ML20079P202
| ML20079P202 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 01/26/1984 |
| From: | Gucwa L GEORGIA POWER CO. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM NED-84-038, NED-84-38, TAC-44445, TAC-44446, TAC-53478, NUDOCS 8401310123 | |
| Download: ML20079P202 (21) | |
Text
.
l Geor;;pa Power Cottery 4
333 Piedrnont Avenue At anta, Georg-a 30308 Telechone 404 5:tM5:'6 Ma Wg Address-Post Ottce Box 4545 Attanta. Georg a 3030?
Georgia Power L T.cuewa re sewux s.wm Meager Nuc: ear Eng:neenn9 NED 84-038 and Cu Nuclear Entneer January 26, 1984 Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 Division of Licensing U. S. Nuclear Regulatory Camission Washington, D. C. 20555 NRC DOCIG'rS 50-321, 50-366 OPERATING LICENSES DPR-57, NPF-5
~
EDWIN I. HA101 NUCLEAR PIANP UNITS 1,2 NUREG-0737 I'ITM II.B.3, POST-ACCIDENP S7MPLING CAPABILITY Gentlenen:
A meeting was held on January 11, 1984 b<. tween members of the Nuclear Regulatory Cmmlssion (NRC) staff and representatives of Georgia Power Cmpany (GPC) to discuss the need for an extension of the implenentation schedule for NrIREG-0737 Iten II.B.3.
In the course of that meeting Mr.
George Rivenbark, Hatch Licensing Project Manager, requested that GPC make a.
subnittal detailing the extent to which the criteria of Iten II.B.3 were presently satisfied at Plant Hatch.
For those criteria which were not yet satisfied, GPC was requested to subnit an optimum schedule for their cmpletion.
Pursuant to that request, we herein subnit a description of the current post-accident sapling and analysis capabilities at Plant Hatch.
The description of the Post-Accident Seple Systen (PASS) in Enclosure 1 is based on the systen design criteria and factory acceptance tests. We PASS is currently functional and undergoing site validation tests. While the systen is functional, a number of tasks renain to be cmpleted prior to declaring the systen fully operational. These tasks, including validation tests, are identified in the Enclosure along with a scheduled empletion date.
On Novenber 30, 1983, Georgia Power Capany requested an extension of the ordered empletion date for iten II. B.3 to August 1985.
%is request was based on the potential
- need, at that
- time, to do extensive re-engineering of the PASS.
We now believe that resolution of the major problens with the PASS can be - achieved without a major re-engineering effort, thus allowing empletion of the PASS and related outstanding itens on an improved schedule. Accordingly, we are hereby revising our extension request to a cepletion date of August 1,1984.
8401310123 840126 fDRADOCK 05000321 PDR
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Georgia Power A Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission January 26, 1984 Page 'No Please contact this office if there are any questions.
Very truly yours, gf* C g - -
L. T. Gucwa JH/jh Enclosure xc(w. encl.): J. T. Beckhan, Jr.
H. C. Nix, Jr.
P. D. Rice J. P. O'Reilly (NRC-Region II)
Senior Resident Insptor 700775
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I ENCIOSURE 1 LIMIN I. HATCH NUCLEAR PIMff UNITS 1 AND 2 A
IOST-ACCIDENT SAMPLING CAPABILITY l
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INIRODUCTION
%e purpose of this report is to dmonstrate the extent of cmpliance of Plant Hatch Units 1 and 2 with the criteria of NURE-0737 Iten II.B.3.
Post-accident smpling and analysis capability is being provided at Plant Hatch by previously existing in-line drywell hydrogen and oxygen analyzers in conjunction with the Post-Accident Smple Systen (PASS).
%e PASS is an in-line, autmated system capable of obtaining and analyzing representative reactor coolant and drywell atmosphere saples during normal or accident conditions. A systen flow diagre is provided as Figure A.
The FASS can be used for smpling and analysis of either Hatch unit at one time.
Each unit has a single reactor coolant ample tap located in a jet pmp flow-sensing instrment line and a single drywell atmosphere saple tap located in a Fission Products Monitoring Systen smple line.
Smple flow is routed frm these sources to the Post-Accident Smple Room (PASR) which is located within the Hot Machine Shop, adjacent to the Unit 2 Reactor Building.
