ML20083M714
| ML20083M714 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 01/28/1983 |
| From: | Devincentis J PUBLIC SERVICE CO. OF NEW HAMPSHIRE, YANKEE ATOMIC ELECTRIC CO. |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| References | |
| SBN-439, NUDOCS 8302010435 | |
| Download: ML20083M714 (15) | |
Text
.
SEABROOK STATION
% Ofma:
1671 Worcesser Road M " M * * ""'01701 Pub 8c Service of New Hampshire (617) - 872-8100 January 28, 1983 SBN-439 T.F. B7.1.2 United States Nuclear Regulatory Commission Washington, D. C. 20555 Attention:
Mr. George W. Knighton, Chief Licensing Branch No. 3 Division of Licensing
References:
(a) Construction Permits CPPR-135 and CPPR-136, Docket Nos. 50-443 and 50-444
Subject:
Additional Information for the Locked Rotor / Shaft Break Event:
(SRP 15.3.3; RAI 440.89 and 440.127; Reactor Systems Branch)
Dear Sir:
The enclosed Additional Information for the Locked Rotor / Shaft Break Event supplements our previous responses which were submitted in letters dated March 12, 1982, September 21, 1982 and December 7, 1982.
This open item was discussed with representatives of the Reactor Systems Branch and Accident Evaluation Branch in meetings conducted on January 10-12, 1983.
The enclosed response will be included in OL Application Amendment 49.
Very truly yours, YAhTEE ATOMIC ELECTRIC COMPANY b
- t/
J. DeVincentis Project Manager Ol ALL/fsf 1
cc: Atomic Safety and Licensing Board Service List 8302010435 830128 PDR ADOCK 05000443 A
PDR 1000 Elm St., P.O. Box 330. Monchester, NH 03105 Telephone (603) 669-4000. TWX 7102207595 1
ASLB SERVICE LIST Philip Ahrens, Esquire Assistant Attorney General A
Department of the Attorney Gene: 51 Augusta, SK 04333 Representative Beverly Hollingworth Coastal Chamber of Commer:c 209 Winnacunnet Road Hampton, NH 03842 William S. Jordan, III, Esquire Harmon & Weiss 1725 I Street, N.W.
Suite 506 Washington, DC 20006 E. Tupper Kinder, Esquire Assistant Attorney General Office of the Attorney General 208 State House Annex Concord, NH 03301 Robert A. Backus, Esquire 116 Lowell Street P.O. Box 516 Manchester, NH 03105 Edward J. McDermott, Esquire Sanders and McDermott
~
Professional Association 408 Lafayette Road Hampton, NH 03842 Jo Ann Shotwell, Esquire Assistant Attorney General Environmental Protection Bureau Department of the Attorney General One Ashburton Place, 19th Floor Boston, MA 02108 m
e eese e a e e
+ Mg e see gpe e.g..
Addit ional Informat ion for the Locked Rotor /Shaf t Break Event In the event of a locked rotor / shaft break of a Reactor Coolant Pump (RCP), the remaining three RCPs will continue to run.
Analysis of the breaker coordination (see response to RAI 430.55) shows the following.
Under all postulated operating conditions, including maximum load of one of the 13.8 kV buses (2 RCPs, 2 Circulating Water Pumps and the 13.8 kV Substations) and minimum bus voltage, failure of one RCP (with incipient locked rotor amps) will not result in tripping of the incoming breaker to the 13.8 kV bus.
Because of the separate power supply to the other 13.8 kV bus (see FSAR Figure 8.3-1), this event will have no ef fect on the power supply of this bus.
Off site power will not be lost as a consequence of the event.
FSAR Sect ion 8.2.2.3 provides the results of stability studies showing that the loss of one or both of the Seabrook Units would not cause a loss of off site power.
FSAR Figure 8.3-1 is a one-line diagram of the Electrical Distribution System showing the generator circuit breaker used for isolating the generator without af fecting the normal supply to the 13.8 kV bus.
During this postulated event, the maximum number of fuel rods with a minimum transient DNBR less than 1.3 (rods in DNB) occurs near the time of maximum clad temperature.
Furthermore, the number of " rods in DNB" is related to the calculated maximum clad temperature.
From the results shown in FSAR Sections 15.3.3 and 15.3.4, it is seen that the maximum clad temperature occurs about 2.8 seconds after turbine trip.
The FSAR analysis assumes no loss of off-site power.
