ML17221A322
| ML17221A322 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 08/05/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17221A321 | List: |
| References | |
| NUDOCS 8708190054 | |
| Download: ML17221A322 (11) | |
Text
. R
~ps REGS, (4
Np I
/p >>>>*w>>
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 ENCLOSURE s
,SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PELATING TO STEAM LINE BREAK ANALYSIS FLORIDA POWER AND LIGHT COMPANY ST.
LUCIE UNIT 1 DOCKET NO. 50-335
1.0 INTRODUCTION
By letter dated November 8,.1985, the Florida Power and Light Company the licensee for the St, Lucie Unit 1 power plant, submitted information on a
steam line break confirmatory reanalysis fn compliance with the requirements of Amendment 63 to account for asymetric thermal-hydraulic and neutronic system behavior (Ref.
- 1).
The-reanalysis*as-beeTI performed-by 'the Exxon Nuclear Company and is included in the report XN-NF-85-85(P) (Ref. 2).
Four scenarios are included in this reanalysis to account for initiation from Hot Full Power (HFP) and Hot Zero Power (HZP), with availability of offsite power for the operation of the primary system pumps and with loss of offsite power after. the break and one RCS pump tripped.
The limiting transient scenario in this analysis is different than the previous analysis, in large measure due to the different methodology utilized.
The transient scenarios were evaluated from the standpoint of fuel failure.
The HFP with offsite power available is the most limiting case for which fuel damage and offsite dose rates were evaluated.
This review is limited to the estimation of the fuel damage and does not consider environnental doses or dose rates.
.2.0 EVALUATION 2.1 The Transient Scenario A break in the main steam system increases the rate of heat extraction by the steam generators and causes cooldown of the reactor coolant.
Mith a negative 8708190050 870805 PDR ADOCK 05000335, P
moderator coefficient of reactivity, the cooldown will produce a positive reactivity addition.
The decrease in. main steam pressure will initiate a reactor trip (low steam generator pressure) and close both main steam isolation valves.
Additionally the auxiliary feedwater system will initiate automatically.
If the break occurs between the steam generator and the isolation valve, blowdown of the affected steam generator continues.
Flow from the intact steam generator stops with closure of both isolation valves, either of which is capable of stopping flow.
Since the steam generators are designed to withstand reactor coolant system operating pressure on the tube side with atmospheric pressure on the shell
- side, the continued integrity of the reactor coolant system barrier is assured.
A description of the course of events following a steam line break follows:
a)
Following a steam line break, reactor trip is initiated by low steam generator pressure (high power level provides a backup trip).
b)
The fast cooldown following a steam line break results in rapid contraction of the reactor coolant.
After the pressurizer is emptied, the reactor coolant pressure is assumed to be equal to the saturation pressure corresponding to the highest temperature in the reactor coolant system.
c)
The safety injection actuation signal is actuated when the pressurizer pressure drops below the setpoint.
Pump acceleration, valve opening, and
'flushing of the unborated piping require additional time'".: Manual 'trip -of
-" -.-=:
the reactor coolant pumps is required on initiation of safety injection.
d)
The rate of temperature reduction in the reactor coolant system increases with break size and with steam pressure at the point of the break.
Therefore, as a worst case it is assumed that a circumferential break of a 34-inch (inside diameter) steam line occurs at the steam generator main steam line nozzle.
The steam line break outside containment is considered to be less severe than inside because the steam flow rate from the break is less and feedwater isolation in both cases will be initiated on MSIS.
Critical flow is assumed at the point of rupture, and all of the mass leavina the break is assumed to be in the steam phase.
This assumption results in the maximum heat removal from, the reactor coolant per pound of secondary water, since the latent heat of vaporization is included in the net heat removal.
It is assumed that steam from the other steam generator flows through the intact steam line and out of the break until the main steam line isolation valves are closed.
2.2 Anal sis Bases and Parameter Conservatisms The following is a discussion of the choice of the parameter.values for the steam line break which assure a conservative evaluation.
The limiting break should have the largest cross sectional flow area and produce the most rapid cooldown and the highest rate for return to power.
Such a limiting break is a double ended guillotine, inside containment between the steam generator outlet and the steam line venturi.
The break mass flow rate is calculated with the Ransom-Trapp model which yields higher values than the Moody model.
