ML20031H419
| ML20031H419 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 09/18/1981 |
| From: | Stancavage P HOUSTON LIGHTING & POWER CO. |
| To: | |
| Shared Package | |
| ML20031H319 | List: |
| References | |
| NUDOCS 8110270465 | |
| Download: ML20031H419 (5) | |
Text
'
Septcmbar 18, 1981 UNITED STATES OF AMERICA y
NUCLEAR REGULATORY COMMISSION 2
Bl: FORE THE ATONIC SAFETY AND LICENSING BOARD 3
In the Matter of S
4 S
HOUSTON LIGHTING & POWER COMPANY S
Docket No. 50-466 5
S (Allens Creek Nuclear Generating S
6 Station, Unit 1)
S 7
DIRECT TESTIMONY OF PETER P.
STANCAVAGE
.AND STEPHEN A.
HUCIK REGARDING:
(1)
DOHERTY CONTENTION NO. 5 - SUPPRESSION POOL UPLIFT (2)
TEXPIRG CONTENTION 40 - HYDROGEN MONITORING g
10 Q.
Mr. Stancavage, have you reviewed your prior 11 affidavit on Doherty Contention No.
5, which affidavit is
~
12 attached hereto as Attachment PPS-l?
13 A.
Yes, I have.
Q.
Are the statements contained therein still true 14 and correct?
15 A.
Yes, they are.
Q.
Mr. Stancavage, what are the dynamic capabilities of the HCU modules during LOCA pool swell loads?
18 A.
The HCU.Lodules are designed to withstand loads 19 associated with responsu spectra peaks in excess of 15 g 20 vartically and 5.9 to 11.9 g horizontally.
As indicated in 21 the testimony of Dragos A. Nuta, the HCU modules will not be 22 damaged by the hydrodynamic forces associated with the 23 vertical water swell postulated to occur during a LOCA..
24 Q.
Mr. Hucik, have you previously given testimonE in l
8110270465 810918 PDR ADO"% 05000466 T
PDR 1
this proceeding?
i 2
A.
Yes, I presented testimony in connection with 3
Doherty Contention 17, regarding the reliability of safety r' lief valves.
4 Q.
Is the sta temer t of your professional qualifications s
5 attached to that prior testimony still cotrcet?
A.
Yes.
7 Q.
Mr. Hucik, directing your attention to page 32 of 8
the Board's Order of 7.cptember 1, 1981, can you state whether 9
tnere is a possibility for simultaneous actuation of s.-
.ty 10 relief valvec on pool swell?
11 A.
The Allens Creek Nuclear Generating Station BWR l'
uses a General Electric sixth generation, boiling water 13 reactor nuclear steam supply system o. quipped with 19 safety 14 relaat valves.
The purpose of these valves is to relieve 15 pressure from the reactor pressure vessel venting steam to 16 the suppression pool where it will be condensed by the pool 17 water.
The valves open after receiving a signal that the reactor pressure is higher than normal.
18 A sudden break of a high energy pipe in the reactor g
coolant pressure boundancy of the nue, lear steam supply 20 system will cause the pool swell phenomenon if the break size is large enough.
Small breaks do not release sufficient 22 energy into the drywell to cause pool swell.
23 For a break large enough to pro 91ce the pool swell 24 l
1 phenomenon, the pressure in the reactor vessel decreases 2
rapidly due to the flow of high energy fluid from the break 3
in the reactor coolant pressure boundary.
This drop in 4
reactor pressure ensures that the safety relief valves remain closed throughout the first few seconds when the pool swell 5
phenomenon occurs.
Thus, we do not consider the actuation of 6
safety relief valves at the same time as pool swell.
t Q.
Mr. Hucik, at page 21 of its September 1, 1981 Order, the Board asked several questions regarding the hydrogen 9
monitoring system far Allens Creek.
Could you please addrer: 3 10 those questions?
11 A.
Most of the questions have been thoroughly 12 answered by Mr. visingart's testimony; however, I can add 13 certain information from GE's perspective.
