ML20083P684
| ML20083P684 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 01/27/1983 |
| From: | Maurin L LOUISIANA POWER & LIGHT CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| W3I83-0022, W3I83-22, NUDOCS 8302030570 | |
| Download: ML20083P684 (11) | |
Text
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142 DELARONDE STREET POWER & LIGHT P O BOX 6008 NEW ORLEANS. LOUIslANA 70174 * (504) 366-2345 i ES SYS E L V. MAURIN Voce Presodent Nuclear Operations January 27, 1983 W3183-0022 Q-3-A29.20 Mr. Harold R. Denton Director, Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555
SUBJECT:
Waterford 3 SES Docket No. 50-382 Clarification of Transient Analyses with Potential for Fuel Damage and Feedwater Line Break Confirmatory Items ENCLOSURES:
(1)
Evaluation of the Effect of Break Size Upon the Consequences of Steam Line Breaks with Concurrent Loss of Offsite Power - Waterford 3 (2) Analysis of Feedwater Line Breaks
Dear Mr. Denton,
Supplement Number 1 to the Waterford 3 SER contains a confirmatory issue concerning clarification of transient analyses with potential for fuel damage.
(Section 15.3.1).
The concern of this item was that we had not addressed the consequences of a small steam line break concurrent with a loss of offsite power. Enclosure (1) describes this analysis and demonstrates that no steam line areak with concurrent loss of offsite power yields consequences more adverse than the limiting consequences reported in the FSAR.
Supplement I to the SER also contains a confirmatory issue requesting con-firmation that small feedwater line breaks ceabined with the limiting single failure and with offsite power available da not exceed the 110% design pressure criteria. This analysis is described in enclosure (2). The NRC staff reviewer also requested a probability analysis for feedwater line breaks which is included in Section I of enclosure (2).
It is hoped that this evaluation will alleviate the staff's concern on these confirmatory issues.
If you have any questions or require further information, please feel free to contact either myself or R. W. Prados.
Very truly yours, go 7d i L. V. Fburin LVM/ DEB:keh cc:
J. Wilson, W. M. Stevenson, E. L. Blake, J. Guttman Er302030570 830127 POR ADOCK 05000382 E
PDR J
O Evaluation of the Effect of Break Size Enclosure (1)
Upon the Consequences of Steam Line Breaks with Concurrent Loss of Offsite Power Waterford Unit 3 Introduction Supplement No. I to the Safety Evaluation Report for Waterford Unit 3 requires that the consequences of small steam line breaks (SLBs) with concurrent loss of offsite power (LOP) be addressed in order to demonstrate that the SLB analyses presented in the FSAR are the limiting cases (p.15-1 of Ref.1). The effect of break size on the consequences of SLBs with concurrent LOT has been evalutted. No SLB, of any break size, with concurrent LOP yields consequences more adverse than the limiting consequances for SLBs reported in~the FSAR.
The consequence of concern for SLB's is offsite dose. The contribution to offsite dose that can be affected by break size is fuel failure. The possible break sizes of concern range from zero up to the maximum flow area for outside containment breaks, the area of the flow venturis.
Inside containment breaks result in much less offsite dose. Degradation in fuel performance (fuel failure) can occur during SLB initiated events either during the portion of the transient prior to and during reacter trip (henceforth referred to as the pre-trip portion) or during the post-trip return-to-criticality, or approach-to-criticality, portion of the transient (henceforth referred to as the post-trip portion).
Pre-trip fuel failure For cases initiated from a power operating limit, and where loss of offsite power is assumed to occur concurrent with the SLB, there will be a CPC trip on projected DNBR within the first 0.6 seconds of the initiation of the event.
The power operating limit is determined such that the CPC trip will prevent the transient =inimum DNBR due to a loss of flow (LOP) from being less than 1.19.
