ML19319D383
| ML19319D383 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/24/1972 |
| From: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Rodgers J FLORIDA POWER CORP. |
| References | |
| NUDOCS 8003160216 | |
| Download: ML19319D383 (6) | |
Text
- _ _
'?
'.y, v, q;y;.;;
,4,
,3.,7;
?%ISTRIBUTIC
_; L'-M MI!
[
,.* ~';'o e
'k ~
~" '
Local PDR PWR Branch Chiefs j.'A j.' ~.
,HDeat'on y f,\\(,,.
Docket
-RWKlecker RP Reading OGC PWR-4 Reading RO (3)
SHHanauer HJFaulkner JUL 2 ; y2 RSBoyd VHWilson (2)
Docket No. 50-302 RCDeYoung DSkovholt FSchroeder RRMaccary Florida Power Corporation DKnuth ATTN: Mr. J. T. Rodgers RTedesco Assistant Vice President &
-Nuclear Project Manager 7
P. O. Box 14042
'A St. Petersburg, Florida 33733 t,*-
a Centlemen:
On the basis of our continuing review of the Final Safety Analysis Report for Crystal River Unit 3 Nuclear Generating Plant, we find that we need additional information to complete our evaluation. The specific information required is listed in the enclosure.
Because of the potentially afsnificant effect of these items on continued construction, we will need your reply by August 18, 1972. Please inform us within seven (7) days after receipt of this letter of your confirmation of, the schedule or the date you will be able-to meet.
Please contact us if you desire any discussion or clarification of the
material requested.
Sincerely, Original signed W J dt.D'5 R. C. DeYoung, Assistant Director for Pressurised Water Reactors a~-
Directorate of Licensing o
~
Inclosure:
Request for Additional Information ccr Florida Power Corporation ATTN: Mr. S. A. Brandimore
/
Vice President (i General Counsel
/
P. O. Box 14042 St. Petersburg, Florida 33733
~
OUICE >
. I.
!".h
.h
.4
_.LIAD y
k
<[N SURNAME >
.. ef,1fAR,45 h m m W'
h
,,, _,, g p
DATE>.?k 12.........7.he D2 3.f.N.J. m......
, \\{'~ j)y Forum AIC-3tS(Rev 9 53[ABCM 0240 Ig
'e**
. Ea-asLstaae-: 4..-efs
.),
- '). 0 ~
.atu.
t*x..
80 03160 21 f k-
s l
I l
REQUEST FOR ADDITIONAL INFORMATION FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302 2.0 SITE AND ENVIRONMENT 2.14 Your reply to our request 2.1 is not responsive. From our review of the PSAR of the design bases for protection of safety related facilities against hurricane !nduced flooding, we concluded that upon final establishment ok probable maximum hurricane criteria by ESSA (now NOAA) further evaluation of the Crystal River Unit 3 site would be required. On this subject the staff safety evaluation to support construction permit issuance states "the applicant has stated, and we will require, that the plant protection will conform to the applicable portions of revised ESSA criteria."
The FSAR reports the use of criteria that we and our consultant, the U. S. Army Coastal Engineering Research Center, do not believe would cause the worst flooding conditions at the Crystal River Unit 3 site. Our preliminary estimate leads us to believe that the surge could reach approximately elevation 123.4 f t. (Florida Power Corporation datum) or 35.4 ft. above mean low water (MLW) as compared to your estimate of 24.6 ft. MLW. Correspondingly higher estimates of wave action could occur coincidentally with such an event. Protection of plant safety related facilities is not con-j sidered adequate in view of (a) the discrepancy in surge level, (b) your earlier commitment to provide protection against a probable maximum hurricane, (c) the necessity of providing a consistently high degree of protection to safety related facilities, and (d) recent hurricane experience and the formulation of more stringent probable maximum hurricane criteria.
Consequently, we require that you reply in entirety to our request 2.1 in the letter of R. C. DeYoung to Florida Power Corporation dated January 17, 1972.
g 5.0 CONTAINMENT SYSTEM AND OTHER SPECIAL STRUCTURES 5.30 Your reply is not responsive to our request 5.5.
A fully responsive answer is required, and in particular you should state the methods used to obtain information on the in-place concrete and provide the criteria used to evaluate the adequacy of this concrete.
5.31 Recent experiences have indicated significant construction problems relating to steel placement and detailing, and concrete placement in critical areas of dhe containment. Describe the provisions that will prevent the occurrence of such problems for this facility.
