM THE TECHNICAL SPECIFICATIONS ON JULY 31, 1985
DAEC-1 550*F range. The differential is due to the sluggish temperature response to the flange metal and its value decreases for any lower heating rate or the saine rate applied over a narrower range.
The coolant in the bottom of the vessel is at a lower temperature than that in the upper regions of the vessel when there is no recirculation flow. This colder water is forced up when recirculation ptsnps 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 have been placed on the operating conditions to which it can be subjected.
The operating 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 includes 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 temperature (RTNDT)ofthe beltline materials, possibly to the point where the beltline reg' ion impacts the pressure-temperature limits for the vessel. However, for Figure 3.6-1, effective to 12 EFPY, the beltline which has an Amendment No.)f, Jff 124 3.6-17 o
gFFECTIVE ON J!!LY 31, 1985.
DAEC-1 removh3 and tested according to 10 CFR 50' Appendix H.
Results of these analyses eill be dsed to adjust Figure 3.5-1 as appropriate.
As described in paragraoh 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 respe: to material f atigue. The results of these transients are compared to al10wable stress limits.
Requiring the coolant temperature in an idle recirculation loop to be within 50*F of the operating loop temperature oefore a r_ecirculation pumo is started assures that the changes in coolant temaeraj' re. at the rea: tor ' vessel nczzles and bottom head region are acceptable.
_g
....;c Amendment No.124 3.6-18 REMOVE _FROM THE TECHNICAL SPECIFICATIONS ON JULY 31, 1985.
DAEC-1 RTMDT 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 (N10) is the highest RTNDT of any component in the non-beltline region.
The closure flange region, with RTNDT = 14*F, has a bolt preload 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.
Neutron flux wires and samples of vessel material are installed in the 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 o
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 ptsnp is started assures that that changes in coolant temperature at the reactor vessel notzles and bottom head region are acceptable.
Amendment No.f>f, Jd 124 3.6-18
^
KFFECTIVE 0N JilLY 31. 1985.
DAEC-1 l
associated installation and maintenance records (newly installed snubber, seal replaced, spring replaced, in high radiation area, in high temperature area,edc...). 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 date may be incomplete.
113 The records will be developed from engineering data available.
If actual installation data is not available, the service life will be assumed to comence 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
- 1) General Electric Company, Low-Low set Relief Looic System and Lower MSIV Water Level Trio for the Duane Arnoio incroy Center,-hEDE-30021-P, January, 19eJ.
- 2) " General Electric Boiling Water Rea: tor increased Safety / Relief Valve.
115 Simer Margin Analysis for Duane Arnold Energy Center," NEDC-30606, May, 1984
)
3.6-40 Arnendment No. MS'l24 Ob\\ALOilDL LLCtUt. &[ 4 - M M D
-eauematemLAaunlus
DAEC-1 I
associated installation and maintenance records (newly installed snubber, 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 i
of the snubber service life monitoring program after several years of plant operation, the historical records to date may be incomplete.
4 The records will be [eveloped from engineering data available.
If actual i
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 1)
General Electric Company, Low-Low Set Relief Logic System and Lower MSIV Water level Trip for the Ouane Arnolo Energy Center _, NEDE-30021-P, January, l
1953.
2)
" General Electric Boiling hater Reactor Increased Safety / Relief Valve Simmer Margin Analysis for Duane Arnold Energy Center," NEDC-30606, May, 1984.
- 3) General Electric Company, Duane Arnold Energy Center Reactor Pressure Vessel Fracture Toughness Analysis to 10 CFR 50 Apoendix G, May 1983, 1
NEDC-30839, December,1984.
r Amendment No. Jf[ 124 3.6-40 EYFECTIVE ON JilLY 31, 1985.
FIGURE 3.6-1 i-1 ;g....
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- 124 3.6-41.
L Amendment No.
a i
FEIBEE FROM THE TECHNICAL SPECIFICATIONS ON JULY 31, 1985.
l e
I I
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I CURVE A - HYDROSTATICPRESSURE TESTS CURVE O - NON-CRITICAL MEATING AND COOLDOWN CURVE C - CORE CRITICAL OPERATION p
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ITING RTNOT*
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O CURVE A,
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/
w f
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CURVE 3, e00
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l BOLT PRELoao 400 TEMPERATURE = M*P PLANGE REGION RTNOT*'**E
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d*
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/
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900 300 m
l MINIMUM VESSEL MSTAL TEMPERATURE (*P)
Figure 3.6-1.
Pressure versus Minimus Temperature Valia to Twelve Full Power Years, per Appendix C of 10CTR50 Amendment Nog J27124 3.6-41 0
$ FECTIVE ON JilLY 31, 1985.
-