ML20248B795
| ML20248B795 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 09/22/1989 |
| From: | Roche M GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 4410-89-L-0097, 4410-89-L-97, NUDOCS 8910030294 | |
| Download: ML20248B795 (65) | |
Text
-
s-GPU Nuclear Corporation nU Nuclear
- ='rs =8o e
Middletown, Pennsylvania 17057-0191 717 944-7621 TELEX 84-2386 Writer's Direct Dial Number:
l (717) 948-8400 September 22, 1989 4410-89-L-0097/0356P 1
US Nuclear Regulatory Commission Washington, DC 20555 Attention: Document Control Desk j
Thme Mile Island Nuclear Station, Unit 2 (TMI-2)
Operating License No. DPR-73 Docket No. 50-320 l
SNM Accountability
]
Dear Sirs:
By NRC letter dated October 17, 1985, GPU Nuclear was granted exemption from certain requirements for periodic inventory and reporting of the special nuclear materials (SNM) balance for Three Mile Island Unit 2 (TMI-2). As a condition of the exemption, GPU Nuclear is required to conduct an assessment of the SNM remaining at TMI-2 following the completion of the defueling effort.
This assessment is referred to in the exemption as the j
" post-defueling survey." GPU Nuclear letter 4410-88-L-0162 dated September 30, 1988, submitted the initial Post-Defueling Survey Reports I
(PDSRs).
As stated in that submittal, the PDSR documents the GPU Nuclear assessment of the amount of residual StN and describes the methodology utilized to determine the quantity of SNM in each case. The attached PDSRS transmit the post-defueling survey results for the Pressurizer, the Reactor Building basement, and a revision of the Letdown Cooler Room survey results originally provided in the September 30, 1988, submittal.
)
Additional PDSRs will be submitted as they are completed. A compilation of the individual PDSRs will form the basis for the final assessment of the quantity of residual StN at TMI-2 for accountability purposes.
Sincerely, 1
l 1
h M. B. Roche Director, TMI-2 l
.333/ emf Acof GPU Nuclear Corporation is a subsidiary of the General Pubhc Utilities Corporation
@$36 ME2b j
h.
Document. Control Desk
. September 22, 1989 4410-89-L-0097 Page 2
- Attacienents cc:
W. T. Russell - Regional Administrator, Region I L
J. F. Stolz - Director, Plant Directorate I-4 L. H. Thonus - Project Manager, TMI Site 1
F. I.- Young - Senior Resident Inspector, TMI i
1
V l<
l T111-2 POST-DEFUELING SURVEY REPORT FOR THE PRESSURIZER l
)
SUMMARY
The estimate of record of the amount of uranium dioxide (UO ) remaining j
2 in the TMI-2 pressurizer after June 1988 defueling is 0.3 kg +_52% (at one sigma, defined as one standard deviation), distributed as follows:
Gravel-like and fine material 0.267 kg Film 0.001 kg Total
~ 0.3 kg A video-examination of the debris material at the bottom of the pressurizer was used to determine the volume of remaining residual core material s.
A sample of this material was removed from the pressurizer for analysis.
Neutron interrogation and gamma spectrometry methods were used to calculate the amount of U0 in the sample by direct comparisons 2
with a uranium standard and standard Ce-144 and Eu-154 sources. The total amount of residual U0 remaining in the pressurizer was 2
calculated by scaling the sample for the total solids observed.
The quantity of UO remaining in the pressurizer is < 0.04% of the 2
maximum allowable residual UO inventory for the entire TMI-2 facility 2
in Mode 2 (Reference 1).
9
J TM1-2 POST-DEFUELfNG SURVEY REPORT FOR THE PRESSURIZER
1.0 INTRODUCTION
This report presents the analysis of the Three Mile Island Unit 2 (TMI-2) pressurizer uranium dioxide (UO ) inventory.
It is one in a series of 2
reports generated to fulfill the requirements of the TMI-2 SNM Accountability Program (Reference 2). All statistical uncertainties are expressed as + one sigma limits (defined as one standard deviation).
Section 2, " Background", describes the physical attributes, location, and intended functions of the pressurizer.
Its relationship to the accident-and subsequent cleanup activities is also discussed.
Section 3, " Methods", discusses the two methods used based on radioisotope fuel analogs and the more direct assay by neutron interrogation.
Section 4, " Analysis", explains how the estimate of record of the amount of fuel in the pressurizer was calculated, and discusses supporting data, assumptions made, and the assigned uncertainty.
Section 5, " Conclusion", states the estimate of record and uncertainty for the amount of residual U0 remaining in the pressurizer and 2
supporting rationale leading to the conclusion that the estimate is reasonable based upon the available data and analysis performed.
2.0 BACKGROUND
The TMI-2 pressurizer was intended to establish and maintain reactor l
coolant system (RCS) pressure within prescribed limits, and to provide a steam surge chamber and a water reserve to accommodate RCS density changes during operation.
It is located in the south end of the ' A' _ -__
- w. i
- l
- D-ring next to the once-through steam generator ' A'.(0TSG ! A').
The-
-c
. bottom of its lower head is at approximately elevation 312', with the top of its upper head at approximately elevation 354'.
The pressurizer is an approximately 13.0 m vertical, cylindrical vessel with upper and lower hemispherical heads.
It has an inside diameter of approximately 2.1 m in its cylindrical portion, with a wall thickness of approximately 0.2 m.
The upper and lower heads have a radius of approximately 1.1 m and a minimum wall thickness of approximately 0.1 m.
The upper head has three nozzles which connect to the relief system, and one approximately 0.1 m ' schedule 120 nozzle which cor.nects to the pressurizer spray line.
It has an approximately 0.03 m schedule 160 vent nozzle and an approximately 0.4 m inside diameter manway opening. The center of the lower head has an approximately 0.3 m schedule 140 nozzle which connects to the surge line.
A surge diffuser is connected to this nozzle, extending approximately 0.3 m into the lower head. The surge line nozzle ' connects the bottom of the pressurizer to the ' A' RCS loop hot leg.
Vent safety and relief valve lines are attached to the top of the pressurizer and connect to the reactor coolant drain tank (RCDT) through a drain line. Three removable 480-V AC electric heaters are located at the bottom of the pressurizer, approximately 1.5 m above the base. They are designed to replace heat lost during normal operation and to raise the pressure to normal operating pressure (approximately 14.4 MPa) during RCS heatup.
The heaters also function to return the system pressure to normal following transients. The pressurizer spray provides an introduction of relatively cool water from the RCS ' A' loop cold leg to pressurizer steam space, which condenses some of the steam to reduce RCS pressure.
The boundaries of this report encompass the pressurizer, the electric heaters, and the internal ladders of the pressurizer. The location of
.the pressurizer in relation to other RCS components is shown in
~
Figure 1.
Figure 2 illustrates the position of the relief, block, and
' safety valves in relation to the pressurizer. Figure 3 illustrates the pressurizer as defined for this report, which excludes all of the pressurizer's incoming and outgoing pipes and valves.
During the TMI-2 accident, the pressurizer water level was raised and l.:
lowered to thema11y control the reactor pressure. Core material was reloedted to the pressurizer, most probably as a ' result of this process.
f This postulate is supported by the existence of a trace of core material in the RCDT and in the. Reactor Building basement.
The general method used to quantify residual U0 in the pressurizer was 2
l to estimate the' volume of residual core debris after defueling and to-obtain and analyze a sample of the material. The results of the sample analysis were used to infer the total UO content of the remaining 2
material. Two independent analytical methods were used due to the relatively large errors associated with the methods. The analytical methods are based on radioisotope fuel analogs and the more direct assay by neutron interrogation.
Radiological Environment Various valves and joints located on the top of the pressurizer leaked intensely radioactive RCS coolant during the accident.
As a result, the general-exposure rates near the surfaces of the pressurizer are between 0.5 and 3 R/hr, with a f requent value of approximately 1 R/hr. The 0.2 to 0.3 R/hr exposure rate inside the pressurizer is significantly less due to lower film activity on the stainless steel liner.
The external carbon steel surfaces are more chemically active and able to secuester more RCS activity in a tightly adherent film nomally temed fixed activity.
A loosely adherent film also exists leading to easy transfer of contamination to personnel and equipment. Typical levels of 2
approximately 1 pCi/cm2 (2.0 E+06 dpm/cm ) on and near the pressurizer require substantial radiation protection controls that further complicate detailed assay programs.
As a result of this hostile environment, it is not considered ALARA to perfom additional measurements to reduce the uncertainties below the values Quoted herein.
1 l
7.__
in 3.0 METHODS Sample : Collection In March of 1988, technicians used Mini-Rover to collect approximately 1 liter'of core debris from the pressurizer. From this amount, a 100 g
. sample was taken consisting of one large consolidated mass, three smaller masses, and two scoops of fine material.
Defueling In June of 1988, a robotic submersible device called Mint-Rover was used to remove most of the debris (13.5 liters) at the bottom of the pressurizer. - Calculations showed that an estimated volume of 0.9 liters + 35% of sediment remains in the pressurizer (Reference 3).
Due to the inaccessible, underwater location of the core debris', the estimated volume was calculated by examination of videotapes recorded via
-the Mini-Rover camera system after the June 1988 defueling of the pressurizer. Almost all of this amount was at the bottom of the pressurizer.
Since the surge line nozzle is not flush with the bottom of the pressurizer, the sediment collected around this nozzle (Figure 3).
Very small quantities of sediment were also deposited on heater assemblies and in shallow pockets at the base of each heater bundle. The material at the bottom of the pressurizer looked very light and was moved easily by Mini-Rover thrusters.
From observations of videotape, it was detemined tnat the small crust-like material that fomed on the heater bundles was due to corrosion and therefore discounted.
Pressurizer dimensions and geometry, and the depth of the sediment (gauged by comparison to parts of the standpipe diffuser nozzle and the scoop that was installed on the Mini-Rover for sediment collection) were
'used to calculate the volume of the sediment. The average depth of sediment around the pipe was estimated to be 1.6 cm, tapering to 1.0 cm (Reference 3). _ _ _
Gamma Spectroscopy A gamma spectroscopy analysis was performed on the 100 g sample by using a calibrated HpGe spectroscopy system (Reference 4). Gamma spectroscopy was used because it is a passive, non-destructive method of assaying fuel. Gamma radiation associated with the decay of fission products can be used to indicate the presence of fuel. The signal strength of a specific energy peak in a gamma spectrum corresponding to a radioisotope that is chemically similar to fuel can be used to determine the amount of j
fuel. A measurement of that isotope activity should be directly proportional to the amount of fuel present, which is expressed as V0
- 2 Fuel sample examinations and code calculations at TMI-2 support the use of Ce-144 and Eu-154 as fuel tracers (Reference 5). Ce-144 and Eu-154 have been shown to be an acceptable analog for transuranic material.
