ML20153D775
| ML20153D775 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 02/14/1986 |
| From: | Amy Hull, Mark Miller, Musolino S, Paolino R, Shanbaky M, Sherbini S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20153D760 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-2.F.1, TASK-3.D.3.3, TASK-TM 50-219-86-01, 50-219-86-1, NUDOCS 8602240246 | |
| Download: ML20153D775 (26) | |
See also: IR 05000219/1986001
Text
.
.
l
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-219/86-01
Docket No. 50-219
'
License No.
OPR-11
Category
C
Licensee: GPU Nuclear Corporation
P.O. Box 388
' Forked River, NJ 08731
l
Facility Name: Oyster Creek Nuclear Station
Inspection At:
Forked River, New Jersey
Inspection Conducted: January 13-17, 1986
l
Inspectors:
hl y 7?
_
- /(', [f6
M. ftrlier, hdiation Specialist
'date
$.
\\k
- lIhl 8 (o
S. Sherbini, rat'4 tion Specialist
' dste
NA6da.g--
a -t - f/-
R. K o~ lino, Lead Reactor Engineer
date
3 Nd b
SddG
-
A. P.'HulT, Broo[)1aven National Laboratory
6 ale
3'l W Akt-
2f6kG
S. V. Musolino, Brpkhaven National Laboratory
'da4.e
Approved by:
WN
~2M
/
M. $ha'nbaky,
f, Fac W ttes
b //'e
dat
Radiation Protection Section
!
Inspection Summary: Inspection on January 13-17, 1986 (Report No. 50-219/86-01)
Areas Inspected: Special, announced safety inspection of the licensee's imple-
mentation and status of the following task actions identified in NUREG-0737:
II.B.3, Post-accident sampling of reactor coolant and containment atmosphere;
i
II.F.1-1, Noble gas effluent monitors; II.F.1-2, Post-accident effluent moni-
!
toring; II.F.1-3, Containment radiation monitoring; and, III.D.3.3, In plant
radioiodine measurements.
The inspection involved 201 hours0.00233 days <br />0.0558 hours <br />3.323413e-4 weeks <br />7.64805e-5 months <br /> by three region-
based inspectors and two contractors from Brookhaven National Laboratory.
Results: No violations were identified. Several areas requiring improvements
and further review were identified.
B602240246
18
'
ADOCM 0
19
0
.
.
.
-
-
-
-
-
-
.
.
- -
-
-
-
.
.
DETAILS
1.0 Persons Contacted
1.1 General Public Utilities
,
J. Anderscavage,
Scheduling Supervisor
i
- W. Behrle,
Director, Start-up and Test
J. Bishop,
Start-up Engineer
I
- G. W. Busch,
Licensing Engineer
!
M. Buday,
Manger Plans & Programs
J. Carscadder,
Consulting Engineer
D. Chandler,
Engineering Process and Instrumentation
,
l
- W.
Duda,
Projects Manager
W. Dunphy,
Senior Chemist
l
- S. C. Gera,
Project Engineer
- P. B. Fiedler,
Vice President / Director
- C, J. Halbfoster,
Manager, Plant Chemistry
R. Hillman,
Senior Chemist
- B.
Hohman,
Licensing Engineer
!
T. Johnson,
Area Supervisor - Electrical
- R. W. Keaton,
Director Engineering Projects
A. Lewis,
Document Control Supervisor
i
M. Littleton,
Manager, Radiological Engineering
R. Parshall,
Administrative Support Supervisor
'
- M. J. Radvansky,
Manager, Technical Functions
- G. J. Sadauskas,
Manager, Instrumentation & Controls
- G. J. Simonetti,
Audit Manager
- J. Solakiewicz,
Manger, Quality Assurance and Systems
- J. Stevens,
Frocess Instrumentation
R. Stoudnour,
Senior Engineer
- J. L. Sullivan Jr. , Plant. Operations Director
- R.
L. Sullivan,
Mana er, Emergency Preparedness
- J. Thorpe,
Dirt.ctor, Licensing and Regulatory Affairs
- D. Turner,
Radiation Control Director
M. Wineberg,
Technical Functions Engineer
l
1.2 Nuclear Regulatory Commission
W. Pasciak,
Chief, Effluents Radiation Protection Section
B. Bateman,
Senior Resident Inspector, OC
J. Wechselberger,
Resident Inspector, OC
- denotes attendance at exit interview on January 17, 1986.
2.0 Purpose
(
The purpose of this inspection was to verify and validate the adequacy of
,
the licensee's implementation of the following task actions identified in
l
-
-
-_. -___ _ -
- _ _ _ .
_-
._ _
_ _ .
_. ..
._ _ - _ _ _ _ - _ _ _ _ _ - _ - _ - _ _ _ _ _ - _ _
.
"
3
NUREG-0737, Clarification of TMI Action Plant-Requirements:
Task No.
-Title
II.B.3
Post-Accident Sampling Capability
-II.F.1-1
Noble Gas Effluent Monitors
II.F.1-2
Sampling and Analysis of Plant Effluents
II.F.1-3
Containment High-Range Radiation Monitor
III.D.3-3
Improved Inplant Iodine Instrumentation under
Accident Conditions
As part of the inspection, a review was performed to verify and validate
the adequacy of the licensee's design and quality assurance program for
the design and installation of the Post-Accident Sampling System (PASS).
3.0 TMI Action Plan Generic Criteria and Commitments
The licensee's implementation of the task actions specified in Section 2.0
were reviewed against criteria contained in the following documents:
NUREG-0578, TMI-2 Lessons Learned Task Force Status Report and '
Short-Term Recommendations, dated July 1979.
Letter from Darrell G. Eisenhut, Acting Director, Division of
Operating Reactors, NRC, to all Operating Power Plants, dated
October 30, 1979.
NUREG-0737, Clarification of TMI Action Plan Requirements, dated
November 1980.
Generic Letter 82-05, Letter from Darrell G. Eisenhut, Director,,
Power Reactors, dated March 14, 1982.
Letter from Darrel G. Eisenhut, Director, Division of Licensing,
NRR to Regional Administrators " Proposed Guidelines for Calibration
and Surveillance Requirements for Equipment Provided to Meet Item
II.F.1, Attachments 1, 2 and 3, NUREG-0737" dated August 16, 1982.
