ML19256A182
| ML19256A182 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 10/17/1978 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19256A179 | List: |
| References | |
| NUDOCS 7811030208 | |
| Download: ML19256A182 (9) | |
Text
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uninD STARS NUCLEAR REGULATORY COMMISSION y-g g
WASHINGTON. D. C. 20555
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IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LIGENSE Amendment No. 46 License No. DPR-49 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Iowa Electric Light and Power Company, Central Iowa Power Cooperative, Corn Belt Power Cooperative (the licensee) dated November 3, 1976 and April 14, 1977, 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 (1) 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 comon 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.
787.1030303
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license snendment, and paragraph 2.C(2) of Facility Operating License No. DPR-49 is hereby amended to read as follows:
(2)
Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 46, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Thomas A.
ppolito, Chief Operating Reactors Branch #3 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
October 17,1978
ATTACHMENT TO LICENSE AMENDMENT NO. 46 FACILITY OPERATING LICENSE NO. D."o.-49_
DOCKET NO. 50-331 Replace the folicwing pages of the Appendix "A" Technical Specifications witn the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Pages 3.2-21 3.2-31 3.7-14 3.7-14a 3.7-41 3.7-49
TABLE 3.2-P SURVEILLAtlCE INSTRUMENTATIOli Hinimtuu tio.
of Operable Type s
Instrument Intitrument Indication Action Channels and Range 2
Reactor Water Level
- Recorder, (1) (2)
Indicator (3) 0-60" 2
Reactor Pressure
- Recorder, (1) (2)
Indicator (3) 0-1200 psig Indicator 2
Drywell Pressure
- Recorder, (1) (2) 0-00 psia (3)
}d U
Indicator N
u 2
Drywell Temperature Recorder (1) (2) n 0-400 F
/,.
Indicator 2
Suppre ssion Chambe r' Tdmperature
- Recorder, (1) (2) 0-400 P (3)
Indicator Suppression Chamber Water Level Recorder (1) (2) 2
-10"/0/+10" M 0 (3) 2 l'
Control Rod Position Process Com-puter, Full Travel 1
Neutron Honitoring SRM, IRM, (1) (2)
LPRM.0 to
_ (3) (4) 100% power 2
Drywell/ Torus AP Alarm Indicator, 10 paid 1
Drywell Pressure Indicator,* 0-100 psia 1
Torus Pressure Indicator,* l-100 psia
- capable of 10.1 paid
TABLE 4.2-P MillIMUM TEST AllD CALIDRATION PREQUENCY FOR SURVEILLANCE INSTITUMENTATION Instrument channel Calibration Frequency Instrument check 1)
Reactor Level Once/6 months Once Each Shi f t 2)
Reactor Pressure Once/6 months Once Each Shift 3)
Drywell Pressure Once/6 months Once Each Shift 4)
Drywell Temperature Once/6 months Once Each Shift Y
O g
5)
Suppression Chamber Temperature Once/6 months Once Eaen Shift gt b
6)
Suppression Chamber Water Level Once/6 months Once Each' Shift e
7)
Control Rod Position NA Once Each Shift 0)
Heutron Honitoring Prior to Reaching 20%
Once Each Shift Power and once per day (When in Startup when in Run Hode (APRH or Run Hode)
Cain Adjust when in Run Hode) 9)
Drywell/ Torus' A,P Alarm once/6 months Once Each Shift 10)
Drywell/ Torus AP Indicator once/6 months Once Each Shift 11)
Drywell Pressure Once/ Operating Cycle Once Each Shift 12)
Torus Pressure Once/ Operating Cycle Once Each Shift
DAEC-1 1.IMI ING CCNDITIONS FCR OPE?A* ION SURVEII.1ANCE REOUIRDENTS c:ust be taken out of power functionally tested once per operation.
operating cycle in conjunction with specification 4.7.A.6.a.
Should one of the two y or 02 analy:ers serving the dryvell or suppression pool be found inoperable, the re=afnfrg analyzer of the same type serv-ing the same eccpartment shall be tested for operability once per week until the defective analyzer is made operable.
7.
Drvvell-Sueeression Cha=ber 7.
Drvvell-Sueeressien Cha=her Differential Pressure Differential Pressure a.
Differential pressure between a.
