IR 05000324/1990028
| ML20059J027 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 08/24/1990 |
| From: | Decker T, Seymour D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20059J023 | List: |
| References | |
| 50-324-90-28, 50-325-90-28, NUDOCS 9009190139 | |
| Download: ML20059J027 (20) | |
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[p 8cE UNITED STATES
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. q'g NUCLEAR REGULATORY COMMIS$10N
HEoloN H
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101 MARIETTA STREET.N.W.
's ATLANTA. CEoRot A 30323
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me 2 pm Report Nos.:
50-325/90-28 and 50-324/90-28
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Licensee:
Carolina Power and Light Company
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P. O. Box 1551
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Releigh, NC 27602
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Docket Nos.:
50-325 and 50-324 License Nos.:
DPR-71 and DPR-62 Facility Name:
Brunswick I and 2
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Inspection Con ucted: July 23-27, 1990 Inspectors:
M F/A
IT. A.
eymo,
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Approved by:
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TUR. ' Deck'er, Chief
~Date ' signed
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g Radiological Effluents and Chemistry Section e'
Emergency Preparedness and Radiological Protection Branch Division of Radiation Safety and Safeguards-
$0MMARY Scope:
This routine, unannounced inspection was conducted in the areas of radiological effluents, plant chemistry, and confirmatory measurements.
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Results:
There were no significant radiological consequences! attributable to the operation of Brunswick in -1989 noted from airborne, waterborne, aquatic, ingestion or direct exposure pathways (Paragraph 4).
Inspector Followup Item 88-28-01 concerning the inoperable condition of the radwaste liquid effluent flow measurement device was closed (Paragraph 2).,
' Inspector Followup Item 90-10-01 concerning1t'he determination of'the main stack flow rate was closed (pParagraph 2).
An audit of the radiological environmental monitoring program at Brunswick'was thorough. The licensee's response to the audit-contained adequate commitments i
by management for corrective actions for the nonconformances identified (Paragraph 5).
9009190139 900824 PDR ADOCK 05000324 O
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i Brunswick is cont,inuing shipments of spent fuel-to Harris, but has stopped
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back-washing the filled casks to eliminate the crud in Harris' spent fuel pool (SFP) and in the shipping casks. A corporate task force has been formed to i
determine the best' method for dealing with this issue.
The inspector
considered the fact that this_ situation was being dealt with as a CP&L concern:
I and not a. specific-site's problem; and the fact that Brunswick and Harris were
'i working together as a. team to resolve this issue, as a licensee strength
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(Paragraph 6),
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Reactor' coolant conductivity and chlorides.were. maintained below Technical.
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Specification limits.and BWROG action -levels.
The' Hydrogen injection System l
was still in the. testing phase. - The electrochemical' potential (ECP) for i
both units was being maint&ined below - 230 millivolts (mV).(Paragraph 11),
i A review:of the chemistry counting room quality control indicated :that this
program was adequate to ensure accurate and reliable analpical results.' This I
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was support.by good agreement between the licensee's chemistry and health
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physics count rooms and the NRC mobile laboratory gamma ' spectroscopic l
measureinents (Paragraph 3).
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REPORT DETAILS 1,
Persons Contacted
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Licenseo Employees
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- K. Altham, Manager,' Regulatory Compliance
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- S. Callis, On-Site Representative, Licensing
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- A. Cheatham, Manager, Environmental and Radiation Control (E&RC)
- W. Dorman, Quality Assurance / Quality Control
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- S. Fitzpatrick, Senio. Specialist, E&RC D. Geddings, Outage Management and Modification
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J. Gurgainous, Foreman, Environmental and Chemistry (E&C)-
- J. Harness, Site General Manager D. Holden, Health Physics Foreman
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- W. Nurnburger, Unit 2 Chemistry Foreman -
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- E. Scharff, Principal Engineer, On-Site Nuclear Safety
- R. Smith, Manager, Radiation Control G. Worley, Radiation Control Foreman, Radwaste Shipping
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Other licensee employees contacted during this inspection included engineers, technicians, and administrative personnel.
Nuclear Regulatory Commission
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- D. Prevatte, Senior Resident Inspector
- Attended exit interview i
Acronyms and initialisms used throughout this report are listed in - the last paragraph.
2.
Licensee Action on Previously Identified Inspector Follow-up Items (IFIs)
(92701)
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(Closed)
IFI 50-324/88-28-01:
Inoperable condition of'the radwaste I
l liquid effluent flow measurement device.
t This item was discussed in Inspection Report 90-10. dated April 13,
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During the current inspection the inspector reviewed the
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progress of this project and discussed pertinent details with i
cognizant licensee personnel and determined that the budget package for this project was completed through the Nuclear Engineering Department.
The budt.t package described the project, the necessity
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l of the project, the consequences of nonapproval, 'and roughly defined I
the scope of the project.
A plant modification package will be developed after the budget package is distributed and reviewed.
Prior to this the scope of the project is planned on being expanded to include replacement of pipe for both units,
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The budget package will go through a long-range planning phase and a design phase, where the modification package will be developed.
If the modification package meets budget approval,' the actual work will begin.
For this projection there is a tentative implementation date of May 1991.
Based on this scheduling this item is considered i
closed..
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(Closed) IFI 50-324/90-10-01 and 50-325/90-10-01:
Determine main
. stack flow with a reasonable degree of accuracy in order to meet FSAR commitments for isokinetic sampling.
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As discussed in Inspection Report No. 90-10, dated April 13, 1990, i
the licensee had planned actions to rectify.the uncertainty in their main stack flow measurements. These actions included replaciig the flow measurement device in the stack. Discussions with the licensee during the current inspection indicated that this project had progressed, a plant modification' package had been developed, and a
budget package had been approved.
This work is scheduled to be t
completed by the end of September 1990.