The PASR heatses in-line monitors capable of analyzing for the ratuired isotopic and chenical variables. Diluted and undiluted grab smpling equipnent is provided in the PASR for backup sapling.
Liquid effluent fra the PASS is discharged to the torus during accident conditions or to a radwaste drain during normal plant operation. Gaseous effluent is discharged to the drywell during normal or accident conditions.
%e PASS is designed to meet or exceed the criteria of NUREG-0737.
While the PASS has been installed and successfully tected to a limited extent, it is not fully operable at this time.
We following section provides responses to the NURm-0737 criteria and clarifications which detail the exact extent to which Plant Hatch emplion with Itan IT.B.3.
l l
f JAN 2 6 GB4
II. RESPONSE 'IO NUREG-0737 CRITERIA AND CLARIFICATIONS A.
CRI'IERION (1)
"The licensee shall have the capability to prmptly obtain reactor coolant smples and t9ntalment atmosphere smples.
'Ihe cmbined time alloted for sapling 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 frm the time a decision is made to take a smple."
CLARIFICATION (1)
" Provide information on smpling(s) and analytical laboratories locations including a discussion of relative elevations, distances and methods for smple transport.
Responses to this iten should also include a discussion of smple recirculation, saple handling, and analytical times to denonstrate that the three hour time limit will be met (see (6) below relative to radiation exposure).
Also describe provisions for smpling during loss of off-site power (i.e. designate an alternative backup power source, not necessarily the vital (Class lE) bus, that can be energized in sufficient time to meet the three hour suopling and analysis time limit)."
RESPONSE
In-line monitoring is used as the primary method of analyzing for the raguired variables.
Manual handling of smples is therefore not necessary.
Since the PASS is provide 3 with emergency power, the smpling and analysis process can be performed within the three hour time limit, with or without off-site power.
Plant Hatch is therefore in empliance with Criterion 1.
'i f
, JAN 2 61984
B.
CRI'IERION (2)
"'Ihe licensee shall establish an on-site radiological and chenival analysis capability to provide within the three hour time frahe established above, quantification of the following:
(a) certain radiontrlides in the reactor coolant and contalment abnosphere that may be indicators of the degree of core daage (e.g.,
noble gases, iodines and cesims, and non-volatile isotopes)
(b) hydrogen levels in the contaiment atmosphere; H ),
chloride (time alloted for (c) dissolved gases (e.g.,
2 analysis subject to discussion below), and boron concentration of liquids.
(d) Alternatively, have in-line monitoring capabilities to perfom all or part of the above analyses."
CIARIFICATION (2)
(a)
"A discussion of the counting equipnent capabilities is needed, including provisions to handle saples and reduce background radiation to minimize personnel radiation exposures (AIARA).
Also a
procedure is required for relating radionuclide concentrations to core dmage.
The procedure should include:
1.
Monitoring fut short and long-lived volatile ard non-volatile radionuclides such as 133 e, 131,
X 1
137 s, 134 s, 85 r, 140 a, and 88 r (See Vol C
C K
B K
II, Part 2, pp. 524-527 of Rogovin Report for further information).
2.
Provisions to estimate the extent of core dmage based on radionuclide concentrations and taking into consideration other physical parameters such as core teroperature data and smple location.
(b) Show a capability to obtain a grab smple, transport, and analyze for hydrogen.
(c) Discuss the capabilities to sample and analyze for the accident smple species listed here and in Regulatory Guide 1.97 Rev. 2.
(d) Provide a discussion of the reliability and maintenance information to demonstrate that the selected on-line instrunent is appropriate for this application (See (8) and (10) below relative to back-up grab smple capability and instrunent range and accuracy)." +
J AN 2 61984 i
RESPONSE
Plant Hatch is in partial compliance with Criterion 2 as explained below:
(a) Quantification of radionuclides in the reactor coolant and contaiment atmosphere is performed using an in-line, intrinsic germanim detector with an adjustable collimator, a multichannel
- analyzer, and a
cmputer with a
spectrun stripping code.
Manual handling of smples is not required for radionuclide analysis.