Order confirming Licensee Commitments on Post-TMI Related Issues,
dated June 17, 1983.
Oyster Creek Nuclear Generating Station, Updated Final Safety
Analysis Report, dated December 1984.
Podifications of Confirmatory Order of June 17, 1983 for II.B.3,
Post-Accident Sampling System dated April 29, 1985.
Regulatory Guide 1.3 " Assumptions Used for Evaluating Radiological
Consequences of a loss of Coolant Accident for Boiling Water
Reactors.
s
ok
A
.m
_ _ _ _ _ _ _ ____ . _ _
_'m______._____._._.
___ _ ___.____ ___ _ _____ ._____
-_ . . . _ . _ _ _ . . _ . . _ . . _ _ . _ _ . . _ _ _ . _ _ _ _ _ _ _ _ .
_ _ .
_ _ _ _ _ _ _ _ _ _ .
..
. _
,
..
._
_
_
_~ .
_ _ _ . . . _.
. .
,
I
['
l
3
-
L
I
Regulatory Guide 1.97 Rev. 3, " Instrumentation for Light-Water-Cooled
i
Nuclear Power Plants to Assess Plant and Environs Conditions During
and Following an Accident.
-
Regulatory Guide 8.8, Rev. 3, '!Information Relevant to Ensuring that
Occupational Radiation Exposure at Nuclear Power Stations' will be
As low As Reasonably Achievable".
l
4.0 Post-Accident Sampling System, Item'II.B.3.
4.1 Position
NUREG-0737, Item II.B.3, specifies that licensees shall have the-cap-
ability to promptly collect, handle,'and' analyze post-accident samples
which are representative of conditions existing.in the reactor coolant:
and containment atmosphere. Specific criteria are denoted in commitments
l
to the NRC relative to the specifications. contained in NUREG-0737.
l
Documents Reviewed
The implementation, adequacy and status of the licensee's post-accident
sampling and monitoring systems were reviewed against the criteria 'identi-
fied in Section 3.0 and in regard to licensee letters, memoranda, drawings
and station procedures as listed in Attachment 1 of this Inspection Report.
The licensee's performance. relative to these criteria was determined from
interviews with the principal personnel associated with post-accident
sampling, reviews of associated procedures and documentation, and the con-
duct of a performance test to verify hardware, procedures and personnel
capabilities.
4.2 Findings
I
Within the scope of the review, the following items were identified:
4.2.1
System Description and Capability
The licensee has installed a Post-Accident Sampling System
which is a standard General Electric design.
It has.the cap-
,
!
ability to obtain undiluted and diluted unpressurized liquid
I
samples. They may be drawn from the reactor vessel through the
regular and through the liquid poison sampling lines, from the
shut-down heat exchange system and from the torus via the core-
~
spray system. Atmosphere samples can be obtained from.the dry-
.
well, suppression pool and reactor building (secondary contain-
I
ment). The PASS sampling cabinet and control panel are situated
in a room just outboard 'of the reactor building.
!
Analysis for radioactivity is conducted in an adjacent
l
laboratory using a Canberra Series 85_ high resolution system
l
l
'
.
-
-
,3
-
- i
'
-;
,.
,
W
s
-
'
I
~
4
.
~$' ,
~
,
.
5-
3
'
f~
,
with a Ge-(Li) detector and comput:arized MCA: system. -Ina' lysis
l
for chlorides, boron and hydregen-are also conducted"in an adja-
'
!
cent laboratory,.using an ion-chromatographic method, the
j
carminic acid method and gas chromatography, respectively.
The'
PASS also' includes a capability for on-line-conductivity mea-
l
surement. Analysis for pH is conducted using micro-electrode-
l
and a small aliquot of a 10-ml , undiluted sample.
x
-
r.
l-
The licensee.was originally committed in the Confirmatory Order
dated June 17, 1983, to having ;he PASS or,erational within 61
'
months after startup from the Cycle 10 refueling outage. Sub -
"
sequently, the licensee discovered leakage of a valve (40-29)
in the recirculation system sampling line, which required -
irolation in accordance with Technical Specifications, so was un-
able to fully test the system at that time.
)
i
A modification of this Confirmatory Order was made on April 29,
1985 to extend the date to no later than the planned shutdown
for October 1985. The valve in question had been repaired and
_the licensee completed operational testing on November 18, 1985.
Reactor coolant and drywell sampling have been conduded in-
which samples from the PASS have been compared with-these from
.
the normal sampling locations.
Flow tests through other sample
lines were not compared due to lcw levels of' radioactivity.
However; all. sample pathways were tested by. physical. techniques
(i.e. ficw ~cf demineralized water or ' freon -injected under pres-
sure at special test taps in sample lines).
.i
l
4.2.2
performance Test
f
,
i
Grab samo.l.es of reactor coolant and of the drywell atmosphere
f
were obtained in a performance test' for this inspection on
January 15, 1986.' During-the test, licensee personnel ver'ified
l
the integrated ability <to' collect and analyze samples within'the
tim,e constraints of NUREG-0737, II.B.3.
3
4.2.3
Sampling
4.2.3.1
ReactbrCoolant
l
The reactor coolant sampling system is designed to obtain
samples of liquids and dissolved gases during all modes of oper-
atton. The folicwing findings were noted:
The volume actually delivered by the ball valve in the
small-sample dilution procedure, which is specified by the
s
j
vendor to be 0.1 m1, has not been verified by.the licensee.
I
,I t
r'
i,
.e
f
'\\
rg
- \\
'
~
u
. ----- ---
-
?
-
- - -
-
,
.
.
6
The procedure for the drawing of.a sample of dissolved gas
(8.10, Appendix 3, Step 3.55.3) calls for.the operator to
" grab the knurled portion of the needle when removing' the
,
syringe".
It does not contain'a precaution against contact
wi.th the. portion of the needle which may become contamin -
ated during'the test.
Guide marks' have been improvised in pencil on the wall.
below and behind the PASS sampling cabinet to guide the
positioning of the large cart bearing the shield which ~is -
utilized for the undiluted sample procedure (8.10, Appendix.
2 and Appendix 3).