Based on the progress ofE this project, and on the scheduling that has been developed..this item is considered closed.
3.
Confirmatory Measurements (84750)
Pursuant to 10 CFR 20.201(b) this area was inspected to verify the
licensee's ability to conduct precise and accurate measurements.
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During this inspection, samples of reactor coolant-and selected-liquid and gaseous process streams were collected and the resultant-sample matrices r
were analyzed for radionuclide concentrations using the. licensee's counting laboratory and the NRC Region 11 mobile' laboratory gamma spectroscopy system. The purpose of these comparative measurements was to
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, verify the licensee's capability to accurately measure quantities ;of
radionuclides in various plant systems. Analyses were conducted using the licensee's three intrinsic germanium gamma spectroscopy systems in the count room; and for selected = samples, the two health physics Lintrinsic germanium gamma spectroscopy systems.
Sample types and counting geometries included the following:
reactor coolant,.50-milliliter bottle ~; liquid weste, one liter marinelli; main stack gaseous effluent,1250 milliliter marinelli; Unit No. I charcoal cartridge; a Unit No.'2 particulate filter;-
a NRC spiked charcoal cartridge, and a NRC spiked particulate. filter.
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Comparison of licensee and NRC results are listed in Attachment 1, Table I with the acceptance criteria listed in Attachment 2.
The results were.in agreement for all sample types analyzed.
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i The inspector observed the licensee obtaining the main stack gaseous
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effluent sample and;the Unit No. 2 particulate filter tample.. Proper
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sampling techniques and health physics practices were observed.
The
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inspector reviewed selected portions of E&RC Procedure No. 2002, titled NMC Monitor Particulates and Iodine Sampling, Volume VIII, and selected-
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portions of Appendix D, Rev. 9, E&RC; Procedure No. 2002,' Wide Range Gas
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Monitor Grab Gaseous' Sampling, and Appendix G. Rev.11, General Atomic
^ Wide Range Gas Monitor Particulate and. Iodine Sampling.
The portions.
reviewed were adequate for.the intended purpose.
Quarterly, the licensee -participated in an extensive' split gamma spectroscopic, tritium. gross alpha, and gross. beta analyses program with
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an outside vendor. The inspector reviewed the results of this cross-check'
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program for 1989, and for the first quarter'of 1990. The licensee and the
vendor were in agreement for_a11' isotopes.
No violations or deviations were identified.
4.
Radiological Environmental Monitoring'(84750).
TS 6.9.1.6 requires the submittal of a' routine. Radiological Environmental operating Report. This report summarizes the results of the. Radiological-
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Environmental Surveillance Program, ; which.. measures; accumulation = of
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radioactivity in.the envi ronment,-
and ' determines whether the t
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radioactivity detected is due to the7 operation of.=the; Brunswick Plant, i
This program also assesses the dose' to off-site populations from plant i
effluents.
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Pursuant to these requirements, the inspector reviewed the report for
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1989, i
The average gross beta concentration for, air particulate samples. for
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1989 was 1.55 E-02 picoeuries per cubic meter (pC1/m3). This is down
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f rom the 11988 average ' of.1.60 E-02 pCi/m3 and '.the preoperation average of 8.7 E-02 pCi/m3
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The radionuclides"in'dicative of: plant effluents were less than the
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iower limit of-deletion (LLD) for-the particulatnifilter analys s.
The concentrations of I-131 based on the analyses of air. cartridges were less than the LLD for all indicator and' control locations.
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.One vegetation sample out of 92 indicator samples contained Cs-137
(1.74 E-02 pCi/m ).
Seven out of thirteen control samples contained
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detectable concentrations of Cs-137 (range 1,56 E-02 pC1/m3 to 7.30 E-02 pC1/m3).
Brunswick's report stated. that this' prubably-l originated 'from world-wide > fallout and was not indicative of plant
effluents.
The shoreline sediment -samples all had radioactivity concentrations
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less than LLD.
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Two of the 24 monthly surface water samples. indicated concentrations
greater. than LLO for three nuclides. - ' The January,1989 surf ace :
waters sample from the Stilling pond _ indicated a - concentration 'of i
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1.60.10.74 E+03 pCi/ liter of trituim (LLD=1.20 E+03). The February
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l 1989 sample f tom tne stilling pond contained l1.53 E+01 pC1/ liter.
i of Mn-54 (LLD=3.0 E+0) arid 1.30 E+01 pCi/ liter.of Co-60 (LLD=4~.0- E+0).,
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This wes attributed to a high_ sediment content which occurred during-
this time-frame. The report stated that these sediments are. effective t
scavengers of radionuclides f rom. nor. mal plant' effluents.. These
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concentrations were within regulatory limits.-
f The fish and invertebrate sample activities were all.less than LLD.
- The average external dose rate based on'the environmental dosimetry-
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was 0.75 mrem per Week, compared to 0.86 mrem per week for 1988. The-
preoperation data _ indicated 1.02 mrem per week..
In summary, no significant radiological consequences'to the environment
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waterborne, aquatic, ingestion or. direct exposure. pathways.-
No violations or deviations were identified.
5.
Quality Assurance (QA) (84750)
TS 6.5.5.2h requires that the Performance Evaluation _ Unit (PEU) of the Corporate Quality Assuranco Department shall function to perform an_a'udit of the radiological environmental monitoring program' and; the; results thereof at least once per 24 months, to assess the. effectiveness 1of-this -
program.
Pursuant to this requirement the inspector reviewed the Quality Assurance Audit of Brunswick Nuclear Project Environmental ^ and Radiological
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Contro1' Program dated May 14, 1990. This audit.was conducted April-2-23, 1990.
Four nonconformances were identified -in the E&RC Program, two of '
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which-dealt with the areas covered bylthis inspection. The inspector also reviewed BSEP's responses to the two pertinent audit findings.