The oxides of uranium, cerium, and europium have similiar water solubility, therefore the Ce-144 and Eu-154 in the reactor fuel matrices were transported through the reactor coolant system during the accident.
The fuel value corresponding to the film layer on the interior surfaces of the pressurizer was determined using the total inside surface area of the pressurizer, the pressurizer ladders, the heater bundles, and the average fuel value determined by measurements on the pressurizer manway cover (Reference 6).
Active Neutron Interrogation In April and May of 1989, an antimony-beryllium (Sb-Be) photo-neutron
( c, n) source was used to conduct a series of active neutron interrogation measurements.
The interrogating neutrons are generated by inserting a Sb-124 source into the Be cylinder where neutrons are produced via Be ( c, n) reaction. These neutrons are thermalized by a polyethylene [(CH )n p
cylinder which covers the Be cylinder.
Thermalized neutrons from this standard or TMI-2 source system were impinged on the normal U03 residual fuel samples to initiate fission in U-235 contained in UO3
" l
L a
a samples.
Fast fission neutrons emitted during the fission process were
)
detected by a He-4 detector that is sensitive to fast neutrons. A schematic diagram of the top and side view of this system are shown in Figures 4 and 5, and the details of the work can be found in Reference 7.
The validity of the standard to unknown comparison method depends on J
small or similiar neutron absorption for both. This was determined to be the case in Reference 8.
4.0 ANALYSIS Gamma Spectroscopy Analysis of Samples The isotopic activity recorded during the spectroscopy analysis of the 100 g sample taken in March of 1988 is shown in Table 1.
Measurements of the Eu-154 activity were uncertain by approximately + 8.7%, whereas the uncertainty of the Ce-144 based activity was 141.3%. The errors quoted are based on counting efficiency, calibration source activity, and geometric uncertainties.
Using Ce-144/ fuel and Eu-154/ fuel ratios (82.8 pCi/g 153% and 40.4 pCi/g + 86% UO, respectively, obtained from more than 100 2
in-vessel and ex-vessel samples as of April 4,1988), the amount of UO in the sample was calculated to be 0.0051 kg 167% for the 2
Ce-144/ fuel ratio and 0.007 kg i 86% for the Eu-154/ fuel ratio (Reference 3). The uncertainties quoted are + one sigma values of the respective data set and not of the means of the sets.
Active Neutron Interrogation Method The active neutron interrogation method required three measurements:
1.
Background for the system was determined while the Sb-Be source was active, in absence of any samples or a normal UO3 standard.
i E
-- J
i 2.
' A normal UO standard (0.5 kg UO ) was placed in the 3
3 measurement location (see Figures 4 and 5) and measurements were repeated.
- 3.
The pressurizer sample'was placed in the same location as the two above.
Results of background measurements were subtr. acted from the results of samples and 00 standard measurements, and then background-subtracted 3
results of samples and UO were compared to one another.
3 Since the U-235 content of the U0 standard (0.711% by weight) and 3
average U-235 enrichment of TMI-2 fuel (2.5% by weight) are known, residual fuel content of samples can be estimated by ratioing the results of samples and standard measurements.
These series of neutron measurements resulted in a UO estimate of 0.0097 kg +_48% in the 2
pressurizer sample.
The U-235 content of both the standard and the sample allowed lower comparisons to be made.
The weighted average of the Ce-144, Eu-154, and neutron based determinations, resulted in. the final UO sample value of 2
0.0067 kg + 38% for the 0.1 kg sample (Reference 3).
The fractional U0 content of the sample was therefore 0.067 kg/kg.
2 Final Analysis The average bulk density of 4.4 g/cm (Reference 3) corresponding to TMI-2 core debris was adopted for the remaining material within the pressurizer.
The amount of UO in the pressurizer, excluding the fuel-2 value corresponding to the film layer on the interior surfaces of the pressurizer, is estimated to be approximately 0.267 kg of UO as 2
follows:
Volume x Density of material x Weight fraction of reactor fuel in the sediment = Amount of UO in pressurizer (Reference 3).
2 3
3 I
908 cm x 4.4 g/cm x 0.067 = 0.267 kg. -
~.
p..
?
2 Using.the tstal 'inside ~ surface area of the pressurizer (8.12 E+5 cm ),
2 2
pressurizer ladders (1.32 E+4 cm ), and heater bundles (4.27 E+5 cm )
.?
2 (Reference 3),'and using the average U0 value (8.2 E-7 g/cm 51%)
2 determined by measurements on the pressurizer manway cover (Reference 4),
the fuel value corresponding to the film layer on the interior surfaces of the pressurizer is estimated to be 0.001 kg 151% UO2 (Reference 3).
The uncertainty is partly determined from a volume estimation of the residual debris at the bottom of the pressurizer. Five engineers examined the videotape of the pressurizer and estimated the amount of material remaining. The estimate of record is the average of these estimates. The volume uncertainty is derived from the range of these estimates (Reference 3) and is 1 35%.
Analysis Conclusion The estimate of record of 0.3 kg has an overall uncertainty of 152%
(Reference 3), which was calculated as the counting, and volume uncertainties-taken in quadrature.
The goal -of the current defueling program is to remove more than 99% of the original core inventory of approximately 94,000 kg.
In that context, remaining in the pressurizer is <0.04% of the 0.3 kg_ quantity of UO2 the maximum allowable residual UO. inventory for the entire TMI-2 2
facility in Mode 2 (Reference 1).
5.0 CONCLUSION
Using all available information, the estimate of record of the amount of fuel in the pressurizer after June 1988 defueling is 0.3 kg 152%-(at one sigma).
The pressurizer, as it currently exists, is considered to be isolated.
For an extended period of time in the past, approximately 400 ' liters / minute of water was continually added to the top of the pressurizer and discharged from the bottom.
The purpose of this flow was.
1... -
'.to clarify the RCS water in the pressurizer. Video-examinations during that period showed that there was no significant change in the quantity of UOp due to that flow. Therefore, the amount of U0 remaining in the pressurizer 2
is considered to be stable, and is not anticipated to change during subsequent -
~
system draindown at the completion of defueling, as long as the flow rate is approximately 400 liters / minute or less.
If the draindown rate is significantly greater, it may be necessary to reinspect the pressurizer and associated surge line to ensure that material did not relocate.
s r
, l l
l
(________
REFERENCES 1.
R. E. Rogan. Residual Fuel Basis.
GPU Nuclear Internal Memorandum 4410-88-M-0306/4400-88-078. Middletown, PA:
GPU Nuclear Corporation.
24 August 1988.
2.
J.S. Schork.,TMI-2 Core and Special Nuclear Accountability Program.
Administrative Procedure 4000-ADM-4420.02, Rev. 3-01.
Middletown, PA:
GPU Nuclear Corporation. 10 October 1988.
3.
M.H. Haghighi.
Calculation of the Pressurizer Fuel Value.
4550-3220-88-016, Rev. 2.
Middletown, PA: GPU Nuclear Corporation.
19 July 1989.
4.
R.E. Lancaster.
Calculation of the Pressurizer Sample.
4550-3223-88-011, Rev. O.
Middletown, PA: GPU Nuclear Corporation.
7 April 1988.
5.
P.J. Babel. Ce-144, Eu-1S4 and Eu-155 as Tracers for Fuel Debris.
TB-86-41, Rev. 2.
Middletown, PA: GPU Nuclear Corporation.
26 February 1988.
6.
J. Greenborg. Deposition of Fuel on the Inside Surfaces of the RCS.
TB-86-37, Rev. O.
Middletown, PA: GPU Nuclear Corporation.
4 September 1986.
7.
M. H. Haghighi.
Fuel Estimate of OTSG and Pressurizer by Active Neutron Method.
4800-3221-89-037, Rev.1. Middletown, PA: GPU Nuclear Corporation. 19 July 1989.
8.
M. H. Haghighi.
Comparison of UO Standard and Pressurizer Sample 3
Total Absorption Cross-Section. 4800-3321 079, Rev. O.
Middletown, PA: GPU Nuclear Corporation.
25 August 1989.
c-b
[.-
TABLE 1 RESl!LTS OF SAMPLE FROM THE PRESSllRIZER*
Isotope Activity (uCi)
Cs-137 36,653 + 5.2%
Cs-134 490 + 6.8%
Co-60 5,006 + 5.7%
Sb-125 5,235 + 6.3%
Ce-144 420 + 41.3%
Eu-154 284 + 8.7%
l Ru-106 788 + 8.6%
1
- References 3 and 4 l
l
3
)
l J'
FIGURE 1 l
'TMi-2 PRESSURIZER l
IIN RELATION TO OTHER RCS COMPONENTSI l
l I
C#NOT CAN(
e'l-k Sit.am ofNtRafo#
pra ron C00Lauf Pune y
!!L pg r.29 RC M-19 i
f84!#fjf
> :nnr a-
@ldig:n.:._,-
4 o ).
/
l ik Bli; '""""' """'
M i
c
[k
.M I -.
'4n
- 4-#;
l fp*
p> r1p j.,... m,<,.ro.
9g;I 4 (g
~.
R lLi "c *-'a Co,e renem 1. - 3)
I
%(. n-9 C -r >
t i
o-me.
,}
y,(
C gm
- {
.jf r
g.
c.1
- llIlI, i
I
{'l
[
/
w
[
i
.i ill!
kgp h*-
hw u -:. ::
h
.g;, m-,.....e k
m:r """'
\\
4=
NN J LEG 5 Jw-----a_ _ _ _
___m_
FIGURE 2 TMi-2 FRE55URIZER (SHOWING VALVE LOCATIONSI 4
- ell [F WALVE OVf# PPf55bPE PP0f(C710h FDP PwRS T f P t:hL REL!!' vutvf!
Jf watre in fe! Prin4*' LCDP CA## tt? HERf FROM THE REACTDa 70 f at
\\
3rgem stNtsetop RNO TMth ettVPN$ 70 fH[ 9f RC f 09 TO FP{W[NT TH(
Act0*'f0DR?! 5*lk.DIRP PIPE AND MGVE $f f G*
\\
mAff# lh THE L0a* **0P 90 FLING. f at 51 stem 15 *EPf UNDER CON $7Hht FLpm pa'tb 08
\\
engg$ueg ev a PpC550pilEp CONNECTED 70 fME PtiffAPT LOOP.
