ML20090H086
| ML20090H086 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 07/23/1984 |
| From: | Woolever E DUQUESNE LIGHT CO. |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| References | |
| 2NRC-4-109, NUDOCS 8407260107 | |
| Download: ML20090H086 (11) | |
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Nuclear Construction Division Rob 6nson Pleas, Building 2 Suite 210 Telecopy (432) 787-2629 Pittsburgh, PA 15206 -
July 23, 1984 United States Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:
Mr. George W. Knighton, Chie f
= Licensing Branch 3 Of fice of Nuclear Reactor Regulation
SUBJECT:
Beaver Valley Power Station - Unit No. 2 Docket No. 50-412 Response to Draf t SER Open Item 102 Gentlemen:
The response to the NRC Accident Evaluation Branch's Draf t SER Open
-Item No. 102 is provided in Attachment 1.
The associated revisions to FSAR Section 15.6.2 and Tables 9.3-8, 15.0-12, 15.6-2, and 15.6-3 are provided j
in Attachment 2.
DUQUESNE LIGHT COMPANY By qr
(
E.(/J. 'Woolever Vice President l
JD0/wjs
[
Attachments k.
f-cc:
Mr. G. Walton, NRC Resident Inspector (w/a) l-Mr. E. A. Licitra, Project Manager (w/a)
Ms. M. Ley, Assistant Project Manager (w/a)
SUBSCRIBED AND SWO TO BEFORE ME THIS
_ g h0 DAY OF
_ d
, 1984.
0 A /}u b D u L
Notary Public ANITA ELAINE REITER, NOTARY PUBLIC ROBINS 0f t TOWNSHIP, ALLEGHENY COUNTY MY COMMIS$f0N EXP!RES OCTOBER 20,1HE 8407260107 040723
~
PDR ADOCK 05000412 E
-i 00!
(
i f
t 1
-o Unitcd StctOs Nuclect Rigulctory Commiccion
-Mr. G;;rgo W. Knighten, Chief
);
Page 2-I f-COMMONWEALTH OF PENNSYLVANIA )
i-
)
SS:
COUNTY OF ALLEGHENY
)
p-i E.
On this -23 day of
/ff before me, G-a ' Notary. Public in and for said Commonwealth and County, personally
[
appeared E. J. Woolever, who being duly sworn, deposed and said that (1) he is Vice President of Duquesne Light, (2) he is duly authorized to execute 1
and file the foregoing S ubmit tal on behalf of said Company, and (3) the statements set forth in the Submittal are true and correct to the best of his knowledge.
I L
s 7
Notary Public ANITA ELAINE REITER, NOTARY PUBLIC ROBINSON TOWNSHIP, ALLEGHENY COUNTY MY COMMISSION EXPIRES OCTOBER 20,1986 O....,.
ATTACHMENT 1 Draf t SER Open Ites No.102 (Section 15.6.2):
Radiological Consequences of a Small Line Break DBA e
Response
The radiological consequences resulting from a small line break hav e
'~
been analyzed based on 40% of the reactor coolant flashing to steam upon entry into the building atmosphere.
The assumption is made that the
. fraction of dissolved iodine fission products becoming airborne as gas and particulates is equal to the fraction of coolant flashing to steam.
All potential locations for the small line break in the auxiliary build-
)L
'ing are within ventilation zones of the supplementary leak collection and release system (SLCRS).
A small line break in the contiguous areas would be. serviced by the SLCRS after receipt of a high radiation signal L
from a QA Category II radiation monitor. However, the analysis does not P
take credit for SLCRS operation.
The results of the analysis are presented in the revisions to FSAR Section 15.6.2 and Tables 9.3-6, 15.0-12, 15.6-2, and 15.6-3 ( At tachment 2).
These revisions will be incorporated into a future FSAR amender.ent.