5.32 Distress in the done of another prestressed concrete containment led investigators to evaluate the effect of the loss of gross section resulting from the presence of the tendons. As a result of laboratory tests, it was concluded that for this particular arrangement of tendons the membrane stresses would increase by approximately 33%. Provide information regarding the effects of a change of this magnitude on the computed stresses and the effects of such a change on the margins of safety for this facility.
I
10.0 STEAM AND POWER CONVERSION SYSTEM 10.8 For Figures 10-2 and 10-2a of the FSAR, identify by number the
" low-load," "startup," and " block" valves discussed in the text of the FSAR.
10.9 For the normal and emergency feedwater systems, specify by number those valves which are automatica.11y actuated on the following signals : (a) reactor trip, (b) low steamline pressure, (c) low down-comer level. Also specify the actuation setting for each.
10.10 During a steamline break accident, demonstrate that a failure of any feedwater isolation valve will not result in continued feedwater addition to the affected steam generator (s) or that the consequences of such a failure can be tolerated safely.
i
o m
14.0 SAFETY ANALYSIS 14.6 For the steamline failure accident:
14.6.8 Discuss the possibility of any feedvater isolation valve to reopen automatically because of primary system pressure recovery, return of the reactor to power, or Integrated Control System (ICS) response.
14.6.9 In Table 3-8 of the FSAR the original moderator temperature coef-ficient of -4.0 x 104 (Ak/k)/*F has been amended to -3.0 x 104 (Ak/k)/*F. Furthermore, the note accompanying this table indicates that a one-dimensional, isothermal calculation of the moderator temperature coefficient was performed. This calculation could be non-conservative {orthesteamlinebreakaccident. Justify the use of -3.0 x 10 (Ak/k)/*F considering spatial temperature distribution and feedback effects.
14.15 The simultaneous rupture of four steamlines outside the reactor building is briefly discussed in section 14.2.2.1.3 of the FSAR in response to our request 10.7.
Provide the results for the double-ended rupture of all steamlines assuming the reactor is initially operating at hot standby (zero power), 60% power, and 100% poser conditions. Assume the single most reactive control rod is stuck out and use a tripped rod worth equal to 1% Ak/k plus the power defect at each condition. The result should include the following as a function of time:
- a. Core power and fuel surface heat flux,
- b. Primary system pressure and level,
- c. Average reactor coolant temperature,
- d. Outlet temperature and pressures and levels of the steam generators,
- e. Reactivity expressed separately for coolant, fuel, and control rods and total reactivity,
- f. DNB ratio (indicate applicable correlation).
14.16 For the dcuble-ended rupture of all steamlines with concurrent loss of off-site power, provide the following:
s
,,s i
i 14.16.1 A detailed explanation of the actions and their timing that would be taken to remove decay heat and to achieve a controlled reactor cool-down after automatic feedwater isolation. Include such important events as startup of the motor driven emergency feedwater pump and emergency diesel (if required), positioning of important valves, and changes to electrical distribution busses (if required). Discuss the power demand on the station batteries and indicate the maximum time the plant could be maintained in this condition safely.
14.16.2 The results of an analysis which considers the effects of automatic feedwater isolation and subsequent operator action to remove decay heat and achieve cooldown with the reactor coolant pumps tripped off.
In your results provide a plot of core flow rate (from natural circulation) as a function of time. Also consider the effects of any power transient that might occur due to the rapid cooldown of the primary system.
14.17 Because the consequences of the simultaneous rupture of all steamlines have not been reviewed previously, provide the following information with regard to the analytic modeling of the accident:
14.17.1. Describe the thermal-hydraulic models of the primary and secondary system used in the steamline break accident analysis. Include the following:
(a) the equations involved and the method of solution, (b) the number, location, and volume of nodes, (c) the heat transfer correlation used in the primary and secondary sides of the steam generator, (d) the steam generator blowdown model including the break flow model, L/D ratio, discharge coefficient, exit quality or amount of super heat, (e) the feedwater flow history during the accident, (f) the pressurizer level model.
14.17.2 Describe the neutron kinetics model including equations, reactivity feedback mechanisms,.nd appropriate input data such as neutron life-time, delayed neutron precursor concentration, moderator and fuel
~
temperature coefficients, boron concentration and worth, and tripped rod worth.
14.17.3 Describe the manner in which the thermal-hydraulic and neutronic calculations are combined to calculate DNB ratio.
14.17.4 Provide confirmation of the models, techniques, and assumptions to assure that the analytic predictions for this ' type of accident are conservative in relation to the actual physical phenomena which would occur.
, _