First, as to
' 4 the question of incomplete convective circulation, Section 15 6.2.5 of GESSAR II demonstrates that post LOCA conditions in 16 containment promote natural convection such that effective 17 mixing of the containment atmosphere is accomplished.
The 18 principal reasons are as follows:
(1) heat transfer mechanism:
19 heat source (the suppression pool) at the bottom and 20 heat sinks (containment wall) at the top and the sides will create unstable conditions due to 22 buoyancy forces 23 (2) mass transfer mechanism:
24
1 additional density gradient due to changing hydrogen 2
concentration near the pool surface will reinforce the thermally induced convective currents.
3 The convective circulation in the containment, when established, 4
will be directed upwards near the drywell wa]l and downward along the containment wall.
The hydrogen recombiners when in 6
operation will not interfere with this pattern because of 7
their location near the top of the drywell.
In fact, the 8
additional heat source they represent will reinforce it.
9 The calculations presented in GESSAR II show that extremely 10 small temperature and concentratica differences (2.6 x 11
~$*F and 4.3 x 10-6%, respectively) are sufficient to 10 12 create a turbulent free convection regime in the containment.
13 Based on these considerations we conclude that the 14 hydrogen concentration in the air supplied to the hydrogen
~
15 recombiners will be at or very near the bulk concentration and the convective circulation will.not be detrimental to 16 the efficiency of these recombinors.
7 Second, as to the conservatism of the alarm set point, Figure 1 shows a typical hydrogen concentration time history in a Mark III Containment following a recirculation 20 line Design Basis Accident (DBA),
The analysis is based on 21 the very conservative assumptions of Reg. Guide 1.7.
At the time 22 when the containment H concentration reaches 3%
('s l7 days),
2 23 the rate of hydrogen evolution from the suppression pool due, 24 to radiolysis is less than 1 SCFM.
(It,actually drops to 1
~ _,
1 that rate in 3 days).
That tr.~
'la. ; to a H2 concentration 2
[
rise of 0.1%/ day.
With a nominal recombiner warm-up time of 3
3 hrs. there is r;re than enough time for the operator to activate a back-up system in case cce fails.
4 5
6 7
8 9
10 JL 12 13 14 15 16 17 18 19 20 21 22 23 24
~
Attachm:nt PPS-1
.a UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Ma'.ter of r
HOUSTON LIGHTING & POWER COMPANY a
Docket No. 50-466 (Allens Creek Nuclear Generating
)
r-Station, Unit No.1) si 4
AFFIDAVIT OF PETER P. STANCAVAGE i.i State of California gr-County of Santa Clara ji I, Peter P. Stancavage, Manager of Coltainment Engineering, within in the Domestic BWR Projects Department of General Electric Company, of lawful
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age, being first duly sworn, upon my oath certify that the statements contained in the attached pages and accompanying exhibits are true and correct to the best of my knowledge and belief.
.s Executed at San Jose, California, July 29,1980.
g f
D M Co4c W
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-y Subscribea and sworn to before me this J/ al. day of July
, 1980, a
m NOTARY PUBLIC IN ANQ.FOR SAID COUNTY AND STATE e Pu My commission expires '97/ arch if of 19ff.
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Attachm:nt PPS-1 I
L UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD _
In the Matter of 5
S HOUSTON LIGHTING & POWER S
COMPANY S
Docket No. 50-465 5
(Allens Creek Nuclear S
Generating Station, Unit S
No. 1) 5
~
Affidavit of Peter P.
Stancavace My name is Peter Stancavage.
I am employed by General Electric Company as a nuclear and mechanical engineer.
I have been employed in this capacity for 12 years.
A j _
statement of my experience and qualifications is set out in t'.
I.
Introduction
^
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The purpose of this affidavit is to address-Mr.
'.) ;
Doherty's Contention 5 which alleges that the control rod drive mechanism hydraulic control units (HCU) and the er I.
transversing in-core probe (TIP) may be damaged by the hydrodynamic forces of a high vertical water swell in the suppression pool following a less-of-coolant accident (LOCA1-1/
J' If LOCA is the sudden break of a high-energy pipe in the reactor coolant pressure boundary of the nuclear steam supply system.