150.000-250 000 L0ea g
It tg=*gea'u#f e%D CON 5tGUE*T rwt55urf IN Tat PRES $U#12tp OECOME
- 16-31 M0/53
\\
700 nice, twt Pt trv vatvf 5 A'O' YME PRC550# f ff# NILL OPEN TD
\\
PELF RSE Sif aa DNO #CDutt PRES 5URE. fME Sa'tif VRLVE5 PROVIDE TME F I NR. ppgt[;f ggg a%[h$! ggge Pegggyp[. OPf hlNG IN RAP!
T NL t*Mit 5 maE % 5 f t AM FLDa 11 GegnTgp twas Twr ogtttr vetvf 5 tas RCC0aMDM'!
M05' Pu*5 wart 80Tft if rE 5 0'.
W A VES. at f MOUsa ON:.'
C00!5 turFE nef et p*
10 TwPtt x
5Asffr.R.vt3 pot PCMIRE0 Bf p5"i or thra Yvat er vetvts Itsf ALLED 4*B0vt f**E P9f 55URIZE*. BOTH 'f*ES r
or gg er g. anttn f f Pgte.t v anscg FPO* I 70 ( Ff. TAtt 0*th Rh; g g'y ypgw'
-I CLOSE nu'Da'f!; ALLY lh P!5*0h5T 70 PPf 559Pt T#RMS!ENT5. #E6 If r vntst !
V faF( DPEPET[O er as g u tteNet Pp.rg Supegy u3ontty couf ROLLED P' 510%n t s Sarf f vntyts eng 5gte.mt f untfD. OP[h!NG Rf ELECTPt u
ppt 5EitCft.Q P#[$$gP($.
y c.
\\
surf 3 vaat U 7.\\
.\\'
u
- Pica Sarer, vntrgt H
20P,t
~ pp pc SUF I ?EP-i
. m.i-gO j
/-
Dtu. PIPE mec whvt
(
5tra u0. ea
%v
- s:msn.r'r5 O.
.- EPP41 LINE
- te.-
I a
/
i n -2 asm
~,
t
/
/
$l 11
. D at J
/
j [
j J
.e
.-.- CDFE FLOOD Tofa r
p l
/M 9l3+ ?!
9 i
y? J
- . l j
V
,,2 E. E C ! F. FUMP5 j
Ij y- !
f h
l, l'
f" = )
/
j til
~ REACTOR COOLANT STEAM GENEF ATOR -
j
. 4
~,
DRAIN TANK 11 j
j usham ~
EUFDE LINE p
N'
'?
l l
l l
FIGURE 3 TMJ-2 PRE 55URIZEP GE O M E T P AND DIMENSIONS 7.
I
/ENT.N0ZILE eg m
fn f1ANWAY COVER
/%
+
c N' t f x5 Fj
' RELIEF VALVE N0ZZLE t_
pQ ff (TYP. OF 3) 4 J J QT SPR A Y-LINE N0ZILE
~
f LEVEL SENSING N0ZZLE (TYP. OF 31 yjY!N T E RI
i, l
p.:
O.PArF c
w-c l...
4 1
INTERNAL LADDER l
g Jh
- NORMAL WATER LEVEL j'Ti /
i f
l
,- /E55EL SUPPORT 5 l
-1302 1 CM j
i-
/ y'
'5 AMPLE NDZZLE BEHDC 4
)
i M!/
l l
f l Muf '/
~ =;TEF suNoLE w
)
0 p it i
/
p.
. _L E,L /
LEVEL SEN5ING N0ZZLF J,
_ N i /
,7 iTYP.
OF 3!
g j,p hun [ni.s yl.j
\\
r 7 t L.
L, t t5 K e
~_
sggy.
7
/
h Limaddh'/y
.1-tF ce- ' ' r,--
L I g* ;- gg::i ;
4 e
i s
~.e-
_-.mmm_____
L.-
FIGURE 4 1
NEUTRON INTERROGATION SYSTEM IN AX130 (Top View) l l
Pre-amplifier h
He-4 Detector Nf
/
U03 location for outside of 4gt >..
the shield measurements t
^,. _.
- Q.
^"3 Am-Be source location
~
n A
UO-3 in glass bottle or sample - in cave Be-poly Cylinder Tables Q
Lead Sb Source Slug Drawing is not to scale.
r FIGURE 5
+
NEUTRON INTERROGATION SYSTEM IN AX130 (Side View)
Lead to shield detector from Sb-Be source UO-3 in glass bottle
/
or sample - in lead cave Be poly Cylinder
\\
/
Pre amp
- x//mA xxx. x x(x g 7
- simmF/
He-4 Detector Jc
$3:
e cc
- =
{ INUEf9MelijsIIt[% '
\\
os o
o No-Lead Source Pig Tables E
Lead E Sb Source Slug
- Drawing is not to scale.
' k I
o-TM1-2 POST-DEFUELING SURVEY REPORT FOR THE REACTOR BUILDING BASEMENT
p.
N;
SUMMARY
The estimate 'of record of the amount of uranium dioxide (UO ) remaining in 2
the Three Mile Island Unit 2 (TMI-2) Reactor Building (RB) basement is 1.3 kg + 54% (at one sigma, defined as one standard deviation).
The distribution of the fuel is as follows:
Basement Sediment 1.3 kg 154%
Basement Water
<0.0003 kg-Basement Walls
< 0.0001 kg Total Estimate 1.3 kg 154%
The amount of UO remaining in the RB basement sediment was developed using 2
remote sampling and gansna spectroscopy.
The' amount of UO remaining in the RB basement water was estimated by the 2
boildown of a sample and the analysis of the gross alpha emitter content by counting a planchet of the evaporator solids in an alpha scintillation counter.
The amount of UO2 remaining in the RB basement walls was conservatively estimated because none of the core samples taken of the walls contained measurable Ce-144 or Eu-154, or inferred U0
- 2 The 1.3 kg quantity of UD remaining in the RB basement is <0.2% of the E
maximum allowable residual UO inventory for the entire TMI-2 facility in p
Mode 2 (Reference 1).
.s
^s l
'TMI-2 POST-DEFUELING SURVEY REPORT ~
FOR THE REACTOR BUILDING BASEMENT
1.0 INTRODUCTION
l This report presents the analysis of the Three Mile Island Unit 2 (THI-2)
Reactor Building (RB) basement uranium dioxide (UO ) inventory.
It is 2
one in a series of reports generated to fulfill the requirements of the L
TMI-2 SNM Accountability Program (Reference 2). All statistical uncertainties are expressed as + one sigma limits (defined as one
.I standard deviation).
Section 2.0, " Background", describes the physical attributes of the RB basement and its relationship to the accident and subsequent cleanup activities. The boundaries of this report are also discussed.
)
i i
i Section 3.0, " Methods", describes how existing sample data were used to produce the estimate of record. Where gamma spectroscopy data were available, these data were used in lieu of calculated sample estimates.
l
.This is considered an overestimate, since much of the sediment volune estimation was _done prior to desludging.
1 Section 4.0, " Analysis", explains how the estimate of record of fuel in L
the RB basement was calculated based on the sample analysis data which were collated for each source category (sediment, water, and walls).
Section 5.0, " Conclusion", states the estimate of record and uncertainty for the amount of UO remaining in the RB basement, and supporting p
rationale leading to the conclusion that the estimate is reasonable based upon the available data and analysis performed. 1
2.'O BACKGROUND-j
.The RB basement consists of the space between the floors of elevation l
282'-6' and elevation 305' of the Reactor Building. AlthoughL the RB
]
b'asement houses several major components,' this PDSR does not include the
-l a
amount of fuel retained in the various system components below the 305' elevation.- These components will be discussed in other PDSRs. This PDSR j
i does, however, include the entire floor area of the 282'-6" level, l
in'cluding sediment, water, 'and walls. The interior walls and floors are
]
constructed primarily of poured reinforced concrete. The RB containment is poured reinforced concrete with a carbon steel liner.
1 During the THI-2 accident, reactor coolant was discharged from the reactor coolant system (RCS) into the RB basement via the pressurizer pilot-operated relief valve (PORV) and the reactor coolant drain tank (RCDT). The discharged reactor. coolant increased pressure in the RCDT l
- causing.its rupture. This vented the RCDT to the RB basement. The reactor coolant that was discharged into the RCDT and RB basement contained reactor fuel fines from the damaged reactor core. The l
locations of the RCDT and the discharge are shown in Figure 1.
The reactor coolant that was discharged into the RB sump became mixed with sediment-bearing river water, building spray water, and additional l
-leakage from the RCS. The RB basement remained flooded with this water l
i for approximately two years.
During this time, the reactor fuel fines settled out onto the RB basement floor (Figure 2).
Radiological Environmec '
..The Reactor Building basement was the repository of most of the RCS water released as a result of the March 1979 accident. The accident water l
remained until removed durina the period starting September 1981 and ending March 1982. During this period, a fraction of the soluble fission products that contaminated the water were absorbed into porous surfaces As a consequence, present area radiation levels vary from a low of 1-2 R/hr to a high of 50 to 100 R/hr. The general area exposure rate tends to be near 10 R/hr.
) I
_ ____ - _ a
_7 y
2 Significant loose surface contamination existed (>l uCi/cm ) and contaminated water (1-3 pCi/ml) was present to a depth of 15 to 20 cm on the RB basement floor as of September 1989. The methodology covered by this report is considered to be consistent with keeping radiation exposure as low as reasonably achievable (ALARA) due to the extremely hostile nature of the basement environment.
3.0 METHODS The estimate of record of the amount of UO remaining in the RB 2
basement was determined by sample analysis of the RB basement sediment, the remaining RB water, and the walls of the RB basement and gama l
spectroscopy data.
Samples have been taken and analyzed for uranium content since 1979. The fuel analysis data were collated for each source category (sediment, water, and walls), and the data were then applied to the appropriate area and source category to arrive at the estimate of record.
For sediment analysis, remote sampling and gamma spectroscopy were performed.
Initial remote samples (from 1979 to 1983) were acquired by lowering sample acquisition devices into the RB basement at various locations. These initial samples were suspect due to an insufficient mass of solid materials. When a robot became available, it was possible to obtain larger samples. Teachability, trace metal, radiochemical, particle size, and morphology analyses were perfonned (Reference 5).