Although the indications of.the radiationLmonitor for-
liquid samples (RI-665) are utilized .in the procedures for
sampling'(8.10, Appendix 2 and 3) to assureLthat flushing
has taken place, it is not specifically referred to as an.
indication for the operator that a high activity sample has
been collected.
4.2.3.2
Containment Air
.
Atmosphere samples can be obtained from the.Drywell, Reactor
Building and the Torus. The following findings were noted:
The licensee procedure for taking an iodine and particu-
late sample (831.19, Appendix 5) calls ~for the determin -
ation of the flow by means of the reading of a rotometer
(FI-725, which is incorrectly designated in the procedure
as PI-175 and which also incorrectly calls for a flow-
reading of GPM, instead of SCFM). However, in correspon-
dence to another utility (T.A. Green,. Manager, Servicing
and Auxiliary Equipment Retrofits GF to T.H.-Wyllie,-
Manager Brunswick Engineering CP&L, dated April 6,1984), .
it is stated that this rotometer is used " strictly to
verify gas purge flow as the critical flow orifice is used as
the accurate flow measurement device during particulate and
iodine sampling".
The indications of the radiation monitor (RI-704 for .the '
particulate / iodine cartridge) are.not specifically called
out in the procedure to alert the operator that a high
activity sample has been collected.
The iodine sampling cartridge depends on a metal to metal l
contact of four individual in-line iodine filter canisters
under modest compression to prevent streaming past them.
-
-
-
.
- l
.
.-
7
4.2.3.3
Recommendations for Improvement
A.
Verify the volumetric-delivery of the ball valve for a
diluted sample.
B.
An appropriate caution against contact with the needle
should be added to the procedure for the drawing of a
sample of dissolved gas.
C.
Clearly visible guidance should be provided and described
in the procedure for the positioning of the large cart for
undiluted samples.
D.
The indications of the radiation detectors installed adja-
cent to the liquid sample and the particulate / iodine
sampling cartridge should be utilized to alert the operator
to the collection of high activity samples.
-E.
The procedures for the determination of the flow in
collection of particulate / iodine samples should be based
on the flow through the critical orifice, with an appro-
priate precaution that the appropriate pressure differen-
tial (approximately 0.5 cfm) is observed.
F.
The use of 0 rings between the canisters in the iodine
sampling cartridge should be considered, unless it can
otherwise be demonstrated that by pass leakage cannot
occur.
This item will be reviewed during a subsequen, inspection
(219/F6-01-01).
4.2.4
Analytical Capability'
The licensee's commitment relative to rcnge, sensitivity and
type of analytical capability as indicated in Appendix B, were
contained in its submittals of March 6 and July 13, 1984.
4.2.4.1
Preliminary screening for chlorides is performed using ion
chromatography.
In the event of interference due to a high
ratio of baron to chloride, a separation would be performed
and the turbidimetric method utilized. Backup off-site analysis
capability would be available through an agreement with B&W's-
Lynchburg, VA Laboratory. A shipping cask is available.
However, a certificate of conformance relating to the quality
assurance program for the cask was not documented.
.
S
,2
.
M
'
4
.
~
8
Chloride analysis was' satisfactorily conducted with the ion
-chromatography method. The results are contained in Attachment
2.
However, the. license stated in Procedure 824.9,
" Chemical
Instrumentation:
Ion Chromatography" that the-lower limit-for
detection was 0.1 ppb. This value could not be demonstrated
and apparently was a typographical error.
,
4.2.4.2.
Boron analysis of PASS: samples is performed using~the carminic
acid method. Mannitol' titration vould be used for low concen-
tration samples.
B'oron analyses was satisfactorily' conducted with the carminic
acid method.
The results'are contained in Attachment 2.
4.2.4.2
pH Analysis
Analysis for pH is conducted in a hood adjacent to the PASS
sampling unit using a micro-electrode that can utilize samples
as small as 0.1-0.3 ml. .The licensee demonstrated the cap-
ability of the micro-electrode using 0.1 m1 sample size. The
results are contained in Attachment 2.
4.3.3.4
Gross Gamma and Isotopic Analysi.i.
Gamma analysis of PASS samples'is performed using a Canberra
Series 85, computer based, high-resolution system with a
shielded. Ge-(Li) detec:or. An extensive library is utilized
which is sufficient to detect the nuclides of interest.
By the
use of dilution and small shielded ~ sample transport containers
with bottom apertures and an adapter on the detector shield,
the full range of anticipated concentrations can'be evaluated.
An isotopic analysis of the undiluted reactor coolant sample was
satisfactorily conducted.
The results of the comparison of the
.
PASS sample and the normal sink sample compared'within a factor
of two. The licensee had not completed its' site specific core
damage estimate procedure.
However, a methodology based on the
-GE core damage estimate procedure was available for iterim use.
4.2.4.6
Hydrogen and Dissolved Gas
Dissolved gas is determined by the GE PASS expansion method.
Hydrogen and/or oxygen content are evaluated by gas chromato-
graphy.
The licensee satisfactorily demonstrated the acility to collect
a dissolved gas sample and to perform hydrogen analysis with gas
chromatography. The results are contained in Attachment 2.
,
.
o-
1
.
9
,
4.~2.5
Additional Findings-
A.
Calibration and Maintenance
'According to the-licensee submittal of March 6,'1984
-(Criterion 10, Item 7), " Equipment used for post-accident-
sampling and analysis will be calibrated or tested approxi-
mately. every six months". .However, it could not be veri-
fied that a schedule for calibration or testing had in' fact
been established. 'It was noted.by the inspector that
several instruments on'the' PASS. panel had calibration
stickers'with dates a year or more old. Calibration
stickers for its radiation monitors were not evident. 'The
inspector was informed by licensee personnel that since
the PASS was used only infrequently,-regular calibration
was.not required and that a schedule would be established'
on the basis of experienced reliability.
Although the-inspector was informed'that'some spare ~ parts
for the PASS were available, a' list.of:them could.not be
provided by licensee personnel during theLinspection.
B.
Radiation Monitors
The value of and the basis for the alarm and warning set
points of the radiation monitors (Eberline RIIA) could not
be determined during the inspection. Also, initially after
the radiation monitors are energized, a "nc mal"-indication
is illuminated. However,'in low background fields,.it dis-
appears shortly thereafter (due' to the infrequency of the
pulses which trigger it).
