ML17286B257
| ML17286B257 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 01/03/1992 |
| From: | Bocanegra R, Louis Carson, Yuhas G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17286B256 | List: |
| References | |
| 50-397-91-40, NUDOCS 9201220076 | |
| Download: ML17286B257 (17) | |
See also: IR 05000397/1991040
Text
U ~
S.
NUCLEAR REGULATORY COMMISSION
REGION V
Report
No.
50-397/91-40
License
No.
Licensee:
Washington Public Power Supply
3000 George Washington
Way
Richland, Mashington
99352
Facility Name:
Mashington Nuclear Plant
2 (WNP-2)
Inspection at:
WNP-2 Site,
Benton County, Washington
Inspection
Conducted:
October
28 to November 1, 1991
and
Decemb r 2-5,
1991
Inspection
by:
ocaneg
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a ia ion
cia is
a
e
igne
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on
p cia is
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Approved By:
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ie
Reactor Radiological
Protec ion Branch
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igne
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e
igne
Summar
Routine
announced
confirmatory measurements
inspection
and review
o
an inspection followup item.
Inspection procedures
84750
and 92701 were
used.
, Results:
This inspection
was started
on October
18, 1991;
however,
due to
anal
ed electronic
component in the Region
V mobile laboratory, the
inspection
was completed
on a subsequent visit to the site
on Oecember 2-5,
1991.
The results of the confirmatory measurements
intercomparisons
showed
that'he'icensee's
ability to accurate
sample
and measure radioactivity was
generally good.
The licensee's
quality assurance
program for radioactivity
measurements
was .adequate.
An item was identified for followup regarding
weakness
in the implementation of the Corporate
Chemistry Committee's oversight
responsibility.
Three other minor followup items were also identified,
involving: (1) intercomparison
results that showed
a slight bias,
(2) a liquid
effluent sample to be split and analyzed
by the licensee
and the
NRC contract
laboratory and (3) a questionable tritium lower limit of detection.
No
violations or deviations
were identified
PZOl22007b
92500039
ADOCK 05
8
,
I
[
0,
Persons. Contacted
Licensee
DETAILS
J.
Baker, Plant Manager
L. Morrison, Plant Chemistry Supervisor
R, Graybeal,
Health Physics/Chemistry
Manager
D: Pisarci k, Assistant Health Physics/Chemistry
Manager
, L. Mayne, Chemistry Operations
Supervisor
The above listed individuals that were at the exit meeting held on
December
5, 1991.
Fol.l owu
(92701)
(Closed)
Unresolved
Item 50-397/85-20-04:
By letter dated October 18, 1991,
e
licensee
prov>,
e
o
e
e p armed actions for improving the
Reactor Building Post Accident Grab Sampling System
(REA-SR-48).
The
need to improve the licensee's
post accident
sampling capability
was
previously identified as
a concern requiring a written response
by the
licensee
in inspection report 50-397/90-29,
as part. of followup on
this unresolved
item.
The licensee's
response
included the fo'}lowing items:
The licensee
agreed that the grab sampling system
needed
improvement.
The licensee
performed
an engineering
evaluation
and decided to
install
an in-line post accident
sampling
system with gamma
spectroscopy
capability.
The
new system is scheduled
to be operational
by the end of 1993.
In the interim,
a correction
factor will be used
when using the
cuirrent system.
By letter dated
November
1, 1991, the
NRC responded
to the licensee's
planned actions for improving the Reactor Building Post-Accident
Grab
Sampling System.
The
NRC found the plan to be acceptable
and will
inspect the system
as part of the routine inspection
program.
This
unresolved
item is considered
closed.
Confirmator
Measurements
84750
The Region
V mobile laboratory
was brought on-site to perform gamma-ray
spectroscopy
intercomparisons
with the licensee's
laboratory.
Sample
types
commonly analyzed to demonstrate
compliance with regulatory
requirements
were analyzed
by the
NRC and the licensee.
Table 1 contains
the complete results of the intercomparisons
performed during this
inspection.
The results
were compared
using the
NRC verification criteria
found in
NRC Inspection
Manual
Procedure
84750.
Attachment
A describes
the
acceptance
criteria.
I
'Sam ling
The inspectors
reviewed sampling proc'edures
and accompanied
licensee
technicians
when samples
were retrieved
and split.
The following .
weaknesses
were noted:
The procedure
used
by the technician to process
a liquid waste
sample did not contain
enough detai
1 to ensure
a
representative
sample
was obtained.