ATTACHMENT 2 BVPS-2 FSAR
-TABLE 9.3-8 (Cont)
Tube side Design pressure (psig) 2,735 650 Design temperature (*F)
Austenitic stainless Material steel operating parameters
-Shell side Maximum Normal Purification Flow (lb/hr) 29,826 59,700 Inlet temperature-('F)
Ef5.f 549
@ 287
. Outlet temperature ('F) g3 Tube side Flow (lb/hr) 22,370 52,250 Inlet temperature (*F) 130 130 Outlet temperature (*F)'
4 8 9..t 446
'}
La @.<a orifices General 2,485 Design pressure (psig) 650 Design temperature (*F)
Normal operating inlet pressure 2195 (psig) 290 Normal operating temperature (*F)
Austenitic stainless Material of construction steel no ya Orifice 2
Quantity 29,826 Design flow (lb/hr)
Differential pressure at design 1,900 flow (psig) 0.242 Diameter (inches) 45 gpm Orifice 1
Quantity 22,370 Design flow (1b/hr) g Differential pressure at design 1,900 flow (psi) 0.215 Diameter (inches)
Amendment 3 2 of 8 October 1983
X f
SVPS-2 FSAR TABLE 15.O-12 POTENTIAL DOSES OUE TO POSTULATED ACCIDEletS (Ree)
Exclusion Area Boundary Low Populatton Zone
- FSAR Whole Body Sota whole Body Beta Postulated Accident Section Thyroid Gamma Skin Thyroid Gamma Skin Main steam line break 15.t.5 Pre-accident todine spike 5.5 6.2xtO-'
2.5xtO-'
3.1xtO-*
3.6xtO-*
t.7x10
- Fatted fuel 2.3xtO' 2.2xtO-'
9.3=10-8 9.4
't.8xtO
8.3=10 '
15.2.6 1.5xtO-*
5.tx10-*
4.Outo-*
2.tx10-'
6.5xtO-*
6.8xtO-*
Loss of nonemergency ac power to the station auxtItarles Locked rotor 15.3.3 2.5x10-8 9.5xtO-*
5.txtO-*
1.FxtO-'
4.2xtO~*
1.7xtO
- Rod ejection 15.4.8 Containment leakage 4.0=10' t.SulO 8 6.3xtO '
2.0 9.2x10-'
3.2x10-*
Secondary sfoe 2.tuto-*
5.Ox10-*
3.6x10-*
1.1x10-'
2.5x10-'
t.8x10-'
Smait itne creak - loss-of-15.6.2
. Su tO 9
. 9= 10 - ')
.7xt0 3
[s.'On t o-3 d 5x10 9
- 11. t x tO _p
/.S X/0
[s.6 sf /O 1.3 K AO
'B,2.%io"
'3.4 *80'
- 2 N
~
Steam generator tube rupture 15.6.3 Pre-accident todine spike 6.2 S.SutO-'
7.OxlO-'
G.9xtO '
4.8xtO '
3.7xtO-'
Concurrent todene spike 6.7 f.txfO '
7.4xtO-'
1.2 7.4x10 '
4.3x10 '
Loss-of-coolant 15.6.5 Centainment leakage 2.7xtO' 4.7 2.1 1.3xtO' 2.4x10-'
t.txtO '
ECCS teakge 6.3xtO
- 4.5=10 '
l.2xtO-'
4.9x10 '
4.9xtO-'
t.GxtO '
waste gas system rupture 15.7.t t.7x10 '
t.6x10-*
Fuel handting 15.7.4 2.4x10' 3.6 6.*
?.S.1D
- For earatton of accident 9 of 1
^
BVPS-2 FSAR
+
L s
Ibnit'value~'throughout the _ transients
- thus, the _ departure from nucleate boiling (DN8) design-basis as. described in Section 4.4 is
- met.
15.6.2 Failure of Small Lines Carrying Primary Coolant Outside Containment
- 15.6.2.1 Identification of Causes and Accident Description Lines connected to the RCS and penetrating the containment, as well as isolation provisions are identified in Table 6.2-60.
