ML20083N882
| ML20083N882 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 04/16/1984 |
| From: | Bayne J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20083N885 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM JPN-84-23, NUDOCS 8404190317 | |
| Download: ML20083N882 (9) | |
Text
123 Ma.n Street h
Vvhte Plains. New York 10601 j
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& Authority April 16, 1984 JPN-84-23 s
Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Attention:
Mr. Domenic B. Vassallo, Chief Operating Reactors Branch No. 2 4
Division of Licensing
Subject:
James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 NUREG-0737 Item II.B.3 Post-Accident Sampling System (PASS)
References:
1.
NRC letter, D. B. Vassallo to L. W. Sinclair, dated July 27, 1982.
,o 2.
NYPA letter, J. P. Bayne to D.
B. Vassallo, dated October 5, 1983 (JPN-83-8 5) 3.
NEDC-30088 " Response to NRC Post
-r-Implementation Review Criteria for Post-Accident Sampling Systems," dated April, 1983.
Dear Sir:
~
Reference 1 transmitted the criteria and guidelines for NUREG-0737 Item II.B.3.
Reference 2 provided responses for criteria 1, 2d, 3, 6, lla'and llb and committed to provide the remaining responses within six months.
This letter provides responses for the remaining criteria and includes a revised Table of Integrated Dose Rates.
The following attachments are included:
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B404190317 840416' 4-PDR ADOCK 05000333
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1.
Attachment A lists NRC criteria 2a, 2b, 2c, 4, 5, 7, 8,
9, and 10, and provides the Authority's responses.
2.
Attachment B is the " Core Damage Estimation Procedure" for the FitzPatrick Plant.
3.
Attachment C is the " PASS Containment Radiation Monitoring Procedure" for the FitzPatrick Plant.
4.
Attachment D is a letter from General Electric (GE) to the NRC on the " Accuracy of Dissolved Gas Measurement for GE Post-Accident Sampling Systems."
5.
Attachment E provides a revised Table of Integrated Dose Rates to replace Table 2 of Reference 2.
If you have any questions, please contact Mr.
J. A. Gray, Jr. of my staff.
Very truly yours, s!
-! j' -
c J./P'.Bhyne xecutive Vice President Nuclear' Generation cc:
Office of the Resident Inspector U. S. Nuclear Regulatory Commission P. O. Box 136.
Lycoming,>N.Y.
13093
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5 NEW YORK POWER AUTHORITY f!
t' JAMES A~. FITZPATRICK NUCLEAR POWER PLANT I
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>il Criterion:
(2) The licensee shall establish an on-site
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radiological and chemical analysis capability to f
j provide, within a three-hour time frame
)
established above, quantification of the j
following:
3 J
al certain radionuclides in the reactor coolant
]
and containment atmosphere that may be 4
indicators of the degree of core damage i
(e.g.,
noble gases, iodines and cesiums, and j
non-volatile isotopes) ;
h b hydrogen levels in the containment atmosphere; k
c) dissolved gases (e.g., H ), chloride (time 2
allotted for analysis subject to discussion below), and boron concentration of liquids.
Response
(2a) The FitzPatrick Plant on-site radiological analysis of post accident samples is performed using a Ge(Li) detector / multi-channel analyzer (MCA) spectrophotometer system with a counting h
cave of sufficient shielding to remain operable
(
in the worst case background radiation levels
[
within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after an accident-(based on g
projected Reg. Guide 1.3 and 1.4 source' terms).
. Provisions have been made to handle samples in f.
such a manner as to minimize personnel exposure through the use of transport casks and sample i
dilution caves.
In addition. remote handling
(:
tools will be available for hendling highly
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radioactive samples.
Reduction in exposure levels is also attained with an approximate 100:1 dilution (internal to the PASS sink) of-i reactor coolant.
Further sample dilutions will be performed in the radiochemistry lab under a fume. hood.
A Plant specific core damage estimate procedure (AttachmentlM has been prepared based on the generic procedure submitted by the BWROG.
