ML19295C023
| ML19295C023 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 07/21/1975 |
| From: | Rusche B Office of Nuclear Reactor Regulation |
| To: | NRC COMMISSION (OCM) |
| Shared Package | |
| ML19295C022 | List: |
| References | |
| NUDOCS 8010150726 | |
| Download: ML19295C023 (11) | |
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
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In the Matter of'
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COMMO:n?EALTH EDISON COMPAh"l-
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Docket No. 50-10
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(Dresden 1:uclear Power Station Unit 1) )
COMMENTS OF THE DIRECTOR OF' NUCLEAR REACTOR REGULATION 01; *11!E REQUEST FOR EXF.MPTIG:: OF THE LEESDEI! i.UCLEAR PO'<!ER STATION UI:IT 1 FROM TliE REQUIRE!!ENTS OF 10 CFii~ SECT 105 50.4Y '
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on June 18, 1975, Commonwealth Edison Coopany requested an exenation from 10 CFR 50.46.
The excaption would relieve them of the obligation to operate Dresden Unit 1 in conformance with 10 CFR 50.46 until cor.ple-tion of certain modifications.
Commonwealth Edison indicates that the earliest completion date for these nodifications is estimated to be December 31, 1977.
The Of fice of the Director of 1:uclear Reactor Regulation submits the following information in support of the request.
On January 4,.1974, the Commission published in the Federal Register an amendment of 10 CFR Part 50~ which added a new section 50.46 to 10 CFR Part 50 (39 FR 1001). This section sets forth the acceptance criteria for Emergency Core Cooling Systems (ECCS) for light-water nuclear power reactors and requires that cach light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy cladding shall be provided with an acceptable ECCS, determined in accordance with an acceptable evaluation model.
The required and acceptable feature of such nodels are set forth in 10 CFR Fart 50, Appendix K.
The new regula-tion also provided that, with respect to reactors for which operating 80101507 M 8
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Licenses previously have been issued, each 'af fected licensee was required to the Director of Regu'.ction an evaluation of the ECCS perf or-to submit mance capability for the reactor involved and proposed 1icense amendments including changes in the Technical Specifications as may be necessary to bring reactor operation in conformity with the new regulation and to thereciter operate the reactor in accordance with the specifications.
Compliance with these new criteria was required by August 5, 1974, unless either (1) an extension of time for submission of required ECCS performance evaluation had been approved by the Director of Regulation to 10 CFR Section 50.46(a)(2)(iii), or (2) an exemption from pursuant the operating requirement of 10 CFl: Section 50.66(a)(2)(vi) had been grcated by the Co: mission.
An extension of time for submittal of th required evaluations for Dresden :ucicar Power Station Unit I was granted to Commonwealth Edison 6,1974 until. April 4,1975, by the Director of Regulation (CE) from August in his " Determination with Respect to Variance from the Interim Acceptance Criteria and Extension in Submitting Evaluations from the Acceptance Criteria for Emergency Core Cooling System" (10 CFR 550.46(a)(2)(iii)),
dated August 5, 1974 (39 FR 29611).
A further extension was granted on April 3, 1975 extending the date f or submission to August 2, 1975 (40 FR 16371, April 11, 1975).
Ey the August 5, 1974 determination, the Director of Regulation also granted a variance from the requirements of the Interim Acceptance Criteria (IAC) until September 1,1976 to provide additional modificat ions
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t to reduce the vulnerabili,ty of the Dres' den Unit 1 ECCS to failure of a single onsite power source.
The Determination required continued inservice inspections at triple the frequency required by the Technical Specifications and that CE diligently pursue compliance with the IAC.
In fiay 1972, CE submit ted an analysis which showed that as the result of an engineered safety system (the core spray system), in s t al le d af ter publiccLion of the IAC and systems which are not specifically designed as engineered safety systems (emergency condenser and the primary feedwater system), the fuel clad temperatures and metal water reactions would remain belou the limits specifi_J In th e IAC for any size primary system pipe break at any location.
Cocling for small and intermediate breaks would be provided by the primary feedwater system.
Cooling for large breaks, including a double ended rupture of the largest pipe below core level, would be provided by the new core spray system.
In February 1973, the licensee submitted additional analyses based on the use of the same systees and accounting for the ef fects of fuel densification which again indic c L ed that the IAC limits hcd been met.
The Regulatory staf f cvaluated the analyses and in a letter to the licensee dated February 22, 1974, it concludcJ that even with dcnsification effects, Core IX (the present core) meets the IAC lie:.t of peak fuel clad tempera-ture of less than 2300 F.
Analyses submitted by the licensee also indicated that for a spectrum of large break sizes below the core, including the doubic ended break of the largest line, the IAC limits could be met without reliance on the fe ed ua t e r system or offsite power.
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Therefore, in the event of a. loss-of-coolant accident (LOCA), it is highly unlikely that the IAC limits for fuel clad temperature and metal water reaction would not be met.
