ML20087C850
| ML20087C850 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 03/06/1984 |
| From: | Fiedler P GENERAL PUBLIC UTILITIES CORP. |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20087C852 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM NUDOCS 8403130199 | |
| Download: ML20087C850 (35) | |
Text
,
2 GPU Nuclear Corporation Nuclear
- ';;r388 Forked fhver. New Jersey 08731-0388 609 971-4000 Writer's Direct Dial Number
March 6, 1984 Mr. Dennis M. Crutchfield Operating Reactors Branch #5 Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555 l
1
Dear Mr. Crutchfield:
Subject:
Oyster Creek Nuclear Generating Station Docket No, 50-219 Nureg-0737 Item II. B.3. Post Accirient Sampling System
'Ihis letter is in response to your correspondence of June 30, 1962. You will find attached, your 11 criteria relating to the Post Accident Sampling System and our responses.
In addition, drawings are provided where applicable. Should you require further information do not hesitate to contact me or Mr. Michael Laggart, Oyster Creek Licensing Manager at (609)971-4642.
l l
Very truly yours, r
e A
3 Peter B. Fiedler Vice President and Director Oyster Creek PBF:BH: dam Enc.
cc:
Dr. Thctras E. Murley, Administrator f
Region I
, U.S. Nuclear Regulatory Comrission 631 Park Avenue King of Prussia, PA 19406 NRC Resident Inspector Oyster Creek Nuclear Generating Station Forked River, NJ 08731 8403130199 840306 PDR ADOCK 05000219 fhkh GPU Nuclear Corporation is a subsidiary of the General Public utilities Corporation
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? Criterion:'-
(1)
'%e licensee shall have the capability to promptly obtain reactor coolant sanples and containment atmosphere sansples. he combined time allotted for sampling and analysis should be 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or less from the time a decision is made to take a sanple.
E Clarification::
Provide information on sanpling(s)-and analytical labora-tories locations including a discussion of relative
- elevations, distances and methods for sanple transport.
Responses to this item should also include a discussion lof sanple recirculation, sanple handling arx1 analytical times to demonstrate that the three-hour time limit will be met:(see (6) below relative to radiation exposure).
JAlso describe provisions for sanpling during loss of off-site power (i.e. designate an alternative backup power source, not necessarily.the vital (Class IE) bus, that can bs. energized in sufficient time to meet the
. three-hour sanpling and analysis time limit).
" Response:.1);
Jaboratory-Sanple System location
.7 Se Post Accident'Sanpling System.(PASS) at Oycter Crcck
~
Nuclear Power. Generating Station will be located ~in its Jr own segregated room adjacent to the Reactor Building.
- his room was the. radiochemistry counting room and.is shielded from the reactor with 4 inches'of permanent lead
- shielding ~ as well as the containment and reactor building walls.. his location will minimize potentially high
. radioactive piping runs outside containment, and due to the nearness of both the. hot chemistry laboratory and the projected counting room, sample transport distances will also be minimized.
%e' Hot Che.nistry Iaboratory, PASS and Radiochemistry Counting Room will be located on the second floor i :
- (elevation 35'. 0") at the north end of the Office Building.
(See Reference Drawing (1) We physical 6
dimensions of the rooms are as follows:
Chemistry Lab (Hot) 23' 10" x 15' 1"
. PASS Room 13' 6"
x 15' 1" Counting Room 16'- 2"
.x 15' 1" The nearness of.the laboratory and the counting room to the PASS assures easy and' quick sample transport.
a
-Dependent upon the disposition of the sample, quick access may be attainable from corridor #4, either to the
- inmediate right or left of the PASS room portal. The walking distances from the PASS to either the hot chemistry laboratory or the radiochemistrv counting room are 15 feet and 20 feet respectively. 1
i
[
In addition to the above, the laundry area, which is located directly across the corridor from the projected PASS room, will be relocated to provide additional laboratory space. This space shall be utilized for new instrumentation to assist plant chemistry in maintaining the stringent chemistry control of plant systems during normal operation. Post accident considerations went into the type of instrumentation which was purchased to provide the necessary timely data required by Reg. Guide 1.97 Rev. 2 for the initial estimate of the severity of an accident and to provide operators with useful information during the course of an accident.
- 2) Sample Recirculation
'Ihe wall mounted G.E. Post Accident Sampling System will be located on the east wall (facing the Reactor Building) of the PASS room (Reference Drawing #2). This system shall have the
. capability to sample the reactor coolant from either the "A" Reactor Coolant Recirculation Inop or the Liquid Poison process line, should the "A" Loop Recirculation Pump not be in operation. Liquid samples may also be attained from the Shutdown Cooling System and/or the Core Spray System ('Iurus) at the G.E. PASS; dependent upon the accident situation and plant conditions. The sample flushing times are listed.
Sample
- Time Reactor Coolant Sample (Liquid Poison).
438 sec.
Reactor Coolant Sanple (RB sample station) 42 sec.
Shutdown Cooling Sample 82 sec.
Core Spray Sanple ('1brus) 76 sec.
- 0ne line volume For non-accident situations (ie, training or sampling during
-normal operation) the liquid sample may be routed to the Reactor Building Equipment Drain Tank (RBEDT). Following an accident, all liquid samples will be routed to the torus
-through the Core Spray pump suction.
The liquid sanpling unit of the G.E. PASS is designed to obtain a 0.1 and a 10 m1 sample. The dissolved gas portion of a defined sample (70 ml) may also De obtained at the gas stripping section of the PASS. Each sample taken at the PASS will be lowered into a lead shielded transport cask for personnel protection during transfer. The liquid sample piping of the G.E. PASS is completely shielded from personnel with 6 inches of lead shielding and the gas sampler contains 2 inches. The station is designed to limit personnel exposure to 0.1 Rem whole body and 0.5 Rem for extremities, based on a 10 Curie / milliliter sample activity level.
The design of the PASS will include gas sampling from the following sources.
(1) Drywell atmosphere sample from either the Drywell Hydrogen or Oxygen Monitoring system.
(2) Wetwell atmosphere sample from the '1brus oxygen Analyzer.
'Ihe flushing times to purge 1 complete line volume is given
-below:
l' Drywell Atmosphere Sample (Hydrogen Monitor) 49 sec.
Drywell Atmosphere Sample (Oxygen Analyzer) 19 sec.
Torus Atmosphere Sample (Oxygen Analyzer) 20 sec.