The pressure differential be-the drywell and suppression tween the dryvell and suppres-cha=ber shall be =aintained at sion cha=ber shall be recorded equal to or greater than 1.30 at least once each shift.
psid except as specified in (1) and (2) below:
(1) within the 24-hour period sub-sequent to placing the reactor in the Run Mode following a shutdown, the differential shall be established.
The differential may be decreased to less than 1.30 psid 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> priop to a scheduled shutdown.
i (2)
This differential may be de-creased ta less than 1.30 psid for a max 1=um of four hours during required operability testing of the EPCI system purp, the RCIC system pu=p, the dryvell pressure suppres-sien cha=ber vacut=1 breakers, and the suppression.cha=ber to reacter building vacuum breakers.
t
.(3)
If the differential pressure of specification 3.7.A.7.a cannot be maintained, and the differen-tial pressure cannot be restored within the subsequent six (6) hour period, an orderly shut-down shall be initiated and the reactor shall be in the Cold Shutdown condition within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
DAEC-1 LIMInNC CCNDITIONS FOR OPERATION SURVE!!IANCE RECUIREME'iTS 8.
If the specifications of 3.7.A.1 through 3.7.A.5 cannot be net, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
9 e
3.7-144
DAEC-1 Due to the nitrogen additicu, the pressure in the contai==ent after a LOCA could possibly increase with ti=e.
Under the worst expected conditions the contaicnent pressure will reach 30 psig in apptcxi=ataly 70 days. If and when that pressure is reached, venting from the con-tai==ent shall be manually initiated.
The venting path will be through the Standby Gas Treatment System in c der to ed 4-dze the offsite dose.
Tollowing a LOCA, periodic operation of the drywell and corus sprays may be used to assist the natural convection and diffusion mixing of hydrogen and oxygen.
In conjunction with the Mark I Containment Short Term Program, a plant unique analysis was performed which demonstrated a factor of safety of at least two for the weakest ele =ent in the suppression chamber support system and attached piping. The maintenance of a drywell suppression chamber differential pressure of 1.30 psid and a suppression chamber water level corresponding to a downcomer submergence range of 4.30 to 3.96 feet vill assure the integrity of the suppression rhneher when subjected to post-LOCA suppression pool hydrodynamic forces. Design details are described in References 5 and 6.
7.
Standby Gas Treatment System and Secondary Contaf ment The secondary contain=ent is designed to =1-d d:e any ground level release of radioactive =aterials which =ight result f c= a serious accident. The react : building p cvides secondary contai=nent during reactor operation, when the drywell is sealed and in service; the reactor building provides pri=ary conta1==ent when the reactor is shut down and the drywell is
DAEC-1 3.7.A & 4.7.A REFERENCES 1.
Section 14.6 of the FSAR.
2.
ASME Boiler and Pressure Vessel Code, Nuclear Vessels,
Section III, maximum allewable internal pressure is 62 psig.
3.
Staff Safety Evaluation of DAEC, USAEC, Directorate of Licensing, January 23, 1973.
4.
10 CFR 50.54, Appendix J, Reactor Containment Testing Requirements, Federal Register, August 27, 1971.
5.
DAEC Short-Term Program Plant Unique Analysis NUTECE Dec. No.
ICW-01-065, August 1976.
6.
Supplement to DAIC Shert-Term Program Plane Unique Analysis, NUTECH Doc. No. ICE-01-071, October 1976.
..r e
3.7-49
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UNITED STATES
'4 NUCt. EAR REGULATORY COMMISSION g-g
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WASHINGTON, D. C. 20665
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 46 TO LICENSE NO. DPR-49 IOWA ELECTRIC LIGHT AND POWER COMPANY CENTPAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DUANE ARNOLD ENERGY CENTER DOCKET NO. 50-331 Introduction In conjunction with the Short Tenn Program (STP) evaluation of Boiling Water Reactor facilities with the Mark I containment system, the Iowa Electric Light & Power Company ()icensee) submitted a Plant Unique Analysis (PUA) for the Duane Arnold Energy Center (DAEC). This analysis was perfonned to confirm the structural and functional capability of the containment suppression chamber and attached piping to withstand newly-identified suppression pool hydrodynamic loading conditions which had not been explicitly considered in the original design analysis for the plant. As part of the STP evaluation, specific loading conditions were developed for each Mark I facility to account for the change in the magnitude of the loads due to plant-specific variations from the reference plant design for which the basic loading conditions were developed.