The requirement that the core temperature be reduced and decay heat be removed f or an extended period of time, as required by the long-lived radioactivity remaining in the core, is fulfilled by the existing core spray system operated in a mode which recirculates water within the containment building and powered by either onsite or offsite power.
Although systems are available which can maintain fuel clad temperatures below 2300 F af ter a LOCA, the redundancy and level of reliability of installed systems should be improved in consideration of long-tern operation, i.e., the system can be substantially reduced in effectiveness by a single failure independent of but coincident with a LOCA.
Tne feedwater system would not be operabic to provide coolant for small and intermediate size primary system pipe breaks if offsite power were lost ce "
ident with a LOCA.
Also the controls for the feedwater system vere not specifically designed to meet single failure criteria and have not been evaluated for their vulnerability to single failure.
The low pressure core spray system would not be available for large size prinary system breaks if the LOCA were accompanied by loss of all offsite power and if a single f ailure were to render the single onsite emergency diesel generator inoperable.
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Eccause the design of the presently installed ECCS when considering
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single failure does not meet the criteria of the IAC, it, is similarly deficient in regard to the core conservative temperature limits (2300 F versus 2200 - i stated in the criteria of 10 CFR 50.46.
CE has previously committed to modify its emergency core cooli,ng system by installing a High Pressure Coolant Inject ion (HPCI) system.
This system would supplement the existing low pressure core epray to provide protection against I OCAs over the entire spectrum of possible break sizes.
The licensee had scheduled the installation of the 1;PCI f or September 1976 as required by the August 5,1974 IAC variance.
CE also has committ ed to make additional modifications to the onsite emcreency electrical power supply system to remove the vulnerability of the present system to failure of single onsite power sources.
The IAC va riance, although granted until September 1, 1976, expires at such time as operation in conformity with the acceptance criteria of 10 CFR 50.46 is required, unless an exemption is granted.
Die staff believes that the justification that was adequate to permit a variance to the IAC is still valid and is equally applicable for the purpose of supporting an exemption to the accept ance criteria of 10 CFR 50.46.
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t In addition, CE submitted a preliminary evaluation of ECCS performance on November 1,1974, and further supplemented that evaluation on November 13, 1974.
The preliminary evaluation concluded that with present operating limitations the calculated peak clad temperature using the Appendix K models would b'c below 2200 F and that no additional operating limits are required.
To provide additional assurance that n' portion of the f eedwater system will be operational during a LOCA, the Technical Specifications have been modified to require that during pouer operation one fe ed wa t e r pump will be powered from the auxiliary transformer powered directly from the offsite power.
This configuration climinates the need to rely on an automatic transfer from the auxiliary transformer powered by the Unit I generator and provides added assurance that this feedvater pump will operate continuously throughout-a LOCA.
The Technical Specifications also have been modified to require that during reactor power operation the emergency reactor feedwater pump will be operable, and the surveillance requirement s have been modified to include a test of the operability of the emergency feeduater pump every 30 days.
The emergency feedwater pump can be operated by onsite power and can provide makeup for a small line break leaking up to 100 gallons per cinute while the reactor is being depressuriecd.
Die operability of this g.
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e pump provides additional assurance that adequate cooling would be available if of fsite power were lost coincident with a pipe break.
In consideration of (1) the e'xtremely low probability of a LOCA occurring simultaneously with a loss of all of f site power, (2) the capa-bility of primary coolant leak detection and inservice curveillance to discover leaks or potential Icaks before cracks can propagate appreciably, (3) the preliminary 10 CFR 50.46 analysis and (4) the acasures taken to increase the reliability of the feedwater pump, the staff has concluded
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that there is reasonable assurance that granting an exemption will not adversely affect the health and safety of the public.
In response to conditions imposed on the licensee as part of the IAC variance, CE has been submitting quarterly reports to the URC st af f regarding the status of the HPCI design and installation.
Report No. 1 dated October 25, 1974, indicated that cencral Electric had in f ormed them that the HPCI modifications would require the use of a pump that would supply emergency coolant at 5200 gpm at 100 psia in addition to the previous requirement of 1000 gpm at 1250 psia.
CE, upon contacting pump vendors determined that delivery of a pump that satisfied the new design criteria could not be accomplished before aid 1977.
In an attempt to ecet the previously committed date of September 1976 for llP.CI installation, CE evaluated whether the new pump performance criteria could be met by more readily available pumps and at the same v.
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V time issued new bi'd requests for culti-stage vertica1 dicsal driven pumps that would meet the new CE design criteria.
On February 4, 1975, CE submitted its second cuarterly progress report on llPCI status.
This report indicated that CE had modified its original specification for multihle-stage HPCI pumps by requesting
-bids for motor driven pumps powered from diesel generators rather than the diesel driven pumps previously specified.