Therefore, the remaining three Reactor Coolant Pumps (RCPs) are assumed to continue operating.
Now, based on the grid stability studies described above, it has been demonstrated that off site power (and the remaining RCPs) would not be lost and would be available well beyond the time of maximum clad temperature.
So, Loss of AC at any time after the maximum clad temperature is predict 3d (about 2.8 seconds after turbine trip) would yield the same number of " rods in DNB" as the existing FSAR analysis.
If the assumption is made that all the rods with a minimum transient DNBR of less than 1.3 become failed rods, then the fraction of failed fuel is predicted to be 8% for this bounding event.
However, it is the applicant's position that " rods in DNB" is an overly conservative criteria for determining t
fuel failure.
When the evaluation procedures of NUREG-0562 are followed, no l
fuel failure is predicted.
The attached material is submitted in support of this position, although no credit was taken for it.
i l
With a conservatively assumed 8% fuel clad failure fraction, the off site presented in FSAR Section 15.3.3.3, are well within the design l
doses, as l
limits (i.e., 10% of 10CFR Part 100 guideline values).
In the event of a failed open atmospheric steam dump valve, the off site I
dosea are well within the design limits specified in 10 CFR Part 100.
The off site thyroid doses, in the event of a failed open valve concurrent with j
the locked rotor event plus the assumed 8% fuel clad failure and the assumed loss of off site power, are a maximum of 100 times the values presented in i
i
{
FRAR Section 15.3.3.3 or 230 rems and 240 rems at the EAB and LPZ, respectively.1 With the failure of a steam dump valve in the open position, plant procedure would call for isolation of the affected steam generator's feedwater flow with subsequent drying out of the steam generator and loss of l
any iodine partitioning afforded by the water volume contained above the point of primary :o secondary leakage. Although some finite period of time is required for the considerable amount of water contained within a typical Westinghouse steam generator to boil off, it has been conservatively assumed that this time is small and no partitioning of iodine occurs within the steam generator.
l The failure of an atmospheric steam dump valve, discussed above, is considered to be the limiting single failure associated with the locked rotor event.
In addition to the sequence of events mentioned above, it is also conservatively assumed that the entire primary to secondary allowable leakage of 1 gpm occurs in the steam generator that is venting through the stuck open valve.
Corrections to the FSAR are attached and will be placed in the FSAR with the next amendment.
1 l
i I
I 1 The off-site whole body gamma doses for this event are calculated to be 1,4 rem and 1.9 rem at the EAB and LPZ, respectively.
i 1
r APPLICATION OF NUREG-0562 TO THE EVALUATION OF FUEL FAILURE AS A CONSEQUENCE OF A LOCKED ROTOR ACCIDENT IN A W PWR INTRODUCTION Analysis of locked rotor (LR) accident under the assumption that rods in DNB fail produces overly-conservative dose rates.
One way to obtain more realistic results is to demonstrate that no fuel failures will occur during a LR accident, even though some of the rods may go into DNB.
The material in the following Please note that in no way sections can be used to support this conclusion.
does this conclusion replace the need to verify by means of peak clad temperature analysis that the fuel maintains a coolable geometry for the LR accident.
LOCKED ROTOR AND FUEL FAILURE' MECHANISMS The design requirements for such faults The LR is a Condition IV occurrence.
do not state that the plant must be capable of returning to operation with the Thus we will only' address the effect of DNB during a LR on the existing core.
ability of the clad not to fail during the accident, during shutdown and during We will not possible removal and inspection of the fuel after the accident.
address the effect of DNB during a LR on subsequent operatior) of the fuel rods.
These Several mechanisms for cladding failure due to DNB have been identified.
(a) clad melting, (b) fuel pellet melting, (c) clad ballooning / bursting, are:
(d) clad collapse, (e) clad recrystallization, and (f) excessive clad oxide.
A discussion of whether any of these is a failure mechanism due to DNB caused by a locked rotor accident is discussed below.
(a)
Clad Meltina The melting pointU) of Zircaloy-2 and -4 is 3360*F. This is well above any clad temperatures conservatively estimated to occur during LR for a typical plant (Figure 3). Thus, LR does not result in con-l ditions in which clad melting is a clad failure mechanism.
l l
i (b)
Fuel Pellet Melting The melting pointU)ofUO during a DNB incident would cause the 2
fuel pellet to expand. This could strain the clad to failure.
