ML20056F463
| ML20056F463 | |
| Person / Time | |
|---|---|
| Issue date: | 07/01/1993 |
| From: | Lyon W Office of Nuclear Reactor Regulation |
| To: | Rosalyn Jones Office of Nuclear Reactor Regulation |
| References | |
| TAC-M54946, NUDOCS 9308270277 | |
| Download: ML20056F463 (13) | |
Text
_
g
/
%g UNITED STATES
~
NUCLEAR REGULATORY COMMISSION
- ["
,, (
g WASHING TON, D. C. 20555 g
cp
/
July 1, 1993 MEMORAt1DUM FOR:
Robert C. Jones, Chief Reactor Systems Branch Division of Systems Analysis
/
THRU:
,i Timothy E. Collins, Chief 6
BWR Reactor Systems Section Reactor Systems Branch Division of Systems Analysis FROM:
Warren C. Lyon, Sr. Reactor Engineer BWR Reactor Systems Section Reactor Systems Branch Division of Systems Analysis
SUBJECT:
MEETil4G REPORT - MEETIf1G AT B&W ROCKVILLE OFFICES ON MAY 11 AND 12, 1993 TO DISCUSS MITIGATION OF STEAM GENERATOR TUBE RUPTURES IN B&W NUCLEAR STEAM SUPPLY SYSTEMS AND THE EMERGENCY PROCEDURES GUIDELINES (TAC M54946)
The staff reviewed steam generator tube rupture (SGTR) information pertinent to the Technical Basis Document (TBD) ("The B&W Owners Group Operator Support Committee Emergency Operating Procedures Technical Bases Document," December 31, 1991.).
Sufficient information was obtained that the staff can proceed with final review.
The staff indicated that it plans to complete the review by September 30, 1993.
B&W's " preferred path" mitigation of SGTR is continued steaming during cooldown unless tube rupture alternate control criteria (TRACC) limits are reached.
The objective is to:
decrease system cooldown time minimize vulnerabilities due to additional f ailures reduce operator burden by keeping the plant in a normal configuration minimize steam generator (SG) stress keep both SGs in service in case a SG becomes unavailable The TRACC limits provide controls on:
offsite dose borated water storage tank (BWST) inventory SG filling i
The TRACC dose limit selected is the 10 CFR 20 Part 10Sa limit of 0.5 rem whole body, the permissible level of annual dose to an individual in unrestricted areas.
Thyroid dose is limiting for SGTR and a maximum thyroid j
dose of 1.5 rem is used.
(Some B&W licensees have a technical specification I
of 1.5 rem to the thyroid that is intended to meet 10 CFR 20.)
The Standard Review Plan requires < 10% of 10 CFR 100 for an accident initiated coolant
/{,gp,l contamination spike (30 rem whole body) and < 10 CFR 100 for a pre-accident
.n, s., S w W '.
1700 '1 u.
e
" n: -
i: Ed)
- vw z.ag y m,ps 9308270277 930701 bli# f lin M2 ; Q Lg,p.,
E 6f4 6 }, l q ('d.y; <,_
g O
1 2J PDR TOPRPEMVByDR
( g) yg c
j
O O
w Robert Jones,
spike (< 300 rem).
In practice, the B&W guidance is realistically calculated to almost never approach the TRACC limit with a loss of offsite power (LOOP).
Dose with offsite power is usually negligible because the release is to the condenser with a condenser decontamination factor of 10'.
Interestingly, the calculations predict that most of the available iodine is often released within the first few hours and centinued steaming results in little addition to the total dose.
Meeting information was provided verbally or was obtained from reference material that has not been released to the staff.
Consequently, complete staff notes are provided in t'.e Enclosure t document the B&W representations.
$1'W Warren C. Lyon, Sr. Reactor Engineer BWR Reactor Systems Section Reactor Systems Branch Division of Systems Technology
Enclosure:
As stated cc:
J. Hopkins W. Swenson J. Arildsen L. Cunningham PDR
O o
ENCLOSURE ADDITIONAL INFORMATION PERTINENT TO STEAM GENERATOR TUBE RUPTURE MITIGATION i
ANALYSIS OVERVIEW AND APPROACH Analyses were based upon the following computer programs:
An in-house personal computer code that conserves mass was used to obtain time-dependent activity release information.