A procedure has been developed for estimatiivj the extent of core dmage based on radionuclide concentrations and other plant parmeters.
This procedure is based on the generic methodology developed by GE in the docment NEDO-22215.
(b) Hydrogen concentration in the contaiment atmosphere is measured independently of the PASS by in-line monitors which utilize thermal conductivity bridges.
(c) Dissolved hydrogen in the reactor coolant is quantificd with an in-line monitor which ures a thermal conductivi:y bridge to teasure the hydrogen content of the stripped gas.
Boron acd chloride concentrations in the reactor coolant are measured by in-line monitors which utilize specific ion electrodes.
%e capability to saple and analyze for Regulatory Guide 1.97, Rev. 2 variables will be addressed
'n GPC's response to NUREG-0737, Supplanent 1.
(d) %e in-line monitors are expected to be reliable with low maintenance rajuirefnents.
Certain PASS functions will be performed routinely by the plant operating staff, which will result in faniliarity with the systen and rapid identification of operational problens.
B;uipnent location and flushing / purging provisions facilitate access for repair.
An adninistrative progre is being developed to ensure proper testing and maintenance of the PASS.
Plant Hatch will be in cmpliance with Criterion 2
upon implenentation of the administrative progra referenced above.
- J AN 2 61984
C.
CRITERION (3)
" Reactor coolant and contalment atmosphere sampling during post-accident conditions shall not require an isolated auxiliary systm [e.g., the letdown systs, the reactor water cleanup system (RWCUS)] to be placed in operation in order to use the smpling systs."
CIARIFICATION (3)
"Systs schaatics and discussions should clearly dmonstrate that post-accident smpling, including recirculation, fr m each s a ple source is possible without use of an isolated auxiliary systs.
It should be verified that valves which are not accessible after an accident are enviromentally qualified for the conditions in whfch they must operate."
RESEONSE (Refer to Figure A.)
Reactor coolant smple taps are located in jet puap flow-sensing instrment lines outboard of the excess flow check valves.
Contaiment isolation valves are provided in the smple lines. To initiate smple flow it is necessary to override the isolation signals to these valves.
It may also be necessary to reset the excess flow check valve.
Drywell atzaosphere smple tepr are located in Fission Products Monitoring Systs smple lines outboard of previously existing contaiment isolation valves.
To initiate this smple flow it is necessary to override the isolation signals to these valves and close a downstrem valve which isolates the Fission Products Monitor.
ne Primary Contaiment Isolation Valves are enviromentally qualified. All other valves are capable of withstanding the enviroment in which they must operate.
Since reactor coolant and drywell atmosphere sapling can be performed during post-accident conditions without the use of an isolated auxiliary systs, Plant Hatch cmplies with Criterion 3. J AN 2 61984
D.
CRIERICN (4)
" Pressurized reactor c0olant smples are not reluired if the licensee can quantify the amount of dissolved gases with unpressurized reactor coolant smples.
The measurenent of either total dissolved gases or H2 gas in reactor coolant smples is considered adequate.
Monitoring the 02 contentration is recomended, but is not mandatory."
CIARIFICATION (4)
" 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 ppn, verification that dissolved oxygen is less than 0.1 ppo is necessary.
Verification that dissolved exygen is < 0.1 ppn by measurenents of a dissolved hydrogen residual of 210 cc/kg is acceptable for up to 30 days after the accident. Within 30 days, consistent with minimizing personnel radiation exposures (AIAPA),
direct minitoring for dissolved oxygen is recamended."
RESPCNSE As discussed in the responsa to Criterion 2, an in-line monitor uses a themal conductivity bridge to measure the hydrogen content of gas renoved from the reactor coolant.
A calculation is performed to relate this hydrogen content to reactor coolant systen concentration.
Dissolved oxygen measurenent is not required by NURB3-0737 and is not presently provided.
Plant Hatch therefore emplies with criterion 4.
, JAN 2 6 $84
E.
CRI'IERION (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 contaiment systens 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 smple being taken.
For all other cases, the licensee shall provide for the analysis to be empleted within 4 days. 'Ihe chloride analysis does not have to be done on i
site."
CIARIFICATION (5)
"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 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 up to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> to perform a chloride analysis.