Gamma spectroscopy was used in the RCDT discharge area and in other locations that were identified as the most probable locations for reactor fuel. Gamma spectroscopy had the advantage of being able to remotely examine large areas of the RB basement.
In three locations in the basement (see Figure 3), gamma spectroscopy was performed with sodium iodide detectors (References 3 and 4).
The sodium iodide detectors were lowered into the basement and were used to measure the 2190 kev gamma ray signature of the praseodymium 144 daughter of cerium 144. Using these data, the fuel amounts in those areas were determined by back-calculating using a cerium-to-fuel ratio __
y
'(Reference 6). The results of.these spectral measurements. provide a
' measure of the total fuel content integrated over the' field of. view of-the detector and have been used when appropriate, y
For the water samples, most of the analyses involved the boildown of a sample of basement water and the analysis of the gross alpha emissions by counting a planchet of the evaporated solids in an alpha' scintillation counter.
Gamma spectroscopy measurements of the RB basement concrete core'did not show the presence of. Ce-144, which is the' analog for reactor fuel. The fuel content of the walls are assumed to be due to intrusion and 'plateout-from the basement water. Since the quantity of dissolved U0 in the 2
water is low, it can be assumed that the walls also contain minimal reactor fuel.
4.0 ANALYSIS
.The following sections describe the analyses used for each of the major.
-source categories.
4.1
- RB Basement-Sediment The UO content for the sediment has been estimated using two 2
samples and one. spectroscopy scan taken in the Reactor Building basement. Although mrtny samples have been taken in the RB basement from August 28, 1979 to present (see Figure 3), most are suspect due to insufficient volumes to provide a representative sample.
The largest of the samples was 0.7 g.
A sample taken in the RCDT (see Figure 3) did not represent the basement floor and was rejected. Two samples taken from the impingeme'nt area (see Figures 1 and 2) had enough mass (10 to 20 g) to be used in the calculation. The uranium to sediment ratio was determined by averaging the results of the samples. The average of the analysis i
results was 3.7 x 10-6 g UOp per gram of sample (Referen:e 7).. _ _ - -.. ___- _ _ - ___ _ _ - _
,9 The standard deviation associated with the sediment samples was approximately 181%. This UO / sediment ratio was used in all 2
areas of the basement except the RCDT discharge area.
In that area, the results of gamma-ray spectroscopy were used.
For the gamma-ray spectroscopy scans, a total of three measurements were made. Two measurements were eliminated due to the non-observance of a Ce-144 peak and a high minimum detectable level (MDL) value caused by high water levels in the basement. The gamma spectroscopy scan of the RCDT discharge area was used to characterize this area. The result was an estimate of 1.2 kg +_ 58%. The uncertainty in the measurement is based on the counting error and the Ce-144 to reactor fuel ratio.
The amount of sediment, previously estimated from video-inspections of the floor area (Reference 7), was reduced by the quantity of sediment removed by the desludging operations (Reference 8) to derive an estimate of the remaining sediment volume.
The sediment mass is given in Reference 7 and summarized in Table 1 of this report.
The ndxt step was to multiply the total grams of sediment in each area by the UO / sediment ratio and add the quantity of reactor 2
fuel for the RCDT discharge area. Table 2 shows the estimated final area tallies. The result was an estimate of 0.05 kg 2 801 -of UO in the sediment that was located outside of the gamma scanned 2
area. To this amount was added the 1.2 kg 158% of UO2 identified from gamma scan conducted at the RCDT discharge area.
Addition of these terms yields a total of 1.3 kg +_54% of UO '
2 Desludging operations to remove the sediment removed approximately 0.2 kg of UO and 4890 kg of sediment, leaving 9710 kg of 2
sedimer' in the basement (Reference 7).
Therefore, the quantity of rear,,
Je1 on the floor was 1.1 kg + 64%. The sediment represents the major source term in the RB basement, since the dense insoluble fuel fragments and fines are assumed to have remained with the sediment. -
This estimate does not account for U0 that has been deposited in 2
the drainage system. However, most portions of the drainage system have been flushed by waste water from the gross decon experiment and other work via the floor drain risers from the upper elevations. Assuming that a portion of the drainage piping not subjected to this flow has approximately 0.3 cm of sediment 3
distributed along its bottom, there was 7.1 x 10 g of sediment in the piping (Reference 8), equating to 0.03 g of UO using the 2
impingement sediment fuel concentration of 3.7 E-6 grams of U0p per gram of sediment.
The drain contribution to the sediment term was, therefore, insignificant.
In addition to the sediment in the drain system, reactor fuel may have been deposited in part of the drain system due to the decontamine. tion of reactor defueling tools at the 347' elevation in the tool decontamination facility. The quantity of reactor fuel was estimated from the number of tools decontaminated (approximately 500), the surface activity of the tools (30 pg 2
UO /cm of micron size particles), and an average surface 2
area.
It was assumed that all surface activity on the tools was washed into the sump. Larger fuel particles may have been transferred to the building drain piping that was connected to the tool decontamination sink.
This drain will be assayed for fuel and reported upon separately. The result of this analysis is that there is, conservatively, 0.2 kg of fuel in the floor drain system /RB sump from tool washing (Reference 8). Therefore, the estimate of record of the amount of UO in the RB basement from i
2 the sediment and the drain system is 1.3 kg + 54%.
The cmount of fuel contained in the RB basement is expected to remain static, since it is not anticipated that further reactor coolant will be deposited on the floor. This estimate also assumes that future desludging operations will not remove additional material.
y r.
4.2-RB Basement-Water
. The contribution of UO fr m the water presently in the RB 2
basement is ~ insignificant in comparison to the sediment estimate.
To determine this estimate, the results of-17 basement water 1
samples taken in 1988 have been retrieved from the GpU Nuclecr Laboratory Data System (NLDS) database for use in this estimate.
Initially,. the samples were not analyzed for the purpose of quantifying the U0 e ntent. However, the boildown residues were 2
counted for 20 minutes in a SAC-4 alpha scintillation counter to obtain a gross alpha analysis. Based on the average water. level in.
the RB basement, as reported by Plant Operations on'the l
Plan-of-the-Day printouts from May 1. to May 24,1988, there were approximately 15 cm of standing water in the RB basement, ecuating 0
to approximately 1.5 x 10 m1 of. liquid.
At an alpha activity.
of <5.3 x 10-6 Ci/ml (the mean MDL activity), there were 2
<7.88 x 10 uti (Reference 10) of the calculated alpha activity In the basement water. Using the ratios of alpha-emitting isotopes present in the reactor fuel and the alpha activity of the water,
< 0.3 g of U0 was estimated to be in the water, which indicates 2
that negligible fuel Quantities were present in the RB basement water.
.The estimate of record of the combined total for the sediment and water therefore remains to be 1.3 kg +_54% of UO '
2 4.3 RB Basement Walls, Gamma spectroscopy measurements of the concrete cores of the RB basement concrete did not show the presence of Ce-144, which is the analog for reactor fuel. The fuel content of the walls was assumed to be due to intrusion or plateout from the ba'sement water. As shown above, the assay of the RB basement water yields negligible a>
UO content. Since the Quantity in the water is low, it can be 2
assumed that the walls also contain minimal fuel due to the low probability that particulate fuel would seep into the concrete. ____ _ _ __ _ ____- - _ ___-_-___- _ -
6:
The quantity of UO in the concrete was calculated from the gross 2
alpna activity in the.RB basement water and the volume of water i.
.that diffused'into the concrete. From numerous concrete core bores of the various' concrete walls and the floor. it is known that the radioactive material penetrated approximately 5 mm into the concrete walls and floors and completely. through the block wall.
The volume fraction of water absorbed into the concrete block. and
. pou'id concrete is 0.21 and 0.05, respectively. The total amount of affected concrete is 6.3 x 106 c,3 of concrete block and 6
3
- 8. 8 x '1 0 cm of poured concrete. The activity is therefore the product of the volume of affected concrete, the fraction of.:
absorbed water, and the gross alpha activity. The gross alpha activity is related to the quantity of UO as described in-2 section 4.2 of_ this report.
If it is assumed that the walls hold a quantity of UO equal to the maximum concentration in the water, 2
the result is an additional'0.003 g of UO '2 The estimate of record of the combined total for the sediment, water, and walls therefore remains to be 1.3 kg + 54% of UO
- 2 The goal of the current defueling program is to remove more than-99% of the original core inventory of approximately 94,000 kg.
In
' that context, the 1.3 kg quantity of UO remaining in the RB 2
basement is <0.2% of the maximum allowable residual U02 inventory for the entire TMI-2 facility in Mode 2 (Reference 1).
5.0 CONCLUSION
The estimate of record of the amount of UO remaining in the RB 2
basement is 1.3 kg + 54% (at one sigma). This uncertainty is based on the square root of the sums of the squares of the individual uncertainties associated with each component of the analysis.
This estimate of record.is derived from existing sample analysis resul ts.
It is expected to remain static since it is not expected that - - -
}
additional quantities of water from the RCS system will be added to the RB basement.
The UO content is not expected to change significantly i
2 if the basement is dewatered, but may be reduced if a more efficient method of desludging can be employed.
I
-9 1.
l
\\
E, REFERENCES 1.
R. E. Rogan. R_esidual Fuel Basis.
GPU Nuclear Internal Memorandum 4410-88-M-0306/4400-88-078. Middletown, PA: GPU Nuclear Corporation.
24 August 1988.
2.
TMI-2 Division. TMI-2 Core and Special Nuclear Material Accountability.
Administrative Procedure 4000-ADM-4420.02, Rev. 3.01.
Middletown, PA:
GPU Nuclear Corporation. 10 October 1988.
3.
P.J. Babel. Reactor Building Basement Fuel Estimate. TB 85-08, Rev. O.
Middletown, PA: GPU Nuclear Corporation. 11 March 1985.
4.
P.J. Babel. Reactor Building Fuel Estimate (A Supplement to TB 85-08, Rev. 0). TB 85-08, Rev. 1.
Middletown, PA: GPU Nuclear Corporation.
13 May 1986.
5.
K. L. Wright. Radioanalytical Report.
Rockville, MD: Science Applications International Corporation.
11 August 1986.
6.
P.J. Babel. Cerium 144 as a Tracer for Fuel Debris. TB 86-41, Revs. 0 - 2.
Middletown, PA: GPU Nuclear Corporation.
