ML20071P683
| ML20071P683 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 05/31/1983 |
| From: | Devincentis J PUBLIC SERVICE CO. OF NEW HAMPSHIRE, YANKEE ATOMIC ELECTRIC CO. |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM SBN-514, NUDOCS 8306080125 | |
| Download: ML20071P683 (21) | |
Text
'
a 1671 Wortesser Rood Framinohom, Massachueens 01701 g%ggg (617) - 872 - 8100 May 31, 1983 SBN-514 T.F. B7.1.2 United States Nuclear Regulatory Commission Wa shing ton, D. C. 20555 Attention:
Mr. George W. Knighton, Chief Licensing Branch No. 3 Division of Licensing
References:
(a)
Construction Permits CPPR-135 and CPPR-136, Docket Nos, 50-443 and 50-444 (b)
PSNH Letter, dated December 1, 1982, " Revised Response to RAI 281.6; Post-Accident Sampling; (Chemical Engineering Branch)", J. DeVincentis to G. W. Knighton
Subject:
Open Item Response (SER Section 9.3.4.2; Chemical Engineering Branch)
Dear Sir:
In response to the Open Item included in the Safety Evaluation Report (Section 9.3.4.2) regarding NUREG-0737, Item II.B.3 (Post-Accident Sampling Capability) we have enclosed a report which addresses the status of our I
compliance with each of the Item II.B.3 criteria. We have included l
commitments to complete the development of sampling and analysis procedures, j
shielding studies, and sampling system time and motion studies not later than six months prior to fuel load.
The enclosed report supplements the information included in Reference (b) and OL Application Amendment 48.
Note that the method of obtaining a containment recirculation sump sample has been modified.
Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY 8306080125 830531
[I N
/g%
h:
PDR ADOCK 05000443
[J.DeVincentis Project Manager llnpl
{
ALL/fsf Enclcsure
/
cc: Atomic Safety and Licensing Board Service List fr 1000 Elm St., P.O. Box 330. Manchester. NH O3105 Telephone (603) 669-4000 TWX 7102207595
CRITERION: (1)
The licensee shall have the capability to promptly obtain reactor coolant samples and containment atmosphere samples. The 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 sample.
RESPONSE
The post-accident sampling system provides the capability to obtain liquid samples from reactor coolant loops 1 and 3 (hot legs), containment recirculation sumps, pressurizer relief _ tank, ECCS pump room sumps and gas samples of the containment atmosphere within three hours from the time a decision is made to take a sample. All electrically powered equipment (i.e., solenoid valves and sample pumps), whose operation is required to perform post-accident sampling, is powered from an emergency backup power source.
Specific details concerning time spans to enter and exit the sample panel area, operate the sample panel manual valves, perform manual sample dilutions, ar.d transfer sample to the shield cart for analysis will be performed when construction activities allow an securate appraisal but no later than six months prior to fuel load.
i l
l
s
\\
r 4.
w
. * ?,.
~
~"
s_
A h*
s
't}
Criterion:
(2) The licensee shall establish an onsi e radiological and N N chemical analysis capability to provide, wir.hin three-hour' time ' Y frame established above, quantification of th's following:
(a) certain radionuclides in tNe reactor co'olant and contain-
- ment atmosphere that any'be indicators pf the degree of core damage (e.g., noble
"', I non-volatile ' isotopes)(
- gases; iod2.nes\\and cesiums, and N
[
y s.
s i
e s.
4 (b) hydrogen levels in the containment atmosphere; _
r dissolved gases (e.g.,s,H ); chloride (time allotted for
{'
- (c) 2 analysis subjec{ to discuss {on;below), and boron con-l t
centration of liquids.
~
I l
v'%
rt c Alternativelyptave inline monitoring capabilities to
.(d) s perforyallor[partoftheaboveanalyses.
i
,)
Clarificati~ni:2 (a),s'A discussion of the Counting equipment capabilities is o
tneeded, including provisions to handle samples and reduce backgrourd1 radiation to minimize personnel radiation exposures (ALARA). Also a procedure,is required for relating radionuclide concentrations to core damage. Tne,
procedure should inch:de:
^
N 1.