There are r.o instrument lines connected to the RCS that penetrate the containment. There are, however, the sample lines from the hot and cold legs of reactor coolant loops and the steam and liquid space of
. the pressurizer, and the CVCS letdown and excess letdown lines that penetrate the containment. The sample lines.and the CVCS letdown and excess letdown lines are all provided with normally open containment isolation valves on both sides of the containment wall.
In all cases, the containment isolation valves are designed in accordance with the containment isolation requirements of General Design Criterion 55 (Section 6.2.4).
The most severe pipe rupture with regard to radioactivity release during normal BVPS-2 operation is a complete severance of the 2-inch letdown line at a location outside containment, upstream of the letdown heat exchanger. This event would result in a loss-of-reactor 3
coolant at. the rate of approximately 160 gym based on a density of
~ ;
57 lbs/fta and on the flow restriction provided by two of the three letdown line orifices in service (the 45-gjm orific and one of the 60-gpa orifices), shown on Table 9.3-8 and Figure 9.3-24.
15
.The time required for the operator to identify the accident and isolate the rupture is expected to be less than minutes.
Diverse - g.f/,s instrumentation-in the form of letdown line pressure / downstream of
~
the ~ postulated break location, volume control tank level and pressurizer level with indication at the main control board will. 2 "8##f *
allow-detection of the failure by the operator.g The operator would isolate the letdown line rupture by closing /Ehe letdown orifice} [ff 6'
}
fisolation valves 2CHS*A0V200A, 8, and C followed by closing the a n fpressurizer ' low level isolation valves 2CHS*LCV460A and B.
Thel,g,JInferi 0 operator would.also close the letdown line containment isolation l (g.15.4 -43.)
tvalve 2CHS*A0V204, to isolate the rupture.;/All valves are provided
'with control switches with indicating lights at the main control board. and at 'the emergency shugdown panel. All valves are air-operated and' designed to fail clos ( on loss of air or electrical 1 power.
There are no single failures that would prevent isolation of the letdown line rupture.
Amendment 3 15.6-4 October 1983 n s
.. ~,, m m e w
-m,
,wmwn,-,,-,-
m 4-INSERT "A"
. In addition, a control room operator can determine specific plant areas which are experiencing high radiation after receiving-plant high radiation annunciation.
INSERT "B" one: _of. the _ pressurizer low level isolation valves, 2CHS*LCV460B, and by closing the letdown line containment isolation valve, 2CHS*A0V204. Following
- isolation. of the r eupture, the operator would also close ' 2CHS*LCV460A and the letdown - orifice isolation valves, 2CHS*A0V200A, B, and C.
n 15.6-4a
/
V >
e
?';'
BVP5-2 FSAR K
_15.6.2.2 Analysis of Effects and Consequences
- g Method of Analysis The amount of primary coolant released is conservatively estimated by assuming critical flows in the ruptured letdown line.
The mass of fluid released from the postulated break was calculated using the Zaloudek correlation in WCAP-8312A (Westinghouse 1975a) for subcooled liquids and the theoretical model developed by Moody fcr saturated conditions.
Immediately after the rupture, the Moody model is used
_for -a saturated liquid until the liquid in the letdown line between the orifices _and rupture point is depleted.
After the liquid is depleted, zaloudek's sub goled correlation is used at the orifice en continues ter av minutesiuntil isolation occurs p.hese critical flow correlations are in accordance with WCAP-6312A (Westinghouse 1975a).
'The assumptions used for the analysis are summarizea in Table 15.6-2}.
The auxiliary building pressure transient for the postulated break iy/)
shown on Figure 3.11-3Q. f "4
-15.6.2.3 Radiological consequences The failure outside the containment of small' lines carrying primary coolant is postulated to occur in the letdog line to the letdown heat enchanger. an the enumww nuildina.T The rupture of this line will' result in the loss of primary coolant. with isolation occurring
/g within qgPminutes.