The only significant additional effect of the SLB, over that of the LOF, up to the time of minimum transient " pre-trip" DNBR will be a reduction in RCS pressure. Therefore for SLBs with cencurrent loss of offsite power the transient minimum " pre-trip" DNBR will be only incrementally lower than 1.19.
Further, the rate of reduction of RCS pressure due to the SLB will be maximum for the maximum break area.
For the Waterford Unit 3 NSSS a conservative evaluation of the decrease in DhoR due to the maximum area, outside containment SLB yields a minimum transient DNER greater than 1.17.
The resultant calculated fuel failure would be less,
than 0.05%.
The consequent 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> exclusion area boundary thyroid dose for the event (including secondary releases) would not be greater than 10 rem.
This is less than the design basis radiological consequences,f the limiting main steam line break of the FSAR. Thus no SLB, of any break size, with concurrent LOP yields consequences due to pre-trip degradation in fuel performance which are more adverse than the limiting consequences for SLBs reported in the FSAR.
Post-trip fuel failure Degradatien in fuel performance during the post-trip portion of SLB initiated transients can only occur if there is a return-to-power (R-t-P).
Therefore the primary consideration for maximizing post-trip degradation in fuel performance is to select those paraneters and conditions which will maximize R-t-P.
The magnitude of R-t-P is primarily determined by the value of the maxinum post-
4 trip reactivity, the timing of this reactivity, and the duration of the reactivity peak. The timing of the maximum post-trip reactivity has an important effect on the post-trip R-t-P, the same reactivity will produce less R-t-P later in a transient.
The maximum R-t-P results from the maximum break area. A smaller break delays the time of maximum post-trip reactivity and therefore decreases the magnitude of the R-t-P generated. Additionally, for the slower transients due to smaller breaks, more time is available for the transfer of heat from the metal of the RCS walls and structure and of decay heat from the fuel to the coolant. This reduces the moderator cooldown and consequently the maximum post-trip reactivity.
Therefore the minimum " post-trip" DNER occurs for the maximum break area. The SLB vith maximum break area and with concurrent LOP has been presented in the FSAR.
Thus no SLB, of any break size, yields consequences due to post-trip degradation in fuel performance which are more adverse than those reported in' the FSAR.
Conclusion SLBs with concurrent LOP have been evaluated for the effect of break size upon event consequences. The conclusion of this evaluation is that the events reported in the FSAR are the most limiting events.
?.eference 1.
NUREG-0787, Supplement No. 1, " Safety Evaluation Report related to the operation of k'aterford Steam Electric Station, Unit No.
3",
- USNRC, October, 1981.
ANALYSIS OF FEEDWATER LINE BREAKS Enclosure (2)
I.
PROBABILITY OF FEEDWATER LINE BREAKS The feedwater line break analyzed in FSAR Section 15.2.3.4 is postulated to occur in the piping between either of the two steam generator nozzles and the containment wall. The methods and data contained in WASH-1400 can be used to estimate the recurrence frequency of such a break.
WASH-1400 (Appendix III, Table III 6-9) provides a summary of pipe rupture rates per plant year for various sized "LOCA sensitive" piping. The median LOCA recurrence frequency for large piping ( 6" diameter) is given as 1 X 10-4 per plant year.
For the Waterford design, the total length of "LOCA sensitive" piping with respect to the event analyzed in Section 15.2.3.1 is 232 feet.
(See Appendix A).
Therefore, using the methodology discussed in WASH-1400, Appendix III, Section 6.4, the estimated recurrence frequency for the above post-ulated Main Feedwater Line Break is 3.2 X 10-5 PER PLANT YEAR.
Thus it is shown that the initiating event which is analyzed in Section 15.2.3.4 is in fact a very low probability event that is highly unlikely to occur in a plant's lifetime.
Additionally, the high primary system pressures reported in FSAR Section 15.2.3.1 are due to the conservatively assumed coincident occurrence of a loss of normal a/c power. Using WASH-1400 value of 1 X 10-J for conditional loss of normal a/c power, the estimated joint recurrence concurrentlossofnormala/cpowerislessthan1X10~yeakwith frequency tor the above postulated main feedwater line b per plant year.