Monitoringfpr short and long lived volatile and non '
131 e,137 s,,
volatile -radionuclides such as.133Xee 1
C Cs, 85 r, 140
, and 88Kr (See Vol. II, Part 2'.,
\\Y 134 K
3
,C pp. 524-527 of. Rogovin. Report for further information).
s, 4
2.
Provisions to estimate the extent of core damage based on radionuclide concentrations and taking into consideration other physical parameters such as core
, temperature data 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 accuracy).
g\\
m7 -
' N., I 1
S-l
. \\
\\
l
\\.
N 8
y
\\f.
y
- . w v, 7:3
.' \\
' l(
C s,
1 v_.
s i
b
t g.
Respoose.2
" Onsi.t e' radf.)1ogic)al and chemical analysis capability will be Pk in D,
- e's' tab'lished to meet thi three-hour time frame. The post-accident
\\
[,
sampling' subsystem ~provides the capability to obtain liquid samples
/.
from recetor coolant loops 1 and 3, ECCS pump room sumps (RHR/CBS J N Susps"Ahnd"B"andPABSump"A"),thepressurizerrelief' tank,and the RHR pump discharge (containment recirculation sump sample).
Gas g
samples of the contah ment atmosphere under post accident conditions
- g.,
- t. -
can be drawn 'from the installed hydrogen analyzer system.
i b
g%
.N Radionuclides 'will be measured on grab samples by gamma ray
- jy
" spectroscopy tising ger:2anium detectors. Hydrogen concentration in containment will be ' measured on grab samples of the containment
"~
catmosphere. Dissolved gas will be determined by degassing an aliquot
<>f liquid and. obtaining a grab sample of gas for hydrogen analysis.
t i
' Chloride and boron concentration on dilute liquid samples will be determined by analysis of grab samples.
Background levels will be reduced in the counting room through the use of a shielded cave. Personnel radiation exposure will be maintained ALARa through the use of lead shield carrying devices and remote handling devices where appropriate.
A procedu're will be developed for r-lating radionuclide con-centrations to core damage levels taking into account core temperature and sample location. Specific radionuclide identification and con-centration will be accomplished through the use-of-a full spectrum--
scan.
Sampling procedures and core damage level determination proce-dures are currently under development and will be available six s.onths prior to fuel load.
4 J
(:12 CRITERION:
(3)
Reactor coolant and containment atmosphere sampling during postaccident 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.
RESPONSE
Neither the-reactor coolant / sump nor containment atmosphere sampling system requires an isolated auxiliary system to be placed in service for the purpose of sampling.
RCS Sampling In the case of the RCS sampling system, samples can be supplied to the panel directly from reactor coolant loops 1 and 3 (hot legs), the discharge of either RHR pump (containment recirculation sump sample), the disci.arge of the pressurizer relief tank sample pump, the discharge of the IAB cump "A" sample pump, and the discharge from either RHR/Chd equipment vault A" or "B" sump sample pump. See revised FSAR Figure 9.3-5c, attached, and FSAR Figure 9.3-Sa.
The system interface valves include both manually-operated and remotely-operated valves. All manually-operated valves required for system alignment are equipped with accessible handwheel reach-rod extensions, which are located on a wall adjacent to the sample panel.
All remotely-operated valves are environmentally qualified for the conditions in which they need to operate and are cycled from either the Control Room or local panel.
The post-accident sampling system will provide a means to override the safeguards signals that automatically close the sample isolation valves.
Containment Atmosphere Sampling l
l The containment atmosphere sampling system draws samples from the hydrogen analyzer suction line and returns tSe sample flow to containment through the hydrogen analyzer return line. Inside containment, both of these lines are open-ended to the containment atmosphere.
j The sample' supply and return taps are located in the rain steam and feedwater pipe chase building and tie into the l
hydrogen analyzer suction line upstream and downstream of j
Valve CGC-V13, for Train "A" (CGC-V35 for Train "B") see revised FSAR Figure 6.2-95, attached.
The system interface valves are manually operated.
l I
l
Y
(
Criterica:
(4) Pressurized reactor coolant samples are not required if the licensee can quantify the amount of dissolved gases with unpressurized reactor cociant samples.