C.
PASS Panel Indications
The licensee's procedures for the operation ofithe PASS
(831.10) instruct the operator to -verify. that selected
illuminated valve and. pump status indicators on the. PASS
panel lo~gic diagram have energized or de-energized. Other'
steps which also cause'a change in one or more indicators
that could be useful fo.- diagnostic ~ purposes are not called
out in the procedures.
4.2.5.1
Recommendations for Improvement
A.
Revise Procedure 824.9 to address l'ower limit of detection
'
capability.
B.
Ensure PAS-cask has been maintai.ned in accordance with-
Quality Assurance Program for ~ transport packages prior to -
use.
s
m
-
.
. .
.
. . . .
.
T
-.
,
_
10
~
-
C.
Complete site specific Core Damage Estimate Procedures.-
D.
A_ defined calibration and maintenance' schedule should be
,
devised.
~
~
E.
A spare parts inventory should be' documented.
F.
.The basis for the alarm and warning set points of'the PASS-
radiation monitors should be documented. ,,Also', the proce-
dures for the PASS should specify that-the operator observe
that they are operational when the PASS isLinitially-
er.ergized.
G.
To.' aid operators, the proper indications of theflights'on
the PASS control panel. logic diagram should be called ~to-
the operator's attention at-appropriate procedural steps-
where they should be energized or de-energized.
This item will be reviewed during a subsequent inspection.
(219/86-01-02).
,
l
5 .' 0 Noble Gas Effluent Monitor, Item II.F.1-1
5.1 position
NUREG-0737, Item.II.F.1-1 requires the installation of noble. gas monitors'
with an extended range designed to function during normal' operating and-
accident conditions. The criteria, including'the design. basis range of-
monitors for individual release pathways,' power supply, calibration and
other design considerations are set forth in Table II.F.1-1 of NUREG-0737.
l
Documents Reviewed
The implementatio.n, adequacy, and status of_the licensee's monitoring
systems were reviewed -against the criteria ' identified in Section 3.0
and in regard to licensee letters, memoranda, drawings'and station pro-
cedures as listed in Attachment 3.
The licenseek performance relative to these criteria was. determined by
interviews with the principal persons associated with the design, testing,
operation, installation and. surveillance of the high range' gas monitoring
systems, a review of the associate'd procedures and documeritation, an
_
examination of personnel qualifications and direct. observation of.the
system.
5.2 Findings
Within the scope of this review, the following.was identified:
.
>
,
~
.
.
_
,
-
,
%
y
-
- .
.
- 11-
.
~. -
~
5~.2.1
System-Description-
-The licensee purchased ~and . installed two Radioactive Gaseousi
-Effluent Monitoring' Systems (RAGEMS) supplied _ by Science Appli-
cations Inc. (SAI), Lone was to monitor ltheLeffluent1 released ~
from:the plant stack'and:the other was to monitor.the effluentsL
from the turbine building. TheyLwere originally: designed'and)
-intended to-perform _ monitoring ~and _ sampling:of plant effluents
L
in routine concentrations.
Due: to technical problems' - they ',
,
never became fully operational.
Following the promulgation of-
'
NUREG-0737 functions the system was modified byLthe' licensee;to1
,
perform-the'high range: functions.
1
The original system had been. designed to perform continuous;
on-line analysis 'of integrated samples of radioparticulates :and
radioiodines and to continu'ously-monitor and analyze forfradio
gases,-so as to determine;the' precise amount of.each isotope'
released.
It-includes three stagesr
a~ particulate: filter
.a.
. halogen _ filter and a~ noble gas channel. ;They were arranged'in
~
_
series with three high purity germanium detectors (HPGE).to-
perform the analyses. 'Both systems are controllsd by one
PDP-11/34' computer with a central terminal?for readoutfof
data.
The licensee modified the system'by deactivating the HPGE cap '
ability and switched over to assessment of the' noble gas activ-
ity using an ion chamber viewingEthe 6000 cc sample volume.
5.2.2
Findings
Credit for dilution has been assumed'in the licensee's~
- ontention that an upper range of 103pCi/cm3 (plant vent)
.s sufficient to meet the, requirements >of II.F.1-1.
However, the licensee.has not demonstratedLthis concen-
tration could not be exceeded.
It was not demonstrated that_the installed high and low
range monitors can provide range overlap.
The only calibration of the ton chamberithat has:been
~
performed to date has bein for one point, using Xe-133: gas.
An upper range of 103 ~pCi/cc was extrapolated from that
point.
Currently the data obtained from the ion ~ chamber,_
independent"of the time after shutdown, isireportedfas-
Xe-133. The energy responseEfunction of the detector for
higher photon. energies has not been determined. Those'.
responsible for:using the data'for dose-assessment are not,
'
cognizant of the' calibration method.
~
'
_
-
d
' a
' -
'
- '
.
.,
__._,,
j
( .
-
m
.
a
12
High concentrations of post-accident Noble gases may
" burn-out" the photo. multiplier tubes of the low-range
detectors, so that the system would not be able to follow
subsequent decreases.
It has not been demonstrated that the stack low range
monitor is adequately shielded from cross-talk, due to
other possible local radiation sources (see Section 6.0
for detailed description of other sources) such as by pass
filters, unshielded piping in the monitoring shack, shine
from the adjacent main stack, etc.
The turbine building RAGEMS does not include a low-range
monitor.
Only limited training on the operation and readout.of the
RAGEMS noble monitor has been provided. Currently only
four persons are trained to query the computer terminal
for data. During off hours a delay up to an hour is
possible before trained personnel would be available to
obtain data from the system.
Routine calibration and maintenance of the RAGEMS have not
been implemented, nor have procedures been developed.
5.3 Acceptability
Based on the documentation discussed during the inspection, the installed
system does not meet the requirements for high range noble gas monitoring
as contained in NUREG-0737, Attachment II.F.1-1.
Further documentation
and/or improvements are required as follows:
A.
The licensee should demonstrate that the current upper range cap-
ability of the installed gas monitors would not be exceeded in a
worst case accident.
B.
Calibration over multiple decades using transfer sources of varying
energy should be performed. The results should be incorporated into
the dose assessment function.
C.
A low range capability should be installed on the turbine building
monitor or it should be demonstrated that it is not required.