II.
REANALYSIS OF SMALL FEEDWATER LINC BREAKS Introduction The following responds to NRC's request for demonstration that small feedwater line breaks (FWLB) meet 110% design pressure criterion when combined with the limiting Jingle failure, and with offsite power available. The existing FilLB evaluation in FSAR Section 15.2.3.1 demonstrates that all FWLPs, even when combined with a loss of offsite power, are well below 120% design pressure. As indicated above the probability of a FWLB in pipes with diameters greater than 6 inches is sufficiently low to allow application of the 120% design pressure criterion. Therefore, the following FWLB reanalysis addresses break sizes which are less than 0.2 ft2 (i.e.,
6 inch diameter).
Methodology The original FWLB evaluation methods (1) which are conservative for application to the full spectrum of break sizes have been modified for the small FWLB reanalysis.
The modifications include the treatment of reactor trip on low water level in the " ruptured" steam generator, and the enthalpy of the fluid discharged from the break.
As discussed in Reference 5, the original FWLB method credited low water level trip in the ruptured steam generator only after its liquid inventory decreased to approximately 9000 lbm. This assured conservative treatment of low level trip even if the FWLB (i.e., large breaks) caused rapid steam generator depressurization and consequently swelling of the downcomer level due to flashing of the downcomer liquid.
However, for breaks less than 0.2 ft2.the average steam generator pressure remains constant or increases prior to reactor trip and no downcomer level swell will occur due to flashing. Therefore, in the reanalysis of samil FWLBs steam generator low water level trip is credited with a larger liquid inventory remaining.
The low level trip setpoint corresponds to a downcomer liquid level of approximately 27 feet above the tube sheet and a liquid inventory of over 70,000 lbm. However, the reanalysis of small FWLBs conservatively delays low level trip until 7.5 feet above the tube sheet (approximately 22,000 lbm of liquid).
To expedite the reanalysis of small FWLBs the enthalpy of the break discharge fluid is assumed to be saturated liquid until all liquid is depleted from the ruptured steam generator. This differs from the original FWLB steam generator method employed for Waterford 3 FSAR (see Reference 5).
Analysis Input 2
A spectrum of small ( 0.2 ft ) FWLBs with the limiting single failure and offsite power available were analyzed with the new method. As for the FWLB analysis presented in the Waterford FSAR, the CESEC-ATWS code was used to model the FWLB transient.
As a result of the evaluation method applied to the FWLB analysis, the only mechanisms for mitigation of the reactor coolant system (RCS) pressurization are the pressurizer safety valves, the reactor coolant i
flow and the main steam safety valves. The last two mechanisms influence the RCS to steam generator heat transfer rate.
There are no credible failures which can degrade pressurizer safety valve or main steam safety valve capacity. A decrease in RCS to steam generator heat transfer due to reactor coolant flow coastdown can only be caused by a failure to fast transfer one half the electrical loads to i
offsite power or a loss of offsite power following turbine trip (i.e., two or four pump coastdown respectively). The FWLB analysis of Subsection 15.2.3.1 considers the worst of the two, the loss of offsite power.
The WSES-SER requires that small feed line breaks with the limiting single failure and offsite power available should meet 110% of design r
pressure, in accordance with the SRP requirements. Therefore, the limiting single failure assumed in the reanalysis of feed line breaks is the failure to fast transfer.
- L
In order to determine the limiting small feedwater line break, the l
initial pressurizer pressure and steam generator inventories were adjusted within their operational ranges to obtain a coincident trip on high pressurizer pressure and low steam generator water level. This coincidence ensures that the RCS pressure at the time of trip is at its maximum value, maximizing the RCS pressurization potential of the FWLB.