The largest possible break is the break of a main steam line.
eew e e
e
.4 4
p II.
Description of the Mark III Containment and Pool Swell Phenomena f
The Allens Creek Nuclear Generating Station design uses a General Electric' sixth generation boiling water reactor nuclear steam supply system with a third generation pressure
- j suppression containment system.
(This combination bears the name BWR/6--Mark III.)
The basic Mark III containment design is shown in the retached diagram (Exhibit 1).
The reactor primary system is surrounded by a cylindrical concrete drywell structure which is in turn surrounded by the primary contain-i ment.
At the base'of the drywell a series of horizontal j;
open-ended pipes (vents) in three rows connects the drywell to the containment.
The vents are submerged in an annular if pool of water that is retained by a weir wall insins the drywell.
Any steam released in the drywell from a postulated
..til.
pipe break will be forced through the horizontal vents into
- ~
the suppression pool where it will be condensed by the pool i~m water.
Almost immediately following a postulated LOCA, the-drywell is pressurized by reactor sceam, and a mixture of steam and air is directed te the suppression pool through the horizontal vents.
The rapid increase in drywell pressure
,s...
will accelerate the water initially standing in the weir annulus and horizontal vents.
Immediately fellowing the
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r-clearing of standing water in any vent, drywell air and steam will form a bubble at the vent exit.
This bubble will expand and depressurize to the local hydrostatic pressure.
These bubbles cause an apper displacement of the pool water above
~
tne vents.
The bubbles rise relative to the pool water, i'
reducing the thickness of the water ligament or film above the bubbles.
When the bubbles braak through the water surface, a froth is formed which rises further before falling back into the suppression pool.
The initial motion of the water film and the subsequent motion of the froth create impcet and drag loads on equipment and platforms located above the poolisurface.
The entire process is referred to as
" pool swell."_/
2
.The pool swell loads on structures and components above the suppression pool have been evaluated in more than i,
1 fifty full-scale and sub: scale experiments as part of the
~
2/
Safety relief valve (SRV) actuation also introduces air into the pool as the released steam displaces the smaller air volume occupying the blowdown lines.
However, SRV pool swell does not exist.
Extensive in-plant tests, laboracory tests and an under-standing of the phenomena involved in SRV discharge demonstrate that there is no pool swell due to 'his discharge.
An under-standing of the phenomena is acquired from scaling laws and analytical models of the SRV discharge.
Full-scale in-plant tests were conducted at Monticello, Caroso, Tokai, KKB, KKP and Fukushima-6.
Laboratory tests were also conducted by General Electric, KWV and CNEN.
All these tests confirm that SRV pool swell does not occur.
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Mark III test program conducted by the General Electric 1"
Company.
From this information, loads are selected and used in the design of the ACNGS plant by the architect-engineer and in General Electric's analysis to qualify equipment
- i Eupplied by General Electric.
III.
Mark III Test Program Immediately following the introduction of the li BWR/6--Mark III, the General Elect ric Company started aA extensive experimental and analytical effort to confirm the
.:l-Mark III design.
The purpose of the Mark III Confirmatory i
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Test Progrmm was to ecnfirm the analytical methods used to i
predict the drywell and containment responses following a
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LOCA and to obtain information on the hydrodynamic loads that
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are generated in the vicinity of the suppression pool during a;;
a LOCA.
.T The General Electric Mark III containment pressure a
]
suppression testing program was initiated in 1971 with a series of smal.:;cale tests.
The test apparatus consisted of small-scale simulations of the reactor pressure vessel, drywell, suppression pool and horizontal vents.
A total of
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I' sixty-seven blowdown runs were made.
The parpose of these tests was to determine the behavior of the horizontal vents and to obtain data for determining the acceleration of the w
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P water in the test section vents during initial clearing.