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i The first finding. dealt with the fact that the total release volume for the Main Stack for the week of March 6-12.,1990 was based on data obtained
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from'an inoperable stack flow-rate integrator. The'E&RC technicians were
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not made aware that the instrument had been placed in a Limited Condition
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of Operation. This could potentially cause erroneous gaseous activity to
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be reported.
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It was noted in the response that _the technical specification action statement for an inoperative flow indicator does not' direct Operations to inform E&RC of. the -flow indicator's inoperable status..The licensee's
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corrective actions included having the Operations Real. Time Training Group develop and schedule training for-the operators.on the technical
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specification instruments that are used by.E&RC to verify compliance; and on the fact that E&RC must be notified.when these instruments f ail.
.In addition, 01-03.2, Unit 1 Control 0perator Daily Surveillance Report, and 01-03.2 Unit! 2 Control' Operation Daily Surveillance Report, will i
be revised to alert the operators 'of the requirement to. notify E&RC of
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inoperable flow instrumentation.
E&RC revised the weekly-and monthly
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gaseous affluent reports' to reflect estimated values for the inoperable f
time period.
The training and procedure changes are scheduled to be completed by August' 31, 1990.
The second nonconformance involved the fact that the main condenser off gas system hydrogen analyzer was inoperable' for a period of time greater than.30 days, but that the monitor was not included in the latest semiannual radioactive effluent report (July - December.1989) as inoperable equipment as required by TSs. A discussion was_ held with NRC
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Region 11 representatives when this was discovered, 'and it was decided
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that an amendment would be included in the : next semiannual report.
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corrective actions included having Operations notify Regulatory Compliance
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about the expired LCO, and having Regulatory Compliance notify E&RC, in
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writing. This written notification wil.l be maintained to ensure that this information will be included in the semiannual' report.
The inspector discussed the audit findings and corrective actions with
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The audit response contained commitments by
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management to effect corrective action. for the nonconformences that had
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been identified.
No violations or deviations were identified.
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Spent Fuel (SFP) Shipments (84750)
Brunswick had made five spent feel shipments to-Harris in 1989.
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shipme.its were continuing in 1990.
The spent fuel had a large amount of.
iron oxide corrosion products adhering to their'outside surfaces.
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corrosion products, or crud, had deposite on various surfaces and the
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bottom' of the SFP and transfer canals at Harris. The activity of - the crud, as measured by an outside-contractor, was approximately 14 millicuries per gram.
The crud easily becomes airborne when dry;
-when wet it is sticky, tenacious, and difficult to flush out of-lines; and
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poses an ALARA concern in terms of clean up (see Paragraph 8 for details).
The fuel is shipped to Harris in specially designed casks, which 'are required to be inspected once per year.
During the annual inspection for 1990 at Harris, an empty cask was discovered to have radiatiun readings on the.inside bottom surface of the cask of approximately i
450 roentgens per hour.
This was-due to a buildup, or " heel", in the cask, of crud which had' f allen off.the spent fuel during shipping.
The fuel is ~ shipped dry,. is thermally hot, and undergoes vibrations during
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the shipping process, all possibly exacerbating the crud shedding process.
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~ Harris personnel flushed the crud from the cask into their SFP. The.
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Brunswick fuel continues to shed crud. into Harris's SFP af ter storage in the pool.
A hypothesis is that the' borated water in Harris's SFP acts as an acid wash.
As a result of the high activity levels discovered in the bottom of-the cask, and as a result of the crud deposition in Harris' SFP, it was decided that Brunswick would perform a cask backflush prior to shipping a
the cask.to Harris. The beckflush would be an attempt to remove som_e or all of the crud from the fuel. and the bottom of the cask..The inspector
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reviewed selected portions of E&RC' Procedure 0582 Vol. VIII, Rev. 6, dated July 23, 1990, titled Handling the 1F-300 Cask. This procedure had j
been amended to include optional-steps for backflushing the cask. At the j
time of this inspection the back flushing procedure had been performed at i
Brunswick one time.
Through conversations with the licensee, it ' was determined that this procedure removed significant amounts of crud from
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the cask.
The transfer line used for the flush read'approximately 250 milliRoentgens per hour on contact for the first ten minutes of the flush.
The flush was terminated af ter approximately thirty minutes. The flushed material was sent to the Waste Neutralizer Tank, part of Brunswick's radwaste system. Because of the nature ~of the crud, hot spots were created in the transfer lines.
The radwaste system at Brunswick,
because of ALARA concerns, was not designed to handle the levels of
activity present in the crud. Because of these and other concerns, at the
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time of this inspection an interim ~ decision had been made.to continue to-ship spent fuel to Harris without back flushing the casks.
As of August 10, 1990, an officialicorporate task force had been formed to determine a permanent resolution to'this complex problem. This task force consists of eight people and includes personnel from Brunswick, Harris and the Corporate office.
This task force is etartered with determining a
- solution which would minimize the ALARA impect, the radwaste shipping l
impact, and cost of resolving this issue.
The inspector considered the fact that this situation was being dealt with as a CP&L concern and'not as a specific site's problem, and the fact that Harris and Brunswick were working together as a team to resolve this issue, as a licensee strength.
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Post-Accident Sampling System (PASS) (84750)
TS 6.8.3.c provides for the establishment, implementation,' and maintenance
of a ' post-accident sampling program.
The inspector discussed PASS.-
operation and maintenance experience with licensee personnel.
The inspector. reviewed the operability tests of the PASS for Unit 2 dated April 1,1990 and for Unit I dated April 2,,1990. Within the scope of the review the inspector determined that the PASS operability. was adequate.
No violatiorrs or deviations were identified.
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8.