The measurement of either total disnolved gases or H2 gas in reactor coolant samples is considered ade-quate. Measuring the 02 concentration is recom-mended, but is not mandatory.
l Clarification:
Discuss the method whereby total dissolved gas or hydrogen and oxygen can be measured and related to reactor coolant system concentrations. Additionally, if chlorides exceed 0.15 ppa, verification that j
dissolved oxygan is less than 0.1 ppe is necessary.
Verificati6n that dissolved oxygen is <0.1 ppm by measurement of a dissolved hydrogen residual of > 10 i
cc/kg is acceptable for up to 30 days after the acci-dent. Within 30 days, consistent with minimizing
{
personnel radiation exposures (ALARA), direct muni-
. toring for dissolved oxygen is recommended.
Response (4)
The amount of dissolved gases in reactor coolant will be deter-mined by extracting a gaseous sample from-the post-accident sampling panel using a shielded syringe if necessary. This sample will be ana-lyzed. for hydrogen and gamma spectrum only. The procedure for this analysis is currently under development and will be available six months prior to fuel load.
i l
l l
l^
l l
0:
1 1
Criterion:
(5) The time for a chloride analysia 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 pro-vide 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 licen-see shall provide for the analysis to be ccepleted within 4 days. The chloride analysis does not have to be done onsite.
C1srification:
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 have only a single barrier protection between the ruactor coolant are required to analyze chloride within 24 hottrs.
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 pps-C1 (the licensee should establish this as
'value; the number in the' blank-should be-no greater s
than 10.0 ppm C1) in the reactor coolant system and (2) that+ dissolved. oxygen can be verified at <0.1 npa, c.onsistent_with the guidelines above in clarifi-t cation no. 4.. Additionally, if chloride analysis is l
performed on a diluted sample, an undiluted-sample-i need also be taken and retained for analysis within 30 days, consistent with ALARA.
Response (5)
Grab sample analysis' for chloride on a diluted liquid sample will be completed within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> of drawing the sample.
Seabrook Station
-utilizes seawater for cooling water but the design incorporates a double barrier between primary containment systems and the cooling
- water. Samples-may be diluted up to a factor of 1000 at the sample statioe. However, the analysis employed will provide for a minimum detectable thceshold of 10 ppm C1.
The post accident sampling system will provide for che capability of taking an undiluted sample consistent with ALARA principles. This undiluted sample will be retained for analysis within 30 days.
Procedures for drawing both the diluted and undiluted chloride samples and for the analysis of the diluted chloride sample are under development and will be available six months prior to fuel load.
G
- CRITERION: (6)
The 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 individual exceeding the criteria of GDC 19 (Appendix A, 10 CFR 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).)
RESPONSE
A shielding analysis will be performed, no later than six-months prior to fuel load, to ensure that operator radiation e.posure from reactor coolant / containment x
atmosphere sampling and analysis is within the acceptable limits of 5 rem whole body and 75 rem extremities. The operator exposure will include entering and exiting the sample panel area, operating the sample panel manual valves, performing manual sample dilutions, and transferring sample to shielded cart for analysis.
l l
l r
l 1
g a -
Criterion:
(7) The analysis of primary coolant samples for baron is required for PWRs.
(Note that Rev.1 of Regulatory Guide.l.97 specifies the need for primary coolant boron analysis capability at BWR plants).
Clarification:
PWR's need to perform boron analysis. The 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.
Rysponse (7)
Boron analysis will be conducted on a diluted liquid grab sample.
Procedures are under development and will be available six mcrths prior to fuel load.
t e
y
9 cz Criterion:
(8) If inline_ monitoring is used for any sampling and analytical capability specified herein, the licensee shall provide backup sampling through grab samples, and.shall demonstrate the capability of anlayzing the ssaples.
Established planning for analysis at off-site facilities is acceptable. Equipment provided for backup sampling shall be capable of providing at least one sample per day for 7 days following onset of the accident, and at least one sample per week until the accident no longer exists.
Clarification:
A capability to obtain both diluted and undiluted backup samples is required. Provisions to flush inline monitors to facilitate. 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)
The Seabrook Station post-accident sampling system does not utilize any inline monitoring ccpabilities.
i I
4
......n.._,
0 y
Criterion:
(9) The licensee's radiological and chemical sample ana-lysis capability shall include provisions to:
(a)
Identify and quantify the isoptopes of the nuclide categories discussed above to levels corresponding to 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 analysia capability should be such as to permit measurement of nuclide con-centration in the range from approximately 1 pCi/g to 10 Ci/g.