10 October 1986, 15 September 1987, and 29 February 1988.
7.
P.J. ~' bel. Reactor Building Basement PDSR Reactor Fuel, GPUNC Calculation 4530-3220-89-013, Rev. 3.
Middletown, PA: GPU Nuclear Corporation. 19 July 1989.
8.
D.E. Ferguson. Reactor Building Basement Sludge. GPUNC Memorandum 4550-87-0218. Middletown, PA: GPU Nuclear Corporation. 18 November 1987.
9.
R. E. Lancaster.
Estimate of the Amount of Fuel Transferred to the RB Sump. GPU Nuclear Calculation 4800-3220-89-013. Middletown, PA: GPU Nuclear Corporation. 16 May 1989.
10.
D. E. Ferguson.
Reactor Building Basement Water SNM.
GPU Nuclear Calculation 4530-3220-88-018, Rev. O.
Middletown, PA: GPU Nuclear Corporation.
15 July 1988. l
~
TABLE 1 MASS INVENTORY SHEEl 4.R REACTOR BUILDING ELEVATION 282'-6" SEDIMEN1*
ESTIMATED ORIGINAL' AREA 0F BASEMENT SEDIMENT MASS (g)
Enclosed Stairwell 5.05 E+6 OTSG Leakage-Clr 3.44 E+5 Impingement Area 6.85 E+6 Northwest Quadrant 1.47 E+5 Inside of D-Ring 9.37 E+5 Leakage Coolers 1.26 E+5 RC Drain Tank 1.23 E+5 RCDT Discharge 1.45 E+5 RB Sump Area and 6.95 E+5 Letdown Coolers Incore Cbl Chase 1.45 E+5 STARTING MASS:
1.46 E+7 REMOVED MASS:
4.89 E+6 REMAINING MASS:
9.71 E+6
- Reference 8.
-_mu_w-___-___
u: e c ;m..
<a cy
. lfg +
.;g 2
} ]!?
\\!! >
- 3 r-( it h
a~
l{
'YABLE 2
'W.
' ESTIMATE OF RECORD OF-REACTOR FUEL IN' SEDIMENT
+
i Reactor Fuel Area Of Basement (kg of 002)*
Stairwell '
O.0187-0TSG Leakage Clr 0.0013 Impipyard Area 0.0250 Northwest Quadrant 0.0005 Inside of D-Ring 0.0035 Leakage Coolers 0.0005
.RC Drain Tank 0.0005 m
. RCDT Discharge 1 2000**
. Letdown Coolers Incore Cbl Chase 0.0005' Initial UO2 Material-1.253 Removed U02 Material
.176 Remaining Floor U02 1.077 Material Based on fuel content of 3.7 E-6 grams of UO2 per gram of sediment and total estimated mass of sediment.
Fuel content is based on direct gamma spectroscopy measurements.
61 FIGURE 1 REACTOR BUILDING SASEMENT FLOOR PLAN r
0 Os
- ::~" g 3
h?
==
y =,
- i gH
- (=-)...
H h~
\\
l N
=g r
O gg
- ~, -
a
,i o
9' y
I
,m zg x.
z H
N t* 5 5 lhj
'- ]
H w
$ AS T D
_~._ s/
~~
_a m
l l-l
__--_ 7 _________
4 i
FIGURE 2 I
REACTOR BUILDING BASEMENT FLOOR (View Looking North) i 1
,Y
~%.y.
9 N,tv -
4 TMI-2 Reactor Building i
i i
i I !
- En q#:
+
(Mi,,
T%3 1
i
/
i i
i j
l l l
~
I i
j e
I i
r, l
l Ws i
i E1. 305'
- e
~
l l
V C
J.-
i i
1 a_4 m
l l l
',y d,
I El. 282'-6"
]
( g%i r g==s ~q t y %
/
y 4
\\;;;qw T
as y
s s
%n.
f Basement
w.
y ;.
1',
.t l
n
-g d,h E..,
s 5
5 E
=
a
.ma E b
g i
5 W
8
(
5 Eh
$n =-g
<e e
6 b
Sh
=
a 81
=
~
~
<a r*
5 5
E3 E3 I
c e
c.
E 5
h s e h o.
E" h.
h-E!
a n
w
'w W
z a
5 y
N.
u N
!d 8
E B
B 3
4 56* 5 5
1 P
2 2
Eg
$p$
$3 g
a 5
b B
- 5 3
IE U
o E
0 8 u
s' "5
Esi a o
e os e
a B
B a
=
a ga 5
t t
R-a s
E at 5
3 g
egg a G
=
ig e e' gelibbb*=*n c
c g
5 5
i a
a o
=
o
- *' *g g
- ~
y 14 Hhi E
b~
z
@ a% i i Di's e
m s
m a
.3 c.
a
,O,.
k
~
g e
/a, -
z 4
V k
a l
g k,'r--
h Q. Qs iT.
a.
a OO g
gj v@
8 f
i g
t ;
t' i
ig 5__
's. _ /
E E
E k
1 i
j
{
\\
W -
x E
il o
0-G og M, N
- D i
x w
fx N4'/
ma
/s Y
H H
Y g'
TMI-2 POST-DEFUELING SURVEY REPORT FOR THE LETDOWN COOLER ROOM (REVISION 1)
L
-s-TMI-2 POST-DEFUELING SURVEY REPORT FOR THE LETDOWN COOLER ROOM (REVISION 1)
REVISION
SUMMARY
The TM1-2 Post-Defueling Survey Report (PDSR) for the letdown cooler room has been revised due to revisions in the analyses that formed its basis.
The estimate of record of the amount of uranium dioxide (UO ) reactor fuel 2
remaining in the letdown coolers and piping in the letdown cooler room has.
been increased from a minimum detectable level (MDL) value of < 3.2 kg to an MDL value of < 3.7 kg with an uncertainty of approximately + 53% (at one sigma, defined as one standard deviation) for the following reasons:
1.
Revision in the determination of the total number of gamma-ray events occurring in the region of interest (ROI) for cerium 144 (Ce-144) for the detector efficiency and the letdown cooler room measurement.
2.
Recalculation of the detector efficiency.
3.
Revision ir +he calculations for the calculated gamma-ray fluence rate from the letdown coolers to the detector.
4.
Revision in the selection of the location of the reactor fuel within the letdown coolers for the estimate of record value.
Revision in the Number of Gamma-ray Events in the ROI for Ce-144 The revision in the number of gamma-ray events in the ROI for Ce-144 resulted from changing the method of determination.
For the detector efficiency, the i
leh
. original PDSR method to detenttine: the numbar of counts in the ROI for~ Ce-144 was. taken to be-those counts in 48 channels,above the 2.19 MeV peak (half-peak i
analysis).
In Revision 1 of the PDSR, the ROI wds determined from'the full l.
width at half. maximum (FWHM). 'The original PDSR method used the number of counts in the'same' ROI channels where the Ce-144 peak should have occurred.
In Revision 1 of this PDSR, the number of counts in the ROI channels r
corresponding to the FWHM were used. The net effect of these changes was to provide a 60% increase in the detector counts during efficiency determination and a 2% increase in the the number of ' measured events.
Revised Detector Efficiency Three parameters were adjusted.to provide a more accurate iretrinsic detector efficiency. The parameters were the standard Ce-144 source activity, the method of peak integration, and the thickness of the lead shield. The overall effect increased the intrinsic detector efficiency by a factor of 1.8.
Details can be found in Reference 8.
Revision of the Reactor Fuel Cagulation Three changes in the calculation of the gamma-ray fluence rate from the assumed letdown cooler sources to the detector were made. First, the letdown coolers' source region was found to have been incorrectly located and a mathematical error was found in the calculated amount of the reactor fuel.
Second, the original calculation was based on preliminary source term data that were the best available data at the time of the measurement. This revision uses the most current data. These changes offset the effect of an increase in the detector efficiency described above. Revision 1 of the PDSR corrects this.
Finally, a revision was made in the method of calculating the MDL value to account for all possible errors.
ii
m,.m
} jip -.g Y n
' Revision of Reactor Fuel Location 7
In the' original ~ calculation. reactor fuel was assumed to be located uniformly l
in the-bottom' ouadrant of the coolers.
This model is now believed to be too conservative. The current model assumes that the reactor fuel is uniformly located in the inlet pipes in the coolers and the initial' tube spiral' in the lower bottom ouadrant.
Letdown Flow Rate The original calculations reported in the original PDSR'showed the letdown flow rate to be 1.6 liters per second.
However, a detailed review of plant.;
logs showed that, in September 1986, the block orifice was removed, which resulted in an increase in the. flow rate to a maximum of 3.0 liters per second.
This, however, did not affect the analysis. Additional analysis has been perfonned to show the flow rates through various sections of piping and the letdown coolers. This also does. not change the results of the ' analysis.
~ Conclusion Tnese changes result in an increase in the estimate of record from an MDL value of < 3.2 kg to an MDL value of < 3.7 kg with an uncertainty of approximately + 537, (at one sigma).
iii f
1 3
SUMMARY
The estimate of record of the amount of uranium dioxide (UO ) reactor fuel j-2 remaining in the-letdown coolers-and piping in the letdown cooler room is
$ 3.7 kg with 'an uncertainty of approximately + 53% (at one sigma) based on J.
the chosen'model and associated minimum detectable level (MDL)* measurements and calculations.
This amount includes reactor' fuel in all forms.
It is not
.possible, at this time, to determine the foms of the reactor fuel (e.g.,-
films, loose debris) since the internals of the system are not accessible and l
'the coolers reside in a radiologically hostile environment.
A collimated lead shielded sodium iodide detector was inserted into a letdown cooler room access penetration to measure Ce-144, which was used as a tracer l
for U0. Calculations were made by use of computer codes to model the 2
associated piping, coolers, and detector configurations.
Since the presence of Ce-144 was not identified, an MDL calculation was performed.
remaining in the letdown cooler room is <0.4% of the The quantity of U02 maximum allowable residual UO inventory for the entire TMI-2 facility in p
Mode 2 (Reference 1).
MDL value is defined as the maximum ouantity that can be present and still not be observed due to interactions caused by background radiation. The true value can range from 0 to the stated MDL quantity.
r -. -
TMI-2 POST-DEFUELING SURVEY REPORT
.FOR THE LETDOWN COOLER ROOM
1.0 INTRODUCTION
remaining in the l
This report presents the analysis'of the amount of UO2 TMI-2 letdown cooler room. The boundaries of this analysis are the makeup.and purification (MU&P) system's two letdown coolers (MU-C-1 A and MU-C-1B) and a
.part of the letdown' piping (Figures 1 through 5 and References 2 through S).