Radioactive Liquid Wastes and Liquid Effluent Treatment System (84750)
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TSs 3/4 11.1,1 through 3/4 11.1.3 define the operating requirements, the
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radioactive ' ef fluent limits.. and the surveillance requirements for the liquid radwaste treatment systems',
' Pursuant to these requirements, the inspector' discussed operation of the liquid radwaste systems with licensee representatives, and toured the radwaste area examining components of the system and - the associated control room in order to gain familiarity with system operability.-
The inspector determined that at Brunswick they do not regenerate. resin or use waste evaporator tanks. They do have several systems to handle liquid
radwaste.
Tne Detergent Drain System (DDS) handles low activity water i
from the showers and laundry area. This water is passed.through a single cartridge filter prior to release to the environment.. Samples for analysis are acquired prior to filtration.
The Waste Neutralizer System
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(WNS) handles chemical wastes and higher activity wastes.. Cask washdown
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normally is directed to the WNS as is floor drain inleakage.
A deep bed demineralizer is used to process this liquid waste stream.
This waste stream was used for the back flush of the shipping cask prior to the shipping of spent fuel to Harris (see Paragraph 6 'for additional details). The licensee stated, during discussions with ' the inspector, that this system was not designed to handle the high'. level of activity that the back flush generated; i.e.,
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associated with the WNS tanks are located in the same room as the tanks (unshielded), and these instruments need periodic maintenance; the
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pipes and pumps associated with these tanks are-located outside the
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shie'ided room containing the = tanks, next to a general-use passageway, creating an ALARA concern; and the piping to the deep bed demineralizer is i
located in an unshielded location.
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Two other liquid radwaste systems at Brunswick are designed to handle
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higher activity liquid radwaste.
These are the Reactor Water Clean-Up
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System (RWCU) and' the' Fuel Pool. Clean-Up System (FPCU). The RWCU system recirculates reactor water and would not be practical to use for the
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backwash of the shipping casks. The FPCU system has filers which are in -
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enclosed (shielded) cubicles and which can be backwashed and precoated remotely.
However, this system uses pumps and heat exchangers outside shielded areas that could potentially' develop hot spots. Also, this system was already in constant use.
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The Floor Drain System (FDS) was modified in the 1980's to include-an l
etched filter disk instead of a precoated filter.
This change was supposed to be an improvement, but small particles would wedge into the
openings of the etched filter disk. The disk would have to be backflushed to remove these particles, and af ter backflushing the particles would still be present in the system and would cause the same problems.
The repeated backflushing to keep the system running'would generate more waste
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water to be processed then originally needed to be processed. A plant I
modification to change back to the precoated filters had been initiated with a requested completion date in 1991. Currently, the FDS is connected to the WNS to alleviate this problem, j
.The Equimnt Drain Process at Brunswick handles valve leakage, and pump-l leakage, ihe liquid is processed and reclaimed for.the condensate storage system, which uses deepbed demineralizers and precoated filters.
Another plant modification to the liquid radwaste; system involves the auxiliary surge filters. This filter which is part of the equipment drain process is rarely used because of inherent design problems with the piping and valves.
This modification is scheduled for completion in 1992.-
The inspector. determined that there was approximately 50: gallons per minute (gpm) inleakage into the radwaste system from the plant and from i
backflushing of demineralizers.
The original design. for the plant was 30 gpm. Pre-1981, at which point Brunswick assigned a dedicated _ radwaste crew, the inleakage was 100 gpm.- The licensee representative : indicated i
that Brunswick had a maintenance program to identify inleakage, and that l
high priority is given to reduce'or eliminate the sources of.inleakage.
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The inspector spoke briefly with licensee representatives about a j
situation in their Phase Separator Tank Room, which is part of the-l radwaste system.
There is resin 'on the floor of this room, which is shielded and is locked as a high radiation area. This room contains only
tanks and piping, no pumps or valves, so there is not any required
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maintenance or reasons for personnel to enter. this room.
The licensee
stated that they did not have plans to. clean up the resin, due. to.
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prohibitive cost and ALARA concerns.
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The inspector reviewed graphs and data detailing the volume of radioactive liquid effluent released per month.
The monttly goal was 300,000 t
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The historical monthly average for If 87,1988 and 1989 was 1052,000 gallons, 1037,000 gallons and 825,000 gr.11ons, consecutively, l
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i The average for 1990 up to.and including May 1990 data,'was 311,000 gallons per month, with the May's discrete value being 47,000' gallons.
The inspector also reviewed information covering the number of curies released in the liquid effluent, including the tritium contribution. The monthly goal was 0.86 ' curies. per month.
The historical monthly average
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for 1987, 1988 and 1989 data was 0.63 curies per mont.h. The discrete value for May 1990 was 0.37 auries.
The inspector determined that -
Brunswick had established' goals and had worked at reducing the volume and total curie content released.in the liquid radwaste.
. The inspector also reviewed. five.1990 liq'uid waste re. lease permits. The examined packages appeared complete per procedural requirements.-
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No violations or deviations were identified.
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Contamination Control (84750)
During this inspection 11t was noted that, of the two entrances to the chemistry laboratory / count room facility only one of the entrances was equipped with a whole body standup frisker. The inspector questioned the validity of requiring workers -exiting from one-door-to f risk, while workers could exit the other door (the backdoor) without frisking.
The door without the frisker exited into an area used by the general plant
population to report to their work sites'and also was the pathway to the
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cafeteria. The licensee explained to the inspector that the door without the frisker, (the back door) was not intended for general passage, but as a means to allow technicians carrying potentially contaminated sampling equipment to exit the lab.
The door with the frisker exi+s irto an
office / desk area which exits into a hallway where radioactive material is not allowed.
The licensee alco explained to the. inspector that if a l
technician had been working with materials which.could cause them-to become contaminated, or if they-had been working in a hood, the technicians
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were expected to use the frisker before exiting"the lab. (i.e. - use the
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frisker and then walk back-through the lab with sampling apparatus and
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exit through the back door). Also, the licensee indicated that if there was a spill or other event which could cause the spread of contamination that there would be a high probability that it would be discovered by the use of the frisker since the majority of the. traffic exits through that
door.