All primary containment gaseous samples are routed back to the torus through the discharge of the containment particulate monitor. This system may also be flushed with bottled nitrogen to reduce the radiation levels created by sampling.
All gaseous samples will be retrieved from the PASS through the use of gas septum bottles which are inserted and removed with an extended handling tool to minimize personnel exposure. After removal from the PASS, the gas septum bottle will be placed into a shielded transport carrier for transfer to the counting room or instrument room.
- 3) Sanple Handling and Analytical Times Additional modifications to the PASS room will include the installation of a radiological fume hood, clean sink, and associated work benches to provide sample preparation space and dilution-work space. All sampling may be performed with special handling tools to maintain 3 feet distances between the operator and high radiation sources to aid in minimizing personnel exposure. Highly radioactive samples will be transferred from the GE sampler to the desired area for analysis or sample preparation with the use of lead shielded
- transport casks. Five - six inches of lead shall shield the sampler operator at all times when obtaining liquid samples.
Two inches of lead shall shield the sampler operator at all times when obtaining gas samples. A four wheel dolly cask positioner will be supplied to manipulate the sampling casks.
. 2
Note:
The particulate filters and iodine cartridges are control sampled. Se quantity of activity which is accumulated is controlled by flow orificing and time i
sequencing. 'Ibe amount of collected iodine is monitored during sampli:xj through the use of an installed radiation detector. These samples having limited activity will not require shielded sample carriers.
It has been concluded that the PASS system and its location
. discussed above shall satisfy the criterion established in NUREG 0737; that sample recirculation, handling and analysis as described below can be performed in the three-hour time limit imposed by NUREG 0737 (Reg. Guide-1.97 Rev. 2).
- 4) Provisions for sampling during loss of off-site power.
%e G.E. PASS at Oyster Creek will not be designed to sample during. loss-of-offsite power. The' Oyster Creek Plant has demonstrated its ability to recover from loss of offsite,wr within 15 minutes. This would still provice post accident saapling and analysis to be performed in the required three hour limit.
. t
3 Criterion: (2) he licensee shall establish an onsite radiological and chemical analysis capability to provide, within three-nour time frame established above, quantification of.'the g
following:
(a) certain radionuclides'in the reactor coolant and l containment atmosphere that may be indicators of the
. degree of core damage (e.g., noble gases; iodines and
+
cesiums, and nonvolatile isotopes);
'(b). hydrogen levels in the containment atmosphere;
+
dissolved gases (e.g., H ), chloride (time allotted
(
(c) 2 for analysis subject' to discussion below), and boron x.;.
. concentration of liquids.
(d). Alternatively, have inline monitoring capabilities to perform all or part of the above analyses.
- LClarification: 2'(a)
' A discussion of the counting equipment capabilities is needed, including provisions to handle samples and reduce background radiation (ALARA). Also a procedure is required for. relating radionuclide
. concentrations to core damage. Se procedure should include:
1.
Monitoring for short and long lived volatile and non volatile radionuclides such as 133 ei X
131,137 s,134 s, 85 r,140 a, and 1
C C
K B
88 r (See Vol. II, Part 2 pp. 524-527 of K
Rogovin Report for further information).
2.
" Provisions to estimate the extent of core damage based on radionuclide concentrations and taking into consideration other physical parameters such as core temperature dats and sample location.
2 (b)
Show a capability to obtain a grab sample, transport
- and' analyze for hydrogen.
- 2 (c)
Discuss the-capabilities to sample and analyze for the accident sample species listed here and in Regulatory Guide 1.97 Rev. 2.
2 (d)-
Provide a discussion of the reliability and maintenance-information to demonstrate that the selected on-line instrument is appropriate for this application.
(See (8) and (10) below relative to back-up grab sample capability and instrument range and accurancy).
e '
h --
v 3
Response: -1) 2a - Radionuclide Analyses Se Oyster Creek Plant utilizes a Series 85 Canberra (high LResolution Ganma-Ray Spectroscopy) 1411tichannel Analyser Nuclide Identification and Quantification System for the analysis of the isotopes listed in Clarification 2(a)1.
' 2 e required radionuclide analyses can be performed in 30
- minutes from the receipt of samples. Background radiation is somewhat reduced at the analysis station through the use of lead shielding. Were is 4 inches of lead shielding inside the
-Reactor Building, opposite the PASS Room which also shields a portion of the Counting Room. The germanium detector used by this counting system is shielded from the counting room environment with 6 inches of lead shielding.
Personnel Exposure
.Further dilutions of highly radioactive samples (0.1 ml) taken from the GE PASS will be necessary to reduce sample activity.
Se facilities provided in the PASS room, such as lead chielded sample transfer casks and a 4 inch tnick brick pile in the ventilation hood, will limit direct exposure to these sanples to a few seconds.
. Core Damage Procedure he Oyster Creek Plant plans to utilize the methodology developed by the BWR Owners Group.(BWROG) to develop a procedure for estimating the extent of core damage under severe accident' conditions. The generic procedures for estimating core damage using the PASS, developed by the BWROG, will be made site specific and submitted to the NRC. Rese generic procedures were reported by the NRC to be acceptible at a BWROG/ PASS committee meeting held on July 20, 1983.
- 2) 2b - Hydrogen Analysis ne reactor containment (drywell) may be sampled with the G.E.
PASS. The drywell sample may be routed to the PASS via the drywell oxygen analyzer or the drywell hydrogen analyzer piping. A small volume sample may be obtained from the PASS in' a septum-type sample bottle. A positive displacement vacuum pump within the G.E. PASS shall draw the drywell sample through -.
_-~
1 tha sampler and return it to the torus atmosphere. Sample flow E
may be verified with the inline flow sensing device provided g
with the G.E. PASS.
E 2 e gas septum bottle may be inserted and removed from the y
shielded gas sampler with an extended gas bottle handling
~
tool. W is provides separation of about 3 feet between sanpling personnel and the sample while tre.nsferring the sample r
from the PASS gas sampler to the lead shielded gas sample transport carrier.
L Hydrogen analysis may be performed Utilizing a Gas Chromatograph.
3) 2c - Accident Sample Species Reg. Guide 1.97 Rev. 2 I
F R e G.E. PASS will have the capability to provide the required samples at the onset of an accident. The Primary Coolant and Sump (Torus) can be sampled from the PASS and analyzed for the
.following; method of analysis is also listed.