(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).
This can be accompolished through the use of sufficient shielding around samples and outside sources, and by use of a ventilation system design which will control the presence of airborne radioactivity.
Clarification:
-(9) (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 sufficier.tly to perform the required analysis.
Discuss the range of radionuclide 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 the contribution from s
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.
w
3-Response (9)
(a)
Isotopes of the nuclide categories of noble gases, iodines, cesiums and non-volatile isotopes will be identified and quantified to levels corresponding to the source terms given in Regulatory Guides 1.4 and 1.7.
There will be provisions for a 1000 to one dilution of the sample at the sampling station. This will be sufficient for transporting a small aliquot to the counting room using lead shield carrying devi-ces and remote handling devices. If necessary, the sample can be counted using collimated counting geometry. Liquid sample measurement capabilities will permit measurement of radionuclides concentration in the range from approximately 1 pCi/g to 10 Ci/g.
Procedures for this analysis will be available six months prior to fuel load.
(b)
Bac kground levels of radiation will be restricted in the counting room and the primary laboratory through the use of shielding and ventilation to provide results within an acceptably small error. The counting room has 30 inch concrete walls and roof.
Ventilation will be controlled to both areas to limit the ingress of airborne radioactivity. All wet chemical analysis will be performed in an operating fume hood. Samples will remain behind shielding during both storage and analysis.
A
J Q
^
Criterion: (10) Accuracy, range, and sensitivity shall be adequate to pro-vide pertinent data to the operator in order to describe radiological'and chemical status of the reactor coolant systems.
Clarification:
The recommended ranges for the required accident sample analyses are given in Regulatory Guide 1.97, Rev. 2.
The necessary accuracy within the recommended ranges are as follows:
- Gross activity, gamma spectrum:
measured to estimate core damage, these analyses should be accurate within a factor of two across the entire range.
- Boron: measure to verify shutdown margin.
In general this anlaysis abould be accurate within 15% 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 1 50 ppm).
For concentraticas below 1,000 ppm the tolerance band should remain at 1 50 ppm.
- Chloride: measured to determine coolant corrosion potential.
i For concentrations between 0.5 and 20.0 ppm chloride the analysis should be accurate within,1-10% of the measured value. At concentrations below 0.5 ppm the tolerance band remaine at 1 0.05 ppm.
- Hydrogen or Total Gas: monitored to estimate care degradation and corrosion potential of the coolant.
An accuracy of 110% is desirable between 50 and 2000 cc/kg but i 207 can be acceptable. For coccentration below 50 cc/kg the tolerance remains at 1 5.0 cc/kg.
- Oxygen: monitored to assess coolant corrosion potential.
For concentration between 0.5 and 20.0 ppm oxygen the ana-lysis should be accurate within i 10% of the measured value. At concentrations below 0.5 ppm the toicrance band remains at 1 0.05 ppm.
pH: measured to access coolant corrosion potential.
BJ veen 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.
~
4-Q To 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. This can be accomplished by per-forming 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 REACTOR COOLANT SAMPLES IN A POST-ACCIDENT ENVIRONMENT Nominal Constituient Concentration (ppm)
Added as (chemical salt)
I-40 Potassium Iodide Cs+
250 Cesium Nitrate Ba+2 10 Barium Nitrate La+3 5
Ammonium Cerium Nitrate Cl-10 B
2000 Boric Acid Li+
2 Lithium Hydroxide NO -
150 3
NH +
5 4
K+
20 Gamma Radiation 104 Rad /gm of Adsorbed Dose (Induced Field)
Reactor Coolant NOTES:
- 1) In,trumentation and procedures which are applicable to diluted samples only, should be tested with an equally diluted chemical test matrix.
The induced radiation environment should be adjusted commensurate with the weight of actual reactor coolant in the sample being tested.
- 2) For FWRs, procedures which may be af fected by spray additive chemicals must be tested in both the standard test matrix plus appropriate spr ay additives. Both procedures (with'and without spray additives) are required to be available.
