IR 05000255/1989009
| ML20245K690 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 04/21/1989 |
| From: | Axelson W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| NUDOCS 8905050133 | |
| Download: ML20245K690 (1) | |
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Docket No. 50-255 APR 211989
l Consumers Power Company ATTN:
David P. Hoffman Vice President Nuclear Operations
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1945 West Parnall Road Jackson, MI 49201 Gentlemen:
This refers to the routine safety inspection conducted by Messrs. E. R. Swanson, J. K. Heller, and M. A. Kunowski during the period of March 14 through April 10, 1989, of activities at the Palisades Nuclear Generating Plant authorized by NRC Provisional Operating License No. DPR-20, and to the discussion of our findings with Mr. G. B. Slade and others of your staff and at the conclusion of the inspection.
The enclosed copy of our inspection report identifies areas examined during the inspection. Within these areas, the inspection consisted of a selective examination of procedures and representative records, observations, and interviews with personnel.
During this inspection, certain of your activities appeared to be in violation i
of NRC requirements, as specified in the enclosed Notice. A written response is required.
The responses directed by this letter and the accompanying Notice are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511.
In accordance with 10 CFR 2.790 of the Commission's regulations, a copy of this letter, your response and the enclosed inspection report will be placed in the NRC Public Document Room.
We will gladly discuss any questions you have concerning this inspection.
Sincerely, W.
. AR i s il t Projects Branch 2
Enclosures:
1.
Inspection Report
No. 50-255/89009(DRP)
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See Attached Distribution
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Consumers Power Company
Docket No. 50-255
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As a result of the inspection conducted on March 14 through April 10, 1989,.
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and in a
REGION III==
Report No. 50-255/89009(DRP)
Docket No. 50-255
License No. DPR-20
Licensee: Consumers Power Company
212 West Michigan Avenue
Jackson, MI 49201
Facility Name:
Palisades Nuclear Generating Plant
Inspection At:
Palisades Site, Covert, Michigan
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Inspection Conducted: March 14 through April 10, 1989
Inspectors:
E. R. Swanson
J. K. Heller
M. A. Kunowski
Approved By:
.L.
urggts, Chief
f/ 2//'77
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Reactor Projects Section 2A
Da'te
Inspection Summary
Inspection on March 14 through April 10,1989 (Report No. 50-255/89009(DRP)
Areas Inspected:
Routine unannounced inspection by the resident inspectors
of: actions on previously identified items; operational safety verification;
radiological controls; maintenance; surveillance; security; safety assessment /
quality verification; reportable events; Engineered Safety System walkdown;
and allegation followup.
Several SEP open items (Safety Issues Management
System (SIMS) items) were reviewed.
Results: Of the areas inspected, one violation was identified (leak testing
of the containment escape lock - Paragraph 6.f).
The inspection did not
disclose any notable weaknesses in the licensee's programs. The inspection
noted strengths in the areas of operator turnovers and the licensee's ability
to maintain black board conditions during most of the period.
No new Open
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Items and/or Unresolved Items were identified.
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DETAILS
1.
Persons Contacted
Consumers Power Company
G. B. Slade, Plant General Manager
+J.
G. Lewis, Technical Director
R. D. Orosz,-Engineering and Maintenance Manager
R. M. Rice, Operations Manager
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+W. L. Beckman, Radiological Services Manager
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+R. E. McCaleb, Planning and Administrative Director
+R. A. Fenech, Operations Superintendent
+K. E. Osborne, Projects Superintendent
+K. M. Hass, Reactor Engineering Superintendent
+R. P. Margol, QA Administrator
R. M. Brzezinski, I&C Superintendent
L. J. Kenaga, Staff HP
+C. S. Kozup, Licensing Engineer
+J. R. Brunet, Licensing Analyst
D. J. Malone, Licensing Analyst
R. J. Frigo, Operations Staff Support Sur,ervisor
U.S. Nuclear Regulatory Commission (USNT.CJ
+E. R. Swanson, Senior Resident Inspector
+J. K. Heller, Resident Inspector
+ Denotes those present at the Management Interview on April 12, 1989.
Others members of the plant staff, and several members of the Contract
Security Force, were also contacted during the inspection period.
2,
Actions on Previously Identified Items (92701, 92702)
a.
(Closed) Open Item 255/04S09-00(DRP): Effects of charging line and
letdown system pipe breaks. The NRC documented acceptance of the
licensee's analysis, submitted May 31, 1985, which demonstrated that
the subject pipe breaks need not be postulated in a Safety Evaluation
dated February 4,1987.
b.
(0 pen) Open Item 255/04S10-00(DRP):
Completion of evaluation of
specific components for seismic design.
Consumers Power responded
to the NRC's questions on April 30, 1987.