Procedure
12.5.3, "Liquid
Effluent Discharge Determination," instructs the technician to
"place 450 ml of sample into a 450 ml Harinelli beaker,
add
gael
agent,
mix thoroughly and seal
the container."
The-
snspectors
noted that the technician transferred
450 ml of
sample in small aliquots while continuously adding gel agent
and mixing.
Based
on discussions
with the technician, this
method
was
used
because
the sample did not gel uniformly when
the sample
was quantitatively transferred to the beaker
and
then gel agent
added.
The licensee
stated that the procedure
will be revised to clarify the sample preparation
steps.
The inspectors
accompanied
a technician to collect an offgas
'ample
using chemistry procedure
12.5.23,
"Recombiner
and
Charcoal Train Offgas Sampling
and Analysis."
The .technician
discovered that identification tags
on the offgas sampling valves
did not match the procedure
being used.
The technician
stopped
work and informed the inspector
of the problem.
A procedure
deviation was approved before sampling
was continued.
The inspector did.not identify any significant safety concerns-
regarding the sampling
and sample preparation activities observed.
b.
Gamma Isoto ic Anal sis
Six samples
were collected
and five were analyzed
by the licensee
and the Region
V mobile laboratory.
A pre-treatment
offgas sample
was not analyzed
due to high radiation
dose rate readings
on the
sample container surface.
The samples, analyzed
included:
Floor Drain Maste Tank - Liquid
Turbine Building Exhaust - Air Particulate
Turbine Building Exhaust - Charcoal. Cartridge
Reactor Coolant - Liquid
Reactor Coolant - Suspended
Particles
on Paper Filter
The licensee
generally
showed
good agreement with NRC results for all
sample
geometries
compared.
The large
number of disagreements
on
the first count of the filtered suspended
particles
sample resulted
from the wet paper filter sticking to the petri dish cover during
the licensee's
analysis.
After the problem was recognized,
the
licensee
recounted
the sample yielding good agreement.
The results
of the reactor coolant sample
showed
a slight bias.
The licensee
agreed to verify the calibration curve for this geometry
when they
receive
new calibration standards
in 1992.
The results of the geometry
verification will be followed as
an open item (50-397/91-40-01).
Other disagreements
encountered
included:
MN-56 in the reactor
coolant sample.
Due to the poor shape of
the 1810
keV peak,
the mobile laboratory analysis
software
'sed
the 846.8
keV* peak to quantify .Mn-56.
Mn-56 and I-134
both emit gamma-rays
at approximately
847
keV that.complicates
the quantitative analysis.
The inspectors
performed 'a manual
calculation to quantify Mn-56 using the 1810
keV peak.
The
results,
which are included in Tab'1e 1, resolved the
disagreement.
Ba-139 in the filtered suspended
particles
sample.
This
nuclide
has
an 84.9.minute half-life.
The licensee
recounted
this sample
when the filter stuck to the petri dish cover
during the initial count.
A sizable portion of the Ba-139
activ>ty had decayed
by the second
count giving too few counts
to make
an effective comparison.
W-187 and I-133 in the filtered suspended
particles
sample.
The disagreements
appear to be
due to poor counting statistics
attributable to the low activity levels in the sample.
The licensee
took a liquid radwaste
sample
(FDR-TK-6) on December
16,
1991 to split with the
NRC contract laboratory,
DOE Radiological
Environmental
Sciences
Laboratory
(RESL).
The samples will be
analyzed for Sr-89, Sr-90,
beta activity by
both
RESL and the licensee,
and the results
reported to
NRC Region
V.
This will be followed as
an open item (50-397/91-40-02).
The results of the confirmatory measurements
intercomparisons
indicated that the licensee's ability to sample
and measure
radioactivity was good,
and appeared
to meet the safety objectives
of the program.
ualit
Control for Measurements
The inspectors
reviewed
a sampling of calibration records for the.
licensee's
qermanium detectors.and
verified that they were
calibrated
>n accordance
with established
procedures
using standard
sources
traceable
to the National Institute of Standards
and
Technology (NIST).
The licensee participates
in a quarterly radiochemistry cross-check
program through
a vendor.
The vendor submits
samples
containing
an
unknown amount of radioactivity for the licensee
to analyze.
The
inspectors
reviewed results for the first quarter
1991.
