ML19256F517
| ML19256F517 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/08/1979 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML19256F505 | List: |
| References | |
| ACRS-1648, NUDOCS 7912190311 | |
| Download: ML19256F517 (4) | |
Text
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APPENDIX XXVI Millstone 2: Project Status Report
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WASHINGTON, D. C. 20555 s.,
m ms June 8, 1979 PROJECT STATUS REPORT - MILLSTONE NUCLEAR POWER STATION, UNIT 2 - POWER LEVEL INCREASE Backaround Applicants: The Connecticut Light and Power Company (CL&P), The Hartford Electric Light Company (HELCO), Western Massachusetts Electric Company (WMECO), The Millstone Faint Company (Millstone).
CL&P, HELCO and WMECO are engaged principally in the production, purchase, transmission, distribution and sale of electricity.
Millstone acts as the representative of CL&P, HELCO and 'JiECO with respect to design, construction and operation of the Millstone Nuclear Power Station.
It has no ownership interest in the station.
Outstanding capital stock of each of the four applicant companies is owned by Northeast Utilities. Millstone uses Northeast Utilities engi-neering staff.
Location: The Millstone site is located in the town of Waterford, New London County, Connecticut on the north shore of Long Island Sound.
The 500 acre site is 3.2 miles WSW of the New London town limits and 40 miles SE of Hartford. The nearest population center per 10 CFR 100 is New London which has a population of 31,630 (1970). There are no residences within 1/2 mile of the plant and only 147 people within 1 mile (1970).
Facility
Description:
NSS Vender: Combustion Engineering - PWR A/E: Bechtel Vessel Vendor: Combustion Engineering Containment Constructor: Bechtel A layout of the reactor coolant pressure boundary is Attachrent A.
Pertinent data on Millstone 2 taken from ORNL-NSIC-55 entitled " Design Data and Safety Features of Comercial Nuclear Power Plants" is found in Attachment B.
Although some data have been superseded it is provided for information only. A comparison to other plants with similar character-istics are shown in Attachment C.
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Power Level Increase Previous power level increases in plants reviewed by the ACRS are listed below:
power increase i
NSSS/ TYPE Maine Yankee 2440-2630 7.7 CE/PWR Zion 2760-3250 7.75 W/PWR Indian Point 3 2760-3025 9.6 W/PWR Millstone 2 2560-2700 5.5 CE/PWR The ACRS letters on the above plants that was granted power increases are found in Attachment D.
The Staff evaluation of Cycle 3 operation at power levels up to 2700 MWt has been completed and the Safety Evaluation report is found in Amendment No. 52 to the Facility Operating License for Millstone 2.
Some of the changes needed are as follows:
. modified (slowed and reduced flow) guide tubes for the control element assemblies
. credit taken for charging pump flow in the LOCA analysis
.a new reactor protection system trip from a reactor coolant pump speed sensing signal
. installation of a new neutron shield.
The Appendix A Technical Specifications were revised by:
. Incorporating changes resulting from the analysis of Cycle 3 reload fuel at 2700 MWt
. Adding a containment air recirculation system response time
. Approving operations at reduced powar level with inoperable main steam line safety values.
. Removing the completed special steam generator surveillance requirements.
In addition, the safety evaluation supporting this power level increase addresses ~ the Staff's evaluation of:
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- .New meteorological data
. Fuel handling accident inside containment
. Engineered safety features component leakage outside containment
. Control room habitability of the postulated LOCA
. Environmental qualifications of safety related equipment
. Containment electrical penetrations
. Piping and support systems.
Many of the Chapter 14 type incidents were reanalyzed because of the power level increase. A table showing the reanalysis status is found in Attachment E.
A table of major core parameters assumed in the safety analysis for Cycle 3 is shown in Attachment F.
Additional Information Concerning, the implications of TMI-2 experience with the core damage, I & E Bulletin No.76-06B which applies to CE facilities was issued.
According to the Staff the licensee response "... indicates that the actions taken by NNECO demonstrate understanding of the salient concerns arising from the TMI-2 accident...." A separate safety evaluation will be issued regarding NNECO response to ISE Bulletin No.79-058.
Millstone 2 was identified as using a computer code which incorrectly summed earthquake loads algebraically. Six piping systems affected required reanalysis. The licensee has performed the reanalysis and states that all the affected systems meet the applicable criteria of the plant FSAR. The Staff agrees with the licensee results.
A survey of the plant operating history from January 1978 to April The information on LER's, average daily power, operating 1979 was made.
status and unit shutdown / reductions were obtained from the " gray" book (NUREG-0020). These are shown in Attachment G.
The effects of power level increase was briefly reviewed by the Staff.
but only in a small way are as follows:
Those that are affected,
.II-7 Fuels under abnormal conditions
.IIC-4 Reactor Vessel Supports, Asymmetric Loads vious ACRS letters on plants requesting power level increases, the ACRS requested additional information except in the Maine Yankee review.
In r4 The additional information requested are is fbilows:
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. Zion 1&2 (ACRS letters dated 6/9/76 and 6/17/77) 1 4 ') X X 3 0_t_
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- 1) system interaction
- 2) backfitting
- 3) instrumentation to follow the course of an accident
- 4) loose-parts monitoring
- 5) fire protection
- 6) ATWS
- 7) Reliability of ECCS
- 8) Reliability of the diesel generators
- 9) Quality Assurance
- 10) Industrial Security All the above issues have been resolved except the concern on backfitting although the resolution of the system interaction issue is still questionable.
This issue is now under the purview of the Reactor Operations Subcomittee.
The Zion letter states specifically that the Staff should review the issues with the ACRS at a specified time.
. Indian Point 3 (ACRS letter dated 7/13/78)
- 1) system interaction
- 2) backfitting
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- 3) instrumentation to follow the course of an accident d
- 4) capability for safe shutdown and residual heat removal, using only safety grade equipment.
A recent call to the Staff indicates that a status report is due on July 1979. At that time the Subcomittee may want to plan a meeting with the Staff.
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