ML20128E931

Forwards Request for Addl Info Re Application for CP

ML20128E931
Person / Time
Site:	Millstone
Issue date:	01/14/1966
From:	Doan R US ATOMIC ENERGY COMMISSION (AEC)
To:	Hicock R, Moses M, Switzer D CONNECTICUT LIGHT & POWER CO. (SUBS. OF NORTHEAST, HARTFORD ELECTRIC LIGHT CO., WESTERN MASSACHUSETTS ELECTRIC CO.
Shared Package
ML20125A422	List: IA-92-198, Requests Written Answers Be Provided to Questions Developed During Discussions at Meeting on 660218,for ACRS Review. Lists Specific Questions Which Applicant Should Be Prepared to Discuss ML20125A419 ML20125A429 ML20125A437 ML20125A965 ML20125B006 ML20128E705 ML20128E860 ML20128E913 ML20128E931 ML20128E997 ML20128F017 ML20128F066 ML20128F091 ML20128F095 ML20128F190 ML20128F246 ML20128F283 ML20128F339 ML20128F355 ML20128F370 ML20128F400
References
FOIA-92-198 NUDOCS 9212080214
Download: ML20128E931 (9)

Text

,

W

, 4 i

[/ 'N 1

• l -

I*

UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON, D.C. 20545 m pnm nunn w

'f; ' '1 --

Docket No. 50-245 N&[(b h

(See attached list tse adderr:::e} I 3

/

I W k'

,w' Gentlemen:

This refers to your application dated November 10, 1965, for a construction permit and facility license which would authorize construction and operation of a nuclear power reactor at Millstone Point in the town of Waterford, Connecticut.

As you know, a meeting was held on December 22-23, 1965 with representatives of your company and the Regulatory Staff to discuss your application. During this meeting a number of technical areas concerning the design of the station were discussed, and it was concluded that additional written information would be required to continue the staff review. 'Accordingly, you are requested to provide the information listed in the enclosure.

It will be necessary to continue to conduct our technical review expeditiously to enable final action on your application for a construction permit by your requested date of June 1,1966. To this end, we urge that you place particular emphasis on providing full and complete answers to each of the attached questions, so that further questions covering the same material will not be

-required. The staff, of course, will be _available as may be required to discuss and amplify the meaning of the questions.

The additional information should be submitted as an amendment to your application.

R E C E fi'- ~

,1 *, s ,, /

' 'I o.sTiuMA.EMc~E~'r t"'a /. -

'g h. 1 . . i: .

,, a/ni8Cfa " ' - ' *

///i AM I

eq S ,9.1011121; .,* . ; .

e ,

}

g 20 g 4 920611

~~

2 M , s LAWRENC92-198 PDR Z. CD

I i

=

J;. 1 .

3 As discussed at the December 22 23, 1965 meeting, our review to date has not included the seismic design criteria for the proposed facility.

We are currently evaluating this aspect of the proposed facility, and will inform you if additional information and/or a meeting will be required.

Sincerely yours, 0^ 'redBy R.L ha R. L. Doan, Director Division of Reactor Licensing

Enclosure:

As stated above cc: Mr. C. Duane Blinn Day, Berry & Howard One Constitutional Plaza Hartford, Connecticut 06103 i

l

i 141.'63 MILLSTONE NUCLEAR POWER STATION DOCKET NO. 50-245 The Connecticut Light and Power Company P. O. Box 2010 Hartford, Connecticut 06101 Attention: Mr. Russell Hicock Vice President The Hartford Electric Light Company P. O. Box 2370 Hartford, Connecticut 06101 Attention: Mr. Donald C. Switzer Vice President Western Massachusetts Electric Company 174 Brueh Hill Avenue West Springfield, Massachusetts 01089 Attention: Mr. Marlowe G. Moses Executive Vice President The Millstone Point Company P. O. Box 2010 Hartford, Connecticut 06101 Attention: Mr. Russell Hicock President

MILLSTONE NUCLEAR TOlER STATION The Connecticut Light & Power Company The Hartford Electric Light Company Western Massachusetts Electric Company The Millstone Point Company Additional Infors,ation Required A. General

1. a. Tour attention is directed to a letter from the ACRS to the Chairman, AEC, dated November 24, 1965 concerning reactor pressure vessels. Please discuss the consideration which has been given in the design of your station to the recommendations contained in numbered paragraphs 1 and 2 of the ACRS letter.

b. In addition to and related to the above (a), please refer to the ACRS letter to the Chairman, AEC, dated November 24, 1965 concerning its review of the Commonwealth Edison Company, Dresdca 2 facility. The Committee in paragraph 2 requested " continued studies of pipe whipping and the generation of missiles which might violate the _ containment. . . . .".

Further, the effects of fuel movement upon reactivity were also indicated to be of concern. Consequently, please discuss the consideration which has been given in the design of your station to these areas of concern.

2. Please describe the method by which the NDT (nil ductility transition temperature) of the reactor vessel is it.itially determined. Discuss the basis for selection of the initial NDT of the reactor vessel taking into account the overall uncertainties and variations in the properties of the vessel plates.