D.
The overlap of the high and low range monitors should be demonstrated.
E.
A method to deactivate the low range monitor near the upper bound of
its dynamic range and to reactivate it when the high range monitor
returns to the low end of its range should be devised.
, . .
.
.-
.
'
.
13
F.
A study on'the effects of'other nearby radiation sources on the
response,of the low range monitor should be made.
G.
Additional personnel should be trained so as to provide-
' round-the-clock". readout of the effluent monitors or a' simple -
readout should be provided to the. control room operators.
H.
Routine calibration and maintenance procedures should be provided
and training to the operational personnel be accomplished, such that
normal surveillance of the RAGEMS will be performed.
This item is considered unresolved and will be reviewed during'a subse-
quent inspection (219/86-01-03).
6.0 Sampling and Analyses of Plant Effluents, Item II.F.1-2.
'
6.1 position.
NURGE-0737, Item II.F.1-2, requires the provision of a capability for the
collection, transport, and measurement of representative samples of radio-
active iodines and particulates that may accompany gaseous effluents
following an accident.
It must be performed without exceeding specified
dose limits to the individuals involved.
The criteria including the
design basis shielding envelope, sampling media, sampit.ig considerations,
and analysis considerations are set forth in Table II.F.1-2.
Documents Reviewed
The implementation, adequacy and status of the licensee's sampling and
analysis system and procedures were reviewed against the criteria
identified in Section 3.0 and in regard to licensee letters, memoranda,
drawings and station procedures as listed in Attachment 3.
1
The licensee's performance relative to these criteria was determined by
interviewing the principal persons cssociated with-the design, testing,
operation installation, and surveillance of the systems for sampling.and
analysis of high activity radioiodine and particulate effluents, by
reviewing associated procedures and documentation, by examining personnel.
qualifications, and by direct observation of the systems.
6.2 Findings
Within the scope-of this review the following was identified:
6.2.1
System Description
Sampling of particulates and iodines'is performed sequentially
,
in the first two stages of RAGEMS. .Both stages can be manipu-
lated remotely from the computer terminal when filter changes
are required. However, entry into the sampling. shack is
,
-
- -
-
-
- -
-
-
- -
-
- - - -
- - -
-
_ _ _ _ _
_
_
_
_-
.
_
. - _ _ .
.-
.
14
l
required to initiate sampling thru RAGEMS and later to. retrieve
the filters after they have been automatically ejected from the-
sampling position. According to the licensee procedures (406.6)
RAGEMS would be placed in service by a change -in inlet valve
lineup only in the event of a high indication (> 105 cps) of the
normal gas monitor.
RAGEMS is not presently used for continuous
l
sampling.
RAGEMS is not used for routine sampling and a-flow of
l
1.5 CFM is routed through an unshielded by pass particulate and
iodine filters,.so as to prevent excessive amounts of activity-
to accumulate on the filters and thus making retrieval and
isotopic analysis suspect with regard to exposure constraints.
'
It is the licensee's plan, in the event of an-accident, that
readily retrievable.. filters will be placed on-line for two
seconds, and then retrieved for analysis. Given the sample flow
rate and the maximum concentration assumed by the licensee of
approximately 5 pCi/cc, isotopic analysis of filter cartridges
installed on a shielded and collimated holder, can be performed.
The 1.5 CFM sample. flow is isokinetic and the system has an
active means of adjusting the flow rate to account for changes
in stack flow over a limited range.
Findings
The licensee's capability to shield, transport and analyze
radioiodine particulate and gaseous samples within the
,
i
design basis range specified in Table II.F.1-2 is dependent
on the assumption that the plant effluents will be signi-
ficantly less than 100 pCi/cc.
Methods, training or procedures to perform representative
!
sampling of iodines and particulates, in accordance.with
Table II.F.1-2 were not demonstrated. The proposed two
,
l
second sample time appears inadequate to provide a repre-
'
sentative profile of the stack concentrations at the time
of sampling since it is doubtful that the flow through the
filter and sampling pipe would reach equilibrium in this
short interval. Possible sources of error include insuf-
ficient purge of air in the sample time due to the lack of
correction for valve opening and closing.
It does not.
I
appear adequate to colle:t a sufficient sequence of two
second samples to meet the requirements of II.F.1-2 for
continuous sampling.
Variations in plant parameters that could cause fluctua-
(
tions in stack flow are beyond the dynamic range of the
active flow control of the sampling system.
Procedures for
the resulting non-isokinetic flow condition, with appro-
priate corrections are not available.
l
l
,
. .
.
.
. .
.
.
_ __
-
.
_ _ - _ _
_
_
_ . _ - _ _
_ _ _ _ - -
- _
. - - _ _
_
[
.
[p
.
15
Entrained moisture could degrade the absorber under some
I
postulated accident conditions, since_the sampling lines
are not heat traced within the. sampling shacks, where the'
,
shack heaters serve as the heat tracing once;the lines
enter.
In the event of loss of off-site power the building
l
. heaters are not connected to the vital power bus or reli-
'
able source of backup power, leaving the inside lines
unheated.
A comprehensive time and motion exposure study to demon-
strate the sampling methods could be accomplished within .
the GDC-19 limits had not been made. A number of potential
radiation sources were neglected in the study that had been
performed but is in draft.
The licensee had not considered
the possible contribution of dose due to' shine from the
!
stack, unshield piping and water trap and the build up of-
high levels of iodines and in the bypass filter cartridges.
For example, if a maximum value of 10 pCi/cc is assumed in
the stack (a factor of 10 less flow is stated in NUREG-
0737) and the sample flows through the bypass filter for 30
minutes prior to the sample retrieval, the dose at one foot
from the bypass filter would be approximately 10 R in.three
,
minutes.
Adequate procedures and training to retrieve and analyze
iodine and particulate samples are not available.
The licensee had not implemented an appropriate routine
maintenance and calibration of the RAGEMS particulate'and
gaseous radiotodine sampling stages.
6.3 Acceptability
Based on the documentation discussed during the t'nspection, the licensee
had not demonstrated that the installed system meets the requirements of
'
NUREG-0737, Attachment II.F.1-2 and that samples can be obtained and
transported within GDC-19 limits.