Another possible failure mechanism would be contact of the molten is fuel with the clad. The melting point of unirradiated U02 5080*F with a melting point reduction during irradiation of 58'F per 10,000 MWD /MTU burnup.
Calculations of pellet centerline tem-peratures for LR conditions give maximum temperatures lower than the melting point temperatures.
Thus fuel melting is not a clad failure mechanism.
(c) Clad Ballooning / Bursting For the LR transient, the RCS pressure should increase or remain unchanged during the event.
Consequently, high powered rods in DNB should have internal pressures sufficiently low compared to the system pressure that the rods will not balloon for clad temperatures predicted to occur during a LR accident. As a result, no fuel failure will occur by this mechanism during the LR event.
l (d)
Clad Collapse Siace most of the rods with low burnup and high F will have internal pressures less than system pressure, if DNB were to occur on these rods and decrease the yield strength, the clad would collapse onto the fuel.
Clad collapse onto the fuel pellets is not considered a fuel failure mechanism as long as there is no subsequent operation.
(e) Clad Re~ crystallization When the zire cladding is raised above the temperature range of 950' to 1000*F, the hardening and strengthening due to cold working is removed. This results in lower yield stresses and lower ductility for irradiated material.
1
- ~
Failures due to a enange in the Zircaloy clad properties due to the temperature increases in the fuel. rods during the LR accident should not occur during nor after the accident.
Such changes would only need to be considered in subsequent operations of the core.
(f)
Excess Clad Oxide Failure due to excessive clad oxidation is considered to be a major I2)
This mechanism is addressed cause of fuel rod failure during DNB in Reference 2, and it can be shown that the clad temperatures reached during the time interval of a LR accident are insufficient to cause fuel failure.
An updated discussion of this topic is presented in this section.
Van Houten(2) of the NRC reviewed the available literature on the operation of nuclear fuel rods under film boiling or dryout conditions.
His review was mainly concerned with data on oxidation-related failures. The discussion below is based on his review.
Figure 1 is based on the results of out-of-pile Zircaloy-steam oxidation tests of Kassner and Chung et.al. of ANL.
In these tests the cladding was subjected to a rapid heatup, an isothermal soak and a rapid cooldown.
The ability of the oxidized cladding to withstand the thermal shock of the quench and/or shock of a subsequent sharp impact was then detemined. The data were for a.6 mm thick cladding (H 17x17 cladding thickness is.0225" =.57 m).
The limit line for surviving a.3 Joule inpact at 300*K (80*F) is given.
Studies showed that if the oxidized cladding can withstand this, then the cladding is still tough enough to withstand both operating stresses when at working temperatures and the The line for surviving mechanical stresses of handling when at room temperatures.
This thermal shock corresponds to the quenching of the hot oxidized cladding.
line is roughly equivalent to a limit line for surviving at.03 Joule impact at 300*K(80*F).
The 17% oxidation and 1477'K (2200*F) lines corresoond to NRC imposed LOCA liriitts.
Figure 2 gives the results of inpile test data of fuel rods operated under DNB or dryout conditions. The two limit lines from Figure 1 describing the ability of the oxidized clads to withstand thermal shock or impact are also plotted.
Comparisons of these lines to the data indicates that the limit lines based on out-of-pile data are effectively the same as those based on inpile data.
Figure 3 gives the W fuel clad temperature as a function of time for a typical worst type LR accident.
The temperatures were calculated under conservative assumptions and assumed that the rod was in film boiling. Thus, if it were conservatively assumed that the clad temperature remains at 2000*F (1366*K,
.00073*K-1) for 10 seconds, it is seen that this gives a point on Figures 1 and 2 well below the limit lines. Thus the limiting rod in DNS during a LR accident will have a large margin to withstand both the temperature transient (heatup and quench) that occurs during the accident and the handling of the rod after the accident.
CONCLUSION l
f W maintains that as long as the locked rotor transient results in a time-at-l temperature below the oxidized clad failure limit lines in Figures 1 and 2, no fuel failure will result. This does not replace the need to verify by means of peak clad temperature analysis that the fuel maintains a coolable geometry for the locked rotor accident.
REFERENCES Engineering Department " Materials Property Manual," WCAP-8222.
1.
2.
Robert Van Houten, " Fuel Rod Failure as a Consequence of Departure from Nucleate Boiling or Dryout," NUREG-0562, June 1979.
l
FIGURE 1 I
OXIDATION TEMPERATURE ( C)
(2552 'F)(2472*F) (2192'F)
(2Ol2'F).