The inspectors
found the cross-check
program to be effective in that the licensee
discovered
an incorrect quench curve for determining
Fe-55 in
effluent samples.
The inspectors verified that the licensee
submitted
a correction to their first semester
1991 Semi-Annual Effluent Report
to the
NRC reflecting the corrected
Fe-55 results.
The licensee
started
an in-house Blind Standards
Program in the
fourth quarter
1989.
The inspectors
reviewed the first quarter
1991
results for radiological
analyses
and concluded that the program
appears
to be adequate
for assessing
individual technician analytical
capabilities.
ualit
Assurance
Audits
The licensee's
Nuclear Operation
Standard
NOS-36 Section
6. l.d
tasks
the Director of Licensing
and Assurance with "providing
overview of chemistry through
an appropriate
level of gA
surveillances,
audits,
and technical
assessments
- Durin'g the last
NRC Confirmatory Measurements
inspection
(Report
No. 50-397/89-15,
the inspectors
raised
a c'oncern regarding
a
declining number of,surveillances
for plant process
water chemistry
activities., During the current inspection,
the inspectors
requested
Corporate
Licensing and Assurance
audits of the chemistry
laboratories
since the last inspection.
The inspectors
found that
equality Assurance
(gA) audits
were performed that indirectly involved
the licensee's
chemistry laboratories.
The inspectors
reviewed Plant
equality Assurance
Surveillance
Report
No. 2-90-100 related to a Na-24
.
injection test performed in August 1990,
The report documented
some
good findings, however,
the surveillance did not encompass
aspects
of
laboratory quality assurance
for radioactivity sampling
and measurements.
The quality assurance
program
was found to be generally adequate;
however,
the inspectors
noted that implementation of the licensee's
Corporate
Chemistry Committee's oversight of chemistry activities
appeared
to be weak and
may need additional
management
attention.
This concern will be followed as
an open item (50-397/91-40-03)
Procedures
Review
The inspectors
reviewed
a sampling of plant chemistry procedures
including:
7.4.4.5. 1 "Reactor Coolant Specific Activity Determination"
7.4.4.5.2
"Reactor Coolant Isotopic Analysis for I-131 Dose
Equivalent"
7.4.4.5,3
"E (E-Bar)"
'.4.4.5.4
"Isotopic Analysis of Offgas Sample"
12. 5. 4
12. 5. 3
"Particulate
and Charcoal Filter Analysis"
"Liqui d
Eff1 uent Discharge
Determinati on"
12.5.6
"Determination of E-Bar"
12. 5. 23
"Recombiner'and
Charcoal
Train Offgas Sampling
and
Analysis"
~"
II
12.5. 18
"Gross
Beta Determination"
3.2.5.29
"Samplinp Radwaste
Collector Tanks,
EDR-TD-2 or
FDR-TK-6
12.8.7
Chemistry Ortec
Gamma-Ray Analyzer System"
'12.5.33
"Reactor Coolant Sampling"
12. 11.3
"Lower Limit of Detection
and Data Reporting"
The procedures
reviewed appeared
adequate
to meet the safety
objectives of the chemistry progr'am.
4
e.
Lower Limit of Detection
WNP-2 Plant Technical Specifications
(TSs) require the licensee to
establish
the "a priori 'ower limit of detecti'on
(LLO) capability
for instruments
used to demonstrate
compliance with radioactive
effluent release limits.
The inspector
reviewed sel.ected
records
and verified LLD calculations
performed
by the licensee.
During an
in-office review of records,
the inspector
noted that in
determining the tritium LLD by liquid .Scintillation counting, the
licensee
used
a procedure that, although technically correct,
would
not have met the intent of the TSs for effluent samples
in cases
where
no tritium was detected.
The licensee
stated that at WNP-2,
tritium samples
normally contain activity levels several
orders of
magnitude higher than the
TS LLD.
Although there
appears
to be
no
safety concern,
the inspector will pursue this issue during
a
subsequent
inspection.
The licensee
indicated that they will also
followup on the inspector's
concern.,
This item will be followed as
'an open item (50-397/91-40-04)
No violatiens or deviations
were identified in this program area.
4.
Exit Meeting
One of the inspectors
met with licensee
management
on December.5,
1991,
to discuss'he
scope
and major findings of the inspection.
The
inspector informed the licensee that based
on the results of the
confirmatory measurements
intercomparisons
and document reviews, the
licensee's ability to sample
and measure radioactivity appears
to be
good.