'
An evaluation of representative sampling capabilities whenevei exhaust
flow occurs must be documented and the required improvements co..pieted as
follows:
A.
An appropriate site specific source term for release of.radiotodines
should be documented.
L
B.
The sampling r.;ethod should be redesigned to increase the sample time
to provide a representative sample.
C.
A procedure to apply appropriate correction factors during non-
isokinetic conditions should be provided.
-
-
. -
-
-
-
-
-
- -
r
-
I
=
!
'
'
16
D.
Heat tracing of the sample lines on vital power should be extended to
the sample flow paths within the sampling shacks.
E.
A comprehensive time / motion and exposure study to insure the GDC-19
criteria can be met for the retrieval and analysis of filters.
F.
Appropriate procedures should be provided and the needed training of
personnel conducted.
G.
Routine maintenance and calibration of the particulate and gaseous
radiciodine samples should be implemented.
This item is considered unresolved and will be reviewed during a
subsequent inspection (219/86-01-04).
i
7.0 II.F.1-3 Containment High Range Area Monitor
This system has not yet been installed but the components have been
i
'
purchased and some are onsite.
The inspection consisted in a review of
the design specifications and drawings, the manufacturer specifications,
and discussions with project engineers.
The system was found to conform
to the requirements of NUREG-0737, II.F.1-3 in most respects.
Some items
l
could not be confirmed at the time of the inspection and are as follows:
Documents and tests to certify that the detector cables and junction -
--
connections in the drywell are environmentally qualified for drywell
l
conditions during a postulated accident.
System drawings and layouts to verify that the proposed detector
--
locations are not close to any equipment or piping that may contain
radioactive fluids during an accident.
Such components may cause
interference in the detector's ability to respond to activity in the
t
drywell atmosphere.
Verification of the nature of the signal sent out by the detector
--
when the radiation field is below the lower limit of detection of
the system. This signal is to be used to
cation upon detector failure (219/86-01-05) produce a failure indi-
.
i
8.0 II.D.3.3 Airborne Iodine Sampling During an Accident
The ability to sample for iodine and to count the samples during an
l
accident were reviewed.
The onsite assembly areas reviewed include the
Operations Support Center and the Technical Support Center. Although the
capability to collect samples during an accident appears to be adequate,
!
some concerns were not resolved during the inspection and must be-
'
addressed in a later inspection.
These items are as follows:
!
9
h
_-
.
._
.
.
'- .
'
17
The' ability to count the air samples collected. Questions.in this
--
area relate to the availability of sufficient' counting systems to
handle the expected large volume of samples, as well as the suscepti-
bility of such systems to being disabled by high ambient' radiation
fields.
The exact lines of authority during an accident, including the mech-
--
anisms that would be used to initiate sample collection and the
assignment of priorities in counting those samples during an
accident (219/86-01-06).
9.0 Quality Assurance and Design Review
.
9.1 As part of the inspection effort a review was performed to verify and
validate the adequacy of the licensee's design and quality assurance pro-
.
gram for the installation of the Post-Accident Sampling System.
Documents Reviewed
The implementation, adequacy and status of the licensee's Post-Accident
Sampling and Monitoring System were reviewed against the criteria.identi-
fled in Section 3.0 and in regard to licensee correspondence, Specifi-
cations, Functional lests, Vendor Drawings and station procedures as
listed in Attachment 4A.
The licensee's performance relative to these criteria was datermined by
interviews with principal personnel associated with the installation and
Testing of The Post-Accident Sampling System.
9.2 Findings
The Post-Accident Sampling System has been classified by the licensee as
Nuclear Safety Related requiring installation in accordance with the GPU
Nuclear QA plan. Sample piping up to and including the second isolation
valve is designed and installed to seismic class.1 requirements.
Sample
piping beyond the second isolation valve is designed and installed in
accordance with ANSI-831.1 requirements.
Electrical power to the Post-
Accident Sampling System Control panel ER-19 comes from one of two thirty
ampere circuits in distribution panel PNL-PD-8.
Power to panel PNL-PD-8
is derived from the Safety Substation 182 throughLdistribution panel "D".
Panel PNL-PD-8 includes an undervoltage trip device to prevent re-activa-
tion of its electrical load on loss of off-site power, requiring manual
activation to put the Post-Accident Sampling System back on line.
The Post-Accident Sampling System, a Generic BWR LOCA Sampling System, has
incorporated all the changes / modifications identified by the manufacturer
and users of similar equipment at other installations.
Within the scope of this inspection, no violations or unresolved items
were identified.
.
4
.
.
-
.
.
.
. . -
m.
.
- .
. .
. .
-
.
.
.
_ _ _
__
_____ ___
_ _ _ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _
___ _ _ _-__-____.-__-_-_ -_ ___ _
'
.
i
18
10.0 Exit Interview
- _
'
The Post Accident Sampling'and Monitoring Team met with the licensee's
representatives at the conclusion of:the inspection on January 17, 1986.
.The Team Leader summarized the purpose, scope and findings of the
inspection. Dr. W. Pasciak informed Mr. P. Fiedler during r, subsequent
.
telephone discussion on January 24,;1986, that the findings,'as discussed-
'
during the exit meeting and as documented in Sections 5.3 and 5.6 of this
.
!
Report are considered unresolved items.
At'no time during the inspection was. written material provided.to the
"
,
licensee.
.
..
M
I
!
1
I
!
T
f
I
i
i
.
l
'
l
!
!
6
. .,
,
. . .
.
.
.m
.
.
.
.
.
. .
.
.
.
.
.
.
.
.
.
.
. - -
.
.
..
. .
..
.
m
..
-.
.
Attachment 1
Oyster Creek Nuclear Generating Station Procedures
823.1,
" Chemical Analysis:
pH"
--
823.2,
" Chemical Analysis: Conductivity"
--
823.7,
" Chemical Analysis: : Boron"
---
823.7.1, " Chemical Analysis: Boron"
--
824.1,
" Chemical Analysis: pH Meter"
--
824.2,
" Chemical Instrumentation Conductivity Bridge and Cell"
i
--
824.6,
" Chemical Instrumentation: Spectrophotometer, UV/VIS
--
(Perkin Elmer Lambda 1)"
824.8,
" Chemical Instrumentation: Gas Chromatograph"
--
824.9,
" Chemical Instrumentation:
Ion Chromatograph"
--
826.1,
" Radiochemical Instrumentation: Canberra Analysis System"
--
l
831.3,
" Post-Accident Sampling and Analysis Preparation and
--
Analysis", Revision 4, dated November 25, 1985.