The results of the staff's review of the hydrodynamic load definition techniques and the Mark I containment plant unique analyses are described in the, " Mark I Containment Short Term Program Safety Evalva-tion Report", NUREG-0408, December 1977. As discussed in this report, the NRC staff has concluded that each Mark I containment system would maintain its integrity ard functional capability in the unlikely event of a design basis loss-of-coolant accident (LOCA) and, thereforo, that. licensed Mark I BWR facilities can continue 10 operate safely, without undue risk to the health and safety of the public, during an interim period of approximately two years, while a methodical, comprehensive Long Term Program is conducted.
7 8 f.1 0 3 O Nf
As discussed in Section III.C of NUREG-0408, of all of the plant parameters that were considered in the development of the hydrodynamic loads for the STP, only two parameters are expected to vary during normal plant operation; these are (1) the drywell-wetwell differential pressure; and (2) the suppression chamber (torus) water level.
Subsequent to the submittal of the PUA, the licensee was requested to submit proposed Technical Specifications which assure that the allowable range of these two parameters during facility operation would be in accordance with the values utilized in the PUA.
The licensee has been operating this facility with differential pressure control to enhance the safety margins of the containment structure since early 1976. This evaluation provides a more detailed basis for establishing the allowable range of drywell-wetwell differential pressure and torus water level, in order to quantify containment safety margins.
This amendment incorporates these parameters into the Technical Specifi'-
cations with the associated limiting conditions for operation and surveillance requirements.
By letters dated November 3, 1976 and April 14, 1977, the licensee proposed changes to the facility Technical Specifications to incorporate limiting conditions for operation and surveillance requirements for differential pressure control and torus water level. Our evaluation of these proposed changes follows.
Evaluation The 'l'icensee has proposed certain Technical Specification requirements for the purpose of assuring that the normal plant operating conditions-are within the envelope of conditions considered in their PUA.
These Technical Specification changes establish (1) limiting conditions for operation (LCOs) for drywell to torus differential pressure and toius water level, and (2) associated surveillance requirements.
All other initial conditions utilized in the PUA are either presently included in the Technical Specifications or are configurational conditions which have been confirmed by the licensee and will not change during nonnal operation.
Differential pressure between the drywell and the suppression chamber will result in leakage of the drywell atmosphere to the lower pressure regions of the reactor building and to the torus airspace.
This leakage from the drywell will cause a slow decay in the differential pressure. Therefore, surveillance requirements for the differential pressure have been included in the Technical Specifications.
Surveillance frequency of once per operating shift for the differential pressure was selected on the basis of previous operating experience.
The torus water level is not expected to vary significantly during normal operation, unless certain systems connected to the suppression pool are activated. The torus water level would normally be monitored whenever such systems are in use. Therefore, we find that inclusion of periodic torus water level surveillance requirements in the Technical Specifications is not required.
We have reviewed the differential pressure and torus water level monitoring instrumentation systems proposed by the licensee with regard to the number of available channels and the instrumentation accuracy. This type of instrumentation is typically calibrated at six-month intervals. To assure proper operation during such intervals, two monitoring channels for. aoth differential pressure and torus water level have been provided, such that a comparison of the instrumentation will indicate when one of the channels is inoperative or malfunctioning. The errors in the instrumentation are sufficiently small relative to the magnitude of the measurement (i.e., a maximum differential pressure measurement error of 0.1 psid in a measurement of 1.0 to 2.0 psid and a maximum torus water level measurement error of 10% of the difference between the maximum and minimum torus water level) that they may be neglected, based on the expected load variation with differential pressure and torus water level.
There are certain periods during normal plant operations when the differential pressure control cannot be maintained. Therefore, provisions have been included in the Technical Specification to relax the differential pressure / control requirements during specified periods. The justification for relaxing the differential pressure control during these specific periods and the basis for selecting the duration of the periods are discusse'd in detail below.
A.
Startuo and Shutdown During plant startup and shutdown, the drywell atmosphere undergoes significant barometric changes due to the variation in heat loads from the primary and auxiliary systems.
