ML17037C369
| ML17037C369 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 03/12/1971 |
| From: | Brosnan T Niagara Mohawk Power Corp |
| To: | Morris P US Atomic Energy Commission (AEC) |
| References | |
| Download: ML17037C369 (8) | |
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LS, ATOMiC,ENERGY ComusSiOIi eVKM(
Regulator v File Cy.
Dr. Peter A. Morris, Director' i v i s ion of Reactor Li cens i ng
'nited States Atomic Energy Commission Washington, D. C.
20545
Dear Dr. Morris:
DOCKETED USAEC MAR151971.~
5 REGUULTORY Ql NIL SECTION DOCKET CmK 6'O Re:
Docket No. 50-220 Your letter of February II, 197I forwarded a copy of the Advisory Committee on Reactor Safeguard's report dated February 6, l971 pertaini-ng to Niagara Mohawk's request for an increase from l538 to I850 thermal megawatts in the licensed power level of the Nine Mile Point Nuclear Station.
The recommendations and understandings discussed in the ACRS letter have been carefully reviewed by our personnel and the current status of each is set forth below:
Additional Safet Valve
~ One additional valve identical in all respects to the original l5 valves has already been installed on the reactor vessel head.
New Reactor Scram Tri s The Technical Supplement To Petition To Increase Power Level proposes modifications to Technical Specification 2.I.2 to include the turbine stop valve closure and turbine control valve high rate of closure trips described in the ACRS letter.
This specification requires both trips to be operative at all power levels above,45 percent of the I850 thermal megawatt rating.
Component installation is essentially complete; final interconnections and testing will be done before power level is raised above l538 thermal megawatts.
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l Timer Settinqs On The Emer enc Power S stem Timer settings governing initiation of will be modified to ensure a maximum core spray These modif'i'cations will be incorporated bef'ore thermal mega'watts.
In addition, it is proposed Specification 3.I.4 be changed accordingly.
emerge'ncy core spray pumps Initiation time of 35 seconds.
power is raised above l538 that the bases for Technical Refinement In Models* For= Evaluation Of Peak Clad Tem eratures General Electri.c has in preparati'on a topica'I report describing the technical bases and model assumptions used for evaluation of peak clad temperatures reached during postulated loss of coolant accidents.
It is anticipated that this report will be made available to the Regulatory Staff within a few weeks.
In addition, General Electric intends to continue its review'f the analytical methods and possible model refinements thereto..
Within approximately one year, the Regulatory Staff will be informed of the results of this review and any significant changes to previous Nine Mile Point analyses.
Reduction Of The Allowable Containment Leak Rate The Technical Supplement To Petition To Increase Power Level proposes a modification to,Technical Specification 4.3.3 incorporating this reduction.
Further Studies Of The S ent Fuel Pool Studies of possible corrective measures to assure adequate integrity of the spent fuel pool in the postulated event of dropping a fuel cask into the pool are progressing.
The results of these studies will be reported to the Regulatory Staff when completed.
It is our intention that appropriate modifications be made to the Station before first use of the cask over the pool.
In-Service Ins ection Of The Main Steam Lines It is our.intention,to begin the expanded program for in-service'nspection of the main steam -lines both inside and outside the containment at the next major outage.
In addition, it is proposed that Technical Specification 4.2.6 be modified to include this expanded program.
Inte rity Of The Biolo ical Shield Analyses of the integrity of the biological shield upon postulated failure of a reactor vessel safe end have been reported by our letters of July 2, l970 and November 23, l970 and in the Second Addendum to the Technical Supplement To Petition To Increase Power Level.
It is our.understanding that these analyses demonstrate the adequacy of the as-built shield; consequently, no further action is contemplated.
Im roved Leak Detection It is our intention to install, during the next major Station outage, an atmospheric radioactivity monitoring system which will recirculate a portion of the containment atmosphere through an external loop and an air monitor.
Continued Assurance Of Reactor Pressure Vessel Inte rity
'Our letter of September 24,
- 1969, responding to an earlier ACRS report, indicated that throughout the first five years of Station operation, inspection experience will be carefully evaluated'nd pertinent industry-wide development
- efforts, such as those sponsored by the Edison Electric Institute, will be closely followed to improve means for continued assurance of pressure vessel integrity.
In the course of these efforts, we intend to be particularly alert to possible improvement in access to reactor vessel surfaces for inspection purposes.
It will continue to be our intent to implement to the degree practical any significantly improved inspection technique which may develop in the course of this program.
H dro en Control In The Containment At the December I970 ACRS meetings, we described active studies underway for control of the buildup of hydrogen in the containment which might follow the unlikely event of a loss of coolant accident.
These studies include further review of basic data and analytica I assumptions, efforts to develop a suitable containment vent procedure, and refinement of concomitant dose calculations.
It is our intent to review our progress in this area with the Regulatory Staff within approximately one year.
Common Failure Modes And Failure To Scram We are continuing our review of General Electric studies for preventing common failure modes from negating reactor scram action and consequences of failure to scram during anticipated transients.
It is our understanding that a
further General Electric study of a generic nature on this topic will be forth-coming within two to three months.
It is our intention to review this study upon completion and to remain current on the possible ramifications of these analyses and any design features they might suggest to make tolerable the consequences of such an unlikely event for the Nine Mile Point Nuclear Station.
This matter will be reviewed again with the Regulatory Staff within approximately one year.
We have previously responded to ihe items mentioned in The ACRS reports of Apri'I l7, l9'69 and June l6, l970 by our letters of September 24, l969 and July 2, l970 respectively.
Very truly yours,
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Brosnan Vice President and Chief Engineer
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