The inspector determined that the chemistry' laboratory /countroom area was
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posted as a radiation area, except' for the hoods which are considered contaminated areas. Radiation areas do not require frisking upon exiting.
S e inspector reviewed E&RC Procedure 0110, Volume VIII, Revision 13 dated February 22, 1989, titled Monitoring Personnel for Contamination. This procedure required personnel to perform a whole-body frisk at the
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first frisker encountered af ter exiting a contaminated area.
However, personnel exiting f rom radiation control areas located in the outside areas of the plant (which included the Service Building which contains the
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chemistry laboratory /countroom), are not required to perform a frisk at
the immediate exit of these areas, as it may not be possible (i.e. no
frisker available). In this case a frisk is to be performed at the first
available frisking station.
The inspector determined that since the lab /countroom area was-considered a radiation area this procedure would apply, but that even if this area was considered a contamination. area, this procedure would allow personnel to exit without frisking.
The inspector also interviewed health physics personnel and determined
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that a radiological and smear survey were performed daily on the chemistry laboratory /countroom area.
The inspector reviewed the Daily Smear Survey. Record for several randomly
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selected days for the time period March through July 1990 and determined that the smear results recorded were all less than 200 disintegrations per minute per 100 square centimeters (dpm/cm2). The limit for contamination =
was anything greater than 2000 dpm/cm.
Verbal discussions with the
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i licensee revealed that, to the best of their knowledge, no contamination events had occurred at Brunswick as a result of material originating from i
the chemistry laboratory /countroom area.
The inspector discussed this situation with the licensee and indicated
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that while this did not seem to be a regulatory issue it did not appear to
be a good practice. The licensee indicated that-they would investigate actions to strengthen or improve the frisking practices for the. lab; including clarifying the procedures, adding " warning signs on the back door, and investigating the need to add additional frisking capabilities
in the lab area.
This situation will be reviewed by r9gional inspections during subsequent inspections.
No violations or deviations were identified.
10.
Counting Instrumentation (84750)
The inspector discussed a recent incident at a waste facility where an incorrect detector efficiency had been used due to source strength assumption errors for tha strontium-90 (Sr-90) radioactive standard.
The certificate of activity supplied by the source manuf acturer did not include the activity contribution from the strontium daughter product, yttrium-90 (Y-90), which was equal to that of the Sr-90. This resulted in an instrument beta efficiency that was twice the correct value and subsequently underestimated the beta activity by a f actor of two.
The inspector det.trmined that inplant count rooms did not use Sr-90 as a -
calibration source.
The licensee noted this -information for possible future use.
No violations or deviations were identified.
.
s-
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. 7
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.
.9
11.
Plant Chemistry (84750)-
t TS 3.4.4 specifies the limits that the reactor coolant system has to be maintained within for conductivity and chlorides. ' TS 3.4.5 specifies the limits for the specific activity of the reactor coolant.
These
parameters are related to corrosion resistance and fuel. integrity.
,
-
i Pursuant to these requirements the inspector reviewed water chemistry data
~
logs and data plots for June 1990, for Unit 1 and discussed various plant
chemistry items with licensee personnel.
This review' included E&RC i
Procedure 1000 Vol. VIII, Rev. : 021, dated June 15,.1989, titled the
'
Sampling and Anabsis Schedule for Technical Specifications Related
'
Radioactive and Nonradioactive chemi stry.
The results-of these conversations rad data review are discussed below.
Chloride concentrations in Unit I reactor, water had been maintained below five parts per billion (ppb) for the period.
This was well below the
,
20 ppb achievable value recommended by the BWR Owner's Group (BWROG).
during power operations..
Chlorides' accelerate intergranular stress-corrosion cracking (IGSCC) of austenitic stainless steels.
Reactor water conductivity in Unit 1 varied between 0.17 and 0.29
micromho/ centimeter (cm) for the period.. Although periodically exceeding the achievable value of 0.20 micromho/cm, conductivity had been maintained consistently below the BWROG's Action Level 1 recommended value of 0.30 micromho/cm and the levels specified by TSs and by the fuel warranty.
During June 1990, Unit I reactor coolant ' gross -activity ranged between 6.3 E-4 and 1.31 E-2 microcuries/ milliliter (uC1/ml).
Dose Equivalent Iodine-131 (DEI-131) at steady state power operations ranged from 2.86 E-06 to 2.46 E-03 uti/ml.
The inspector-determined that these various coolant parameters were within TS limits.
Chemistry
'
personnel indicated that there were no. indications of fuel leakers this
cycle for both Unit I and Unit 2.
There were four leakers
,
identified during the last cycle.for Unit 1.
l The licensee had projected operability dates for the Hydrogen Injection
'
System as on or before June 30, 1990, for both units.
As of this inspection, though hydrogen was being injected into the feedwater for both
'
units, the complete Hydrogen Injection System was still in the testing phase.
As a result, the hydrogen gas monitoring instruments in the
,
Augmented Offgas System were technically considered inoperable and the I
-
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,
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-
.
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12.~
!
licensee continued to remain in. an ' Action Statement of - the TS.
The inspector _ discussed. current projected schedules with the licensee. The
.
!
testing should be completed for this project by.the end of September 1990,-
although delays may be incurred due to' t;ie refueling outage on Unit 1.
!
'
An electrochemical, potential (ECP) of..-230 millivoltsL (mV)s is the recommended level for completely retarding crack; growth' in selds.
The:
licensee had determined - that, as core ;11fe increased,. the amount of-a a
-
hydrogen injected into the system needed to be_ increased to maintain-the J
ECP at or below - 230 mV. The inspector determined that<the ECP for both'
units was less ~ than -230 ' mV.