The high pressure safety injection system will be activated to mitigate the return to power by pumping borated water into the core.
It is conservatively assumed that the injection line volume initially contains no boron and that one of the required two high pressure pumps will not function.
'n order to maximize the cooldown of the primary system for HZP scenarios the auxiliary feedwater flow is initialized to match the pump heat but not the decay heat.
For HFP scenarios the main feedwater is initially set to match
-'the steam flow but after the break-is -ramp'ed'-to"zero in'bout-"60'-"s'eco'nds..'-The "
safety injection system,. main feedwater valves and main steam isolation valves trip delays are such as to account for uncertainties.
The core neutron
kinetics input assumed the minimum required control rod shutdown worth of the end-of-cycle and end-'f-cycle values for the reactivity feedback.
Different values were utilized for H2P and HFP scenarios to conservatively account for moderator density reactivity feedback.
The Doppler feedback was treated conservatively by assuming the temperature at the bottom of the affected core segment for H2P scenarios.
For HFP cases again a conservative reactivity vs temperature curve is employed.
The decay heat which will reduce the cooldown rate is conservatively set to zero at the initiation of both the HZP and the HFP scenarios.
The nodalization treats all major NSSS components and subcomponents as discrete elements.
(With the exception of the secondary side of the steam generator.)
In order to simulate the asymmetric thermal-hydraulic and reactivity feedback effects the core is nodalized into two halves one of which is directly connected to the affected steam generator.
This maximizes the reactivity feedback.
The assumptions made in this analysis are conservative and assure that the cases considered constitute a limiting analyses for the steam line break.
For example the SRP requires that the loss of offsite power and reactor coolant pump trips should be investigated to determine the most severe scenario.
In this case it has been found that the HFP with the RCS pumps on scenario is the limiting case.
This conclusion is drawn from the study of HFP with one pump tripped and all pumps on, where it is clear that the severity of the transient is determined by the amount of cooling of the core due to the steam line break and the ensuing loss of energy.
Thus tripping 'of a primary pump any time during the transient will result in a less severe transient as when the pumps are on.
2.3 Fuel Res onse Calculations The final objective of the calculation is to.determine fuel -failure,.conditions='=.-.
which are:
(a) MDNBR-1.135 and (b) fuel melt.
However. three intermediate calculations are required (a) the NSSS response which is computed with RELAP5, (b) the detailed core and hot assembly power distribution is estimated using
the XTG code, and (c) the detailed core and hot assembly flow is calculated using the XCOBRA-IIIC code.
A summary of the results of the analysis is shown in Table 1.
The scenario which results in the highest post scram power level, fuel damage and the largest LHGR is from HFP with offsite power available for the operation of the primary coolant pumps.
The event sequence for this case is shown in Table 2.
3.0 SUViMARY AND CONCLUSIONS l'e have reviewed the information submitted by Florida Power and Light on a
confirmatory analysis for a steam line break for the St. Lucie Unit 1 plant.
The reason for the reanalysis was to confirm that the non-uniform core cooling and the resultant non-uniform feedback during a steam line break are acceptable.
The acceptance criteria are based on the General Design Criteria (GDC) 27 and 28 to assure that the fuel design limits are not exceeded, GDC 31 to assure that the boundary behaves in a non-brittle manner and GDC 35 which assures that the RCS is designed to provide abundant emergency core cooling.
From our review we concluded that the estimated consequences of the steam line break accident meet the requirements of the GDC 27, 28, 31, and 35.
The resultant fuel damage is limited so that control rod insertability will be maintained and the core will be eoolable.
The MDNBR was estimated at 1.84 however, the maximum linear heat generation rate is 30.3 kw/ft which caused fuel melting to 1.61K of the fuel pins.
This is also the percentage of rods assumed to fail.
The integrity of the primary system boundary is assured because the pressure during the transient will remain below the normal operating pressure.
The submittal demonstrated the adequacy of the emergency cooling system to provide cooling and reactivity control during the
, transient.
The input parameters for the calculation were found =to.-be, conservative.
The codes used for the analysis have been previously reviewed and found acceptable by the staff.
No information has been provided to demonstrate the adequacy of the auxiliary feedwater system to remove decay heat following the steam line break or that the RCS pumps will withstand this accident, as required by the SRP.