This was input into l
CRAC-2 (Ref. 1).
t RELAP-5 (Ref. 2) provided energy released during cooldown which was averaged over the time of interest. This provided olume energy for input into CRAC-2.
(Release velocity was assumed to be zero which has the effect of reducing the plume elevation.)
l CRAC-2 provided dose predictions.
CRAC2 was selected for. the dose calculations following a December 19, 1989 meeting in which the staff indicated CRAC2 calculations would be better received than calculations from an unfamiliar methodology.
(Lyon, Warren C.,
" Meetings on Steam Generator Tube Rupture Mitigation Strategy and Other i
Emergency Procedures Guidelines Topics, B&W Plants, November 16 and December i
19, 1989; Staff Position on Steam Generator Tube Rupture," NRC Memorandum for Chief, Reactor Systems Branch, January 24,1990).
i i
A test case was run based upon one year of site-specific weather data for each r
B&W plant site.
The calculations were based upon no sheltering with the
" person" exposed at the site boundary assumed to continuously move to the location of highest dess ic-the entire duration of the event. -Peak dose values corresper. ding to the vorst weather were used.
TMI-1 was found to.have the largest dose.
Dose to the thyroid at the site exclusion area boundary (EAB) was consistently limiting with the low population zone (LPZ) dose always lower - for example, a factor of four for Davis Besse and a factor of six for Crystal River 3 (comparison basis: integrated dose).
Consequently, the TMI site was used to calculate behavior for the Technical Basis Document (TBD)
(Ref. 3).
i OBJECTIVE AND RESULTS B&W considered it important to minimize system cooldown time and vulnerability to additional failures. A significant difference in cooldown time occurs between use of one SG and two, as illustrated on the following page (TBD Fig.
III.G-14). The 3.2% capacity corresponds to the atmospheric dump valve (ADV) capability in one steam generator (SG) in all operating B&W plants except Davis Besse, which has twice the capability. Note the time to reach the decay heat removal cut-in capability of 275 *F and the 900 hour0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br /> difference between using one SG (3.2%) e csus two SGs (6.4%).
This was a significant contribution to thei
'nclusion that it is best to steam both SGs.
Historical steam generator tube rupture (SGTR) mitigation guidance involves operator determination of tube leak rate. This can be inaccurate and 1
r
O aw FIGURE lit.G - 14 NATURAL CIRCULATION COOLDOWN TIME AND CONDENSATE REQUIREMENTS NATURAL CIRCULATION COOLDOWN PRIMARY SYSTEM TEMPERATURE 550 N'*I, l l l l ll i l l llll.
} l l lll 3a qcv cap l i
l' -
!ll eh%dovcb
,gg il
! ! I N 0i
! i l l llIl I i l l ll
~
ii c
I i i l li't Llilili!
!I Il T-TC
- 450 k
\\
l, -ll l
l l
E i
!!in
\\ l l riLd!
1 l
1
< 400 l
l k
k l!! l l !
l
! lllilli W
l!I i',J llL I
Ei.,,0 C'
! l llll1 hl IT%!d
! l l'l'BU l
l i i iii!!ll l\\'N. i I!lllN I I ll l' -L. I lll
,00 j
j03 FlHR C hit
\\ l l h k, l
l l }
I
! I I IIIII I\\iIIIIIII
'I I I IIIli' I
III 250 0.1 0.3 1
3 10 30 100 300 1000 TIME (HRS)
NATURAL CIRCULATION COOLDOWN AUXIUARY FEEDWATER REQUIRED 2E+06 j
,l, "
32% $Dy CAq
/
_ 1E+06 p Sc+05
,,,,.;!/',
i,-!
o 6E-05
, ;;i
- i i
,ii4 4.bn0M i l i i
,iiii i
i i i i xi
~
j
, j.
] 4E+05
- 7p7.; j;j j!, ;j j
g l-
- I IlIll{
/
lll l
l l
N l } l!
IIII I
3 2E+05
!.l l [!
l![r C
1E+05 A
> BE+04 jKC 1
. ii i
G 6E+04
! ! L...-f% ;, ; ; q
, i l ll!!.
i
I' 3 4E+04
'!N I
i i i lIlli i
i III{
l I
l
}
i i I /g' '
R i
2 I
I MiiI
! ! I l!!il I l l lIllll i
l i
IE + 04
'i j'
l l
j l
I 1E+04 0.1 0.3 1
3 10 30 100 300 1000 TIME (HRS) 1 D00. NO. 74-1152414-06 i
1 O
o i
~
2 l
di f ficul t.