3)
For BWRs if procedures are verified with boron in the test matrix, they do not have to be tested withaut boron.
- 4) In lieu of conducting tests utilizing the standard test matrix for instruments and procedures, provide evidence that the selected instru-ment or procedure has been used successfully in a similar environment.
ff Q
M All equipment and procedures which are used for post accident sampling and anlayses 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
.samp'.ing, analysis and transport. A minimum frequency for the above efforts is considered to be every six months if indicated by testing.
These provisions should be submitted in cevised Technical Specifications in accordance with Enclosure 1 of NUREG-0737. The staff will provide model Technical. Specifications at a later date.
Response-(10)
The accuracy, range and sensitivity of post-accident analyses capabi-lities will be detailed in the analytical procedures currently under deve-lopment..The procedures will be available six months prior to fuel load.
l
[
5 I
i I
l l
9T y
T-9 v
-yesy.-
w.m------ - -,. -
s w
w
,r---e w-
, w wrt,, - - - -
w-
. -- w
---.g.a-w..,~w
- ----+. ----- - - --- - - - - + -
--emw
CRITERION: (11)
In the design of the postaccident sampling and analysis capability, consideration should be given to the following items:
(a) Provisions for purging sample lines, for reducing plateout in sample lines, for minimizing sample loss or distortion, for preventing blockage of sample lines by loose material in the RCS or containment, for appropriate disposal of the samples, and for flow restrictions to limit reactor coolant loss from a rupture of the sample line. ~The postaccident reactor coolant and containment atmosphare samples should be representative of the reactor coolant in the core area and the containment atmosphere following a transient or accident. The sample lines should be as short as possible to minimize the volume of fluid to be taken from containment. The 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.
RESPONSE
(a) RCS Sampling The Post-Accident Sampling System (see revised FSAR Figure 9.3-Sc attached and FSAR Figure 9.3-Sa) is capable of obtaining samples from reactor coolant loops 1 and 3 under accident conditions where the Reactor Coolant System remains pressurized. Under these conditions, natural circulation could be established to remove decay heat and provide mixing of the RCS wa ter.
Loop samples are taken from existing sampling points located in the hot leg of loops 1 and 3.
These sampling points are the sample points utilized for routine primary sampling.
If an accident resulted in the depressurization of the primary system (i.e., a LOCA), long-term, post-accident samples would be provided from the discharge of the RHR pumps. These samples would be recirculation water which is pumped from the containment recirculation sump, through the RHR heat exchanger and back into the reactor vessel. Mixing in the vessel is accomplished as the recirculation water is forced to flow through the core to remove decay heat.
The sampling panel has the capability of obtaining a liquid sample from the pressurizer relief tank when the Reactor Coolant System is either pressurized or
depressurized. This is accomplished through the use of the PRT sample pump.
Equipment / valve leakage will be collected in RHR/CBS Sumps "A" and "B", as well as PAB Sump "A". Samples can be taken from these sumps and supplied to the sample panel from the discharge of the RHR/CBS sump sample pumps and the PAB Sump "A" sample pump, respectively.
The sampling system is purged using demineralized water which is flushed through the system. In the event of a loss of off-site power, the sample system ie parged by establishing flow through the sample line, panel and returning the flow to the containment. See revised FSAR Figure 9.3-5c, attached.
The sample purge flows discussed above are within the region of turbulent flow which should promote mixing in the sample lines and help reduce sample plateout and distortion. Each sample line of the Post-Accident Sampling System will be kept as short as possible to limit the volume of fluid needed to be taken from the system.
The sample panel has two sample return lines. One line can be valved to return samples to the containment. This flow path is used under post-accident conditions when it is desirable to return the sample purge flow to the containment.
The second return line is used for normal operation which includes testing and operator training exercises. This line returns sample flow to the Floor and Equipment Drain System.
Flow restrictions in the sample line are provided by flow restrictors (orifice) and solenoid-operated isolation valves to limit reactor coolant loss from a rupture of a sample line.
Strainers / screens shall be incorporated in the
-sample lines or pump inlets, respectively, to prevent system blockage.