Seples diluted by up to a factor of up to one thousand are acceptable as initial scoping analysis for chloride, provided (1) the results are reported as _
ppn C1 (the licensee shall establish this value; the nmber in the blank should be no greater than 10.0 ppn Cl) in the reactor coolant systen and (2) that dissolved oxygen can be verified at <0.1 ppn, consistent with the guidelines above in Clarification No. 4.
Additionally, if chloride analysis is performed on a diluted sample, an undiluted smple need also be taken and retained for analysis within 30 days, consistent with AIARA."
l l
RESPONSE
Since sea or brackish water is not used as cooling water at Plant Hatch, the 96 hour0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> time limit is applicable.
Measurenent of chloride concentration in the reactor coolant can be performed with the PASS within this time limit.
Plant Hatch is therefore in empliance with Criterion 5. JAN 2 6 1984
F.
CRITERION (6)
"The design basis for reactor coolant and contaiment abnosphere smpling and analysis must assme that it is possible to obtain and analyze a saple without radiation exposures to any individual exceeding the criteria of GDC 19 (Appendix A,10 CFR Part 50) (i.e.,
5 Rm whole body, 75 Re extraities).
(Note that the design and operaticnal review criterion was changed frms the operational limits of 10 CFR Part 20 (NUFM-0578) to the GDC 19 criterion (October 30,19B letter frm H. R. Denton to all licensees))."
CIARIFICATION (6J
" Consistent with Regulatory Guide 1.3 and 1.4 source terms, provide information on the predicted personnel exposures based on person-motion for sapling, try. sport, and analysis of all required parmeters."
RESIONSE Prior to smpling and in-line analysis with the PASS, it is necessary to activate the boron and chloride analyzers at local control panels.
These panels are located outside the FASR, but within the Hot Machine Shop.
'Ihis area would have a low radiation level and be accessible following an accident.
The rmainder of the smpling and analysis process can be rmotely controlled frm the Health Physics area, which would also be a low radiation area.
The grab smple portion of the PASS has been designed such that operator doses would be within the GDC 19 limits, consistent with Regulatory Guide 1.3 source terms.
Plant Hatch cmplies with Criterion 6.
t G.
GRIIEUCN (7)
"The analysis of primary coolant smples for boron is required for PWRs.
(Note that Rcv. 2 of Regulatory Guide 1.97 specifies the need for primary coolant boron analysis capability at BWR plants)."
CLARIFICATION (7)
"PWRs need to perform boron analysis. The guidelines for BWPs are to have the capability to perform boron analysic but they do not have,to do so unless boron was injected."
RESFONSE Boron concentration in the reactor coolant is measured by the PASS. Plant Hatch is therefore in cmpliance with Criterion 7.
p2 6 WB4
H.
CRITERION (8)
"If in-line monitoring is used for any sampling and analytical capability specified herein, the licensee shall provide backup smpling tnrough grab smples, and shall dmonstrate the capability of analyzing the smples.
Established planning for analysis at offsite facilities is acceptable.
muipment prcvided for backup sampling shall be capable of providing at least one smple per day for 7 days following onset of the accident, and at least one smple per week until the accident condition no longer exists."
CLARIFICATION (8)
"A capability to obtain both diluted and undiluted backup smples is required.
Provisions to flush in-line 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
In-line monitoring is used as the primary method of performing the required analyses.
Flushing and porging provisions facilitate access for repair of the monitors.
Diluted or undiluted backup grab saples can be obtained.
Grab smpling sluipnent allows one smple per day for the first seven days to be obtained, then one smple per week.
A Pooled Inventory Management (PIM) licensed shipping cask will be used to ship backup smples off-site.
The PIM cask is scheduled for delivery to the PIM facility by May 1, 1984.
Arrangements for analysis have been made with Oak Ridge National Laboratory.
Plant Hatch will be in empliance with Criterion 8 upon delivery of the PIM shirping cask.
p2 6 1964.
I.
CRITERION (9)
"The licensee's radiological and ch mical smple analysis capability shall include provisions to:
(a)
Identify and quantify the isotopes of the nuclide categories described 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 smples to provide capability for measurment and reduction of personnel exposure should be provided.