The procedure estimates the core damage: based on
. fission product concentrations and integrates other physical parameters for confirmation of-these estimates.
Attachment C,;provides'a Plant'-
unique procedure for estimating the core damage based on the' containment radiation monitors.
4 i
Responset (2bi A gas chromatograph located in a fume hood will 1
3' be used for analysis of hydrogen gas from the y
. *s' primary coolant gas sample.
In addition, the j.
containment atmospheric monitoring system s
f hydrogen analyzer provides in-line monitoring of hydrogen.
Atmospheric samples are available f
from the torus, drywell and secondary x
s containment.
The gas sample is withdrawn from
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the sample vial through a small access port from i
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the gas vial cask in the PASS sink.
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' Response.:
(2c) Chloride concentrations will be analyzed of f-site by Babcock and Wilcox as mentioned in
~
Response 5.
An initial scoping for chlorides j
will be conducted on-site by the turbidimetric i
or spectrophotometric method.
Boron will be g
3 analyzed tur the carminic acid method.
Liquid 1
sample pH will be measured with a flat surface h
combination electrode.
3 ii
, Criterion:
(4) Pressurized reactor coolant samples are not required if the licensee can quantify the amount 1
\\'
of dissolved gases with unpressurized-reactor coolant samples.
The measurement of either total dissolved gases or H2 gas in reactor coolant samples-is' considered adequate.
Measuring the 02 concentration is recommended, but is not' mandatory.
j-
Response
(4)
As described in Attachment D, the plants with the GE PASS system are changing'_the method of measuring dissolved gas.
The method, including L
range and accuracy, is all described in 1
Attachment D.
At the time of this: letter, no J
written confirmation of approval from the NRC
[
for this dissolved gas modification has been h
received tar the Authority.
Therefore the L
FitzPatrick Plant can not confirm the ranges and accuracies'at this time.
The Authority will i.
confirm the entire dissolved gas portion of~the PASS after resolution of this issue and receipt of an approved. Technical, Specification Amendment:
to allow sampling!during power operation with a containment atmosphere' monitor temporarily:
a isolated.
I Criterion:'
(5)l 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-(4 if there is only a single barrier
,between primary containment systems and the.
cooling water.
Under both of the-above
1 conditions the licensee shall provide for a yi 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 i
being taken.
For all other cases, the licensee
/
shall provide for the analysis to be completed
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within 4 days.
The chloride analysis does not g
have to be done on-site.
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Response
( 5)
Since the FitzPatrick Plant does not use seawater or brackish water for plant cooling, a
T the chloride analysis is required to be I
completed within four (4) days.
An on-site f
scoping analysis using the turbidimetric method J
is available for an initial approximation.
Arrangements have been made with Babcock and Wilcox for accurate chloride measurements within 4 days, using an ion specific electrode with a liquid ion chromatograph as an alternate technique.
4 d
In addition, the FitzPatrick chemistry 1aboratory has purchased a liquid ion
- a chromatograph and will investigate the use of 1
this equipment to more accurately analyze for f
chlorides and to further reduce personnel y
exposure.
3 Criterion:
(7) The analysis of primary coolant samples for i
boron is required for_PWRs.
(Note that Rev. 2 7
of Regulatory Guide 1.97 specifies the need for primary coolant boron analysis' capability at BWR f
plants).
i nesponse:
(7)
Boron analysis capability for' liquid coolant samples will be performed by the carminic acid-method in the on-site laboratory.
Criterion:- (8)
If in-line monitoring'is used for any sampling y
land analytical capability specified herein,-the-E licensee shall provide backup sampling through~-
-grab samples,'and shall demonstrate the capability of analyzing thel samples.
- Establishedfplanning for' analysis at off-site a
facilities is: acceptable.
Equipment provided
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for backup sampling shall be capable of-
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.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 _ condition-no longer exists..
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Response
( 8)
The FitzPatrick Plant will not utilize in-line
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monitoring for PASS sampling.