The NRC's review of
this issue is tracked under SEP Topic III-6 and Unresolved Safety
Issue A-46.
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c.
.(0 pen) Open Item 255/04S12-00(DRP):
Evaluation of specific
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structural loads referenced in SEP Topic III-7.B and IPSAR 4.12.
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The NRC is reviewing the licensee submittals.
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d.
(0 pen) Open Item 255/04S15-02(DRP): Consumers Power committed
to submit revised Technical Specifications for primary coolant
leakage detection systems inside containment as documented in the
IPSAR 4.15.2 (NUREG-0820). As committed in their January 24, 1989
letter, the licensee will submit revised Technical Specifications
as part of their restructured Technical Specification effort.
e.
(Closed) Open Item 255/04S28-02(DRP):
Install forced coeling
for inverter and charging cabinets and AFW junction boxes. The
inspector verified that fans had been installed in the inverter
cabinets.
The NRC is preparing a Safety Evaluation of this topic
(SEP IX-5, IPSAR 4.28.2).
f.
(Closed) Open Item 255/86035-08(DRP): Develop a long term solution
to the freezing of the stack gas flow recorder sensing lines. A
monthly periodic activity control (PAC) has been generated to
periodically blow down the sensing lines.
PAC RIA-002 has been
performed on a monthly basis (+/- 25 percent) since May 1988.
In
a letter to File PR12/01/86A-258 the system engineer stated that
this is an acceptable alternative to modifying the sensing lines
or revising the piping.
g.
(Closed) Open Item 255/86035-10(DRPj: The licensee will evaluate
the flow characteristic of Auxiliary Feedwater (AFW) flow control
Valves FCV-0736A, FCV-0737A, FCV-0727 and FCV-0749 to determine the
methodology to improve flow control below 100 gpm.
The licensee
chose to install per FC-789 a 1 and 1/2" bypass flow control valve
for the
"C" AFW pump.
The inspector reviewed the control room
prints and confirmed that the prints reflected the Facility Change.
Also, the inspector discussed the modification with control room
operators and confirmed that they had been trained on the
modifications. An Engineering review team has selected FC-789
for additional reviews.
Items identified by that review are
discussed in Inspection Report No. 255/89007.
h.
(0 pen) Open Item 255/86035-12(DRP):
Replace AFW Valve CV-0521
because internal valve leakage is causing a slow rotation of the
Turbine Driven Auxiliary Feedwater (TDAFW) pump, The valve was
repaired by the vendor per Work Order 24706677.
Internal
Correspondence DAB 89*002 stated the valve was repaired satisfactory
and is leak tight. A tour of the AFW pump room on March 22, revealed
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that the TDAFW still has a slow rotation. The inspector discussed
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this with control room operators and found that at the start of
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operations subsequent to the 88 refueling outage that the pump was
not rotating. The inspector has asked the licensee to revisit this
item and determine if additioral actions are required.
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1.
(Closed) Open Items 255/86035-18, -19 and -20(DRP):
Resolve the
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component cooling water pumps mechanical seal leakage problems.
During the 88 refueling, the
"C" pump was modified at the vendor
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facilities per SC 86-269 to accept a different mechanical seal.
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In addition, modified sleeves and nuts were installed on the pump.
Internal Correspondence GJS 89*002 stated that the repairs appear
successful in eliminating the seal leakage problem.
The licensee
plans to modify'the "A" and "B"' pump during an upcoming refueling
outage.
J.
(Closed) Open Item 255/88020-05(DRP):
The thermal overload on one
phase of the Emergency Diesel Generator starting air. compressor
motor was tripping.
It was found that the overloads were recently
replaced with a new style and one phase required a minor calibration.
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No similar problems have recurred.
No violations, deviations, unresobed or open items were identified.
3.
Operational Safety Verification (71707, 71710, 42700)
Routine facility operating activities were observed as conducted in
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the plant and from the main control rooms.
Plant startup, steady
power operation, plant shutdown, and system (s) lir.eup and operation
were observed as applicable.
The performance of licensed Reactor Operators and Senior Reactor Operators,
of Shift Technical: Advisors,.and of auxiliary' equipment operators was
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observed and evaluated including procedure use and adherence, records
and logs, communications, shift / duty turnover, and the degree of
professionalism of control room activities.
Evaluation, corrective action, and response for off normal conditions or
events, if'any, were examined. This included compliance to any reporting
requirements.
Two events (D-PAL-89-012 and 036) identified and. implemented
corrective action for late, and ' failure to implement, fire watch
requirements.
Observations of the control room monitors, indicators, and recorders were
made to verify the operability of emergency systems, radiation monitoring -
systems and nuclear reactor protection systems, as applicable. A strength
noted was the " black board" conditions maintained during most of the
period except during testing and repair activities. Operators were
well informed as to the reasons'for any alarms which indicates that
good turnovers are being conducted.