3. In regard to the station's 'af ter-heat' removal capability, numerous references are made in the Design and Analysis Report to an " equivalent" system as a substitute for the Isolation condenser. Since this " equivalent" system has not been identified or discussed, please clarify the intent for its inclusion and/or provide necessary information as to its design, desc ription, and functional capabilities for our evaluation.

4. Tha Commission published " General Design Criteria for Nuclear Power Plants" on November 22, 1965. Please review these criteria and provide a comparative tvaluation with the design criteria used for the Millstone Nuclear Power Station. Your evaluation should include a discussion of how the objectives of the Commission's criteria are fullfilled by your design criteria. In addition, identify and analyze those areas, if any, in the station design that do not meet the Commission's criteria.

]

< a q

l B. 0verall Station Design & Operation 1

1. In view of the recent ' blackout', it is apparent that auxiliary power from .

of f-site sources may not be available for all required safety systems. In this situation, full reliance for reeoval of core decay heat and hence for public safety would be placed on the single emergency diesel generator )

proposed for the station. Consequently, please provide the design basis and justification for the adequacy of a single emergency diesel generator.  !

Consideration of an alternate backup system may be provided as part of your evaluation.

8'. It is indicated in Section XIII of the Design and Analysis Report, that - ,

training of responsible personnel would be accomplished at the Yankee '

(Massachusetts and Connecticut) facilities. Since these plants are both i pressurized water systems and the Millstone station is to be a boiling water.

facility, please discuss your overall pre-operational program wherein responsible personnel would receive addtional training on a similar reactor plant.

3. The Design and Analysis Report contains noteworthy discussions on the overall dynamic plant characteristics as related to stability. These discussions include the identification of the interacting parameters and a description of the analytical model used for your investigations; however, specific quantitative results of dynamic anelyres as related to design and 'off-normal' '

conditions have not been included in the report. Consequently, so that we-may reach definitive conclusions concerning reactor stability, please provide and discuss the results of your studies to evaluate andfanalyze the dynamic characteristics of the Millstone Nuclear Power Station. The - in forma tion provided should indicate the overall design criteria for assuring station stability and how the design features will be selected to implement the objectives of the criteria. It should include an identification ot particular parameters (or design variables as given on page IX-2-5 of .the report), their design values, and associated gain and phase margins.. In addition, the changes in these parameters as effected by normal maneuvering transients and

'off-normal' conditions that could further_ change ltheir-values and related stability margin should be-included in your reply. In the latter case the effects of either equipment or control system failure should be considered, i.e., partial or complete _ loss.of recirculation pumps, failure in the Flow Control System, partial or complete loss' of Feedwater heating, f ailure in the pressure regulator, and etc.

4 '. Please identify and discuss the design bases related to the adequacy of the-Station Control Room shielding to assure the maintenance of required watch stations.under post-accident conditions. -In addition, please consider.

potential- exposures that could occur during access - to and egress from the control room under similiar conditions.

i

3 C, Containment & EnRineerinR SafeRuards Systems DesiRn

1. Please provide a complete list of all significant differences between the proposed containment designs for the Millstone and Dresden 11 stations and discuss the purpose and justification of these differences, and their effect on safety.

2. The leakage rate specified for the Millstone Nuclear Power Station is reported in the Design and Analysis Report to be 0.5% of the free volume of the con-tainment per day. Since the leakage rate is a function of various parameters, including containment pressure, temperature, and environment, please identify the attendant design basis accident conditicas related to the specified leakage rate. Assuming that the peak accident pressure is the corresponding pressure for the given leakage rate limit, please indicate how the leakage rate limit would differ if the peak accident pressure changed to approach the design value. An example would be to consider a partial pressurization cf the drywell prior to the postulated large pipe rupture. This could cause the peak pressure to increase from 46 to 52 peig.

In addition, discuss the relationship between leaba<a rates under test and accident conditions and consider how the leaksge rA-a under accident conditions would be extrapolated to the permissible limit under test conditions. Although this information may appear to be premature for the immediate requirements of a construction permit application review, its consideration is necessary at this time since such consideration of testing capability could lead to design changes.

3. Please provide a graph showing the containment leakage rate (under accident conditions) as a function of time following the postulated rupture of the Recirculation Line for the design basis accident. (Curve F on Fig.115)

4. Figure 59 of the Design and Analysis Report is a composite P&I diagram of the proposed Millstone Point Facility. In addition to the Nuclear Steam Supply system, the engineering safeguards identified as the containment spray, standby liquid control, and the core spray systen are also shown on the diagram. A review of the proposed engineerius safeguard systems indicates that although there exists redundancy An the pumps, pipe lines, and valves, ultimate operation of these systems is predicated on administrative control to assure that certain valves, some of which are inside the containment, are open during station operation. Consequently, please provide your justification and bases for using valves in critical safety systems which are manually operated without remote indication. What assurance would be provided to the operator that these valves are indeed open pric~ to securing the containment in preparation for station operation?