Gama Spectrum
- Canberra Series 85 High Resolution (Geli)
. Boron Content
- Carminic Acid Method Chloride Content
- Gloride Screening Method *
=
T Tbtal Gas
- GE PASS Gas Expansion Methcxl PH
- Electrometric y
- Screening methods may include ion diromatography, silver nitrate or equivalent 2e Containment Air (Drywell and Ibrus Atmosphere) may be 7
sampled at the PASSc te following is a list of requirad analyses to be performed on these samples and methods to be utilized at Oyster Creek.
Gama Spectrum - Canberra Series 85 High Resolution (Geli) g i
Hydrogen Content
)
Oxygen Content
) -Ga' mromatography
,E E
All the above analyses shall be performed by trained, qualified g
technicians using station approved procedures.
k Ee close proximity of the Counting Room and the Laboratory to
[
the PASS minimizes the samples transport time to approximately M
3
..x _
'u W^Z
'E
f 1
n 2 minutes per sample. mis, in conjunction with having two or more technicians to sample and analyze fcr the species required in Reg. Guide 1.97, will insure the results are obtained within the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> time limit from initial notification to sample.
4) 2d - Instrumentation ne only in-line instrumentation in the G.E. PASS at the Oyster Creek Plant will be a conductivity monitor in the liquid sample stream. 21s instrument will continually readout on the PASS Control Panel which will be located on the south wall of the PASS room approximately 5-6 feet from the G.E. Sample Station.
n is instrument will be used as a back up to grab sample analysis for verification of water purity in the RCS and p'
torus.
In addition, the initial mixing of the liquid poison f
with the RCS may be trended with this monitor during activation of this system.
The Post Accident Sample Station is equipped with a 0.1 cm -1 conductivity cell. 'Ihe conductivity meter has a linear scale with a six position range selector switch to give conductivity ranges of 0-3, 0-10 0-30, 0-100, 0-300, and 0-1000 us/cm when using the 0.1 cm-1 c, ell. mis conductivity measurement system can be used to determine the primary coolant or suppression pool conductivity. During normal operation the BWR technical specifications requir? maintaining the primary coolant below 10.0 J2S/cm. Conductivity measurements are the primary method of coolant chemical control.
Conductivity measurements are, of course, non-specific, but they serve the important function of indicating changes in chemical concentrations and conditions. Perhaps even more importantly, in the case of the BWR primary coolant, the conductivity measurements can establish upper limits of possible chemical concentrations and cc.n eliminate the need for additional analyses.
Conductivity of Borax: Boric Acid Solution (1.028 ratio)
Generial Electric Specification D50YP1, Rev. 1, calls for the Standby Liquid Control System (SLCS) to be filled with a solution of 1.028 Borax (Na2 4 7 10 H O) to Boric Acid B0 2
(H B0 ) ratio. During normal operation the BWR primary 3 3 coolant has an operating limit of 1 u3/cm.
In the event the SIES were actuated, it is possible to estimate the primary coolant boron concentration from the conductivity, provided the
-primary coolant was within specification at the time of injection and, furtner, provided that higher conductivity water has not also been injected from some other source or extensive fuel damage has not occurred.
i
~l*
A conductivity curve was run on 1.028 Borax: Boric Acid solution in the range of 5.4 to 201 ppn Boron. Se curve was linear between 10.8 and 201 ppm Boron. me linear equation is:
Ppn Boron = 2.345*(uS/cm) + 6.56 (EQ l.0)
Plans are to run this curve to higher concentraticns.
In the meantime, the curve may be extrapolated to higher concentrations with a high degree of confidence as the index of fit for equation 1.0 was 0.99965.
W e primary method for the determination of boron, however, shall be by analysis of a grab sample.
Criterion:
(3) Reactor coolant and containment atmosphere sampling during post accident conditions shall not require an isolated auxiliary system (e.g., the letdown system, reactor water cleanup system (RWCUS)) to be placed in operation in order to use the sampling system.
Clarification:
System schematics and discussions should clearly demonstrate that pcat accident sampling, including recirculation, from each sample source is possible without use of an isolated auxiliary system. It should be verified 1
that valves wnich are not accessible after an accident are I
environmentally qualified for the conditions in which they must operate.
Response: (3) he operation of isolated auxiliary systems is not required to facilitate post-accident samplings. Reactor coolant camples are taken from the recirculation loop and the liquid poison system. S e motive force for the samples is reactor coolant pressure. S e samples are circulated at a rate of 1 GPM. Shutdown cooling and core spray samples are obtained during system operation through system samples valves. (bntainment gas samples are obtained from H2 and 02 analyzer sample lines wnich will be in operation following an accident. All samples are returned to the torus through core spray system return lines. Se various flow paths are indicated on P&ID M0012.
(Reference drawing
- 3)
Valves that are required to operate to facilitate sampling which will not be accessible after an accident have been reviewed to determine their suitability for the expected sampling service conditions.
Containment isolation valves V-40-8, V-40-24, and V-40-12 were purchased to Nuclear Claes 2 and Seismic I to assure operability during and after an accident. Se electrical components for these valves were purchased to meet the environmental requirements of IEEE Standard 323(1974).
Valves V-40-2, V-40-14, V-40-16, V-40-20, and V-40-26 purchased to ANSI B31.1.
S e solenoid actuated coils are class H.
Cbmponents located on the sample station piping rack were reviewed for material suitability for the expected service. S e 2 flon gasket in FET-664 and the Teflon gasket and packing in V-40-75 were identified as marginal for the design basis conditions. Rese materials will be replaced in the near future.
ke E
Criterion:.(4)~
Pressurized reactor coolant samples are not required if the licensee can quantify the amount of dissolved gases with unpressurized reactor coolant samples. 'Ibe measurements of.
either total dissolved gases or H2 gas in reactor coolant samples is considered W9te. Measuring the 02 concent. ration is reconnended, but is not mandatory.
Clarification:
. Discuss the method whereby total dissolved gas or hyurogen and oxygen can be measured and related to reactor coolant
,. system concentrations. Additionally, if chlorides exceed 0.15 ppni verification that dissolved oxygen is less than 0.1 ppa is necessary. Verification that dissolved oxygen is _0.1 ppm by measurement of a dissolved hydrogen residual of.
10 cc/kg is acceptable for up to 30 days after the accTdent. Within 30 days, consistent with ALARA, direct
' monitoring for dissolved oxygen is reconnended.
l Response:'(4)-
Resolution of this item is pending NRC approval of the BWROG's proposal to modify dissolved gas monitoring requirements for BWRs.