Injection rates of 1 app'oximately.~ 11.5'
a standard cubic feet per minute,(sefm) of._ gaseous hydrogen was required for -
'
Unit 2, while Unit-1 required approximately 16.3: scfm-gaseous: hydrogen.
l No violation or deviations were identified'in these areas.
l
,
,
Exit Interview f
'
I The inspection scope and results weret summarized' on July 27, 1990, with those persons indicated ' in Paragraph 1.. The inspectors.- described ' the areas inspected and discussed in detall the inspection results as listed
.,J in the summary. Proprietary information is not' contained in.this report.
~'
Dissenting comments were not received from the' licensee.
,
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.
~ Acronyms and Initialisms
-
1:
ALARAL Ai Low As Reasonably: Achievable:
l
-
2.
BWROG Boiling, Water Reactors Owner's: Group-
)
3.
CFR. Code of Federal Regulations.
4.
DDS : Detergent. Drain Systent >
l 5.
dpm/cm2. disintegrations,per minute p'er-centimeters'squ'ared'
.!
E&C Environmental and Chemis*:ry:.
.
'
7.
E&RC Environmental and Radiation Control-
'i Bi ECP electrochemical potential
'
9.-
FDS Floor Drain: System'
~
)
,10.
FPCU Fuel Pool Cleanup-
,
o,
-
Ell.
FSAR' Final-Safety Analysis Report-
,
'12. =gpm ' gallons per minuter j
>
13.
IFI l Inspector Follow-up Item W
,
' 14.
LC01 Limiting Condition of 0peration
'
'
15.
LLD lower limit of detection
'i
+
,
16.. mrem: millirem,
!
17. 'mV millivolt =
.
.
.
- i
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.
.18.
NRC Nuclear Regulatory Commission ~
.
19. OIL Open" Item List
20. ' PASS Post Accident Sampling. System i
21. LpC1/11ter picoeuries perL11terl,
..
22'
pC1/m3:picocuries per meter cubed
.
23.
PEU Performance Evaluation Unit 1
24. RWCU Reactor Water' Clean-up
'
25.
SFP Spent Fuel Pool.
26. TS Technical Specification
'3
>
27. WNS W&;te Neutralization System e
.
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ATTACHMENT 1
~
'
~
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~ TABLE 1 ~~
. * -
NRC-LICENSEE SAMPLE COMPARISON EVALUATION FOR BRUNSWICK STEAM ELECTRIC PLANT, JU!.Y 23-27, 1990 Concentration f uci/ unit)
.
Ratio'
gamp_le Isotope Licensee.
I!RC Resolution Licensee /NRC Comparl129
- Unit 2 c.
Pa rt icula te
'
Fi l te r Detector No. 49RB Cr-51 1.06 E-12
'1.4410.53 E-17'
3.
0. 7's Ag reement Ba-139 4.75 E-11 4.6910.37 E-11 13'
l1.01 Agreement S r-191
.3.58 E-12 2.6310.63 E-12 8: -
1.36 Agreement _.
- Ba-140 2.78s E-12 2.8tsio.40 E-12
0.96 Agreement
. La-1 ts0 1.86 E-12 2.5710.28 E-12
- O. T2 Ag reement Detector No - 1267
' C r-51 1.20 E-12 1.t:4to.53 E-12
0.83 Ag reement Ba-139 4.60 E-11 4.6910.37 E-11
0.98 Agreement S r-91 3.35 E-12 2.63io.63 E-12 as '
1.27 Ag reement Ba-Ilso -
2.56 E-12 2.88810.h0 E-12
0.90 -
Agreement-
.La-140 1.984 E-12-2.5720.28 E-12^
0.75'
Ag reement Unit No. 2-Reactor Coolant 50 ml Bottle Detector No. is9RB Cr-51-1.69 E-02 1.8210.04 E-02 is6.
0.93
' Ag reement'.
-
Co-58 5.92 E-01:
. 5.8210.883 E-04
1.02 Agreement'
Co-50 5.51 E-Ots '
6.08:10. t:5 E-Oik
0.91
_ Agreement W-18 7.
5.t8 E-04-7.0310.95 E-04.
7..
0.78 Ag reement 9.97 E-04.
1.18to.07 E-03
0.84~
Ag reemen t*
Sr-92.
5. 7'2 E-03:
5.93f 0.08e E-03
.148~
0.96 Ag reement -
Tc-99m-I-132 1.55 E-03 1.63io. M E-03-
0.95 Ag reement 1-133 5.72 E-Oas~
6.1110.31 E-Of:
0.9PJ Agreement Unit 2
-
Reactor Coolant 250 mt Bottle Detector No. 1267
~Cr-51 1.72 E-02 1.8210.08s E-02
0.9:e Agreement
Co-58 3.56 E-Ort
. 5.8210.as 3 E-O's its-0.96
-
Agreemect:
W-187
4.82 E-Ots:
6.0410.45 E-04
0.80 Agreement
_
Co-60 5.Sts E-04 7.0310.95 E-04-T 0.79 Ag reement -
-
Sr-92._.
1.07 E-03:
1.1810.07 E-03 --
11'
O.91 Ag reement Tc-99m-5.78:E-03"
,5.93to.0fs E-03 148 -
0.9T Ag reement
'l-132~
1.46 E-03.
. 1.6310.0C E-03-2T 0.90 Agreement
,11-133 5.5ts-E-04
~6.1110.31 E-04 20!
O.91 Agreement-
.
~
.
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__
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_
_
--.
~
=.
.
'
, i.
'
.
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-
Concentration f uCi/ unit)
.
' Licensee /ssRC
. Comparison Rat 10 Sample'
isottDe.