The inspector also stated that the licensee.'s
quality assurance
program for radioactivity measurements
was adequate,
but corporate
oversight appeared
to be weak.
The licensee
acknowledged
the findings
presented
at the meeting.
0
~
'
TABLE 1
U.
S.
NUCLEAR REGULATORY COMMISSION
CONFIRMATORY MEASUREMENTS
PROGRAM
T
FACILITY:
WNP-2
FOR THE 4th
QUARTER OF 1991
SAMPLE
NUCLIDE
NRC VAL.
NRC
UNC.
LIC VAL.
LIC UNC.
RATIO
RESOL.
RESULT
Waste
Tank
Zn-65
(Tank 9)
3.90E-07
2.52E-08
3.93E-07'.65E-08
1.60E-06
6.26E-08
1,85E-06
1.37E-07
1. 01
1. 16
15. 5
25. 5
47
mm
Filter
(Tea)
47
mm
Filter
(Tea)
2nd Cnt
Ba-139
Sr-91
1-131
I-133
Cs-138
BA-139
Sr-91
1-131
1-133
CS-138
1.87E-ll
1. 68E-12
1. 23E-13
9. 58E-13
1.85E-ll
2.20E-ll
1. 83E-12
l. 46E-13
1. 05E-12
2. 17E-11
3. 98E-13
8. 08E-14
2. 55E-14
4. 92E-14
2. 08E-12
4. 67E-13
8. 79E-14
2. 93E-14
5. 38E-14
2. 44E-12
2. 48E-ll
1. 85E-12
2. 17E-13
1.22E-12
1.78E-11
2. 48E-ll
1. 85E-12
2. 17E-13
1.22E-12
1. 78E-11
3. 22E-13
.9. 12E-14
4. 34E-14
7. 36E-14
4. 32E-13
.
3. 22E-13
9. 12E-14
4. 34E-14
7. 36E-14
4. 32E-13
l. 32
1. 10
1..77
1. 27
0. 96
1. 13
1. 01
l. 49
1. 17
0. 82
47. 1
20. 8
4.8
19. 5
8.9
47. 0
20. 8
5.0
19. 5
8.9
A
A
'
A
A
A
A
A
A
A'harcoal
1-131
(Tea ¹1) l-133
1-135
5. 35E-13
3. 41E-14
5. 26E-13
5. 33E-14
1. 76E-12
5. 59E-14
1. 84E-12
7. 18E-14
1. 39E-12
1. 90E-13
1, 15E-12
1. 69E-13
0. 98
1. 04
0. 83
15. 7
31. 5
7.3
Reactor
Coolant
MN-56
Zn-65
Zn-69m
Sr-91
Sr-92
Y-92
I-132
I-133
I-134
I-135
Cs-138
6.14E-03
1.43E-04
6.26E-02
6.76E-04
4.02E-04'.58E-03
6.37E-03
2.98E-03
1.55E-04
4.73E-03
2.55E-03
2.39E-02
6. 62E-03
2. 55E-03
2 51E-05
5.62E-06
1.53E-03
1.69E-05
7.72E-06
1. 17E-05
9. 07E-04
8.78E-05
6.53E-06
1.76E-05
9.03E-06
4. 61E-03
4. 60E-05
1.60E-04
1. 11E-03
1.52E-04
4.93E-02
5.58E-04
3.58E-04
1.30E-03
4.82E.-03
2.58E-03
2.05E-04
4.34E-03
2, 18E-03
1.67E-02
5.30E-03
2.48E-03
1. 15E-04
2. 33E-05
7. 68E-03
9. 72E-05
3. 36E-05
7.77E-05
1.37E-04
2.26E-04
6.88E-05
7.73E-05
4.77E-05
6.45E-04
1.82E-04
5.28E-04
0.18 244.2
1. 06
25. 5
0. 79
41. 0
0. 83
40. 0
0. 89
52. 0
0.82 135.2
0.76
7.0
0. 87
33. 9
1. 33
23. 7
0.92 268.9.
0.85 282.7
0.70
5.2
0.80 144.0
0. 97
15. 9
A = Agreement
D = Disagreement
~
'[
t
Sample
NUCLIDE
NRC VAL.
NRC UNC.
LIC VAL.
LIC UNC.
RATIO
RESOL.