Instead, B&W uses the known pre-event primary water contamination as a criterion and has correlated this to steaming time as shown in Figures Ill.E-2 and III.E-3 (from TBD volume 3).
Steaming limitation would be updated during the event as information such as new contamination levels, SG level and dose data were obtained.
(To be conservative, the guidance is always changed to decrease the allowed steaming time.)
An interesting observation is that steaming time is unlimited for concentrations to the left of the vertical line.
This occurs because part of the process involves selections based upon j
release of essentially all of the initially available iodine.
Further note that the technical specification limit is 0.7 micro-Ci/cc (~ 0.13% failed feel), and the figures start at 5 micro-Ci/cc (~ 1% failed fuel). Thus, j
" normal" SGTR events would not reach the TRACC limit.
Table 1 provides the leak rates that would fill the SG, release 99% of the initial iodine, and that were used to establish steaming limits.
Note the steaming durations of 1 and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> are terminated by SG isolation due to t
filling, and thus all iodine has not been released.
Table 1.
Limiting Leak Rates Steaming gpm gpm To gpm To Comments l
1263 6400 1313 1313 gpm corresponds to 3 HPI pumps at 1225 psia 4
445 1600 533 12 hr 533 gpm steaming limit i
used because it fills SG; higher value judged overly restrictive i
l 12 292 533 533 24 170 267 267 50 105 128 130 130 gpm selected as bounding to I
reduce number of calculations 120
<100 53 130 130 ppm selected as above
{
A single double-ended tube break will pass 370 gpm at 2200 psig and 580
- F, the average temperature for most B&W plants. On average, this decreases to l
145 gpm with a subcooled margin of about 50 *F.
If on natural circulation, the rate will be about 175 gpm at 2 hrs and 130 gpm at 10 hrs.
If cooling with reactor coolant pumps operating, the pump NPSH becomes restrictive during i
cooldown and subcooling margin will increase, increasing the leak rate.
I
-.._s,,--
,-m
A a
~
j Figure Ill.E-2 Steaming Time of Affected SG(s) to Atmosphere (applicable to all 177FA plants) l 200 100
,,.6...g..
..6..
..l..
..h..h +.
50 T
-:n 20 e
E F
C)
C
'E 10 G
..i..
..p.,..i....
,.. j... i... i..
..i... i..
.O
..i.-
CO
..i i..
..;...i...<..
..i...i...i..
>..i...i...i.
..i...i... i..
l t
.. +....
^
5
.. +...
.. +.
l i
2 i
P l
l 1
-d 0
5 10 15 20 25 30 Pre-Existing 1-131 D. E. Concentration (uCi/cc) 8 Notes 1. Steaming time unkmeed for cone. left of vemcal hne j
J
- 2. ff actual cana excesos 20.9 uCi/cc.
vsolate SG as soon as possion.
j 00C..;3. 74-1152414-36 i
i I
I i
a.
m
)
w I
1 l
Figure Ill.E-3 j
Steaming Time of Affected SG(s) to Atmosphere l
(appilcable to all 177FA plants) 200 i
t i
100
..,...i..
.4
..y..... 9..
..s..
......i..
..q..
j
.4....
..+.4...
.. 4..
.4...
. 4.....
q 50
.g.
...9..
i u) 5 1
20 i
o E
~
i 1
}._
J 3
.E l
E 10 7
g
..;... i... i... ;..
i...;...;...
~l M
....j..,
.,.. [... j...
...j...j..
{
. 4....
5 i
l
.4.....
. ~
i 4
.4..._..
.4....
i Q
2
+
t
^
e 1
30 40 50 60 70 80 90 i
Transient Peak l-131 D. E Concentration (uCl/cc)
{
Notes 1. Steaming time unkmned for cone. left of vemcal line.
l
- 2. tf actual data exceeds 85.5 uCi/cc.
isolate SG as soon as possible.
.l D0C. NO. 74-1152414-06 i
?
T~
m a
i o
o 3
I ANALYSIS ASSUMPTIONS AND DISCUSSION j
The principle assumptions used for the analyses were:
(1)
The SGTR event occurs during an extended operation at 100% power.
f (2)
A power ramp-down is initiated immediately following occurrence and the reactor is tripped eight minutes later.