}
. This report is one in a series of reports to be prepared to fulfill the requirements of the TMI-2 SNM Accountability Program (Reference 6).
All statistical uncertainties are expressed as + one sigma limits (defined as one standard deviation).
- Section 2, " Background", ~ describes the physical attributes of the %tdown coolers and associated piping and' describes their function. The relationship l
of the letdown cooler room to the accident and subsequent cleanup activities is discussed as well as its current status, l
1 Section 3, " Methods", describes how measurements were obtained through the use l.
of Ce-144 as a tracer for reactor fuel.
A calibrated system featuring a shielded sodium iodide scintillation detector connected to a multichannel analyrer was inserted into the letdown cooler reom to a known location. Since the presence of Ce-144 was not identifiable, an MDL calculation was performed.
Section 4, " Analysis", explains the methodology for arriving at the estimate L
of record of fuel in the letdown cooler room using three measurements and discusses supporting data, assumptions made, and calculations used.
Section 5, " Conclusion", presents the estimate of record and uncertainty for the amount of SNM remaining in the letdown cooler room, and states that this conclusion is reasonable based upon the available data.
l
?
2.0' BACKGROUND'
~
LThe letdown coole'rs are located in the Reactor Building-(RB) basement, as
~
4 shown in Figure 6.
The letdown coolers'are counterblow spirhl tube-in-shell heat exchangers which are cooled by the intermediate cooling system (ICS).
'The purpose of' the letdown coolers is to reduce the reactor coolant system (RCS) water temperature from 293 C to 49 C to prevent damage to the MU&P l.
demineralized resins. The normal RCS water flow rate through the letdown
' coolers is 2.8 to 4.4 liters per second. Figure 2 shows a cross section of a letdown cooler (Reference 7). Each cooler shell is carbon steel and is 1.09 m in diameter and 0.6 m long, with a 12.7 m shell thickness and an 83 m minimum end plate thickness.
The design. working pressure is 1.4 MPa. The internals consist'of stainless steel inlet and outlet distribution headers and
- thirty 19 m OD spiral tubes with a working pressure of 17.2 MPa. The letdown coolers are connected to the MU&P system by 76 m schedule 160 pipe which is reduced into common 63 m schedule 160 inlet and outlet headers. Figures 3 to 5 show the piping and coolers in the letdown cooler room.
The 76 mm inlet piping taps off the bottom of the '1 A' RCS cold leg (as shown in Figure 1),
goes down to elevation 282'-11", and rises up to enter the letdown cooler room horizontally from the inside of the RB ' A' D-ring at elevation 299'-6" above a service platfom. The inlet pipe divides, goes vertically downward through inlet isolation valves (MU-V-1 A and MU-V-1B) and expands into the 76 m pipe.
The pipe enters each cooler at approximately the centerline of the cooler.
The 76 m outlet piping is at the bottom of each cooler. The outlet pipes rise through reducers and the outlet isolation valves (MU-V-2A and MU-V-2B),
combining into a 63 m outlet pipe above the service platform at elevation 299'-6".
The pipe exits through the Reactor Building wall into the annulus l
and the Auxiliary / Fuel Handling Building ( AFHB).
The letdown coolers are located in the north quadrant of the RB at elevation 282'-6" (see Figure 6).
They are in a concrete shielded enclosure adjacent to the RB sump. The room is normally accessed through the RB sump room. Since the March 1979 accident, manned access to the room has been prohibited by high radiation exposure rates in the RB basement. Access by wheeled robots has not been possible because of piping, eouipment, and structural congestion. -___. __. -.
LThe RCS flow rate 'through the letdown.coalen during the TMI-2 accident, as
- contYolled by the plant operators, ranged fron, O to approximately 10 liters per second.
.During the post-accident recovery, letdown flow was established at approximately 1.6 liters per second and was. maintained at this level until the-E block orifice was removed in September of 1986. The letdown coolers are currently.in service as part of the letdown path from the RCS to maintain the reactor vessel water level. The current flow rate has been no greater than 3.011ters per.second, limitad by the hydraulic head on the letdown pipe, and is intermittent since it is being used for RCS level control. At '3.0 liters per second, the maximum flow velocity through the piping is about 1.3 meters per second- (Reference 8).
.The letdown coolers have been. examined for residual 00 because they are one-
'l 2
of the low points of the MUSP system.
Radiation Environment The radiation measured in the letdown cooler room is approximately 10 Rem per.
hour near the enclosure ceiling. Normal ingress-to the letdown cooler room required access to the Reactor Building basement. Absorption of soluble fission products into porous basement surfaces p.roduce present day general area exposure rates that are freauently >10 Rem per hour.
Due to the hostile environment, measurement access was limited to a spare penetration at the 298'-6" elevation that could be reached from the more accessible 305' elevation. The size of the penetration requires a shielded measurement system that was less than 25 cm in diameter. The 10 Rem per hour exposure rate required the use of at least 7.6 cm of lead around the sodium iodide detector. The measurements were completed in April 1986.
No measurable Ce-144 activity was observed which resulted in reporting the minimum detectable cuantity. 5
~,
f For the reasons discussed above, it is ALARA to accept the uncertainties associated with the present fuel assay.
Additionally, the value is a small part of the total ex-vessel fuel and is not considered easily transferred by.
accident or design.
- 3..
METHODS ~
Ce-144 was selected as a tracer for reactor fuel because of its long half-life (284 days), relative high abundance, low escape rate coefficient from reactor fuel, and its characteristic 2.19 MeV gamma-ray that is readily identifiable on a multichannel analyzer (MCA). Using sample data, a Ce-144-to reactor fuel (U0 )' ratio was developed (Reference 9). This value was g
5.6 x 10 Bq g'I on August 1,1987.
6 To detect the gansna-rays, shielded sodium iodide scintillation detectors were connected to a preamplifier,' amplifier, and an MCA. The sodium iodide detector crystals were 12.7 mm in diameter and ranged in length from 12.7 to 10.1 mm.
The detectors were shielded with either tungsten or lead to reduce l
detector dead time and pulse pileup due to the high exposure rate in the The detectors were, inserted through a spare penetration in the east-room.
wall of the letdown cooler room from the incore instrumentation chase. This penetration was' one of only three into the letdown cooler room from an accessible location.
The penetration is at elevation 298'-6", as shown in Figures 4-and 5.
The penetration positioned the detector at about 3.66 m above the coolers and about 0.3 m below the horizontal inlet and outlet pipes.
Prior to their insertion into the room, the sodium iodide detectors were calibrated with a standard Ce-144 source for the 2.19 MeV gamma-ray. The detector's efficiency was determined from the net count rate in the ROI on the MCA for the 2.19 MeV gamma-ray, the energy release rate from the standard
.Ce-144 source, and the source-to-detector distance. The detector was then j
inserted into the room to a known distance and spectral data were obtained.
l Due to the detector shielding and the distance from the coolers to the shield, long count times were necessary to obtain statistically significant data.
If an identifiable peak in the ROI at 2.19 MeV for Ce-144 was present, the l L E
~,b measured photon fluence rate at the detector was calculated from the net count
. rate and' the calculated detector efficiency.
If the Ce-144 ' peak was not visually or statistically identifiable, an MDL calculation was performed, as l
described below. -
1 To complete the calculation for an identifiable Ce-144 peak, radiation shielding computer codes were used to determine a calculated photon fluence l
rate at the detector from the coolers an'd piping per kilogram of reactor fuel. The codes modeled the piping, coolers, and platforms as closely as possible in three dimensions.
ISOSHLD-II.(Reference 10) was used to model the piping sources.- QAD-UE (Reference.11) was used to model the coolers because of its ability to model complex geometries and source distributions. The source input to the computer codes was based on the Ce-144 to reactor fuel ratio,'taking into account decay time.
Parametric studies were performed in the engineering calculations for this report (References 8 and 13) using l
various reactor fuel distributions in the coolers to determine-a range of values. From the range of values, an estimate of record was determined. To l
determine the quantity of reactor fuel in the coolers and piping, the measured photon fluence rate was divided by the calculated photon fluence rate per l
kilogram of reactor fuel in each source.
If the Ce-144 peak was not identifiable, an MDL calculation was performed (Reference 12). The gross counts in the ROI for Ce-144, converted to an MDL value, was divided by the calculated detector efficiency, the calculated l
photon fluence rate per kilogram of reactor fuel, and the count time.
l Alternative methods considered for this measurement were gamma spectroscopy by using a high purity germanium (HPGe) detector and neutron interrogation. The HPGe detector has a superior gamma-ray spectrum resolution compared to an Nel detector. The HPGe detector was not used, however, due to space limitations with available equipment.
Neutron interrogation techniques could not be used due to the low emission rate of neutrons from the reactor fuel and the high boron content ( > 4500 ppm) in the RCS water in the letdown coolers and MU&P system piping. _
!L 4.
ANALYSIS On October 25,1985 (approximately 6.5 years after the accident), a preliminary measurement was perfomed in the letdown cooler room. The sodium l-iodide detector crystal was 12.7 mm in diameter and 12.7 mm in lengtr. The detector was shielded with tungsten. The detector was not collimatec. and,
- therefore, measured radiation from the entire system piping and the 1.etdown coolers. The configuration of the detector and shield in the penetration is shown in Figure 7.
The intent of this measurement was to provide an initial l
assessment of the Quartity of reactor fuel in the room. The count time for this measurement was 103,300 seconds.
This preliminary measurement had essentially no spatial resolution as well as no observed peak corresponding to Ce-144. The MDL* results were equally consistent with having 0.02 kg of fuel assuming uniform distribution (at.
2.5 x 10-3
-I) in the inlet und outlet piping, or up to 64.3 kg g mm (Reference 13) in the letdown coolers.
These two configurations could produce l essentially the same radiation signal since the piping was very close (0.3 m) to the detector and constituted a relatively low shielding mass for attenuating the radiation, while the coolers were both more distant (3.66 m) and more massive. Since the spread of values was great and the larger values m
are a significant fraction of the then existing 70 kg admir.istrative limit (now 140 kg) for sub-critical mass (Reference 14), twc additional measurements were performed with the goal of better t.efining locations and cuantities of reactor fuel.