Containment Atmosphere Sampling The Containment Atmosphere Sampling System draws air samples from two existing lines which are open to the containment at the dome area. Sampling from these areas should provide samples representative of the containment atmosphere. See revised FSAR Figure 6.2-95, attached,
g Purging of the sampling lines is accomplished by establishing flow through the hydrogen analyzers using the analyzer sample pump. Removable sample cylinders are installed upstream of the analyzer with bypass' lines to allow continuous purging.
-Redundant sample lines exist (one for each H2 analyzer), therefore, blockage of a sample line will not prevent the capability to obtain samples.
Heat tracing has been provided to both sample lines and will maintain the sample temperature at
'approximately 3000F. This will ensure that moisture in the sample will not condense before reaching the sample cylinder. Sample lines will be kept as short as possible to limit the air volume needed to be taken from containment.
(b) The post-accident RCS sampling panel vacuum pump and vents discharge-into the PAB Ventilation System.
The PAR ventilation flow passes through roll, medium efficiency, HEPA and carbon filters before being discharged to the plant primary vent stack.
l i
g ie w$
T t
N c
83 v *a S
48 i4 E
9 D
3 t1 s I
9 a R e.
C t1 l
ey g n E
s.
C er S
"3 t
u R
n %N A
d :
w t s T
G e
U T
nu en Rs
- .. o S
I Os
- ,s.
- e. M c O
F ma Fi
- .i t
a u.
AJ e
tn
.g i
X N,.
II
- m. iS RA s
AR F
D c 9 %.
e a
W e, t
oE EG
'u*
3 f
4
=
5 LA
. e w
- a. M -"
=
7.S m t..v CI l
f 7
N1 J',
UD M
o W
- m. '8_
C s
T MP
,I
)
I js I
)
l
)
T l
g E
)
)
)
i r
,*v S
u e
o y.,*
s e
6 Y
z f
,a s
s S
e v
v A
.,ca Pgs e w**
A-a a
a'
-,r
';g, e
8 f*.
7 t
ts a
ya
+
es t
n y
3" m,b"
.m*
E
- 4=.e
..r s v'*
I* e.
u L
P r
c u
f \\s P
9' p
/
M wJg x.
.J a
s.
A q
1 n
S
"%r c
N *=- ;m
- 4 e
r
_m
-,.u uy g
+
a s,.
E a
g wl
.E I
ug e
w R
+
uY.
a n
e u
s.
.=
H e
s
- g. 3'~
e
. r.
M S
r 3
a.
- 4 P
E
+e 2
M T
w
^
D__
A &R i
g g
H1 O x.
EI R s
P g**2 WSE 2
h pw NN T
.-s
.:s8
~
ws g4-S 4
1
,~
e FUI m
, '4 OS
. T.
g ~.i.
Y
+
" e
$}
i Y NL
.=
ji f
."b a
NOA w
- qb VTY IN
+
s
- s N
P A A T A
,~
3, i.
9,.
,,l it
.6 m
s.
OST i.
u
.i E
CK F
m is
,S' EOA COS e.
.q h'.<
b =.e wv.
EAN I
V R L.
R Df e
- w**
i, c
S SEI
(
- 4 i.
SF C
D y.
s y
t.
L I
e s
S U
B
[
os
=a m
e.
w t
=-
t.
^..
s t e
m s
s P
N-es.
%e awsn
%e
=-
'w e
e t.e se en
_.~
n
. g.
swe a/
a
=
n.
s.
g w
1 1i T.
e L
_Y.
T-'
T g 1'1 (e.
1<-
4 d
e
' u.
g 1 -
d w
e g1,.,* I e
==
,t
.c.
,=
Y
,;n 3-Ls.
ea
,=
a
,.,. 4=
e l'
w
,.-,t
,2 a
w$ei a
r g,
,. u L.1o,,
c s.
)
. g._'.
.L7
%M* 9
_ " =
e r e.
s,,
i.r. b' Y,
s.
4 A. h"E*
ge e
o L
s g
L n.
J aa
, m,9 n
l' 3n
'p 4
bp L' 0 e.
!4 e
g.
- g s
g 5
H
,v
]p.
Vl q g
Aq e
u s
c
- L c,P.%
s
,x.
dg
=
l_
e.
-v.