- However, this is only a confirmatory analysis limited to the non-sytmetric cooling, therefore, this information is not necessary at this time.
4.0 REFERENCES
1.
Letter from J. Miller, NRC, to J.
W. Williams, Jr., Florida Power and Light Company, dated March 1, 1984.
2.
XN-NF-85-85(P), "Steamline Break Analysis for St. Lucie Unit 1", dated November 1985.
TABLE 1 STEAMLINE BREAK ANALYSIS
SUMMARY
Power Level Power Available Initial Offsite Maximum Post Scram Return to
- Power, MWt MDNBR Core Average Power 9
- MDNBR, MWt Fuel LHGRMax
- Failures, Kw/ft X Core H2P Yes 908 2.06 908 25.4
.92 H2P No 366 1.20 332 17.1 0.
HFP Yes 1192 1.84 1192 30.3 1.61 HFP No 586 1.26 400 23.2
.34
TABLE 2 STEAMLINE BREAK EVENT SEQUENCE - HFP WITH OFFSITE POWER TINE EVENT t=0.
A.
Reactor operating at 2700 %It.
t=0. +
B.
Double-ended guillotine break located between affected steam generator and the flow meter in the downstream steam line.
T=1. 8 T=2. 7 t=6.6 t=8.7 t=l4.3 t=33.9 t=43.2 t=78.
t=180.
t=600.
t=600.+
C.
D.
E.
F.
G.
H.
L.
Reactor trip and main steam isolation signal generated by low steam generator pressure.
Reactor scrams
.9 seconds after low steam generator pressure signal.
All control rods fully inserted except the most reactive one.
Rod worth = 12.24$.
Main steam line isolation valves stop blowdown from intact steam generator 6.9 seconds after low steam generator pressure signal.
Safety injection signal generated by low primary coolant pressure.
Safety injection pumps at rated speed 19.5 seconds after safety injection signal.
Reactor becomes critical.
Thermal power reaches maximum level at 44K of rated power.
Auxiliary feedwater initiated:
254.5 lb/s to affected steam generator.
Auxiliary feedwater isolated manually.
Primary system temperature increase due to 'steam generator dryout and additional safety injection will terminate power excursion..
August 5, 1987 DISTRIBUTION jDocket File w/o encl-.
PD22 Reading w/o-encl-:
D. Miller w/encl.
E, Tourigny w/encl.
DOCKET NO(S).
50-335 and 50-389 Mr. C. 0. Moody Group Vice President Nuclear Energy Florida Power and Light Company Post Office Box 14000 Juno Beach, Florida 33408
SUBJECT:
ST.
LUCIE UNITS 1
AND 2 The following documents concerning our review of the subject facility are transmitted for your information.
Notice of Receipt of Application, dated Draft/Final Environmental Statement, dated Notice of Availability of Draft/Final Environmental Statement, dated Safety Evaluation Report, or Supplement No.
dated Environmental Assessment and Finding of No Significant Impact, dated Notice of Consideration of Issuance of Facility Operating License or Amendment to Facility Operating
- License, dated Qg Bi-Meekly Notice; Applications and Amendments to Operating Licenses Involving No Significant Hazards Considerati'ons, dated 7/29/87 lsee page(s) ]
Exemption, dated Construction Permit No.
CPPR-
, Amendment No.
dated Facility Operating License No.
, Amendment No.
dated Order Extending Construction Completion Date, dated Monthly Operating Report for transmitted by letter dated Annual/Semi-Annual Report-transmitted by letter dated En cl osures:
As stated Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation cc:
See next page L.
I-2 OFFICE)
SURNAME/
DATEP
.P...
cr.;hg.
8/
/87
~ ~
~ ~ ~ ~ ~ ~ ~ ~ ~
~ ~ ~ ~ ~ ~ ~ ~ ~
~ ~ ~ ~ ~ ~
~ ~ ~
~
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
~ ~ ~ ~
~ ~ ~ ~ ~ ~
~
~ ~ ~ ~ ~
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
~ ~ ~ ~ ~ ~ ~ 0 NRC FORM 3IS IIO/BOINRCM 0240 OFFICIAL RECORD COPY
N, 1
I
'P
'I H,
I
'H