831.9,
" Post-Accident Sampling and Analysis PASS' Analytical
--
,
Program", Revision 1, dated. December 12, 1985.
i
831.10,
" Operation'of the GI Post-Accident Sampling", Revision 3,
--
dated January 20, 1986
l
l
831.11
" Post-Accident Sampling and Analysis Cask Transport Off-
--
l
Site", Revision 6, dated November 26, 1985.
l
bysterCreekNuclearGeneratingStationDrawings
l
P&ID 3431-M0012, " Flow Diagram Post-Accident Sampling", dated
-
October 14, 1984.
!
Licensee Correspondence
P. B. Fiedler, V.P. Nuc. GPU, to D. G. Eisenhut,' Dir. 00L, dated
-
April 20, 1982.
'
P. B. Fiedler, V.P. Nuc. GPU, to D. G. Eisenhut, Dir. 00L, dated
-
~
June 15, 1982.
.
,
-
,.
..
_
._ .
_
- -
.
___
-
-
,
_- _.
._ _ ._-__
-9 - r
1.
. -
Attachment.1
. 2
i
--
D.'M. Crutchfield,' Chief ORB #5, to P. B. Fielder, V.P. Nuc. GPU,.
-
. dated June 30, 1982.
D. G. Eisenhut, Dir. 00L,~ to .P. R. Clark, Exec. V.P. GPU, dated
-
July 30, 1982.
.
D. M. Cru.tchfield, Chief ORB #5, to P. B. Fiedler, V.P..Nuc. GPU,.
-
dated October 10, 1982.
P. R. Clark, Exec. V.P. , GPU, to D. G. Eisenhut, Dir. ' DOL, dated
[
-
December 24, 1982.
-
- D. M. Crutchfield, Chief, ORB #5, to P. R. Clark, Exec. V.P.; GPU,-
dated January' 17, 1983.
'
P. B. Fiedler, Dir. 00L, to D. G. Eisenhut, Dir. DOL, dated
-
April 15, 1983.
-
P. B. Fiedler, Dir. DOL, to D. G. Eisenhut, Dir. DOL, dated
-
May 20, 1983.
D. G. Eisenhut, Dir. 00L, to P. B. Fiedler, V.P. Nuc. GPU, dated
-
February 10, 1984.
P. B. Fiedler, V.P. Nuc. GPU, to D. M. Crutchfield, Chie'f ORB #5,
-
dated March 6, 1984.
P. B. Fiedler, V.P. Nuc..GPU, to D. M. Crutchfield, Chief ORB #5,
-
dated March 16, 1984.
P. B. Fiedler, V.P. Nuc. GPU, to D. M. Crutchfield, Chief ORB #5,
-
dated July 19, 1984.
3
W. A. Paulson, Actg. Chief ORB #5,- to P. B. Fiedler, V.P. Nuc. GPU,
-
dated August 29, 1984.
P. B. Fiedler, V.P. Nuc. GPU, to D. G. Eisenhut,'Dir. DOL, dated
-
October 3, 1984.
P. B. Fiedler, V.P. Nuc. GPU, to D. G. Eisenhut, Dir. DOL, dated
-
October 10, 1984.
P. G. Loza, OC Fuel Projects Engineer to G. R. Bond, GPU Nuclear
-
Analysis and Fuels Director, dated December 11, 1984.
o
R. D. Gagliardo, Prog. Mgr., Burns-& Roe, to D. N. Green, Mgr. 0CNCS
-
Eng. Proj., dated March 20, 1985.
.
P. B. Fiedler, V.P. GPU, to J. A. Zwolinski, Chief 0RB #5, dated
-
April 22, 1985.
~
4
Y
E,._t_.-.__l_______
.
m
m.. ..
. .
.m.m
.
m
.
-
m
.
...
. ..
m
.
-
.
m. .
... .
. m
.
m .
.-
.
Attachment 1
3
Other Correspondence
T. A. Green, Mgr., Servicing and Aux. Equipment Retrofits, GE Co.,
-
to T. H. Wyllic, Mgr. Brunswick Engineering, CP & L, dated
April 6, 1984.
Vendor Manuals
NE00-24889, " Post-Accident Sampling System O&M Manual".
-
.
f
r
..
..mu -
-
1
.
.
Attachment 2
4
Attachment 2
Comparison of Chemical Analytical Test Results
Dilution
Meas.
Analyses
Licensee
Standard (ppm)
Factor
Conc. (ppm)
Results& Error (ppm) Commitments (ppm)
997.7
100
10.2
1020 + 22.3
+/-50(0 to 1000)
2981
1
3065
3065 + 84
+/-300(> 1000)
4895
100
4.8
4800 - 95
+/-300(> 1000)
l
Dilaticr.
'S :: .
A921yret
Licensee
l
Standard (ppm)
Factor
Conc. (ppb)
Results& Error (ppm) Commitments (ppm)
l
l
10.04
1000
10.7
10.7 7%
+/-10%(0.5 - 20 ppm)
l
30.11
10,000
3.7
37
<0.5 ppm
+/-0.05 ppm (0-0.5 ppm)
70.06
10,000
7.6
76
9%
+/-10%(0.5 - 20 ppm)
Ed
Standard
Analyses Result
Error
Licensee Commitment
6.198
6.3
+0.1
+/-0.3
Standard
Analyses Result
Error
Licensee Commitment
3.2%
3.1%
.3%
+/-10%
.
- -
.
-
.
.
Attachment 3
Documentation for NUREG-0737, II.F.1-1&2
Oyster Creek Nuclear Generating Station Procedures
831.7,
" Post Accident Sampling and Analysis:
Preparation and
--
Analysis", Revision 4, dated November 15, 1985.
831.4,
" Post Accident Sampling and Operations:
RAGEMS",
--
Revision 2, dated November 22, 1985.
EPIP-9,
"Off-site Dose Projections", Revision 5, dated
--
January 17, 1985
TP-300/0.1 MTX 138.9.2.1,
" Station 3 Ion Chamber Full Scale Range
--
Determination", Revision 0, dated ?