The later measurements replaced the unco 111 mated detector with a co111 mated detector. The tungsten shield was also replaced with lead since it was found that the tungsten was contaminated witn thorium. A 2.6 MeV gamma-ray is produced in the thorium decay chain which interfered with the acquisition of spectral data in the ROI of the characteristic gamma-ray of Ce-144. The effect of the thorium contamination was accounted for in the calculations of the Quantity of reactor fuel in the room...
~1 e
i
- q:
1
,y
(
r I
1The final measurement set was performed from April 16 tc 21,1986 k,
(approximately 7 years af ter the accident). The sodium iodide detector a
Q, crystal was 12.7 mm in diameter and 19.1 mm in length.
A co111 mated lead shielded detector assembly was. used, as shewn in' Figure 8.
The sodium iodide detector was' placed inside a lead cylinder with a hemispherical head. This 3
G :i 1
- assembly was placed irdide an aluminum cylinder lined with 19 ntn-thick lead.
Y A slot cut in the end of the cylinder prov'ied the desired co111mation. The det.ector/ shield asseinbty was designed so that' two different measurements could be made by moving the' thielded detector in the shielded tube.
In the first 8.
ti measurement, the detector was located in. the front'of the shield tube as shown iW Figure 8.-
in this position, ths detector-scanned the coolers and vertical inlet and outlet piping and v&lves, ss shown in Figure 9.
For the second measurement, the' detector. was'meved back slightly in the shield tube so' that the detector scanned the coolers and minimized scanning the, inlet and outlet 3
piping. The field of vied of the ' detector. is shown in Figures 10 and 11<
l s
' />
('
The' measurement times for the final set in the first and second ' measurement s
ppsitions 'were,155,499 -and 240,948 seconds,.respectively, resul ting in gross counts in the ROI for Ce-144 of 170 and 117, respectively. Trie statistical l
Quality of, the data was improved from the first measurement by using longer count times.: Uthough there were counts in the ROI for Ce-144, the results of the measurements showed that there was not a statistically identifia"'e gamma l
peak at.2.19 MeV (nergy location on the MCA.'
This resulted in an MDL calculation for the cuantity of reactor fuel. The sodica iodide detector
~I 2
efficiency was 1.22 10-2 counts gamma cm. The esiculated photon p
l fluence per kilograta of UO using the QAD-UE computer code ranged from 7.2 x 2
h
- 10~4 to 9.7 ' x 10-3
-2
-1 kg'I The estimate of record is
.gamaas cm 3
an MDL value of i 3.7 kg +_53% of 902 in the letdown cooler room. The 53%
. value is a l a error that accounts for all other errors in the calculation i
other than the counting error for the letdown coolers. This estimate of l
record assumes that the reactor fuel is uniformly distributed in the cooler inlet distribution header and in the bottom quadrant 'in the. first spirals of i
L the tubes.
Particles of reactor fuel would most liWy have settled here due to a decrease in the flow velocity from the inlet pipe into the tube bundle.
i-
- l. L lI L
The MDL. quantity of reactor fuel in the letdown coolers was calculated to range-from <l.9 kg to <25.6 kg, using improbable reactor fuel distributions with an uncertainty of approximately 153% (Reference 8). These values take l
ali sources of error into account (i.e., calibration and measurement counting, physical measurements, calibration, and standard source). This range in Quantity of reactor fuel issumed that all the reactor fuel in the room was in the coolers. The low value assumed that the fuel was in the inlet i
distribution header. The high value assumed that the reactor fuel was uniformly. distributed in the bottom of the cooler near the outlet distribution header.- The region where deposition was most likely was in the cooler distribution header and in the bottom quadrant of the first spirals of the tubes.
This region was used to establish the estimate of record of
< 3.7 kg 153%. The form of the reactor fuel is not known since this part of the MUaP system is not available for internal visual inspection or. sampling.
F The piping was calculated previously to have an MDL value of < 0.02 kg (Reference 13), which is insignificant when compared to the coolers.
The quantity of reactor fuel in the two letdown coolers is currently considered static based on analyses of the flow rate from the cold legs to and l through the coolers, particle size distribution of the reactor core debris, data from other plant locations, and the layout of the MU&P system.
The maximum flow velocity in the cold legs is 0.004 meters per second, and through the pipine; and coolers is approximately i.3 meters per second and 0.2 meters per second, respectively. Samples of the reactor core fue*f debri s show that a majority of tne perticles by weight (70%) are greater than 1000 pm (Re%rene 14). The density of the material is assumed to be approximately
[
5 g crf3 (Reference 14). Assuming that the particle size distribution and density has remained the saine from the time of the accident, a flow velocity j
of approximately 0.004 metsrs per second in the cold leg piping is insufficient to transport reactor fuel pt.rticles greater tnan 40 pm (Reference 8) to the MU&P piping and the coolers.
Particles less than 40 pm either remain in suspension or, if settled out on termination of flow through the cooler, become resuspended and are carried out of the cooler when flow is reestablished.
I i i
u
The' samples. from the reactor core debris are conservative, since the sample may have had washout that removed fine fuel debris that was less than 1 pm.
Fine reactor fuel debris of this size would be readily moved. The once-through steam generator (OTSG) 'B' upper tube sheet sample data showed that greater than 92% (by weight) of the particles were greater than 1000 um (Reference-15).
Sampling attempts in the RCS J-Legs showed very. fine debris.
l A 1 xm filter used as a collector was not able to remove debris observed as dark-colored. water flowing through' the sampling equipment. Visual inspections l
-of these areas also'showed fine powder-like reactor fuel debris that would be easily removed.
There are low points in the piping and the. letdown coolers. The MU&P piping to the letdown coolers taps off.the bottom of the RCS 1 A cold leg (Figure 1) and continues down to elevation 282'-11".
After a horizontal run of approximately 8 m, the letdown line. rises to elevation 299'-6".
The piping then drops to the letdown coolers where the center line elevation is 282'-11".
These low points in the piping and the letdown coolers would tend
-to prevent large ( >10,000 um) particles from getting to or leaving the letdown coolers at current flow rates. These particles could have been transported to the letdown coolers when the flow rates were as high as 10.1 liters per second inte the MU&P piping. The current maximum-flow rate of 3.0 liters per second would be insufficient to remove large particles.that were transported to the letdown coolers under the higher flow rate.
Although the latdown r.oolert have not been isolated, the quantity ct reet:tnr fuel the" contain, as stated in this reoort, is considered stable. For approximately tN last three years, the flow rate throxgh the letdown coolers I
has been,at approximately 2.8 liters per second wit 1 a maximum flow rate of 3.0 liters per second.
This continuous flow rate results in a stable equilibrium where reactor fuel will not be added or removeo from the letdown coolers. This flow rate through the cold leg and the MU&P piping is not conducive to moving additional large reactor fuel debris to or from the l
coolers. Small fuel debris particles will remain in suspension or be moved through the coolers.
After RCS system draindown, the letdown cooler system will be isolated and a confirmatory survey may be performed, if deemed appropriate, to ensure that the reported values have not significantly changed. l _ _ _ _ _ - - - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _
m p
l.,C/
.4
' The goal of the current defueling program is to remove more than 99% of the original core inventory of approximately 94,000 kg.
In that context, the
< 3.7 kg quantity of U0 remaining in the pressurizer is <0.4% of the 2
maximum-allowable residual UO inventory for the entire TMI-2 facility in 2
Mode 2 (Reference 1).
5.
CONCLUSION A MDL value of < 3.7 kg with an uncertainty of approximately + 53% (at one
- sigma) is the estimate of record of 00 remaining in the letdown coolers and 2
piping in the letdown cooler room. The piping is calculated to have an MDL value of ' 6 0.02 kg, which is insignificant compared to the MDL value of the letdown coolers. Therefore, the estimate of record for the letdown cooler room is considered the same as the letdown coolers, which is < 3.7 kg + 53%.
Although the' letdown coolers have not been isolated and reactor coolant still travels through them, the estimate of record of the amount of reactor fuel is considered valid. The flow rate is currently low and intermittent as opposed to the high and continuous flow rate during, and for a long period of time following, the accident.
The flow rate is no greater than 3.0 liters per second and is intermittent'since it is being used for RCS level control. The maximum flow velocity through the piping is approximately 1.3 meters per
~
second, which is sufficient to have long ago removed fine fuel debris from the letdown r.oolers, and to have prevented the buildup of fuel debris.
4 i
i u
y p - :r - ----
b-
[
REFERENCES.
l 1.
R. E. Rogan. Residual Fuel' Basis.
GPU Nuclear Internal Memorandum 4410-88-M-0306/4400-88-078. Middletown, PA: GPU Nuclear Corporation.
24 August 1988.
~2.
Burns and Roe Drawing 2024. Flow Diagram, Reactor Coolant', Makeup &
[
Purification.
3.
Burns and Roe Drawing 2168. Makeup & Purification & Intermediate l
Cooling-Reactor Bldg., Plan E1.- 282'-6".
4.
Burns and Roe Drawing 2413. Makeup & Purification, Intermediate Cool.ing &
l-Nuclear Services C.C. Wtr.-Reactor Bldg.-Sections.
5.
Burns and Roe Drawing 2066. General Arrangements, Reactor Building, Plan l-E1. 282'-6".
6.
J.S. Schork. TMI-2 Core and Special Nuclear Accountability Program.
j Administrative Procedure 4000-ADM-4420.02, Rev. 3-01. Middletown, PA:
GPU Nuclear Corporation. 10 October 1988.
7.
TM1 File Drawing 007-20-1, Crahem Manufacturing Co., Inc. Drawing l
30-540-4066. Weldseal Heldficw Letdown Coolar,.
8.
P.J. Babel. Calculation of Letdown Coolars Fuel Debeis. GPUW Calculation 4550-3211-86-056, Rev. 6.
Middletown, PA: GPU Nuclear Corporation. 2 August 1989.
7 9.
P. J. Babel. Jg1J4, Eu-154 and Eu-155 as Tracgs for Fuel Debris.
l TB-86-41, Rev. 2.
Middletown, PA: GPU Nuclear Corporation. 29 February 1988.
-;1 1 -
.i
\\
REFERENCES-(Continued) 10.
ISOSHLD-II: Code Revision to Include Calculation of Dose Rate from l
Shielded Bremsstrahlung Sources. BNWL-235 Supplement 2. April 1969.
- 11. 0AD-UE, Revised Numerical-Integration Option for Gamma-Ray Volume Source l
l Problems in the QAD-CG Point Kernal Shielding Code System. RSIC Computer p
Code Collection CCC-448, 12.