Reviews of surveillance, equipment
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condition, and tagout logs were conducted.
Proper return to service
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of selected components was verified. Activities' reviewed are discussed
below,
a.
Steam Generator Tube Leakage
The unit began the reporting period at a reduced power level of
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approximately 60 percent to maintain primary to secondary leakage
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at an acceptable level, while plans for a steam generator tube
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inspection outage were formalized. On April 3, during a licensee
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initiated conference call, the plant manager informed NRC personnel
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consisting of the resident inspection staff, the NRR project manager,
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section chiefs from Region III Divisions of Reactor Safety and
Reactor Projects, and NRR staam generator specialists, of six
options that were being considered by senior Consumers Power
corporate managers pertaining to operation of Palisades until
the upcoming outage.
The option that was implemented on April 4,
consisted of a power increase in three distinct steps to 90 percent
power with a specified holding period between steps.
Limitations
pertaining to the leakrate and off gas were imposed and
contingencies established if the limits were exceeded. A
90 percent power limit was a previously established administrative
limit to resolve NRC concerns pertaining to tube leakage.
Prior the
power increase, the resident inspection staff was informed of the
option selected. At the close of this inspection period, the power
level had been increased to 80 percent power with no changes in
primary to secondary leakrate.
b.
Post Accident Operation of the Component Cooling Water System
.
On March 27, the licensee became aware of a potentially significant
design deficiency in the Component Cooling Water (CCW) system relating
to the lack of a qualified (safety grade) air system-to support the
operation of the CCW containment isolation valves.
The CCW piping
inside containment is not seismically qualified and has not been
reviewed for high energy line break (HELB) concerns.
Containment
isolation valve CV-0910 does not have an air accumulator and would
fail open on loss of air. Therefore, a HELB could result in a
complete loss of CCW and a containment integrity violation.
Postulating a single failure of the 1-2 diesel generator, all
containment cooling would be lost.
This concern should have been
evaluated under the NRC's Systematic Evaluation Program, but
apparently was not because the system portion inside containn.ent
is not considered safety related.
The issue was identified under
the licensee's Configuration Control Project design basis
reconstitution effort, a part of a continuing corrective action.
It is expected that a LER will be submitted discussing this
discovery.
c.
Off Gas Monitoring
(1) The licensee has installed an additional off gas flow meter
that is calibrated in the 0-3 cfm range.
The other flow meter
was not calibrated in this range and would not provide repeatable
readings in this range.
Primary to secondary leakrate
calculation results dropped by a factor of 10, as result
of the more accurate readings from the new meter.
(2) The inspector observed an auxiliary operator take an off
gas flow reading and implement the procedural requirements of
Paragraph 7.8.1 to S0P 13, " Air Ejector, Gland Steam Condenser
and Condenser Vacuum pump."
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(3) During a Turbine'Buiiding. tour on April 5, the inspector
observed the needle of the.5 to 3 cfm off gas flow meter
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rapidly bouncing from off scale. low to approximately.75 CFM.
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The inspector found that the. instrument root isolation valves
had not been closed subsequent to the last off gas measurerr. ant.
An Auxiliary Operator closed the root isolation valve and the
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reading. returned to zero.
This was discussed with-the Shift
Supervisor and Operation Superintendent.
A review of S0P 13 reveals that this configuration was contrary-
to the restoration steps of paragraph 7.8.1. This was discussed
at the management interview.
d.
During a. tour of the Auxiliary Building on April 10, the inspector
identified. leakage (a small spray) from a main feedwater vent line-
near the containment penetration.
The. leakage was at the cap on the
line. The inspector verified that the line and manual Valve FW-746
were not considered part of the containment integrity boundary and
that a work request was subsequently initiated.
No violations, deviations, unresolved or open items were identified.
4.
Radiological Controls (71707)
During routine tours of radiologically controlled plant facilities or
areas, the inspector observed occupational radiation safety practices
by the radiation protection staff and other workers.
Effluent relea~ses were routinely checked, including examination of
on-line recorder traces and proper operation of automatic monitoring
equipment.
. Independent surveys.were performed in various radiologically controlled
areas. The' inspector witnessed the conduct of a survey being performed
in'the West Safeguards Room.
5.
Maintenance (62703, 42700)
Maintenance activities in the-plant were routinely inspected, including
both corrective maintenance (repairs) and preventive maintenance.
Mechanical, electrical, and instrument and control group maintenance
activities were included as available.
The focus of the inspection was to assure the maintenance activities
reviewed were conducted in accordance with approved procedures,
regulatory guides and industry codes or standards and in conformance
with Technical Specifications.
The following items were considered
during this review: the Limiting Conditions for Operation were met while
components or systems were removed from service; approvals were obtained
prior to initiating the work; activities were accomplished using approved
procedures; and post maintenance testing was performed as applicable.