Correspondence
J. Knubel, Mgr. BWR Lic. , to D. Crutchfield, Oper. Rec. Br. #5, DL,
-
dated February 18, 1983.
P. Fiedler, V.P. to D. Eisenhut, Dir. 00L, dated February 23, 1982.
-
D. Muller, Asst. Dir. Rad. Prot., DSI to T. Novak, Asst. Dir. Lic.
-
DL, dated October 9, 1984.
H. Kister, Chief Proj. Br. No.1, Div of Reactor Proj., to
-
P. Fiedler, V.P. & Dir. DCNGS, dated March 27, 1985.
Licensee Internal Memoranda
J. Stevens to File, dated July 23, 1985.
-
G. Sadauskas to M. Laggart, Dated July 17, 1985.
-
J. Stevens to B. Hohman, dated November 11, 1985.
-
J. Stevens to Mgr. Lic., dated July 26, 1985.
-
J. Cline to S. Gera, dated May 4, 1984.
-
J. Stevens to S. Gera, dated July 9, 1984.
-
United Engineers Correspondence
J. Ucciferro, Proj. Mgr., to D. Chandler, dated January 15, 1986.
-
Scientific Applications Inc. Documents
Assessment of Radiation Dose Rate in the Oyster Creek Stack RAGEMS
-
Building, James Cline, SAI Inc., Rockville, Maryland, GPU Nuclear
Number 990-1214.
d
-
'
.
w,-
- - . - - - -
- -
- .
- - .
.
- - . - - - - .
. - .
. - .
- - -
-
- - -
.__. _
_
.
_ _ _ _ _ _ .
_ _ _ - - _ .
_ _ _ - _ _ _ _ .
_ _ _ .
_ _ _____ -
. -
1
.
.
. Attachment 4
4
Documentation for'NUREG-0737. II.B.3
5.1- Vendor Manuals
1.1.1
Post-Accident Sample Station, General Electric Company No.
NEOC-24889
,
5.2 Drawings
' 5.2.1
Burns and Roe Drawings
BR-M0012,
Revision 7, Flow Diagram Post Accident Sampling
System
BR-E0172,
Revision 4 Miscellaneous Power. Panels ~ Post -
Accident . Sampling System - Power Distribution
BR-E0215
Revision 2, Miscellaneous' Connection Diagram
General Electric LOCA Sampler Panel ER-19.
BR-E0366
Revision 1, Internal Wiring Otagram - General.
t
Electric LOCA Sampler Control Electric Panel
ER-19.
Field Change Request Nos. FCR-016469,
022503, 032919 & 032917.
,
t
BR-M0123,
Revision 2, Post-Accident Sampling tSO - Reactor
'
Building Containment Atmosphere Sample Supply.
BR-M0124,
Revision 2, Post-Accident Sampling ISO Reactor
Building Panel H 21T-A to TIP Room.
BR-M0126,
Revision 4,-Post-Accident Sampling ISO Reactor
Building Liquid Sample. Return to Waste Treatment.
BR-M0127,
Revision 2, Post-Accident Sampling ISO Reactor
Building Liquid Sample Return to Torus.
BR-M0129
Revision 3, Post-Accident Sampling ISO Reactor
- Building Reactor Cooling Sample Supply.
l
BR-M219,
Revision 1, Post-Accident Sampling ISO Reactor
Building Core Spray and Shutdown Cooling Sample
Supply.
BR-M0246,
Revision 0, Type 488 Typical Tubing Supports.
BR-H0254,
. Revision 2, Type 47A&B Typical-Tubing _ Supports
(120*F and up)~
'
n'. .
-
,
,
,
k
. .
m m.m
.m
. m um
.
m
m .
. m.
i
T~
.
.
Attachment 4
2
5.2.2
General Electric Drawings
GE-C5474-E-601,
Revision 2, Generic BWR LOCA Sampler
Electrical Schematic Diagram
GE-C5474-E-603,
Revision 3, Generic BWR LOCA Sampler
Electrical Connection Diagram
GE-C5464-E-607,
Revision 0, Generic BWR LOCA Sampler
Electrical Graphic Panel.
GE-C5474-E-101,
Revision 2, Generic BWR LOCA Gas Sampler
Mechanical Flow Diagram
GE-C5474-E-102,
Revision 3, Generic BWR LOCA Liquid Sampler
Mechanical Flow Diagram.
Field Disposition Instructions FDI No.
367-91700
5.2.3
Oyster Creek Generating Station Procedures
OCNGS Procedure No. 119, Revision 6, " Housekeeping".
OCNGS Procedure No. 120, Revision 10, " Fire Hazards".
OCNGS Procedure No. 112, Revision 15 "0yster Creek Calibration
of Maintenance Test and Inspection Tools, Gauges and
Instruments".
5.2.4
GPU Nuclear Technical Functions Procedures
Revision 0, "Startup and Test Program and Test
Requirements".
Revision 0, " Test Procedure Generation / Approval /
Change".
Revision 0, " Turn Over From Maintenance and
Construction and Test Performance".
5.2.5
Plant Modifications
1.2.5.1
BA-402048, Post-Accident Sampling Systems Phase II
5.2.6
Specifications
Installation Specification for Electrical Installation for
Post-Accident Sampling System - Phase II, OCIS - 402048-005
No. 399.00-9, Revision 1.
-
- -
.
.
.
.
Attachment 4
3
Irstallation Specification for Post-Accident Sampling System
Phase II (Mechanical) OCIS-399-11, Revision 2.
Division II, System Design Description for Post-Accident
Sampling System, 500-0C-555, Revision 0.
5.2.7
Technical Specification
Oyster Creek Nuclear Generating Station Technical
Specifications up to and including Amendment 74.
5.2.8
Test Procedures
l
TP 280/3, " Functional Testing for Post-Accident Liquid Sampling
l
Valves using Clean Water".
l
TP 280/4, " Functional Testing for Post-Accident Gas Sampling
Valves using Clean Gas",
'
i
TP 280/8 " Functional Testing of Post-Accident Sampling System
Miscellaneous Solenoid Operated Valves".
l
l
l
l
l
l
!
i
l
.
..
,
.
. . .
..
m
.
.
. .
. . .