In addition, it is during these periods that the drywell is being either inerted with nitrogen gas or deinerted.
In order to keep the periods during which the differential pressure control is not fully effective as short as is reasonable, we have limited the relaxation of the differential pressure control requirements for the startup and shut-down periods to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following startup and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a shutdown. This time period was selected on a basis simil.f ar to that 787.33
for the inerting requirements, already existing in the Technical Specifications.
The postulated design basis accident for the containment assumes that the primary system is at operating pressure and temperature.
During the startup and shutdown transients, the primary system is at operating pressure and, temperature for only a part of the transient, during which the differential pressure is being established. These time periods have been shown by previous operating experience to be adequate with respect to the startup and shutdown transients, and at the same time sufficiently small in comparison to the duration of the average power run.
Since the principal accident event to which differential pressure control is important to assure containment integrity (i.e., with a factor of safety of two) is a large break LOCA, we have considered whether there is a significantly greater probability of a large break LOCA during the startup and shutdown transients. We have concluded that there is not.
Further, the operation of the plant systems is monitored more closely than nomal during these periods and a finite magnitude of differential pressure will be available during the majority of these periods to mitigate the potential consequences of an accident.
B.
_ Testing and Maintenance During nomal operation, there are a number of tests which are required to be conducted to demonstrate the continued functional performance of engineered safety features. The testing of certain systems will require, or result in, a reduction in the drywell-torus differential pressure. The operability testing of
- the drywell-torus vacuum breakers requires the removal of the differential pressure to permit the vacuum breakers to open.
For the testing of high-energy systems (e'g., high pressure coolant injection pumps) during normal operation, the discharge flow is routed to the suppression pool. This energy deposition will raise the temperature of the suppression pool, resul*' - in an increase in torus pressure and a reduction in the di; ential pressure.
Functional performance testing of engineered safety features is necessary to assure proper maintenance of these systems throughout the life of the plant. Some of these tests (i.e., pump operability and drywell-wetwell vacuum breakers) may require or result in a reduction in the differential pressure. We estimate that not more
than four tests will be required each month which will result in a reduction in differential pressure.
In order to keep the periods during which the differential pressure control is not fully effective as short as is reasonable, we have permitted a relaxation of differential pressure control in order to conduct the tests, limited to a period of up to four hours. Again, we have carefully considered whether the probability of a large LOCA is significantly greater during these testing periods than that during normal operation. We conclude that it is not. Moreover, only the test of the drywell-wetwell vacuum breakers requires complete removal of the differential pressure.
Provisions have also been included in the Technical Specifications for performing maintenance activities on the differential pressure control system and for resolving operational difficulties which may result in an inadvertent reduction in the differential pressure for a sho'/t period of time.
In certain circumstances, corrective action can be taken without having to attain a cold shutdown conditioc. To avoid repeated and unnecessary partial cooldown cycles, a restoration period has been incorporated into the action requirements of the LCO for differential pressure control; i.e.,
in the event that the differential pressure cannot be restored in six hours, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The six hour restoration period was selected on the basis that it represents an adequate minimum period of time during which any short-term malfunctions could be corrected, coupled with the minimum period
.of time required to conduct a controlled shutdown.
The allowable
.-time to conduct a controlled shutdown has been minimized, because the containment transient response is more a function of the primary system pressure than the reactor power level. On this basis, we find the proposed restoration period and action requirement acceptable.
We conclude that the limits imposed on the periods of time during which operation is permitted without the differential pressure control fully effective provides adequate assurance of overall containment integrity, and the periods of time differential pressure control is completely removed are acceptably small.
Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this
detennination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR Section 51.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
Conclusion The proposed Technical Specifications will provide the necessary assurance that the plant's operating conditions remain within the envelope of the conditions assumed in the Plant Unique Analysis (PUA) performed in conjunction with the Mark I Containment Short Term Program. The PUA supplements the facility's Final Safety Analysis Report (FSAR) in that it demonstrates the plant's capability to withstand the suppression pool hydrodynamic loads which were not explicitly considered in the FSAR. We therefore conclude that the proposed changes to the Technical Specifications are acceptable.
We further conclude, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will tte' conducted in compliance with the Comission's regulations and the Tssuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public.
Dated: October 17, 1978