(Letter dated January 18, 1984, from G.G. Sherwood, General Electric, to D.G. Eisenhut, USNRC). We will' update this response upon notification of resolution.
f..
4 '
Criterion:
(5)
The time for a chloride analysis to be performed is dependent upon two factors:
(a) if the plant's coolant water is seawater or brackish water and (b) if there is only a single barrier between primary containment systems and the cooling water. Under both of the above conditions the licensee shall provide for a chloride analysis within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the sample being taken. For all other cases, the licensee shall provide for the analysis to be completed within 4 days. The chloride analysis does not have to be done onsite.
Clarification:
BWR's on sea or brackish water sites, and plants which use sea or brackish water in essential heat exchangers (e.g.,
shutdown cooling) that inve only single barrier protection between the reactor coolant are required to analyze chloride within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. All other plants have 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> to perform a chloride analysis. Samples diluted by up to a factor of one thousand are acceptable as initial scoping analysis for chloride, provided (1) the results are reported as ppn C1 (the licensee should establish this value; the number in the blank should be no greater than 10.0 ppm C1) in the reactor coolant system and (2) that dissolved oxygen can be verified at _ 0.1 ppm, consistent with the guidelines above in clarification no.
4.
Additionally, if chloride analysis is performed on a diluted sample, an undiluted sample need also be taken ana retained for analysis within 30 days, consistent with ALARA.
Response: (5)
In the event of an accident involving extensive core damage, chloride analysis at the site will be limited to a scoping analysis of a 0.1 m1 to 10 ml diluted sample. The G.E. PASS manual has suggested a turbidimetric method. The maximum sensitivity for this method is on the order of 0.1 ppm in the solution to be analyzed. Since the analysis will be performed on a sample diluted by a factor of 100, the initial concentration must be greater than 10 ppm.
More realistically, considering chloride blank and contamination uncertainties, the sensivitity limli. will be closer to the maximum specified concentration of 20 ppn.
For a more accurate analysis, undiluted 10 m1 samples of the primary coolant can be sent to an offsite analytical facility. Babcock and Wilcox, Lynchburg, VA; is under contract to GPUN and other BWR Owners Group members to provide post accident analytical services.
For accidents involving lesser degrees of core damage, other analytical methods will be available at the site. t
-Criterion: -
_ 6) n e design basis for plant equipment for reactor coolant
(
and containment atmosphere sampling and analysis must assume that it is possible to obtain and analyze a sample without radiation exposures to any indivdual exceeding the 3-criteria of 'DGC 19 (Appendix A,10 CPR Part 50) (i.e., 5 rem whole body, 75 rem extremities)..(Note that the design and operational review criterion was changed from the operational limits of 10 CFR Part 20 (NUREG-0578) to the GDC 19 criterion (October 30, 1979 letter from H. R. Denton to all licensees.
~ Clarification:.
. Consistent with Regulatory Guide 1.3 or 1.4 source terms,
- provide information on the predicted man rem exposures based on person-inotion for saiqpling, transport and analysis of all required parameters.
Response: 1(6)
%e sampling and analysis provisions at Oyster Creek have been designed such that it will be possible to obtain and analyze a sample at any time without exceeding the criteria of GDC 19. A detailed time motion study has not been conducted at this titre. A time motion study will be performed after the PASS system has been installed and operating procedures are established. We preliminary time sequence includes the time required to obtain and transport a sample to the sample analysis facility. We dose
' contributions to personnel during sample analysis have not.
been calculated at-this time. The analysis procedures to be used hava not been fully determined. When the PASS station is in place and operational, a procedure covering the sample analysis will be written, then a complete dose Lassessment can be performed. The preliminary conservative time sequences have been used in conjunction with 7
~
conservative dose calculations Reg. Guide 1.3 source terms to verify compliance, for a sample.taken at I hour after an accident. Although a detailed dose rate has not been j.
' determined, it can be seen that the expected exposure will be below the' criteria of GDC 19.
I E
o r
Whole Body Dose Assessment for Sampling Operations
Background
Sample Integrated Time Dose Dose Dose (min.)
(mr/hr)
(mr/hr)
(rem)
Liquid Sanple
.1) Recirculate Sanple 37 72.6 34.2 0.065
- 2) Operate Station 5
72.6 118 0.016
- 3) Handle Sanple 10 sec.
40 137 0.001
- 4) Transport Sanple 2
112.6 6.1 0.004 0.086 Gas Sanple
- 1) Recirculate Sanple 20 72.6 17.7 0.030
- 2) Operate Station 5-72.6 67.2 0.012
- 3) Handle Bottle 1
72.6 25.2 0.002
- 4) Transport Sanple 2
112.6 54.0 0.005 0.049 Particulate / Iodine
- 1) Recirculate Sample 20 72.6 17.7 0.030
- 2) Operate Station 5
72.6 67.2 0.012
- 3) Transport Sample 2
112.6 890.0 0.033 0.075 (1) The Hot Chem Lab for analyzing the samples is situated adjacent to the PACS room.
(2) Sane additional dose to extremities will result from limited handling of sanples in the operations.. Because of the use of samples dilutions, small sanple volumes, shielded casks, lead brick piles and laboratory extension
' devices (e.g., tores), doses to extremities are estimated to be small, at about 300 mr for each set of sampling operations.,
Criterion: (7).
- 2 e analysis of primary coolant samples for boron is
' required for PWHs.
(Note that Rev. 2 of Regulatory Guide 1.97 specifies the need for primary coolant boron analysis capability at BWR plants).
Clarification:
PWR's need to perform boron analysis. We guidelines for BWR's are to have the capability to perform boron analysis but they do not have to do so unless boron was injected.
Raspcase: 1). Boron: Inw Activity
~Large volum samples (10 ml) obtained from the G.E. sample can be analyzed for boron using the sodium pentaborate titration procedure. The boron in the solution shall be first converted to an acid complex through the additic,n of manitol after acidification with sulfuric acid and neutra-lization to pH 7.0 with sodium hydroxide. Se boric acid is then titrated to pH 7.0 with standard sodium hydroxide.
.2)~ Boron: High Activity me titration procedure can only be performed on low activity samples as
= the anelysis'of high activity samples obviously would result in high
. personnel exposure. In the case of high activity samples, a small volume sanple (0.l~ml) from the GE sample system can be analyzed for boron using the Carminic Acid colorimetric procedure. his method quantitatively measures boron by relating the boron concentration to the color intensity produced by the addition of carminic acid - sulfuric acid reagent to the sample.