Sensitivity of onsite liquid smple analysis capability should be such as to permit measurment of nuclide concentration in the range frcra approximately 1 aci/g to 10 C1/g.
(b) Restrict background levels of radiation in the radiological and chmical analysis facility frm sources such that the saple analysis will provide results with an acceptably small error (approximately a factor of 2).
%is can be accmplished through the use of sufficient shielding around samples and outside sources, and by use of a ventilation syste design which will control the presence of airborne radioactivity."
CIARIFICATION (9)
(a)
" Provide a discussion of the predicted activity in the saples to be taken and the methods of handling / dilution that will be employed to reduce the activi'y sufficiently to perform the required analysis.
Discuss the range of radionuclida concentrations which can be analyzed for, including the mount of overlap between post-accident and normal sapling capabilities.
(b) State the predicted background radiation levels in the counting rom, including the contribution frm smples which are present.
Also provide data deonstrating what background radiation levels and radiation effect will be on a saple being counted to assure an accuracy within a factor of 2."
RESENSE Plant Hatch is in cmpliance with Criterion 9 as discussed below:
(a) %e PASS in-line detector provides the capability to identify and quantify the required isotopes to levels corresponding to Regulatory Guide'l.3 and 1.7 source terms. S eple dilution is not required for isotopic analysis.
%e range of the PASS detector is 10-lpCiAl to 10 Ci
%e range for normal isotopicanalysiscapabilityis10gl.
mCiA l to 10 pCi/ml.
(b) %e PASS detector is located in the PASR which is an area of low background radiation.
Sufficient shielding of radiation sources is provided to assure accuracy within a factor of two.
PASR ventilation provisions control the presence of airborne radioactivity.
~ ~
p2 6 W34
J.
CRI'IERION (10)
" Accuracy, range, and sensitivity shall ne adequate to provide pertinent data to the operator in order to describe radiological and chenical status of the reactor coolant systen."
CIARIFICATION (10)
"Tne reconnended ranges for the required accident smple analyses are given in Regulatory Guide 1.97, Rev. 2.
'Ihe required accuracy within the reconnended ranges are as follows:
+
Gross activity, gamma spectrtsn: measured to estimate core dmage, these analyses should be accurate within a facter of two across the entire range.
Boron: measure to verify shutdown margin.
In general, this analysis should be accurate within 15% of the measured value (i.e. at 6,000 ppn B the tolerance is +300 ppn while at 1,000 ppn B the tolerance is 150 ppn). For concentrations below 1,000 ppn the tolerance band should reaain at 150 ppn.
j Gloride: measured to determine coolant corrosion potential.
For concentrations between 0.5 and 20.0 ppn chloride the analysis should be accurate within +10% of the measured value.
At concentrations below 0.5 ppn Ee tolerance band renains at 10.05 ppn.
Hydrogen or Total Gas: monitored to estimate core degradation and corrosion potential of the coolant.
An accuracy of 110% is desirable between 50 and 2000 cc/kg but 120%
can be acceptable.
For concentration below 50 cc/kg *he tolerance remains at i 5.0 cc/kg.
Oxygen: monitored to assess coolant corrosion potential.
For concentrations between 0.5 and 20.0 ppn oxygen the analysis should be accurate within +10% of -the measured value.
At concentrations below 0.5 ppn he tolerance band renains at 10.05 ppn.
pH: measured to assess coolant corrosion potential.
Between a pH of 5 to 9, the reading should be accurate within 10.3 pH units.
For all other ranges 10.5 p1 units - is 5
acceptable.
4 l JAN 2 6 1984
[
CIARIFICATION (10) (CONT'D.)
j To denonstrate that the selected procedures and instrunentation i
will achieve _the above listed accuracles, it is necessary to provide informaticn denonstrating their applicability in the post accident water chenistry and radiation enviroment.
'Ihis can ' be accmplished by performing tests utilizing the standard test matrix provided below or by providing evidence that the selected procedure or instrunent has been used successfully in a similar enviroment.