The GE system employed at the FitzPatrick Plant utilizes a grab sample system which is transferred into a cask and analyzed in the chemistry lab on-site.
For backup capability and primary chloride measurements, the undiluted sample will be inserted in a Post Accident Shipping C a k.
Two casks have been purchased by the Pooled Inventory Management spare parts program (PIMS),
of which the Authority is a member.
These casks, being manufactured by Nuclear Packaging, Inc.
are awaiting final NRC licensing approval and subsequent delivery to the Memphis, Tennessee warehouse.
Emergency withdrawal procedures have been-developed.
i Additional arrangements have been made with Babcock and Wilcox for receipt of the cask for backup sampling capability which meet all of the Reg. Guide 1.97 ranges and accuracies.
l-Criterion: (9)
.The licensee's radiological and chemical sample analysis capability shall include provisions to:
W Identify and quantify the isotopes 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 on-site liquid sample analysis capability should be such as to permit measurement of nuclide concentration'in the range from approximately hgCi/g to 10 Ci/g.
-h Restrict b'ackground levels of radiation inLthe radiological and chemical analysis facility from.
sources such that the. sample analysis will provide results with an acceptably.small error.(approximately.
a factor 2).
This can be accomplished through the use of sufficient sh,ielding around samples and outside-sources, and~by the use'of a ventilation: system design which-will control the presence of airborno-radioactivity.
---4-?
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Response
(9ai As discussed in Attachment B, the source terms g
for FitzPatrick Plant are based on Reg. Guide 1.3 and 1.4.
The predicted activities from the e
4 various samples drawn were based on a reference f
BWR-6/238 Mark III containment (Reference 3).
1 Correction factors were applied (coolant mass, j
containment and drywell volumes, thermal rating) to develop plant unique activities for the j
FitzPatrick Plant.
These activities were j
incorporated into integrated whole body and k
extremity doses (Attachment E), for samples taken 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the accident.
O t
As discussed in response 2A, internal dilution 1j of approximately 100:1 is conducted for highly radioactive coolant samples to minimize 1
personnel exposures.
Also, transport casks are j
used for the same purpose.
r r
Although the maximum expected primary coolant activity for the FitzPatrick Plant is estimated at 2.63 Ci/g, capability for measuring up'to 10 1
Ci/g is available with a dilution of 1X10,
7 l-(9b) The GeLi detector in the FitzPatrick chemistry laboratory is adequately shielded to insure that l
background radiation levels 2 hrs. after an
(
accident (allowing 1 hr. for obtaining, transporting and preparing the sample) do not radically affect the sensi,tivity (f actor of 2).
i I
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.
l
Response
. (10) The accuracy, range and sensitivity for post accident sampling and~ analysis at FitzPatrick
(
are adequate to provide pertinent data to the i
operator for'the radiological and chemical j
status of the reactor coolant systems, as shown t.
below.
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Analysis Range Accuracy j;
Gross Activity AgCi/g-10 Ci/g i 100%
i (Radionuclide)
Boron 100 - 1,000 ppm 1 50 ppm
)
. Chloride (Of f-s ite)
<0.5 ppm i 50 ppb
(
0.5 ppm - 20 ppm
+ 10%
e Total dissolved gas To be addressed following resolution of the j;
dissolved gas issue.
l pH*
1 - 13
-+.03 i
- Assuming conductivity
>l Mpho/cm j
The procedures for the above, have been tested by GE with the effects of radiation and interference from other a
i constituents being negligible.
See Reference 3 for a-1 explanation of these results.
j Training for applicable personnel in the collection,
'y
-transport, and on-site analysis of samples for the PASS will
'be_ conducted annually and documented in'the Indoctrination
-Training Procedure ITP-7 for Radiological and Environmental m.
Li Technicians (chemistry).
Functional testing and calibration frequency will be addressed in the Authority's Technical ~
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. Specification Amendments related to NUREG 0737.
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