RESULT
Crud
Filter
(SJ-S)
Initial
Count
Crud
Filter
(SP-8)
Count 82
Ba-139
Cr-51
Co-58
Zn-65
M-187
Mn-56
Sr-91
Sr-92
Y-92
I-133
Zn-69
Ba-139
Cr-51
~
Co-58
Zn-65
M-187
Mn-56
Sr-91
Sr-92
Y-92
I-133
Zn-69
2. 05E-04
1.65E-05
4.67E-05
5.50E-03
3.48E-06
3. 81E-05
6.55E-06
3.00E-06
8.83E-05
5.02E-05
1.81E-04
1.84E-04
2.56E-06
6.62E-07
8.17E-OG
2.05E-.,04
1,65E-06
4.67E-05
5.50E-03
3.48E-06
3.81E-05
6.55E-06
3.00E-06
8.83E-05
5.02E-05
1. 81E-04
1.84E-04
2.56E-OG
6.62E-07
8. 17E-06
1.33E-05
3 '0E-07
2.79E-06
1.05E-04
2:45E-07
8.30E-07
8.58E-07
2.72E-07
5.29E-OG
5.80E-07
2.40E-04
8.40E-06
2.41E-07
1.75E-07
3.27E-07
1.33E-05
3.50E-07
5.58E-OG
1.05E-04
2.45E-07
8.30E-07
8,58E-07
2.72E-07
5.29E-06
5.80E-07
2,40E-06
8.40E-OG
2. 41E-07
1.75E-07
3.27E-07
1. 91E-06
1.39E-05
'3.77E-05
4.,10E-03
<3.50E-04
2.82E-05
<8. 10E-06
<1. 90E-06
6. 17E-05
3. 47E-05
1. 38E-04
7. 38E-05
<1. 70E-06
<1. 70E-06
6.57E-06
2. 89E-04
1.99E-05
6.32E-05
5.73E-03
3.57E-OG
4. 13E-05
<8. 50E-06
3.46E-08
9. 12E-05
5.47E-05
1. 91E-04
2. 34E-04
4. 61E-06
1.44E-06
8.67E-06
1. 91E-06
1. 39E-07
3. 77E-07
4.10E-05
O.OOE+00
2.82E-07
O.OOE+00
O.OOE+00
6. 17E-07
3.47E-07
1.38E-06
7.38E-07
O.OOE+00
O.OOE+00
6.57E-08
2.89E-06
1.99E-07
6.32E-07
5.73E-05
3.57E-OS
4. 13E-07
0:OOE+00
3.46E-OS
9. 12E-07
5. 47E-07
1.92E-06
2.34E-06
4. 61E-08
1.446-08
8.67E-OS
0. 93
0. 84
0. 81
0.75
0. 00
0. 74
0. 00
0. 00
0. 70
0.69
0.76
0.40
0.00
0. 00
0. 80
l. 41
1. 20
1. 35
1. 04
1. 03
1. 08
'.00'.
15
1. 03
1. 09
1. 06
l. 27
1. 80
2. 18
l. 06
15. 4
47. 2
16. 7
52.4
14. 2
45. 9
7.6
11. 0
16. 7
86. 6
75. 4
21. 9
10. 6
3.8
25. 0
15 ~ 4
47. 2
8.4
42. 4
14. 2
45. 9
7.6
11. 0
16. 7
86. 6
75. 4
21. 9
10. 6
3.8
25. 0
0
D
0
0
D
D-
D
D
A
D
A
A
A
A
A
D
A
A'
A
A
D
A
A
Reactor
Coolant
(Manual
Calc)
Mn-56
1.32E-'3
3.43E-05
1. 11E-03
1. 15E-04
0.84
38. 5
A = Agreement
'0 = Disagreement
l
W,
t
x
ATTACHMENT A
Criteria for Accepting the Licensee's
Measurements
lt
Comparison
I'.
Divide each
NRC result by its associated
uncertainty to obtain the
resolution.
The uncertainty is defined
as the relative standard
deviation,
one sigma, of the
NRC results
as calculated
from the
'ounting statistics.
b..
Divide each licensee result by the corresponding
NRC result to
obtain the ratio.
c.
The licensee's
measurement
is in agreement if the value of the
ratio falls within the limits shown in the following table for the
corresponding
resolution.
2.
Criteria
Resolution
(4
4-7
8-15
16-
50
51 - 200
)200
Ratio
0.50 - 2.00
0.60 - 1.66
0.75 - 1.33
0.80 - 1.25
0.85 - 1.18
1
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If
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II