L. A. Currie. Analytical Chemistry. Vol. 40,1968, " Limits for I
Qualitative' Detection and Quantitative Detection-Application to Radiochemistry", p. 586-593 (Table. II).
13.
P. J. Babel. Calculation of Letdown Coolers Fuel Debris.
l
'4550-3211-85-008, Rev. O. Middletown, PA: GPU Nuclear Corporation.
4 November 1985.
- 14. - Criticality Safety Evaluation for Increasing the TMI-2 Safe Fuel Mass Limits.
Safety' Evaluation Report 4710-3220-89-1, Rev. O.
Middletown,
/A: GPU Nuclear Corporation. February 1989.
15.
D.W. Ackers, et al. TMI-2 Core Debris Grab Samples - Examination and
-l Analysi s.
GEND-INF-075. September 1986.
16.
T.L. Hardt and G.O. Hayner. TMI-2 B-trcp Steam Generator Tube Sheet j
}.oose Debris Examination and Analysis,,
October 1987. _ _ _ _ _ _ _ _ _
FIGURE 1 LETDOWN COOLERS AND REACTOR COOLANT SYSTEM HDT LEG PIPE i
)
Z 4
I
,J_
n i,p
%A
-,, i,,,, m rn,, i RC-P-1A i ~ RC-P-2A p -
um
~
w
=
\\
\\
,. e -
i rm y
i i
'm mM_' 3' u
9 psmg
,,,,g E5=2.
q l
7
-=
..gr,23
~
t
],:10
[
v
-C
', A >
h t;
t-ll
}
s l
l
/9 COLD LEC PIFE--
1 l
l
)--,i_
J-d J L
q
.,&l
- i ){[ ';
Ir 1
s
-W BOUkDARY
~.,
!5=Jg g7Mg
)
..,,1 1m 0F ANALYSIS-~
J
}i M
,,l y
3; -q
_ Y,-
u
.e
, 9 W O
h 4
[..- 5,4 hfh Mu-C-1B MU-t-1 A w
3 ONCE LLETDOWN THROUGH LINE STEAW.
GENERATOP
4
.c :
FIGURE 2 u.
LETDOWN COOLER ROOM, ISOMETRIC VIEW i
-:s%
%~MY"
/_
y g
c w vccccureren
,11,
,11....,
- h.._...._l 2h E ' i-- h i+N"u u m 1 u 1 11 11 1 1 1 m i s
'- -I-k -i-i L+K+= 12 z
uruu1 xuur
.. civ tr w 111..
.m
. _.. 1 w :
- 1. _1,.. g.1 1.gm..
,,,111,
,,,n.
. c c.
..m. g. c.a.
.m..-
XW'ffrXf^~t.M} O^>Xi=^CCOTC ^=^i r n e...
x-
__ _..___ m X C u:_.a. %-* _ T r : : r :~ f r -*:
u i 11_1 1 1 = 1 1}L
= = =_ = = >>i n _ _ - - _'
. '. '... '.'. ' '}_ '. e. L.b '. _* _m _.E =_ H_ * '. _E. '. ' '. '.'. = _ _ _ _ l I {cR uER RHRE YX EEI EI _ o Tu^m^m^ix^u'im^r x^iiN^iiiii ^r^i t. ... _ ;.. m.. ..z_=. . m m... c_ rictcccrecccccccnveccccccr vc,, m..u w_. .c. .c.. .........cc.a....m ln m_ _e n_m_ m_ .ri 5l g - l ' ~q wc,c_c_c_c_e_r_nn_n_v_ _e_ n_ n%_n ,a DE g i o, ,h +. y j 1 l [n. ,j ki Ihl j f.j, . \\ 0 i ! i i s L 1 M~d, h '4 dh M I J {;d,j,$ ,,l A g,',L3 y"a d Ry4 9 ' MVP SYSTEM - ~ ' M. ' ICS OUTLET ' 1htET tws_xusJ m o .s -.. s x 1 t r m d l. .s v.si.a. . _ _ _ _. = _ ' DOCCCCCr M..._. . = -. -. - _ - >_=. 1_v...v. u.. u_.x. v.. -_._. _ v._.v > 4 iO. r> r->X X X=X.r. O..Cu.XOtiv v)Xh,r ' =._ _ _ u _ _ x v _=1,. u.x. v := v _ v _= = = =_ @_ _ e u_m._,,v_. r ua. w.'^.^. ^^. ^.^M... A_^^U_ V. f:!M - ~ DO. _C.erWX.C0.C.. rX. ^.T.. O_.X_..X. M._ W.. ._.2, ._..-._..M. s, . 't E ,E E g ) i g l g y r --i-,. .w. ,, j bM I ICS INLET I J G s.t y% O N NJwyvosvv.6 h.J SYSTEM i L_.__J p.... OJ. <,L ET _ __y K' .s ..w .M 9 y g e m... L l s_;?.'.*,*.5 4 i 6, el A 5 r e
o FIGURE 3 LETDOWN COOLER ROOM, PLAN VIEW FIGURE 4 $'l // / ./ RB WALL // ,/ / OUTLET PI PE ,/, / / '/ /' / / ,,5 / / / ,/s / / i pf,W,/ ~ / ; e , s,k. ?/ rn/w/ sbh- [' J' /;// N ?, / l MU-C-1 B i { 1
--___7-FIGURE 4 LETDOWN COOLER ROOM PENETRATIONS, PLAN VIEW l. t SECTION A c i ESEsg GEREE ~~ l i INLET PIPE t -- SECTION B i j SECTION B = H ill O EEE l H H I5 1 ~ 745 ~ b [ ,. _t gg P y ENETRATIO i 5 zit' 3 ~ yC ~-~~-- INCORE E ~ INTRUtiENT = l CHASE l J 747
- 7 3 <g =--,
j ,l - { e.b ll- -)V- - A A MU-C-1 A r .- a y,N "'_h NT I t-s ~
- -~._g_
~ c. 'llll HVAC q, +... l o / i n/ sE: Irm A l I l L _ _ _. _ -.--. - - - - - - - - - - - - - - - - - - - - - - - - - - -------
o FIGURE 5 LETDOWN COOLER ROOMS, SECTION A-A PENETRATIONS PENETRATION 747 r746 E m fh - --- av ~; __ = il.i ._l n SERVICE PLATFOPM q l f g
- l HVAC q
A-q pF:.- ou -- ~ pp l =t[ i - I 1 s - n. l l 1 i OUTLET PIPE [tdl j 1 [" a e e- =- j E M N ~Q m.! j c.${; ~ Nh4 1l 9W 3IfBi IL j Mll-C-1 A MU-C-1B I L
- r.
FIGURE 6 REACTOR BUILDING BASEMENT, GENERAL ARRANGEMENT L [ 0 6[ s .e ____ __. x pc 3 o.' .r e _ _ 2, n gH 'J s.rv.Tm - g H iXi ' '.i 6L-s = ,,= z O m s - ca ~ '- O mg sz I \\.$'. } ' y... 5 \\*hk_w. sa > 'M \\ ,d ~EE,E fxi l '. CD - H y' s im tit:F$ q d g .ca ca / s h, e!r - / E [.] ' $*'/;1H' t w LETDOWN COOLERS
FIGURE 7 SODIUM IODIDE DETECTOR IN LETDOWN COOLER ROOM WALL ,'I. 66.7 mm 44.45 mm TUNGkiitr4 PVC PIPE \\, l, , SHIELD / CONCRETEWALL f / 6 7 NN $ 111 mm DIA. EL. 90.98 m N L \\ ".8 m m 0.2M m NAl 50 INCORE DETECTOR CONCRETEWALL INSTRUMENTATION gg CHASE PENETRATION NO. 747
f.,o FIGURE 8 S0DIUM IODIDE DETECTOR IN LETDOWN COOLER ROOM WALL LETDOWN COOLER ROOM - LEAD 160 mm SHIELD CONCRETEWALL [ M / / LEAD n s l 4 1 INNER ~ i SHIELD ) 162 mm DIA. 0.254 m--0.219 m EL. 90.98 m Y l a. er 0 1.3716 m e-- I NAl DETECTOR CONCRETE WALL 38.1 mm INCORE' DETECTOR 15' INSTRUMENTATION FIELD OF CHASE VIEW PENETRATION NO. 746 i2 7 Esiuminum shiele 19 I less shtold o . mm OffE top r% A. _69.25 rn g y N N.que nticcioa s ,\\ 105 rnm 6Y g ' c(ig: top \\- t27mmD:Ak 15(enn @ 4 \\Q3 j _;. gsh;jd j ) t, Ax
- s 'J i
is-
r :-- u-- 1 ~ ~ FIGURE 9 LETDOW!i COOLER R00t4 - FIELD OF YIEW, SECT 10t4 B-B i i E - g.
- x j 77
- !f# EXATFORM Xy _ E Rv l i o h 1 {
- (
I s i, FIELD of VIEW M7 m = FIRST MEASUREMENT l y' ^ ' = tocmou j g \\ l L 1 1 n yy A r-1 L. [ b_ ! ..L._. = W+- T k CI,,M 'W j a 0 t il. drs N ! \\ l 1 i : nu-c is-d.!
r-y- -- 6 FlGURE 10 LETDOWN COOLER ROOM FIELD OF VIEW, SEC'10N A-A 1 i l 5 m .,,..q \\p i /T ir'i \\/ ifT fj w . p - = u ,w3 ,q 4 \\ SERVICE PLAT,0RM f i N / \\ . Ap s . m=._-._ L i rr - - - - .~ INEET PIPE b. k' (-\\ ~ ' OUTEET PIPE l { R l / ! 1 I ~ i \\ i g 1 . x =v ~m -v LN D 9 $,$ b ,,EE..,,,E. $$s \\ q)dh %9 , %2m nu> ) l e !?d-C-1 A MU.c.18
i x FIGURE 11 LETDOWN COOLER ROOM FIELD OF VIEW, SECTION B-B e m T @ Y w e k__________,\\ l ,/ M h \\\\ ~ i U7 3ERVil ;iPLATFORM i --j j s y FIELD OF VIEW g E SECOND MEASUREMENT s E LOCATION l;,, ft i M 1 1 i j N 't \\_ d .L,._.! Risil x M c:L. h b h $ if\\ hb 9!!' \\ l 1 I w-i._i bSN ' Ms.C-1B 1.' ~ _. _ _ _ _ _ _ _ _ _ _. _ _}}