(1832*F) 1400 1000
!2CO I10 0 I000 1
I I
I I
I
/
2 4
10 LIMIT FOR SURVIVING THERMAL SHOCK
/
Z 8
~
3 6
l
~
^
w i
g 4
LIMIT FOR SURVIVING
/
W O.15-J IMPACT AT g
2
~
300 K
/
/
10 O
/
17 % OXIDATION -
/
LIMIT 9
6
/
4 t
f H
i gj W
L
-LIMIT FOR SURVIVING l
3 2
/~
O.3-J IMPACT AT 300 K l
i 2
10 8
6 1477 K LIMIT 4
j 2
- 6. 0 6.5 7.0
- 7. 5 8.0 (FIGURE 2-B - NUREG-0562)
I /T x 104 (*K-I)
1 FIGURE 2 Comparison of In Reactor Post DNB Survival / Failure Data for Zircaloy Cladding with FBRS/ANL Zircaloy Dpctile-Brittle Boundary Curve 2000K A
g g
i AA 9
O 1800K A
k b
O A
'Y, 1600K f
aA t
2 a%
s.
,,g i
gas I 1400K
'd q.
O y
[
e-I a
%g 0
5 1200K 2
0 w
Brittle: (Rod Failure)
A PBF/PWR O cR/BWR Ductile: (No Rod Failure)
Q 6 PBF/PWR O CR/BWR 800K D SGHWR/BWR O GETR/BWR O Halden/BWR 1
10 100 1000 10,000 100,000 Time After DNB (Seconds)
(FIGURE 4 - NUREG-0562) 1 i
l
FIGURE 3 ma, l
l t600 16C0 E
wg 1400 Q5 E#
e
~5 12C0 a
i 1000 800 l-l l
600 0
2.0 4,o 6.0 8.0 10,0 IIHE(SECCHDS)
CLAD TEMPERATURE AS A FUNCTION OF TIME FOR A TYPICAL WORST CASE LOCKED ROTOR EVENT (All Loops Operating, One Locked Rotor)
/hebed f N b' /'
5 Y lH he d Rs ad SB 1 6 2 FSAR (e) Prig g t,o y cong 4r gpmg or the 8Mour period during whicfi steam i eYeasesego{a,ngs j
f D
g to the atmosphere may be required. For this 8-hour period, 100% of the noble gases and 1% of the iodines in the secondary coolant system, from the primary coolant leakage, are asstaned to be released with the steam to the atmosphere.
Table 15.3-5 lists the activity released to the environment via the steam generator safety / relief valves.
2.
Realistic Analysis (a)
Initial primary coolant activity prior to the accident corresponds to the equilibritan concentration at 0.12%
clad defects (see Table 11.1-1).
(b) As a result of the accident, 0.02% of the iodines and noble gases of the core activity are released into the primary coolant system.
This release from the damaged fuel rods, plus the primary coolant activity prior to the accident, represents the source term for the realistic analysis of this accident and is lisced in Table 15.3-6.
(c) For the realistic analysis, offsite power is asstaned to be available. The condenser and the secondary coolant system are used to cool down the plant. The time period of plant cooldown is asstaned to be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
(d) The realistic source term in Table 15.3-6 is uniformly mixed with the mass of primary coolant. Table 15.3-7 lists the activity concentration of primary coolant after the accident.
(e) Primary to secondary coolant system leakage rate is 0.009 gpm for the 8-hour period during which the release from the secondary side is via the condenser air evacuation pump.
Decontamination factors of 0.01 and 1.0 for iodine and noble gases, respectively, are assumed for the steam generator. Decontamination factors of 0.001 and 1.0 for iodines and noble gases, respectively, are assumed for the main condenser. During this 8-hour period of time, 100% of the noble gases and 0.15% of the iodines in the secondary coolant system, from the primary coolant leakage, are asstuned to be released with the steam to the atmosphere. Table 15.3-7 lists the activity released to the environment via the condenser air evactntion pump.