This information was used to establish an analytical model for predicting vent system performance in Mark III and the resulting drywell pressure response.
In November 1973, testing in the Mark III Pressure s-Suppression Test Facility (PSTF) began.
The PSTF consists of I
an electrically heated steam generator connected to a simulated drywell which can be heated to prevent steam condensation within its volume during the simulated blowdowns.
The drywell is modeled as a cylindrical vessel having a 10-foot diameter i
and 26-foo*. height.
A 6-foot diameter vent duct passes from the drywell into the suppression pool and connects to the i
simulated vent system.
Pool baffles are used to simulate a o[
scaled or full-scale sector of a Mark III suppression pool.
The full-scale PSTF testing performed between 9
November 1973 and February 1974 obtained data for the confirma-f tien of the analytical model.
In March 1974 pool swell tests en l_
were performed in the PSTF.
These full-scale test' involved i
li -
air blowdown into tne drywell and suppression pool to identify bounding pool swell impact loads and breakthrough elevation, r
i.e.,
that elevation at which the watar slug begins to break 1"
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up and impact loads are significantly reduced.
Impact load
[h-l data were obtain d on selected targets located above the Jb pool.
In June of 1974, after the PSTF vent and pool system h
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was converted to 1/3-scale, four series of tests were performed to provide transient data on the interaction of pool swell with flow restrictions above the suppression pool surface.
The next series cf 1/3-scale testing, which began j
in January, 1975, measured local impact pressures and total loads for typical small structures located over the pressure suppression pool including I-beams, pipes, and grating.
Data from this test series expanded the data base from the full-scale air tests.
A further series of 1/3-scale tests was added in June, 1975, to obtain comparable data on pool swell velocity and breakthrough elevation to the full-scale air j'
tests.
The emphasis in the testing described above was directed at the evaluation of the pool swell phenomena.
Each test run consisted of a simulation of the postulated i
blowdown transient.
Various postulated break sizes up to i
l lE two times the Design Basis Accident for the containment were tested.
Data were recorded at selected locations around the test facility suppression pool throughout the blowdown so that the hydrodynamic conditions associated with each phase of the blowdown are known and are available for selecting 1"
i appropriate design loading conditions.
General Electric has t I" l;,
used this data to develop hydrodynamic loading conditions in
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ths GE Mark III rafar:nca plant proccurs cuppre.ssion containm:nt system during the postulated LOCA.
IV.
Pool Swell Loadings Equipment and platforms, like the HCU, the HCU ors and the TIP, located in the containment annulus region
~
above the pool surface experience pool swell induced dynamic loads, the magnitude of which are dependent upon both the location and the geometry of the surface exposed.
The pool swell phenomenon occurs in two phases:
" bulk" pool swell followed by a " froth" pool swell.
Bulk pool swell imparts two different loads on exposed structures and components:
impact i
loads and drag loads.
The froth stage of pool swell contributes only a drag load.
A.
Impact Loads The PSTF air test data show that after the pool has risen approximately 1.6 times vent submergence below normal 1
pool level (12 feet)., the slug thickness has decreased to 2
)
feet or less and the impact loads are significantly reduced.
For evaluating the time at which imL:act accurs at various elevations in the containment annulus, the maximum water surface velocity of 40 feet /second is assumed because this i
value bounds all the test data and analysis.
The basis for
[
the loading specification is the PSTF air test impact data.
.[-
These tests involved charging the reactor simulator with 1000
'l' M
psia air and blowing down through an orifice.
Instrumented 7-
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targets located over the pool provided the impact data.
For structures above the 18-foot elevation, the conservative froth impingement load is 15 psig based on data generated during the PSTF air test series.
Again, this impingement load is applied uniformly to all structures.
~
B.
Drag Loads In addition to the impact loads, structures that experience bulk pool swell are also subject to drag loads as the pool water flows past them.
Drag loads are calculated assuming a velocity of 40 feet /second between the pool i"
surface and HCU floors.
C.
Design of HCUs for Pool Swell Loads Larg9 platforms or floors will completely stop the rising pool, and thus incur larger loadings.