The rupture will result in the discharge of primary coolant directly into -the auxiliary building with the radioactivity released to the environment at ground level.-Insert "C" (M IE.G mor. Info #aCan+1 The assumptions for evaluating the radiological consequences of the postulated small line failure are summarised in Table 15.6-2.
The conservative analysis assumes primary coolant Technical specification equilibrium activities as presented in Table 15.0-8.
Additionally, a concurrent iodine spike is postulated to occur with iodine release rates into the primary coolant as shown in Table 15.0-10.
The resulting releases to the environment based on the stated assumptions and postulated activities are presented in Table 15.6-3.
The radiological consequences resulting from a postulated failure of
_a small :line carrying primary coolant outside containment are presented in Table 15.0-12.
The offsite doses are determined using the calculated environmental releases for this accident and the atmospheric dispersion values given in Table 15.0-11.
The methodology for calculating the offsite doses is discussed in Appendix 154.
The radiological consequences for this event are a small fraction of the guidelines of 10 CFR 100 that is less than 2.5 Rem whole body and 30 Rem thyroid.
Amendment 3 15.6 5 October 1983
o 1
INSERT "C" i
All potential locations for the small line break in the auxiliary building are within ventilation zones of the supplementary leak collection and release system (SLCRS). A small line break in the contiguous areas would be serviced by the SLCRS. af ter receipt of a high radiation signal from a QA Category II radiation munitor.
Howeve r, the conservative analysis does not take credit for SLCR8 operation.
1 i
c l
15.6-Sa J
m m
(_.
4 SVPS-2 FSAR TABLE 15.6-2 PARAMETERS USED FOR THE SMALL LINE CARRYING PRIMARY COOLANT FAILURE Technical
' Characteristics Expected Specification
- Power (!!Wt) 2,766 2,766 Fra-tion of failed fuel 0.0012 0.0026
. Line failure Letdown line to inlet of letdown heat exchanger Break size (in) 2 2
Time required to detect
.andisolatefailure(R),m
@ l6 h /8 Total amount of primary coolant released (lb
@ 2.0 N 10 Temperatnre of released primary coolant (*F) 287
@ 647 Fraction of iodine assumed airborne from pipebreak 0.1
@ O.4 Surolementary leak
. collection and release system iodine filter efficiency (*5) 95
@O Primary coolant concentrations Table 11.1-2 Table 15.0-8 Iodine spiking - release rates (assumed to occur for duration of accident)
Table 15.0-10 Table 15.0-10 f%y aclad release.
.r ade fen beeab. (l>~ Sec Amendment 3 1 of 1 October 1983
. ~..
--__.,.. -.-. ~
,s BVPS-2 FSAR TABLE 15.6-3 SMALL LINE FAILURE RELEASES TO THE ENVIRONMENT Releases (Ci) f Nuclide 0-2Hrj
{2Hr-30 Days
- Kr-83m 1.9 9 1810
Kr-85m 9.1 4.5 Kr-85 4.8x101. 2.4 x10' Kr-87 5.2 2.6 Kr-88 1.4x101 6.9 Kr-89 4.4x10-1 11xto '
~
Xe-131m 4.7x10-1 2.3 v. s o 8. 3 x10 - 2TM 4.6 2.9x10-1 Xe-133m 1.3x101 Xe-133 1.1x102 5, G 1 lo' 4,0 Xe-135m 4.0x101 3.6 1.6x102 Xe-135 1.6x101
.70 1.0x101 Xe-137 7.1x10-1 3.5 xio '
~
Xe-138 2.9 1.5 I-131 7.1x10-1' l.2.e;0' I-132 3.4
.2.. o x i O' L (s
-I-133 1.5 1 7 *10' I-134 1.4 2..~/a80' I-135
- 2. 3 x t o' m
QDuetodecayofiodinescollectedonSLCRSfilter
,s e
-+
s 1 of 1
_ _.