E None of the chemical species listed in the NRC Criterion (1_0,) matrix are expected to interfere with the Hach carminic acid procedure.
In addition, the Vendor's Technical Manual states that "there are no known
~
interferences in the method". ASTM Standard D3082-79 specifically states that " ammonium, molybdenum, cerium, germanium, chloride, sulfate, magnesium, sodium, and potassium do not interfere". The ASTM Standard does indicate that silica and phosphate may interfere when present in
_ concentrations greater than 25 ppm while fluoride causes negative inter-ference due to the formation of fluoroborate ion. mese conditions are not
. expected to occur, however, in the BWR primary system.
Irradiation tiests conducted by General Electric have shown that an energy absorption rate of 4.4 x 105 Rad /hr during the color develognent phase y'
of the analysis.resulted in an error equivalent to 27 ugm of boron or 270
. ppm for a.0.1 ml coolant sample. Assuming the indication effect is proportional to dose, reducing the exposure to correspond to Reg. Guide
-1.3 source terms and a one hour decay would result in only a 1 ppm error
-for the. analysis of 2 ml of a 100:1 diluted, 0.1 ml primary coolant sample.
M,fi i criterion: -(8)
.If inline monitoring iu used for any sepling and.
analytical capability specified herein, the licensee shall provide backup sanpling through grab sanples, and shall dehenstrate the capability of analyzing the samplies.
Established planning for analysis at offsite facilities is
-acceptable.
Bluipment provided for backup sanpling shall be capable of providing at least one sanple per day for 7 days following onset of the accident, and at least one sanple per week until the accident condition no longer exists.
Clarification:
A capability to obtain both diluted and undiluted backup sanples is required. Provisions to flush inlint monitors to facilitiate access for repair is desirable.
If an off-site. laboratory is to be relied on for the backup analysis, an explanation of the capability to ship and obtain analysis for one sample per week thereafter until accident condition no longer exists should be provided.
Response:'(8)
In line monitoring equipment is not utilized at Oyster y
Creek. Grab samples are obtained from the GE designed and constructed sample station and analyzed in the one-site Hot-01em Lab facility. Arrangements have been made through the BWR owner's group with Babcock and Wilcox Cb.
for backup and supplementary analyses.
-Oyster Creek has arranged entrance into the Pool Inventory Management System which would provide them access to a shipping cask for the PASS samples.
9 L
a j_,gg Criterion:
(9)
We licensee's radiological and chemical sample analysis capability shall include provisions-to:
(a) Identify and quantify the isotopes of the nuclide categories discussed above to levels corresponding to r
the. source terms given in Regulatory Guide 1.3 or 1.4 and 1.7.. Where necessary and practicable, the ability to dilute samples to provide capability for measurement and reduction of personnel exposure should be provided.- Sensitivity of onsite liquid sample analysis capability should be such as to permit measurement of nuclide concentration in the range from approximately 1 uci/g to 10 Ci/g.
t (b) Restrict background levels of radiation in the
-radiological and chemical. analysis facility from sources such that the sample analysis will provide results with an acceptably small error (approximately a factor of 2). Wis can be accomplished through the use of' sufficient shielding around samples and outside sources, and by the use of a ventilation system design which will control the presence of airborne radioactivity.
Clarificacion: (a) Provide a discussion of the predicted activity in the samples to be taken and the methods of handling / dilution that will be employed to reduce the activity sufficiently to perform the required analysis. Discuss the range of radionclide concentration which can be analyzed for, including an assessment of, the amount of overlap between post. accident and normal sampling capabilities.
(b) State the predicted background radiation levels in the counting room, including tue contribution from samples which are present. Also provide data demonstrating what the background radiation levels and radiation effect will be on a sample being counted to assure an accuracy within a factor of 2.
Response: 1) 9a - Predicted Post Accident Sample Activity
%e reactor ficaion products inventory for the Oyster Creek Plant was
-developed from Reference 1.
We following list provides the predicted activity levels of post accident samples taken at various times after reactor shutdown. Se source terms are for a Ioss of Cbolant Accident
'(IDCA)' assuming a fuel release of fission product activity as defined by NUREG 0578 with a'one hour decay.
(See Table 1 - 7 of Appendix A) y, Conc.
R/hr/ gram (doserate)
Sanple C1/g
@l cm
@l0cm
@l00 cm 3 ft RCS 2.17 1.05E4 105E2 1.05 Torus
- 0.18 8.55E2 8.55 8.55E-2 Conc.
R/hr/cc (doserate)
Sanple Ci/cc 1 cm 10 cm 100 cm 3 ft Torus & Drywell 0.036 146 1.46 1.46E-2 Note: The CINDER code which was used to calculate the core inventory of fission products assumed a three year irradiation, 100% availability, and a reactor
~
operation at 102% of rated power.
- The torus sample assumes complete mixing of the torus liquid with the RCS volume (operation of core spray).
Reactor coolant samples obtained from the PASS during accident conditions can be diluted while sampling (1:100) while maintaining shielding of 5-6 inches of lead or approximately 2-3 feet distance from the sample. m is initial sample taken one hour after reactor shutdown could have a maximum exposure rate of 1,040 R/hr at I cm and more accurately 10.4 R/h at 10 cm.
For severe accident conditions which could release core fission products as defined by NUREG 0573, a calibrated syringe would be utilized to serially dilute the original sample for gamma spectroscopy and gross activity analyses. A dilution factor of 1 x 105 would be recuired for samples at the expected reactor coolant activity level of 2 Ci/gm; a 1 x 106 dilution factor would be required for RCS activity level 10 Ci/gm suggested by Reg. Guide 1.97.
Direct counting of the initial (1:100) dilution sample would allow analysis at coolant activity levels down to 1 uCi/cc. In addition, the degassed, undiluted 10 m1 sample available from the sample station could be utilized for analysis for samples in the 10 2 to 10 3 aci/cc range. mus, useful samples may be obtained from the J
i post-accident sampling station for coolant activity levels ranging from design basis accident source terms to well below the maximum level that can be tolerated at the normal reactor sample station.
- 2) -Reference 1.
NEDC-24889, General Electric Co., - Post Accident Sampling Compilation of Technical Information, March, 1981
Response
(1)(96)
Background Invels of Radiation Additional study will be required for this task. -.
e..