15.3-10
TABLE 15.3-5 RELEASE FROM STEAM CENERATOR STEAM DUMP LOCKED ROTOR - CONSERVATIVE Primary Coolant Concentration 0-2 Ilours 0-8 Hours Radionuclide After Accident (//Ci/ gm)
Release (Ci)
Release (Ci)
I-131 3.5 + 0 3*
l.l.14 + 01 4.g' 64 + 01 132 4.8 + 03 g,(, 2<2 + 01 6 7-Scr + 01 2 q )3d + 01 qkId+021 133 7.4 + 03 y,y <5 + 01 4.9 LA + O2-t 134 7.8 + 03 135 7.0 + 03 33 3<2 + 01
$. A l<3' + O 2-I Kr 83m
- 6. 3 + 01 1.df 24 + 01 S f Jefr + 01 85m 9.6 + 02 3,g 4-r4 + O2 t.3 1 r1 + 0 3 85
- 6. 5 + 01 g,l M + 01 g, g' l rt + 02-1 87
- 1. 7 + 0 3
.5~ f Jr7 + 0 2
- 2. 2 3A- + 0 3 88 2.5 + 03 71 Id + Of A
.3 3 4 h + 03 89 3.1 + 03 g,q 1.A + 0 3 4,o 54 + 03 g,3 rr7 + 01 J
Xe-131m 2.6 + 01 g.f 1<2 + 01'O 1-r8 + 0 3 133m
- 1. 0 + 0 3 3.A ArS + 02 4,3
(
id +.0kO3 133 7.0 + 03 2 5.3 c2 + 03 9,p 135m 1.4 + 03 q.r %;4 + 02
,g.2r5 + 03 135 1.5 + 03 A(.f 6 4 + 02
,j.2r7 + 03 g 1.0 + Otr3 138 5.7 + 03 f.( 2<6 + 03 3
- 3. 5 + 0 3 = 3. 5 x 10.
SB 1 & 2 FSAR TABLE'15.3-7 ACTIVITY RELEASED VIA AIR EVACUATION PUMP LOCKED ROTOR - REALISTIC Primary Coolant Activity 0-2 Hours 0-8 '.<ours Radionuclide Concentration (//Ci/ gm)
Release (Ci)
Release (Ci) 1-131 8.8 + Ol*
J 16 - 06 1.o k4 - 05 132 1.2 + 02 g, g.Sd - 06 L8t 210 - 05 133 1.8 + 02 5,8f M - 06
%.I h0 - 05 134 2.0 + 02 f, T Sd - 06 2,. 3.3r3 - 05 135 1.8 + 02 S.4 M - 06 A.o 2,9 - 05 Kr-83m 1.1 + 00 3.1 4-r4 - 03 1.k k. 02 85m 2.5 + 01 7,6 M - 0/A.
J.o L 01 85 5.4 - 01
- 1. 6 2d - 03 4 3 Sr8 - 03 87 4.4 + 01
),3 M - 01 S.4 -712 - 01 88 6.3 + 01 1 7 256 - 01 7,3 A O + 00 -dl 89 7.8 + 01
%. $ M - 01 f.o4.3 00 -Of Xe-131m 6.3 - 01 J.S 2:6 - 03 1.Y -110 - 02' 3 133m
- 2. 6 + 01 76 M - RA.
3 0 4d - 01 133 1.8 + 02 6~.k. M - 01 10 -219 + 0 0
(
135m 3.6 + 01 12 0 M - 01 4.k 5r8 - 01 135 3.8 + 01 1.$. F.6 - 01 Af.'t 4:2 - 01 138 1.4 + 02 4.} L 01
- f. 4 2,.-3 + 00 I
- 8.8 + 01 = 8.8 x 10.
h
1 i
SB I & 2 FSAR TABLE 15.3-8 OFFSITE DOSES, AT GIVEN SITE DUE T_0 LOCKED _ ROTOR ACCIDENT
{
Site
- Thyroid (Rem)
Whole Body (Rem)
Skin (Rem) 4 3,,3[rf00 hfI4/
7.fs-Of Conservative Analysis
- ~
EAB (0-2 Hours) 3C:00*
' 1&G 1 -
1.0E-0G LPZ (0-8 Ilours)
+00-44s:too
'A.3:.i g ot s.6 s-ee 01-1.iE+GG-t Realistic Analysis EAB (0-2 Hours) 1.0E-07 1.9E-05 3.3E-05 LPZ (0-8 Hours) 1.lE-07 2.1E-05 3.6E-05 i
i
- EAB:
Exclusion Area Boundary; LPZ: Low Population Zone
!(
- [r-3E+0C - [. O a ibr
.2.seco = a.s x n* =.q.s/-
i e
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4 l
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