Offsite power is lost at reactor trip. The SGs are steamed to the condenser for the first 8 min l
with subsequent steam release directly to the environment via the atmospheric dump valves (ADVs).
The condenser normally will trap -
5 essentially all of the iodine and little dose results from release during the ramp-down.
If the condenser cannot be used, a typical case may realize perhaps '10% of the dose due to thi.s ramp-down.
(3)
Both SGs are steamed following reactor trip.
I
(
(4)
The event is terminated when the decay heat removal (DHR) system is i
placed in operation at 100 hrs.
(5)
ANS 5.1 (1979) was used for the decay heat with an initial power of 2568 MW for the plume calculations.
(Lower decay heat results in less dispersal of the plume. Decay heat is the only significant contributor to the release characteristics after 2-hours.)
4 (6)
An atmospheric dump valve capacity of 339,000 lbs/hr/SG at 1052 psig was-l assumed.
This corresponds to a release capability of 3.2% of full power l
per SG.
i (7)
One percent of the fuel is defective in a 2772 MW plant (Davis Besse).
Equilibrium activity is provided in Table 2.
l Experience with B&W fuel for the past two years is operation with 0.01%-
- 0.03% failed fuel with one case of 0.05%. As of September 1992 there were 8 leaking fuel pins in the 10 operating plants fueled with B&W' fuel.
(All B&W plants presently have B&W fuel.)
The contamination range for all B&W plants over the last five years is 0.005 to 0.06 micro-Ci/gm (0.7 micro Ci/cc is ~ 1.0 micro Ci/gm). The mean decreased from ~ 0.05 to 0.02 micro-Ci/gm during that time. These values would lead to a dose of a factor of 100 below the TRACC limit.
(8)
An iodine spike occurs for 3 hrs at 100 times the normal equilibrium release rate, a rate selected as consistent with NRC thinking as described in Reference 3.
B&W stated that experience is bounded by a factor or about 5 to 20 or 30.
If by some chance the release were
=!
greater, then this would be found during or prior to the event arid j
compensated for by the steaming time curves.
All cf the available iodine was calculated to be released into the RCS coolant in 2.7. hours as opposed to the 3 hrs used in the calculations.
Y e
oi.
4 y
O v
4 (Two hours is generally assumed for the SRP in conjunction with a factor of 500 for the I spike - values used for SRP-type calculations that are discussed later in this report.)
Table 3 summarizes the spike characteristics.
(Zero values correspond to an approximate time where all of the radioisotope has been released from the fuel gaps.)
Table 2.
Equilibrium Activity Corresponding to 1% Failed Fuel Radioisotope Activity, micro-Ci/gm I-131 5.40 I-132 6.27 I-133 5.67 I-134 0.713 I-135 2.89 Xe-131m 2.47 Xe-133m 3.92 Xe-133 358 Xe-135m 0.455 Xe-135 8.08 Xe-138 0.648 Kr-83m 0.488 Kr-85m 2.22 Kr-85 30.1 Kr-87 1.17 Kr-88 3.61
O A
a w
5 Table 3.
Fuel-to-Coolant Iodine Release Characteristics
- Time, Micro Ci/sec from Fuel to Coolant l-131 1-132 1-133 1-134 I-135 0
1.614E-2 1.868E-2 2.740E-2 3.480E-2 2.740E-2 0
1.614 1.868 2.740 3.480 2.740 3600 1.614 1.868 2.740 3.480 2.740 360]
1.614 0
0 0
0 10800 1.614 0
0 0
0 10800+
0 0
0 0
0 (9)
There is no holdup of radioactive material in the SG (decontamination factor of one).
Realistically, the SG decontamination factor will tend toward one and saturation with respect to iodine can occur.
This will not occur in the condenser due to the amount of water and there is always enough steam to trap most of the iodine in the condenser during condensation.
(10) lio purification of the reactor coolant system (RCS) water occurs during the event.
(11)
Plume energy corresponds to the smallest B&W plant and a zero release velocity is assumed.
The release is 70 ft above the ground elevation (the elevation of the ADV above ground at TMI).
(12) The site boundary dose is the summation of the instantaneous maximum values existing anywhere on the site boundary for the total duration of the event.