Results Four different break sizes were analyzed including 0.01, 0.05, 0.10, and 0.20 ft2 A plot of the results of these analyses is provided~in Figure 1.
In all cases, the maximum pressure remained below 2700 paia which is less than 110% of design.
The worst case small FWLB in terms of primary pressurization is the 0.2 sq ft break. Tables 1 and 2 provide the initial conditions assumed j
and.the sequence of events for the transient. Figure 2 provides a plot i
of RCS pressure vs. time for the event.
Conclusion Small feedwater line breaks with the limiting single failure and offsite power available result in maximum primary pressures of less than 110% of design.
2 J
k I
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R7fer:nces 1.
CENPD-107 Supplement 1, "ATWS model modifications to CESEC", September 1974.
(Section 3.0).
2.
CENPD-107 Supplement 1. Amendment 1-P, "ATWS model modifications to CESEC", November 1975.
(Section 3.3).
3.
CENPD-107 Supplent 3, "ATWS model modifications to CESEC", August 1975.
(Sections 240.8, 240.11 and 240.9).
4.
CENPD-107 Supplement 4 "ATWS model modifications to CESEC", December 1975.
(Sections 1.6, 1.8 and 4.2).
5.
FSAR, Appendix 15A,'Section 15.A.2 Amendment 22, Septewoer, 1981 -
TABLE 1 ASSUMPTIONS FOR THE FEEDWATER SYSTEM PIPE BREAK Parameter Assumption Initial core power, MWt 3,478 Core inlet coolant temperature 560 6
Core mass flowrate, 10 lbm/hr 132 Reactor coolant system pressure, psia 2,250 Steam generator pressure, psia 964 Moderator temperature coefficient,10-4 0.0 Doppler coefficient multiplier 0.85
,CEA worth for trip, 10-2
-8.55 Steam bypass control system Inoperative l
Pressurizer pressure control system Inoperative Pessurizer level control system Inoperative 2
0.2 Feedwater line break area, ft Initial steam generator total inventory, Ibm 132,000 Table 2 SE00ENCE OF EVENTS FOR THE FEEDWATER SYSTEM PIPE BREAK TIME EVENT SETPOINT OR VALUE 0.0 Double-ended rupture of the main feedwater line 0.0 Complete loss of feedwater to both S.G.s 21.9 High pressurizer pressure trip condition (psia) 2474 23.1 High pressurizer pressure trip signal occurs (psia) 2474
~
i l
23.4 Steam generator safety valves open (psia) 1085 23.8 Pressrrizer safety valves open (psia) 2525 24.0 CEAs begin to drop into core 25.8 Maximum pressurizer surge line i
flow (1bm/sec) 1374 26.0 Maximum RCS pressure (psia) 2698 28.6 Maximum steam generator pressure (psia) 1156 30.6 Pressurizer safety valves close (psia) 2525.
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f 2700
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DATA s 2600 BREAK MAX. RCS SIZE PRESSURE -
2580 l
(SQ. FT.)
(PSIA) 2560 0.20 2698 0.10 2638
~
2540
- INCLUDES ELEVATI0M AND REACTOR 0.05 2620 COOLANT PUMP HEADS 0.01 2603 2520 2500 0.22 0.24 0.12 0.14 0.16 0.18 0.20 0
0.02 0.04 0.06 0.08 0.10 BREAK SIZE, SQ. FT.
LOUISIANA POWER & LIGHT CO*
MAXIMUM RCS PRESSURE vs BREAK SIZE 1
Waterford Steam Electric Station
2700 i
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i gwo 3M -
5 E 2400 uf5 C
$2300 w
2200 2100 1
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10 20 30 40 50 TIME, SEC RCS PRESSURE vs TIME FOR THE LIMITING SMALL Rgure PO R
GHT CO-FWLB (DOES NOT INCLUDE REACTOR COOLANT 2
Waterford Steam PUMP AND ELEVATION HEAD)
Electric Station
.