For this reason, the HCU platform is located above the bulk pool swell zone.
The GE Confirmatory Test Program indicates that pure bulk pool l
swell terminates at levels much lower than 18 feet above the suppression pool.
Consequently, General Electric advises the p.
architect-engineer to use 18 feet as the elevation of bulk l
l poo". c;well with a linear transition from water to froth in the space of 18 feet to 19 feet above the normal pool surface.
Therefore, for design application, the impact of water from bulk pool swell is applied conservatively at or below elevations s
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}es of 19 feet above the surface of the suppression pool.
The structures above this elevation experience an impulcive loadi.ng followed by a pressura differential loading.
The impulsive load is cue to the momentum of the froth which is decelerated by the structure.
The pressure differential is based on an analysis of the transient pressure in the space between the pool surface and the HCU floor resulting from the froth flow through the approximately 1500 square feet vent area at this
- 3. g elevation.
General Electric test results are the basis for the froth impingement load of approximately 15 psi lasting for 100 msec.
An 11 psi froth flow pressure differential lasting f"
for three seconds is based on an analysis of transient pressure
- m..
in the space between the pool surface and the HCU floor.
The a
approximate value of 11 psi is from a calculation which assumes
.i that the density of the flow through the annulus restriction il.
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is a homogenous mixture of the top 9 feet of the suppression 3
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pool (i. e., 18.8 lb /ft ).
This is a conservative density m
assumption confirmed by the GE one-third scale test which f
shows an average density of approximately 10 lb /ft3 The m
ji analytical model used to simulate the HCU floor flow pressure I
differential has also been compared with tsst data.
These
(
(L tects indicate HCU ficor pressure differential is more realistically c-in the 3 to 5 psig range.
1!I" Vibratory response of the HCU floor to the froth impingement would subsequently transmit a load to the HCU i :J
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The magnitude of this load for Allens Creek will be computed by the architect-encineer in a plant unique dynamic analysis to assure that it does not exceed the dynamic qualification of the HCUs by General Electric.
D.
Design of the TIP for Pool Swell Loads
]
General Electric PSTF test 3 demonstrate that for structures such as the TIP station, which is located approximately six feet abcve the suppression pool surface, pool swell impact il loads are not experienced.
The TIP station does experience a L'
- i drag load and a " bubble" load.
Bubble pressure lozd occurs
[
when the air in the drywell is driven through the vents and forms air bubbles in the suppression pool prior to bulk pool
. c swell.
The pressure of these bubbles is then exerted on the wetted surfaces around the suppression pool.
(
PSTF data also establish that the TIP station would 0
experience a maximum drag load of 11 psid and a 21.8 psid i.'
bubble pressure load.
The TIP system itself is protectel from the loads by cantilever structures which extend beneath the surface of the suppression pool and are specifically designed ll by the architect-engineer to absorb this loading.
I In a larger sense, the issue of pool swell loading on the TIP station is a red herring.
The TIP is a movable radiation source used to calibrate the Local Power Range lh Monitors when the reactor is shut down.
It is not designed or li ;
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Consequently, its ability to survive a LOCA environment, including pool swell loading, has no itaportance save an economic effect which pales in comparison to the other consequences of such an accident.
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4 ATTACHMENT 1 n,
i PROFESSIONAL QUALIFICATIONS PETER P.
STANAVAGE MANAGER - CONTAINMENT ENGINEERING Mr. Stancavage has more than 13 years of Engineering 7
experience with General Electric in the Nuclear Energy Group.
Mr.'Stancavage is now the Manager of. Containment t
Engineering, a positicn he has held for more than two years.
His fi* t eleven years with GE included a variety of Engineers
!],
ing jobs among whi.ch were three years in Containment Engineer-o-
ing, Radiological Lyaluations and Nuclear Engineering.
Mr. Stancavage received his Master's Degree frcm d.I.T.
in Nuclear Engineering.
He completed his undergraduate 1
e work at U.S. Military Academy (West Point).
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