Licensee feRC Pesolut_io!}. "
-
Detector No. 1216 C r-51 1.66 E-02 1.82io.04 E-02
0.91 Ag reemen t..
Co-58 4.12 E-04 5.8210.43 E-04
0.71 Ag reement Co-60 4.2C E-04 6.0410.45 E-04-13 0.71.
Ag reement
-
,
W-187 7.13 E-04 7.03f0.95 E-04
1.01 Agreement Sr-92 9.70 E-04 1.18io.07 E-03
0.82 Agreement Tc-99=
5.92 E-03 5.9310.04 E-03 148 1.00:
Agreement 1-132.
1.64 E-03 1.6310.06 E-03:
1.01
- Agreement I-133 5.26 E-04 6.11iO.31 E-04
~20 0.86 Agreement-Liquid waste Tank Liquid Marineiti
_
Detector No. 49R8
'Cr-51 7.53 E-05 7.45t0.12 E-05 62.-
1.01 Agreement.
Mn-54 2.19 E-06 2.1210.14 E-06-
-1.03 Agreement.
Co-58 3.99 E-06'
4.03io.17 E-06-
0.99-
~ Agreement Co-50'
1.01 E-05 9.5310.26 E-06
.1.06
' Ag reement Np-239 2.92 E-06
.3.30f0.52 E-06
0.88 Agreement Tc-99e-1.65 E-05 1.61f0.02 E-05 80^
1.02.
_
Ag reement Detector No. 1267 Cr-51 7.72 E-05 7.45i0.12 E-05
'1.04 Agreement Mn-54 2.19 E-06
' 2.1210.14 E-06
1.03:
-Agreement Agreement-4.28 E-06 4.03f0.17 E-06-
1.06
.
.- -
Co-58-
' 9.99 E-06
. 9.5310.26 E-06-
1 0F Agreement
_
' Co-60'-
3.00 E-06-3.3010.52 E-06
-0.91 Agreement
-
'
--
Tc-9?e 1.66 E-05
.1.6110.02 E-05
1.03 Agreement -
7.51 E-05 7.4510.12.E-05 62-1.01-
_Ag reement
. Detector 1216'
Cr-51
. 2.07 E-Of 2.1210.14 E-06-
0.98
' Ag reement Mn-54
- Agreement 1.07'
Co-58 4.30 E-Oi,.
4.0310.17 E-06 24-'
1.07
.
00-60 1.02 E-05 9.5310.26'E-06 37-Agreement -
-
Np-239-3.65 E-06~
3.30f0.52 E-06-6
_1.11 Agreement
. Tc-99m 1.69 E-05 1.61to.02 E-05-
1.05
- Agreement Unit'No. 1 Cha rcos I -
Cartridge Detector _ No.L 49R8 -
I-131 7.23 E-13 5.89fD.59 E-13 to 1.23.
Agreement I-133 2.13 E-12 L1.8610.18 E-12'
1.14~
Ag reement Detector No. 126b l-131
- 7.62 E-13:
5.8910.59 E-13 to 1.29 Agreement t-133 2.03-E-12 1.8610.18E-12l 10.
- 1.09-Agreement
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_oncen%rstion t uC_ffunit]
Ratio
-
C u
,
Samp_l e
_f_S_otope Licensee NRC Resolution 1.Icensee/NRC C_ofpa rf son
.
Main Stack Gaseous Erriuent 1250 Mi t liter Gas
. Ma ri ne l l i
,
Detector No. 49RB Xe-135m 1.04 E-07 9.81 6.06 E-08
1.06 Agreement Xe-135 1.66 E-0T 1.51io.29 E-07
1.10 Ag reement Detector No. 1267 Xe-135m 1.02 E-07 9.81f6.06 E-08
1.04 Agreement Xe-135 1.28 E-07 1.51i0.29 E-07
0.85 Ag reemerit NRC Spiked Cha reca I Ca rt ridge Detector No. 49RB Co-60 5.38 E-02 4. T T 0.06 E-02
1.13 Agreement Cd-109 3.96 E-03 3.38t0.04 E-01
1 17 Agreement Sn-113 3.44 E-03 3.0810.21 E-03
1.12 Agreement Ce-139 2.70 E-03 2.25f0.09 E-03
1.20 Ag reement Co-57 5.23 E-03-4.71fo.11 E-03
1.11 Ag reement Y-88 4.80 E-03 4.7110.25 E-03-19 1.02 Ag reement CS-137
- 4.86 E-02 4.15f0.05 E-02
1.17 Agreement Detector No.
~ '67 Co-60'
5.33 E-02 4.77f0.06 E-02
1.12 Ag reement Cd-209 3.97 E-01 3.3810.04 E-01
1.1T Ag reement Sn-113 3.48 E-03 3.08to.21 E-03-
1.13 Agreement-Ce-139~
2.55 E-03 2.2510.09 E-03
1.13 Agreement Co-57 5.39 E-03 4.7110.11_E-03 43-1.14 Ag reement Y-88-4.86 E-03 4.7110.25 E-03
1.03
. Agreement Cs-137 -
4.68 E-02 4.15fo.05 E-02
1.13 Agreement Decector No. 1216 Co-60..
5.53 E-02 4.7710.06 E-02
1.16-Agreement Cd-109 4.03 E-01 3.3810.04.E-01
1.19-Agreement Sn-113 3.96 E-03 3.08f0.21 E-03
1.28 Agreceent
-
Ce-139 2.96 E-03 2.25f0.09 E-03
1.31 Agreement 00-57 5.46 E-03'
4.7110.11 E-03
1.16 Agreement
Y-88-3.22 E-03-4.71f0.25 E-03-19 1.10 Ag reement -
Cs-137 4.92 E-02 4.15f0.05 E-02-
1.18 Ag reement HP No. 1 Co-60'
.5.09 E-02
~4.7710.06 E-02
1.0T Agreement Cd-109 -
3.65 E-01.
3.3810.04 E-01
1.08 Agreement'
Sn-113 3.33 E-03 3.08f0.21 E-03
1.08 Agreement
'Ce-139 2.48 E-03 2.25fD.09 E-03
1.10 Agreement Co-5?
4.95 E-03.
4.7110.11 E-03
1.05 Ag reement Y-88 4.83 E-03 4.7110.25 E-03
1.02 Ag reement Cs-137 4.50 E-02 4.1510.05 E-02
1.08 Ag reement l
- - - _ _ _ _ - _ - -
--
-
-
~
- _ _ _ _ _ - - _ _ _ _ _ _ _
_
_
5'Z N
"
.i 4 O
_(
3.-
Concentration (uCi/onid
.