1 STAEARD 'IEST MATRIX FOR UNDILUIED REACIOR 00(TM SAMPLES IN A POST-ACCIDENP ENVIROMENT Naninal
_ Constituent Concentration (ppn)
Added as (chenical salt)
I-40 Potasslun Iodide Cs+
250 Cesiun Nitrate Ba+2 10 Bariun Nitrate La+3 5
Lanthanun Chloride Ce+4 5
Anunoniun Cerlun Nitrate Cl-10 B
2000 Boric Acid Li+
2 Lithian Hydroxide j
j NO -
150 3
2-5 4
K+
20 Gatsna Radiation 104 Rad /gn of Absorbed Dose (Induced Field)
Reactor Coolant W2 6 1984 n:
.o s
CIARIFICATION (10) (CONT'D.)
NOTES:
1 1)
Instrmentation and procedures which are applicable to diluted
- smples only should be tested with an equally diluted chenical test matrix.
The induced radiation enviroment should be adjusted canensurate with the weight of actual reactor coolant in the smple being tested.
2)
For IWRs, procedures which may be affected by spray additive chenicals must be tested in both the standard test matrix plus appropriate spray additives.
Both procedures (with and without spray additives) are required to be available.
3)
For EWRs, if proculures are verified with boron in the test i
matrix, they do not have to be tested without boron.
4)
In lieu of conducting tests utilizing the standard test matrix for instrtments and procedures, provide evidence that the selected instrtment or procedure has been used soccessfully in a similar envirornent.
i All equipnent and procedures which are used for. post-accident smpling and analysis should be calibrated or tested at a frequency which will ensure, to a high degree of reliability, that it will be should. receive-initial and available if required.
Operators refresher training in post-accident
- sampling, analysis, and transport. A mininta frequency for the-above efforts is considered 2
to be every six months if indicated by testing.
These provisions should be subnitted in revised Technical Specifications in accordance with Enclosure 1 of NUREG-0737.
The staff will provide model Technical Specifications at a later date."
RESPONSE
Plant Hatch is in partial empliance with Criterion 10 as discussed below:
(a)
Instrument accuracies and sensitivities were. determined by factory tests which were perforned using the reconnended test matrix.
There were no substantial deviations fran the-accuracies specified in Criterion 10.
In-line' analysis'within the specified accuracies is available for -all' required analyses except chlorides.
Off-site chlorido analysis with the specified accuracy within the 96_ hour time limit has been arranged.
Fttrther in situ validation testing is necessary to verify that the installed PASS achieves the necessary accuracies.
4 6 W+ y2
RES NNSE (CONT'D)
(b) An adninistrative progra is under developnent to ensure the capability for post-accident sapling and analysis of reactor coolant and drywell atmosphere.
The progra will include (1) training of personnel, (2) procedures for sapling and analysis, and (3) provisions for maintenance of smpling aN3 analysis quipnent.
When empleted this program will be referenced in a proposed revision to the Administrative Technical Sp mifications.
Upon successful validation testing of the PASS and implmentation cf the above referenced administrative progrm, Plant Hatch will cmply with Criterion 10.
K.
CRITERION (11)
"In the design of the post accident smpling and analysis capability, consideration should be given to the following ites:
(a)
Provisions for purging smple lines, for reducing plateout in smple lines, for minimizing smple loss or distortion, for preventing blockage of smple lines by loose material in the RCS or contalment, for appropriate disposal of the smples, and for flow restrictions to limit reactor coolant loss frm a rupture of a smple line.
%e post accident reactor coolant and contaiment atmosphere samples should be representative of the reactor coolant in the core area and the contaiment atmosphere following a transient or accident.
The smple lines chould be ac chcrt oc pocsible to minimize the voltsne of fluid to be taken frm contaiment.
The residues of saple collection should be returned to contaiment or to a closed systs.
(b) ne ventilation exhaust frm the smpling station should be filtered with charcoal absorbers and high-efficiency particulate air (HEPA) filters." J AN 2 61984
~
~
CLARIFICATION (ll)
(a)
A description of the provisions which address each of the itms in clarification ll.a should be provided.
Such ites, as heat tracing and purge velocities, should be addressed. To dmonstrate that smples are representative of core conditions a discussion of mixing, both short and long term, is needed.
If a given smple location can be rendered inaccurate due to the accident, (i.e. s m pling fra a hot or cold leg loop which may have a stem or gas pocket) describe the backup sapling capabilities or address the maximm time that this condition can exist.