5. Please provide an analysis to show why the presently proposed containment spray system is considered more reliable than that proposed for Dresden II which consisted of two independent systems each redundant in pumping capacity. It-would appear that if one system of the Millstone Plant were disabled, every component of the second system would need to function to oitain the required spray flow. In addition, discuss the considerations, other than test conditions, which led to the requirement for a cpray over the suppression pool.

4

6. Discuss the validity of the rod worth minimiter concept in relation to the case in which a rodia stuck full-in while its drive is withdrawn. Subsequent changes in the surrounding rod patterns are made within the limits permitted by the rod worth minimiter and the indicated position of the stuck rod. What is the highest worth of the stuck rod which could be obtained for this case (1) at hot standby and (2) at power?

7. In assessing the _ capability of the Control Rod Thimble Suppsrts, it is apparent that the communication provided by the collet fingers between the control rod index tube and thimble housing is required to achieve design performance. Please consider the case in whica the collet fingers are released from the control rod index tube in anticipation of rod movement and there is a complete circumferential break in the control rod thimble resultins in an unbalanced force due to the pressure differential across the contrai rod. In this situation we could assuae that the control rod could move downward for at least 12 inches (equivalent to 2 notches) before the collet fingers become engaged and latch the rod. I additional movement of the entire assembly would then occur until the support frame is contacted.

In regard to the foregoing, please provide the following informationt (a) Discuss the design capability of the thimble support to withstand the total downward force of the control rod and thimble housing.

(b) Discuss the effects of the accelerating rod contacting the collet fingers.

(c) Could the circunferential break of the thimble in turn produce a situation in which the control rod could not be engaged and held by the collet fingers upon contact with the support frame? If so, please provide an evaluation of the consequences of a rod ejection accident assuming failure of the collet finEers to couple the control rod to the thimble housing.

D. Jgfety Analyses

1. Pith respect to the design basis accident for the containment:

a. Please provide a discussion of the basis on which the melting point of zirconium was chosen as the termination temperature for.the ricconium-water reaction. Include a discussion of any experimental data in the literature, which might support this contention. What additional amount of zirconium-water reaction would take place if the termination- temperature was the melting point of the zirconium oxide rather than that of zirconium?

b. The calculational model used to obtain the capability of the containment to withstand large amounts of metal-water reaction (Figure 118) has been discussed with the staff in recent meetings. Please provide documentation of the assumptions and model used to obtain this curve. In addition, if the zirconium-water reaction occurred up to the aelting point of the oxide, what would be the pressure in the containn.et.t and what would be the affect i on the containment capability?

g 5-a

c. Please rectify the inconsistency between Figures 114 and 115 with respect; to the pressure plateau reached prior to actuation of containment _ spray.

3. Please clarify the "at power" rod drop accident to indicate whether the peak (

enthalpy in' cal /gm indicsted in Figure 111 is _for the entire reactor _ orc for l the assumed excursion zone only. What is the maximum enthalpy increase which a core hot spot (above 100 cal /gn) could experience from a rod drop in a nearby area of the core (consider low worth rods close to the hot spot' as well as high worth rods at greater distances)?

3. Considering the everall thermal and hydraulic effects of a major pipe rupture, please idantify ti.e design features related to core structural capability that would be available to ensure control rod insertion into the core thereby-preventing a subsequent criticality ennasion upon core reflooding.

4. Please discuss the model differences between the fuel drop and rod drop i accidents. In particular what physical or flux shape parameters vary to ,

explain the fact that the total reactivity inserted and the drop rato !

postulated for the refueling accident cannot_be superimposed on Figure 109.

E. Instrumentation

1. Please discuss the safety system capability to protect the reactor (no fue1~

damage) against excursions induced by the simultaneous withdrawal of any;two rods under all conditions of allowed bypass of protection channels and individual chamoers.

2. How will the design features of the Intermediate and Power Range safety amplifiers protset against overload and possible failure to function during -

those excursions which they are designed to terminate?

We have observed- that actuation of the test twitches in the neutral sides of 3.

the solenoid scram pilot valves has the effect of tying together the.two buses of the otherwise mutually independent dual bus system. One consequence is!that, under. certain conditions, the current from one bus can tend- to hold : closed -

the ?alves in the other bus. Please provide the justification and basis'for this design.

4. Are the circuit breakers which protect _ those lines feeding equipment essential to the- operatior, of engineered safeguards -capable of responding to dead shorts occurr_ing at the lines? Please provide the- justification for your answer.

5.- What provisions are being made to ensure containment isolation in the event of an accident which results in rupture of thimbles used for instrument calibration

- (TIP System)?

I l

=1 l

1

.d- e #

i*

6-F. Site Heteorology

1. The meteorological dispersion models used do net appear to be suf ficiently-related to the _ existing peninsula site in that (a) the assumption of constant stack height to a distance of nine miles does not take into account hills which rise to 100 feet at a distance of 1 1/2 alles, 250 feet at 5 miles and 500 feet at 10-uiles, and (b) the assumed fumigation conditions did not consider the cooling of the lower layers over a cold water surface and then subsequent travel over a warmer land surface which might give significantly longer periods of fumigation. Please discuss the above comments with regard to the doses calculated for the major accidents.

t l

i i