Criterion: (10)
Accuracy, range, and sensitivity shall be adequate to provide pertinent data to the operator in order to describe radiological and chemical status of the reactor coolant systems.
Clarification:
'Ihe recomended ranges for the required accident sample analyses are given in Regulatory Guide 1.97, Rev. 2.
The necessary accuracy within the recomended ranges are as follows:-
Gross activity, ganna spectrum: measured to estimate core damage, this analyses should be accurate within a factor of two across the entire range.
Boron: measure to verify shutdown margin.
~
In general this analysis should be accurate within 1 5% of the measured value (i.e. at 6,000 ppm B the tolerance is 1 300 ppm while at 1,000 ppm B the tolerance is 150 ppm). For concentration below 1,000 ppm the
-tolerance band should remain at i 50 ppm.
Chloride: measured to determine coolant corrosion potential.
For concentrations between 0.5 and 20.0 ppm chloride the analysis should be accurate within + 10% of the measured value. At concentrations belov
'O.5 ppm the tolera de band remains at 1 0.05 ppm.
Hydrogen or 'Ibtal Gas: monitored to estimate core degradation and corrosion potential of the coolant.
An accuracy of i 10% is desirable between 50 and 2000 cc/kg but i 20%
can be acceptable. For. concentration below 50 cc/kg the tolerance remains at i 5.0 cc/kg.
Oxygen: monitored to assess coolant corrosion potential.
For concentrations between 0.5 and 20.0 ppm oxygen the analysis should be accurate within + 10% of the measured value. At concentrations below 0.5 ppm the tolerade band remains at 10.05 ppm.
pH: measured to assess coolant corrosion potential.
Between a pH of 5 to 9, the reading should be accurate within 1 0.3 pH.
units..For all other. ranges 1 0.5 pH units is acceptable. _ - _ _ _ _ - _ _ _ _
~
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~
'Io demonstrate that the selected procedures and instrumen-
-tation will achieve the above listed accuracies,-it is necessary to provide information demonstrating their applicability in the post accident water chemistry and radiation environment. 'Ihis can be acconplished by performing tests utilizing the standard test matrix provided below or by providing evidence that the selected procedure or instrument has been used successfully in a similar environment.
STANDARD TEST MATRIX FOR UNDILUTED REACIOR COOLANT SAMPLES IN A POST-ACCIDENT ENVIRONMENT Nominal 4'
Concentration (ppm)
Added as (chemical salt)
Constituient 9
I-40 Potassium Iodide Cs+'
250 Cesium Nitrate Ba+2 10 Barium Nitrate La+3 5
Amonium cerium Nitrate Cl-10 B'
2000 Boric Acid Li+
2 Lithium Hydroxide No3-150 tel +
5-4
- K+
20 Gama Radiation 104 Rad /gm of Absorbed Dose Reactor Coolant NOIES:
1)
Instrumentation and procedures which are applicable to diluted samples only, should be tested with an equally diluted chemical test matrix.
- 'Ibe induced radiation environment should be adjusted commensurate with the weight of actual reactor coolant in the sample being tested.
~
2)l. For IHRs, procedures which may be affected by spray additive chemicals must be tested'in both the standard test matrix plus apprcpriate spray
-additives. Both procedures (with and without spray additives) are required to be available.
3)
Por BWRs,.if procedures are verified with boron in the test matrix, they do not have to be tested without boron.
4)
In lieu of conducting tests utilizing the standard test matrix for instruments and procedures, provide evidence that the selected instrument or procedure has been used successfully in a similar
~
environment.
.m r
' A111 equipment' and procedures which are used for post accident sampling and analyses should be calibrated or tested at a frequency which will ensure, to a high degree of reliability, that it will be available if required.
.# -Operators should receive. initial and refresher training in post accident sappling, analysis and tran port.
.A minimum frequency for the above efforts is considered to be every six months if indicated by testing.
These provisions should be submitted in revised Technical Sepcifications-in accordance with Enclosure 1 of NUREG-0737. ne staff will provide model Technical Specifications at a later date.
r
Response
(1) Gross Activity, Ganna Spectrum The worst case accident condition at Oyster Creek Plant would result in a reactor coolant sample activity concen-tration of 2.17 Curies / gram (Ci/g). Reactor coolant activities on the order of 1 to 3 curies per gnuns would 5
require dilution factors of 1 x 10 or larger. We expected counting (statistical) uncertainties are 10% or less for the required analyses which would be considerably less than the uncertainties involved with sample dilutions and representiveness of the initial sample. Truly soluble nuclides, however, should not add significant uncer-tainties to the analyses.
- Por severe accident conditions, the 0.1 milliliter of primary coolant diluted to 10.0 milliliter at the PASS would-be used for the initial sample. 2e sample would be retained and transferred to the sample dilution area in a shielded cask containing 3 inches of lead shielding.
Extended saaple handling tools enable personnel to
-maintain about 3 feet distance from samples to also minimize exposure during transport. ' A mirco syringe would be used to extract an aliquot from the sample septum
^
bottle for further dilution.
General Electric 00, has demonstrated that by permitting exposures up to 5 ~ REM, gamma spectral analysis can be performed at the Reg. Guide 1.97 upper limit of J0 Ci/g.
For lower activity reactor coolant, a large volume sample (10 ml) may be used for counting and obtaining the
- required gamma spectrum. ' Counting the degassed 10 m1 i
sample will provide the sensitivity of analysis at 10-2 to 10-3 uCi/gm level. S e diluted small volume sample
. (1:100 dilution) would also satisfy the Reg. Guide 1.97
. lower limit of analyzing at 10 uci/ml.
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2 Using High Resolution Gamma Spectroscopy, micro syringes and trained technicians, it is expected to have accuracies within 20 to 30% over the frill range of expected activities. Se G.E.-PASS to be installed at Oyster Creek will provide the useful samples at coolant activities ranging from the full accident source term to well below the maximum level that is required by NUREG 0737 and Reg.
Guide 1.97.
2)
Boron Boron concentrations will be determined by the Carminic Acid method with the 100:1 diluted sample (small volume PASS sapple 0.1 ml). 21s method is capable of-determining 50 to 1100 PPM boron with an expected accuracy of + 50 PPM. 50 to 1100 PPM Boron is the concentration of e
interest for BWR reactivity control in the event sufficient control rods are not inserted to shutdown the reactor.
No interferences have been identified using this method.
ASTM Standard D 3082-79 specifically states that " ammonia, molybdenium cerium, germanium, chloride, sulfate, calcium, magnesium, sodium, and potassium do not interfere.