BWRs should specifically address samples which are taken frm the core shroud area and demonstrate how they are representative of core conditions.
Passive flow restrictors in the saple lines may be replaced by redundant, enviromentally qualified, rmotely operated isolation valves to limit potential leakage from smpling lines.
%e autmatic contalment isolation valves should close on contalment isolation or safety injection signals.
(b) A dedicated smple station filtration systs is not required, provided a positive exhaust exists which is cubsequently routed through charcoal absorbers and HEPA filters."
RESPONSE
Plant Hatch is in partial cmpliance with Criterion 11 as discussed below:
The PASR is located as close to the smple sources as possible (a)_ while still maintaining it in an area of low background radiation.-
Saple lines are constructed of the smallest practical tubing (3/8 inch dimeter) and are routed in the most direct path. These considerations minimize the volme of Jfluid being removed fr m containment. Liquid saple lines are piirged with deineralized water while gas saple lines are purged with service air.
Saple and purge flow rates are
'udximized to prevent sample line blockage and to enhance purging effectiveness.
He use of large radius bends and large throat valves minimizes saple loss or distortion.
Gas saple lines will be provided with electric heat tracing to s
minimize condensation and plateout.
Potential reactor coolant loss frm a
ruptured smple line is limited by the f167-limiting orifice in the jet pmp flow-sensing instrment line. Both reactor coolant and drywell atmosphere smple lines are ' provided with redundant, enviromentally qualified, N
reotely operated contaiment isolation valves which close on
?
a contaiment isolation signal.
As discussed in Section I, liquid effluent from the PASS is discharged to the torus during accident conditions or to a radwaste drain during normal plant operation. Gaseous effluent is discharged to the drywell during normal or accident conditions.
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RESPONSE (ENT'D)
Reactor coolant smples taken frm the jet pmp instrm ent line will be representative of coolant in the core, because a direct path will always exist frm the core region to the saple point.
If a small break or no break occurs, reactor water level will be maintained at or near normal level.
As long as water is maintained in the upper plenm, natural circulation will occur with a large loop fra the downcmer to the shroud region via the jet pe ps.
With the thermal conditions forcing water up through the core and down past the tap frm which the PASS saple is taken, the saple will be representative of water in the core.
If normal water level can not be maintained, as in the case of a recirculation pmp suction line break, the jet pmp saple point will still provide a representative s mple.
In this case the Core Spray Systs will supply water to flood the core.
The flow will be down through the core then up through the jet pmps to exit through the postulated break.
Again, flow past the saple point directly frm the core will ensure a representative smple. Contaiment atmosphere saples will be representative of conditions throughout the primary contalment due to the small size of th contaiment end the emplete mixing of tne j
contaiment atmosphere.
(b) 'Ihe PASR is provided with ventilation by the Standby Gas Treatznent System, which contains both charcoal adsorbers and HEPA filters.
Upon installation of heat tracing, Plant Hatch will cmply with criterion 11...
III.CDNCWDING SLM4ARY
%e preceding discussion deonstrates that Plant Hatch is in substantial cmpliance with the criteria of NrJRM-0737 Its II.B.3.
The actions necessary to achieve full empliance were noted in Sections II.B, II.li, II.J and II.K and are reiterated with scheduled empletion dates below:
(a) An administrative progra to ensure the capability for post-accident sapling and analysis of reactor coolant and drywell atmosphere must be impleented. (June 1,1984)
(b)
%e PIM shipping cask Inust be available for transportation of grab smples to an off-site facility. (May 1,1984)
(c) Validation testing of the PASS must be successfully cmpleted. (May 1, 1984)
(d) Heat tracing must be installed on the PASS contalment atmosphere smple lines. (August 1,1984)
In addition to the capabilities discussed in Section II, develognent of on-site laboratory analysis procedures is in progress.
Analytical techniques for Boron and pH have been raviewed and necessary procedures, quipment, and training should be cmplete by June 1,1984.
It should be noted, however, that for radiation exposure reasons these laboratory analyses will be performed on diluted reactor coolant saples.
We accuracy and sensitivity of these analyses, based on discussions with other utilities, will likely not meet the NUREG-0737 Clarification 10 criteria.
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