(13)
Doses were obtained for 116 weather conditions and the maximum dose was used.
Calculated results for the maximum dose permitted by the TRACC limits are summarized in Table 4.
W 6
i Table 4.
Calculated Site Boundary Dose Steaming Duration, hrs Mean Thyroid Dose, R Peak Thyroid Dose, R 1
1 0.04 0.38 4
0.09 0.97 12 0.20 1.58 24 0.22 1.49 50 0.22 1.33 120 0.22 1.34 The peak thyroid doses illustrate a maximum at 12 hrs.
This behavior is attributed to variation introduced by the weather sampling methodology.
The fraction of iodine remaining in the primary system ranged from 0.403 for steaming for one hour to 0.0002 with 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> of steaming.
l l
COMPARISON WITH STANDARD REVIEW PLAN METHODOLOG5 The B&W owners group (BWOG) used the CRAC2 code in conjunction with Standard Review Plan (NUREG-0800) guidance to predict dose to assess the influence of i
the CRAC2 code.
Four cases were studied with the assumptions of no holdup of radioactive material in the SGs, steam release through the SG atmospheric release valves following reactor trip, a 2 hr exposure at the site boundary and a 30 day exposure in the low population zone (LPZ):
)
(1)
An iodine spike occurs at accident initiation and the affected SG is i
isolated at 34 min, consistent with the final safety analysis rcport I
(FSAR) assumptions (2)
Item I with no SG isolation 1
(3)
Item 1 with the affected SG returned to service at 8 hrs (4)
A pre-accident iodine spike occurs with no SG isolation The whole body dose was predicted to be less than 0.011 Rem in all cases.
Thyroid dose and SRP acceptance criteria are given in Table 1.
l
i O
O 7
a Table 1.
B&W Thyroid Dose Predictions with CRAC2 and SRP Methodology Case 2 hr Thyroid Dose at 30 day Thyroid Dose in SRP EAB, Rem LPZ. Rem Acceptance Mean Peak Mean Peak
~-
1 0.0016 0.0136 0.0173 0.0355 30 2
0.0225 0.198 0.239 0.444 30
{
3 0.0014 0.0124 0.198 0.341 30 4
0.0245 0.220 0.196 0.374 300 Note the site exclusion area boundary (EAB) dose is smaller than the LPZ dose, the reverse of the TRACC calculations.
This is due to the 2 hr exposure time l
at the EAB when following the SRP in contrast to using the entire event time in the TRACC calculations.
The major differences between SRP calculations and the CRAC2 calculations performed by the licensee were stated to be:
Regulatory Guide 1.4 addresses a diffusion factor for ground release.
The B&W calculation assumes release from the ADVs at an elevation of 70 ft.
Site boundary dose is reduced by a factor of 4 to 8 due to the elevated release.
Plume buoyancy reduces the B&W calculated site boundary dose by a factor of 25 to 100 when compared to a near-zero plume energy.
ALTERNATIVE COOLING METHODS The only viable cooling methods other than using SGs is to use the HPI pumps in a once-through cooling mode.
If the SGs are isolated and one HPI pump is used, it will be about 10 to 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> before the RCS can be cooled while maintaining pressure < 1000 psi (which is necessary to prevent opening the SG safety valves in the affected SG).
Two HPI pumps might be able to provide adequate cooling about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after SGTR initiation.
However, if one were to commit to two cump operation and one was subsequently lost in less than 10 to 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />, it may be impossible to regain use of a SG and RCS pressure would consequently exceed the code valve release pressure.
REFERENCES 1
(1)
" Calculations of Reactor Accident Consequences, Version 2, CRAC2:
Computer Code," NUREG/CR-2326, February 1983.
(2)
"RELAP-5/ MOD 2-B&W," BAW 10164P, December 1987, t
i k
b
~.
O-0;
- ~
i (3)
Lyon, Warren C., " Meetings on Steam Generator Tube Rupture Mitigation Strategy and Other Emergency Proce62res Guidelines Topics, B&W Plants, t
November 16 and Dorember 19, 1989; Staff Position on Steam Generator Tube i
Rt'p tu r e, - NRC Memorandum for Chief, React or Systems Branch,' January 24, 1990.
- i 1
6 e
r t
1 i
I
{
i 4
l i
e T
ar
- ~ -
~
r-