.
. Ratio
,,
_
Sample isotope Licensee NRC Resolution L.icensee/NRC Coccarison
HP No. 2 Co-60 5.17 E-02 4.77 0.06 E-02
1.08 Agreeent Cd-109 3.69 E-01 3.3810.01: E-01
1. 09 -.
Ag reement-Sn-113 3.60 E-03 3.0810.04 E-103
- 1.17-Ag reement Ce-139 2.58 E-03 2.25to.09 E-03-25 -
1.15 Agreement Co-57 5.29 E-03 4.7110.11 E-03
1.12 -
Ag reemen t -
Y-88 4.70 E-03 4.7110.25 E-03
'19 1.00 Agreement Cs-137-4.65 E-02 4.1510.05 E-02'
.83 1.12 Ag reement NRC Particulate f i l te r Sp i l<e
' Detector No. 49R8 Co-60 8.60 E-03 9.2610.26 E-03 36 ~
0.93
~ Ag reemen t -~
-
Cd-109 2.37 E-02 2.2610.10 E-02
1.05 Agreement ~
.C0-57 5.35 E-04 5.06do.35 E-04
1.06'
Ag reement Y-88 5.55 E-Of4 6.48iO.92 E-Of4
0.86 Ag reemen t ';
Cs-137 8.56 E-03 9.3210.21 r -03 -
fala en Agreement.
Detector No. 1267 Co-60 8.73 E-03 9.2610.26 E-03
'36 0.94 Agreement 2.f8 E-02 2.2610.1G E-02
_23 1.10 Ag reement Cd-109
-
5.63 E-Of4 5.0610.35 E-Of4
~ 1 24 1.11 Ag reement -
Y-88 5.45 E-Of4 6.4810.92 E-04
0.84 Ag reement Cs-137.
-9.20 E-03 9.32IO.21 E-03
0.99, Agreement Detector No. 1216 Co-60 -
18.94 E-03-
'9.2610.26 E-03
0.96 Agreement Cd-109 2.25-E-02 2.2610.01 E-02
1.00
- Agreement Co-57 5.83 E-04-
'5.06io.35 E-04
1.15-Ag reement 6.08 E-04 6.fsBi0.92 E-Of4
0.9fa Ag reement
'Y-88
.
8.62 E-03 9.3210.21 E-03
, isf4-0.92 Ag reement-Cs-137 Detector No. HP1 Co-60 8.81 E-03
.9.2610.26 E-03 J36-0.95 Ag reement Cd-109 2.32 E-02.
2.2610.10 E-02
1.03 Ag reement C0-57 5.92 E-of4 5.0610.35 E-Of4 1f4 1.17 Agreement Y-88 7.28 E-Of4 6. f4810.92 E-04 -
1.12'
Ag reement Cs-137 9.14.E-03'
9.3210.21 E-03-
. f4 4 _
0.98 Ag reement f
Detector No. - HP2
'Co-60 8.62 E-03-9.26to.26'E-03
0.93~
Agreement Cd-109 2.31 E-02 2.2610.10 E-02--
1.02 Agreement Co-57 5.48 E-Of4:
-5.0610.35 E-Of4 1f '
1.08 Agreement
'
Y-88 6.08 E-OfJ 6.4810.92 E-04
0.9fs Ag reement
Cs-137 8.72'E-03 9.3210.21 E-03 fsf4 0.9f4-Agreement '
- - - -
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3 ['
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,
ENCLOSURE 2 CRITERIA FOR COMPARISONS OF. ANALYTICAL MEASUREMENTS j
l This enclosure provides criteria for-the comparison of results of analytical I
radioactivity measurements.
These criteria are based on empirical l
relationships which combine prior experience in comparing radioactivity analyses, the measurement of the statistically random process of radioactive emission, and the accuracy needs of this program.
In these criteria, the " Comparison-Ratio Limits"2, denoting agreement or.
~
disagreement between-licenso and NRC results are variable.
This variability d
is a function of the. ratio of the NRC's' analytical value relative to its j
associated statistical and analytical uncertainty, referred to in this. program; t
>
as'"Resolutionna,
,
For comparison purposes, a ratio between the licensee's analytical value and.
the NRC's, analytical value is computed for each radionuclide present in a' given q
sample.. The-computed ratios are then evaluated'for. agreement or di;,ngreementi
!
based on " Resolution."
The corresponding-' values: for " Resolution" 'and the-
" Comparison Ratio Limits" are listed in the Table below.
Ratio values which
!
are either above or below the " Comparison Ratio Limits" are considered to be in disagreement, while ratio values within.or encompassed by 'the " Comparison Ratio
!
Limits" are considered to be in agreement.
TABLE l
NRC Confirmatory Measurements Acceptance Criteria
' Resolution vs. Comparison Ratio. Limits.
q Comparison Ratio-Limits i
Resolution
- for Agreement j
<4 0.4 - 2,5 4-7 0.5
'2.0 8 - 15 0.6 - 1.66 16 - 50 0.75 - 1.33 51 - 200 0.80 - 1.25
>200 0,85 --1.18
'
,
2 Comparison Ratio = Licensee Value NRC Reference Value.
i 2 Resolution = NRC Reference Value Associated Uncertainty
- i
.