Phosphates and silica concentrations above 25 ppm cause positive interference while flouride causes negative interference due to the formation of flouroborate. cus, none of the chemical species listed in the NRC criterion 10 matrix are expected to interfere with this procedure.
Irradiation tests conducted by GE have demonstrated that the design basis source term activity level causes only a 1 ppm error in the analysis. See Criterion (7).
Boron analysis may also be performed on high activity samples with the ion-chromatograph provided at the site.
Iow activity samples may be analyzed for boron content by the titration procedure; utilizing standard sodium hydroxide to titrai.e a 10 ml PASS sample in the presence of manitol.
(3) Chloride Initial chloride analysis for screening purposes can be performed on site using the ion-chromatographic (IC) and/or the turbidimetric method. The only known interference for the method is that boron in the sample results in a peak immediately adjacent to that of chloride. A high ratio of boron to chloride would significantly interfere. The boron may be removed by separation of strong and weak acid with a special resin column so that analysis for chloride can be performed on the strong acid phase of the sample _
Bi r
which will be free of boron (weak acid phase). This analysis is sensitive to 0.001 ppn and the initial screening sample (0.1 diluted to 10 ml) could provide a
'O.1 ppm sensitivity. De analysis provides an accuracy of i 104 over the entire range of the analysis 0.001 ppm to 20 ppm..%e turbidimetric method sensitivity is described
.in the response to criterion #5..
Oyster Creek' Plant is also a member of PIMS (Pool Inventory Management System) which is a group of utilities-
. ithin the BWHOG PASS connittee. Wis group has purchased w
the use of a shipping cask large enough to contain the G.E.' PASS liquid sample cask. Se shipping casks shall be stored at a designated site until their use is required.
.This shall provide means for.offsite shipment of a degassed 10 ml reactor coolant sample for chloride analysis should this be required. One vendor who can provide this analysis is BW at Iynchburg, Virginia, whom have the capability to perform chloride analysis within 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> of ~ sanple receipt. Weir procedure for chloride analysis will be accurate i 10% over the range of 0.5 to
. 20 ppm and 10.05 ppm below concentration of 0.5 ppm.
(4) Disadlved Hydrogen See response to criterion 4
-(5) Dissolved Oxygen See response to criterion 4 (6); pH_
A combination electrode will be utilized to measure the pH of-coolant sanples. Testing performed by GE has verified Lthat expected levels of irradiation result in a shift of
-less than 0.3 pH units.
'The semi-micro and flat. surface electrodes can be utilized
.to perform analysis on samples as small as 0.1 to 0.3 ml.
- % ey have also been successfully tested by GE.to perform
-satisfactorily at the rad levels stipulated in the Standard Test Matrix of Criterion 10. All of the pH electrode systems above have the operability range of pH 1
,1 to 13 as required by Reg. Guide 1.97.
-(7)' Equignent Calibration and Operator Training aquipment used for post-accident sampling and analyses will be ' calibrated or tested approximately every six months.. Personnel training.in the collection and on-site analysis of.;amples will be performed annually.
- r y
lQ Criterion:-(11)
- In the design of the post accident sampling and
- analysis capability, consideration should be given to the following items:
.c (a)- Provisions for purging sample lines, for reducing plateout in sample lines, for minimizing sanple loss or distortion, for preventing blockage of sample lines by loose material in the RCS or containment, for appropriate disposal of the aanples, and for flow restrictions to limit reactor coolant loss from a rupture of the sample line. Se post accident reactor coolant and containment atmosphere samples should be representative of the reactor coolant in the core
-area and the containment atmosphere following a transient or accident. We sample lines should be J
as short as possible.to minimize the volume of fluid to be'taken from containment. We residues of sample collection should be returned to containment or to a closed system.
(b) The ventilation exhaust from the sampling station should be filtered with charcoal absorbers and high-efficiency particulate air (HEPA) filters.
Clarification:.
A description of the provisions which address each of the items in clarification 11.a should be provided.
Such items, as heat. tracing and purge velocities, should be addressed. To demonstrate that samples are representative of core conditions a discussion of mixingj-both short and long term, is needed.
If a given sample location can be rendered inaccurate due to
~ the. accident (i.e. sampling from a hot or cold leg loop p
which may have a steam or gas pccket) describe the backup sampling capabilities or address the maximum time that this condition can exist.
BWR's'should specifically address samples which are W'.,
taken from the core shroud area and demonstrate how they are representative of-core conditions.
LResponse:
' 1)
Purging Sample Lines
- 2e sample system has been designed to allow purging of the liquid and gas lines to lay them up with
- noncontaminated fluid. 'Ibe liquid lines are flushed via the demineralized water valve V-40-39, through the
- cooler in the Traverse Incore Probe Room (TIP), through the liquid sanple panel, in the PASS Room and returned to the torus. We demineralized water flush is capable of flushing the entire liquid sample panel and the sanple return line.
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h e gaseous lines are. flushed-via the nitrogen valve V-40-79 through the sample panel in the PASS room and
-returned to the wetwell. The gaseous flush is capable of
. flushing the entire gas sanple panel and the sample return line to the Torus.
Se capability to flush'the sample panel reduces the radiation levels in the umple area.
- 2)
Reducing plate out and minimizing sample loss or distor-tion.
me containment atmosphere sample lines are provided with electric heat tracing to minimize condensation and plate-out in the sample lines. ANSI N13.1 was used as a guide in the design and layout of the gas sample lines in order
- to minimira plateout.. Distortion of liquid sanples through loss of noncondensible gases is minimized by main-taining the liquid samples under pressure in the sample lines. Distortion can be reduced by keeping sample lines
.as short as possible and minimizing the use of fittings.
i 3)
Flow restrictions he use of flow restrictors is not utilized. (bolant loss
.can be limited in the event of a sample line rupture by using small diameter sample lines and closing isolation valves, V-40-2, V-40-8 and V-40-12.
-4)'
Heat tracing Heat tracing is provided on the gas sample. lines to pre-vent moisture from condensing on the tube walls.
Condensation allows iodine to be absorbed by the water, and prevents obtaining a representative sample. The gas sanple lines are heat traced to 250*F. The heat tracing is controlled with a switch and monitored by digital readout.
5)
Sample Disposal The liquid snaple wastes return to the Reactor Building Equipnent Drain Tank (RBEDT) during normal plant opera-tion. Following an accident, all liquid sample wastes will be returned to the torus through the Cbre Spray pump suction, m is will confine the high radiation levels to the torus and not the radwaste system.