,,=
y 1g
]
o
'u, o.
e 1i a e
J" *_ u, d
s.
m i,
e g
s
)
w iu
.,4N.
s
.s
(
g,!
8
.i+
x-e.
g.d
.e em2L
)
s, 1p
_mi l
=
=
l e
m/
m\\
q]
}
u
- s.
- a e e u.
- c o
- t. a t*
r.
{.pI s
83_
s
!!:l!,Il:i,y l! l 4,,,3 es i
o j
I
- i. slim !l ?Ik
- 9 i
0 Pl 46 94 e
$.! IN !.. !dX 85 5 !
3
,;f 4
s ia; 8
9 m-(
%} ni a t.9, [;
_e c o l e, 5,!
L y
y
$i I
l n
,a
~
3,,i r,I i
i.
(h
,-W+
_ _. _ _ N h
NE 1e:,p
- gi, lieM 'as!@l d
,p
[>
e is d i
li!
1 j)= H f,i 3
)*
i
- m
!!,:,Mi. a e
ess
)
- M)
- b
$$I o>
j l
M Eh' Gk ((
j i
4 g!!
i i
i-i s e m
I oom 5-
(( ] "!
I f
!" A A
.1 A Ni
?;
l l
'"'If I
9
.f l
}.
I
)
n l;
h
("
'Y t
}M'q3 p
i i
ji
.i i
0
.Ak;_k'3$)
h$f a z
e s.
A M
g!
, fg
,il, f '
lg a_
!dfh)h;l'i.
O li ip qT"
~~
l' ! r
- t E
'd 666 d'
m 7:
)
D i
~
-5 P
ii d
i i
'l i~
il ii..
j i
i f
it i
'r l
I y ;
N 06y i
}
L u
11.
i i
ig.
5o
.j l I
l3rl
.75 !
g M
I')
i G
- 4. '
f*
0{
i i
=
a 2; a; p
y v
a i 42 I
6/
I
Rep. Beverly Hollingworth Ms. Olive L. Tash Coastal Chamber of Commerce Designated Representative of-209 Winnacunnet Road the-Town of Brentwood Hampton, NH. 03842 R.F.D.
1, Dalton Road-Brentwood, NH 03833 William S. Jordan, III, Esquire Harmon s Weiss Edward F. Meany 1725 I Street, N.W.
Designated Representative of Suite 506 the Town of Rye Washington, DC.20006 155 Washington Road Rye, NH '03870 Roy P. Lessy, Jr., Esquire
- Office of the Executive Legal Director Calvin A. Canney U.S.
- Nuclear Regulatory Commission City Manager Washington, DC 20555 City Hall 126 Daniel Street
' Robert A. Backus, Esquire Portsmouth, NH 03801 116 Lowell Street
.P.O.
Box 516 Dana Bisbee, Esquire Manchester,-NH 03105 Assistant Attorney General Office of the Attorney General Philip Ahrens, Esquire 208 State House Annex Assistant Attorney General Concord, NH 03842 Department of the Attorney General Augusta, ME 04333
. Anne Verge, Chairperson-Board of Selectmen Mr. John B.
Tanzer Town Hall Designated Representative of South Hampton, NH 03842 the Town of Hampton 5 Morningside Drive Patrick J. McKeon Htmpton, NH 03842 Selectmen's Office 10 Central Road Poberta C.
Pevear Rye, NH 03870 Designated Representative of
-the Town of Hampton Falls Ruthanne G. Miller, Esquire Drinkwater Road Law Clerk to the Board Harpton Falls, NH 03844 Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Mrs. Sandra Gavutis Washington, D.C.
20555 Designated Representative of
- the Town of Kensington Dr. Maury Tye, President RFD 1 Sun Valley Association East Kingston, NH 03827 209 Summer Street Haverhill, MA 01830 Edward J. McDermott, Esquire Sanders and McDermott Mr. Angie Machiros Professional Association Chairman of the Board of Selectmen 408 Lafayette Road Town of Newbury Hampton, NH 03842 Newbury, MA 01950
'Jo Ann Shotwell, Esquire Assistant Attorney General Environmental Protection Bureau Department of the Attorney General One Ashburton Placo, 19th Floor Boston, MA 02108
_, -. - - -. - - - - - -