It was hoped that a shorter delivery schedule could be achieved by this change, but pump vendor response was not encouraging.
In addition, CE reported the results of a survey performed by CE's consultant, Sargent and Lundy, to determine whether other readily available pu=ps could be obtained that would nect the GE design criteria.
'1he survey indicated that there were no core readily availabic pumps meeting the two flow condit ions which would result in an improvement in schedule over the use of multi-stage vertical pumps which were then out for bid.
In addition, GE was requested to determine if the Zimmer or LaSalle llPCS pumps already ordered by CE vould be acceptable taking into consideration the requirements of the Final Acceptance Criteria.
CE's February 1975 report further stated that the GE evaluation establishing the required size of the pumps significantly affects the size of the diesel generator, the size of the HPCI Euilding, and the arrangement of the rest of the system required for compliance with the FAC.
CE has indicated that its larast best estimate of the availability
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e of the required multi-stage vertical pumps, based upon sizing the pumps and placing an order within one conth of receiving the GE Appendix K ECCS evaluation which is needed to define the HPCI flow requirements would result in HPCI system completion in September 1978.
CE has further indicated that it believes that this schedule can be shortened by as much as nine months by paying a premium, exchanging pumps with the CE LaSalle County Station and/or ordering the pumps before final analysis is available.
CE has committed to pursue one or more of these methids in an attempt to have the HPCI system available by the end of 1977.
The URC staff has evaluated Commonwealth Edison's request for exempt ion and the affidavits prepared in its support.
We have revieucd the licensee's proposed schedule for completion of the modifications and the various equipment delivery schedules.
In light of the staff's knowledge of general installation requirements and procurement situation for nuclear reactor grade equipment and its recognition that CE has taken all reasonable steps th a t could result in a shortening of the HPCI installation schedule, the staff is satisfied that the proposed schedule for having all additional components installed and operable by December 31, 1977, represents a reasonable minimum time to complete the major modifications which are planned and required and that no other alternatives can be employed to further improve the schedule.
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staff therefore' concludes that good cause exists f'or an exemption
.from the requireacnts of 10 CFR 50.46 to December 31, 1977.
The steff further believes that it is in the public interest to allow Dresden 1 to operate until the. completion of the HPCI modifications on or before December 31, 1977, bascd on our conclusion that the public
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health and safety will not be adversely affected and on consideration of CE's assest. ment that the unavailability ~of Dresden 1 from August 3, 1975, to December 31, 1977, could necessitate the use of approxinately 260 million gallons of fuel oil or purchased power for replacement capacity representing a dif f erential cost increase of approximately 77 million dollars, the bulh of which would be passed on to the consumer.
For the reasons stated above, the staff recommends that CE's request for exemption from the requirements of 10 CFR 50.46 be granted.
The st a f f notes that the exemption as requested, and as conmented upon by us, relates solely to an exemption f rom the design and diversity of emergency systems or the diversity of emergency power sources which will be provided by the planned HPCI related codifications.
Dresden 1 will be required to meet all other 10 CFR 50.46 limitations and requirements.
FOR THE UUCLEAR REGULATORY C0hSISSION Nmg_
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I Ben C.
Rusche, Director
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Office of Nuclear Reactor Regulation Dated at Bethesda, Maryland, this 21st da'y. of July 1975.
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UNITED STA111S OF AMERICA NUCLEAR REGULATORY COM'4ISSION In the Matter of
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ComiONWEALTH EDISON COMPANY
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Docket No. S0-10
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(Dresden Nucicar Power Station Unit 1)
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CERTIFICATE OF SERVICE I hereby certify that copies of the " COMMENTS OF THE DIRECTOR OF NUCLEAR REACTOR REGULATION ON 'IllE REQUEST FOR EXEMPTION OF THE DRESDEN NUCLEAR PONER STATION UNIT 1 FROM THE REQUIRD4ENTS OF 10 CFR SECTION S0.46,"
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dated July 21,197S, in the above-captioned e.atter, have been served on the following by deposit in the United States Mail, first class, this 22nd day of July,197S:
John W. Rowe, Esquire Isham, Lincoln 6 Beale Counselors at Law One First National Plaza Chicago, Illinois 60670 i..
Commonwealth Edison Company
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Mr. J. S. Abel s.j 2
Nuclear Licensing Administrator -
Boiling Water Reactors v
Post Office Box 767 Chicago, Illinois 60690 Anthony Z. Roisnan, Esquire
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Berlin, Roisman and Kescler 1712 N Street, N. W.
Washington, D. C.
20036 DISTRIBUTION 045h"#'AY Docket 7%'#'*** M ORB #2 Read ng Reba M. Diggs RMDiggs Operating Reactors Branch #2 kg G
Division of Reactor Licensing oF len
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Form AEC.318 (Rev. 9 53) AECM 0240 Tr u. s. novspausm? PAINT 1MG OFFICEa G SM ? RG-166