~ ll drywell and torus atmosphere samples will be returned A
to the torus through the discharge of the containment particulate monitor.
n e secondary contain,mnt (Reactor Building) atmosphere sanple will be taken and returned to the secondary con-tainnent.
6)
. Purge velocities
%e sanple lines have been designed for turbulent flow to allow good mixing of the samples. W e lines were sized to limit transport time to under 10 minutes and to limit the sample volumes needed to obtain representative samples.
7) ventilation An exhaust filtration system will be provided to exhaust from the GE sample station and the fume hood. W e exhaust
-from the sample station and the hood will be combined in a connon duct and directed to the filter unit enclosure building on the west low Turbine Building roof. The airflow will be routed through the building into the filtration unit whenever samples are being taken, or through a filtration unit bypass during normal operation.
.me filtrati ' unit will consist of the following:
(a) A prefilter of 50% minimum efficiency when rated in accordance with ASHRAE Standard 52.
'Q
('b) An electric heater to maintain airstream relative m
humidity below 70%.
(c)- An upstream HEPA filter with an efficiency of 99.97%
rated in accordance with IES CS-1 and ANSI N-510.
(d) An activated charcoal filter with an efficiency of 99% for elemental iodine and 95% for organic iodine rated in accordance with USDOE DP-1082 and ANSI N-510.
(e) A downstream HEPA filter with an efficiency of 99.97%.
me exhaust flow will be drawn through one or two 100% capacity exhaust fans, and will be discharged to the Turbine Building monitored exhaust stack. me fans will be controlled by a control damper to insure that the exhaust flow stays very close to the design flow of 1000 CFM. The filtration unit enclosure will be provided with an electric unit heater, and a ventilation fan, to maintain acceptable en7ironmental conditions in the enclosure.
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-APPENDIX A J
t
' TABLE 2' OMSTER CREEK CORE INVENIORY OF FISSION PRODUCTS (1933 Mwt)
.(Mega-Curies)*
(Reference 1)
Fission Product
- 1 Hour 8 Houra 24 Hours Decay Period
" Halogens; 411.7' 249.4 166.1 Noble Gases 210.7 166.4 133.4 Others.
-2,799 2,003 1,616 i
-
- Data from Table 1 of Appendix A TABLE 3 OYSTER OtEEK FISSION PRODUCT RELEASE (NUNEG 0578)
.(Mega-01 ries)
Fission' Product 1 Hour 8 Hours 24 Hours Decay 504 Halogens 205.9 124.7 83.0 )
1004 Noble Gases 210.7 166.4 133.4 )
RCS 14-Others 28.0 20.1 16.2 t
Mtal 444.5 311.1 232.6 L'~
~254' Halogens J103.0 62.4 41.5 )
Torus 100% Nobel Gases 210.7
-166.4 133.4 )
Drywell r
meal 313.6 228.8 174.9 Atmosphere h
r J";t TABLE 4 OYSTER CREEK.00RE INVENIDRY (IDCA ISOIOPIC RELEASE)
Volume--(Mass) 1 Hour 8 Hours 24 Hours Decay' Reactor Coolant Liquid 2.17 Ci/gm 1.52 Ci/gm 1.13 Ci/gm
.(2.056E8 grams)
-RCS & Torus Liquid 1.83 Ci/gm 0.94 Ci/gm 0.72 Ci/gm (2.428E8 grams)~
Torus & Drywell Atm.
- 0.036 Ci/cc 0.026 Ci/cc 0.020 Ci/cc T
(8.676E9 cc)
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TABLE 6 MeV/sec for 1933 mt (Ofster Creek 1*
1.0 Hr Decay NUREG 0578 Release PRIMARY COOLANP DRYWELL + 'IORUS ATMOS.
(MeV/sec/gm)
(MeV/sec/cc)
ENERGY GROUP HAWGENS NOBLE GASES ALL OTHERS HALOGENS NOBLE GASES (MeV)
(50% Rel.)
(__100% Rel.)
(1% Release)
(25% Rel.)
(100% Rel.)
0.000-0.100 9.67E6 5.86E8 1.59E7 1.14E5 1.38E7 0.100-0.400 1.83E9 1.68E9 1.97E8 2.16E7 3.97E7 0.400-0.900 3.10E10 2.74E9 1.12E9 3.66E8 6.48E7 0.900-1.400 1.30E10 7.75E8 4.21E8 1.53E8 1.83E7 1.400-1.800 5.75E9 1.70E9 4.49E8 6.80E7 4.GlE7 1.800-2.200 1.75E9 3.13E9 1.23E8 2.07E7 7.40E7 2.200-2.600 4.44E8 6.30E9 1.27E8 5.24E6 1.49E8 2.600-3.000 3.37E7 5.76E7 4.93E7 3.97E5 1.36E6 3.000-4.000 1.48E8 4.16E7 3.22E7 1.75E6 9.8035 4.000-6.000 3.29E6 0
6.39E5 3.90E4 0
Total 5.40E10 1.70E10 2.53E9 6.39E8 4.02E0
- Data from Table 5 of Appendix A TABLE 7 Dose Rate Per Unit Volume (Mass)
R/h @ 1.0 CM For Point Source Bauivalent Activity PRIMARY COOLANT DRYWELL + 'IORUS ATMOS.
ENERGY GROUP HAWGENS
. NOBLE GASES ALL UIEERS HAwGENS NOBLE GASES
-(MeV)
(50% Rel.)
(100% Rel.)
(1% Release)
(25% Rel.)
(100% Rel.)
0.00 - 0.10 1.18 71.7 1.94 0.014 1.69 0.10 - 0.40 266 244 28.6 3.14 5.76 0.40 - 0.90 4764 421 172 56.3 10.0
.0.90 - 1.40 1860 110 60.1 21.8 2.60 1.40 - 1.80 753 223 58.8 8.95 5.26 1.80 - 2.20 217 388 15.2 2.56 9.13 2.20~- 2.60 52.4 744 15.1 0.62 17.7 2.60 - 3.00 3.78 3.95-5.54 0.45 0.15 3.00 - 4.00 15.7 4.39 3.40 0.19 0.11 4.00 - 6.00 0.313 0
0.061 0.004 0
'Ibtal
-7933 2210 361 94.0 52.4 R/h/cc at.1 cm: 1.04 x 104 RCS 146 Containment..
-