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f
The following activities were inspected:
a.
Repair of Sigma Indicator TIC-0111H (W,0. 24901095)
b.
Rebuild of Relief Valve RV-0527 (W.0. 24901031)
c.
Auxiliary Feedwater Pump Governor packing adjustment
(W.O. 24901446)
d.
Blowdown flow meter cleaning and repair (W.0. 24901304)
e.
Pump (P-60) alignment and coupling (W.0. 24901341, RWP P890240)
f.
Remove, calibrate and reinstall the 0-3 cfm flow meter
(W.O. 24901663)
g.
Install mechanical tube plugs in "B" steam generator (W.O. 24900478).
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The inspector confirmed the plugs were receipt inspected by review
of Purchase Order CP11-7012Q.
No violations, deviations, unresolved or open items were identified.
6.
Surveillance (61720, 61726, 42700)
The inspector reviewed Technical Specifications required surveillance
testing as described below and verified that testing was performed in
accordance with adequate procedures, that test instrumentation was
calibrated, that Limiting Conditions for Operation were met, that removal
and restoration of the affected components were properly accomplished,
that test results conformed with Technical Specifications and procedure
requirements and were reviewed by personnel other than the individual
directing the test, and that deficiencies identified during the testing
were properly reviewed and resolved by appropriate management personnel.
The following activities were inspected:
a.
00-21
Auxiliary Feedwater System Valve Test (CV-0521)
This test demonstrated the operability of the steam admission
valve and also the turbine control after adjustment of the governor
packing.
The test was successful but also revealed that the steam
pressure control was still not optimum in that the relief valve
lifted several times before steady state operation was achieved.
The inspector verified that long range plans exist to resolve this
issue which include the installation of new air supplies to the
steam valves which will modulate their operation and, if that is
not successful, system design changes.
b.
DWO-1
Daily Control Room Surveillance.
c.
SHO-1
Operators Shift Surveillance.
d.
Escape Air Lock Penetration Leak Test.
e.
DWO-13
Local Leak Rate Test for Inner and Outer Personnel
Air Lock Door Seals"
f.
Personnel Air Lock Penetration Leak Test
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The inspector performed a technical review of S0-4b and DWO-13 using
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10 CFR 50 Appendix J, Standard Technical Specifications and the
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vendor manuals (Licensee Files C-53) as references and referred
the following items to the site licensing engineer.
(1) The inspector found that a between the seal test of the
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escape lock door was not performed after the escape air
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lock penetration test was completed on March 2, 1989.
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Interviews with personnel involved with the test indicated
that the air lock doors were opened a number of times to
restore the air lock to service following the test.
Restoration includes removal of the strong backs and
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fluffing of the seals to remove the seal set induced
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by the strong backs.
10 CFR 50, Appendix J, III.D.2.(b) (iii) requires that air
locks opened during periods when containment integrity is
required shall be tested within three days after being
opened.
For air lock doors that have testable seals, testing
the seals fulfills the three-day test requirements.
In the
event that the testing for the three-day interval cannot be
done at Pa, the test pressure shall be stated in the Technical
Specification.
The current Technical Specifications do not address 10 CFR 50,
Appendix J, III.D.2.(b) (iii) but the licensee does have a
Technical Specification change request that is pending with
NRR. The proposed Technical Specification does require a
reduced pressure between the seal test after use of air locks.
The licensee is implementing this requirement after each use of
the personnel air lock but not for the escape air lock. The
escape air lock vendor's manual does not recommend a between
the seals test be performed at Pa but does state the seal
should be capable of a reduced test pressure.
Failure to do a between the seals test of the escape air lock
after each use of the air lock is a violation of 10 CFR 50,
Appendix J, III.D.2.(b) (iii) (255/P9009-01(DRP)).
(2) The inspector noted that the licensee Technical Specification
change request is for a reduced pressure between the seals
test.
However, the proposed Technical Specification does not
provide an acceptance criteria. A review of Standard Technical
Specification reveals that a general acceptance criteria for
reduced pressure testing appears to be.01 La. The acceptance
criteria implemented in DWO-13 is approximately.025 La. The
inspector asked the licensing engineer to review the information
and assure the appropriate acceptance criteria is used.
In
addition, the inspector discussed this item with the NRR
Project Manager on March 16 for consideration during review
of the Technical Specification change request.
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MO-7A
Emergency Diesel Generator Test.
One violation and no deviations, unresolved or open items were
identified.
7.
Security (71707)
Routine facility security measures, including control of access for
vehicles, packages and personnel, were observed.
Performance of
dedicated physical security equipment was verified during inspections
in various plant areas.
The activities of the professional security
force in maintaining facility security protection were occasionally
examined or reviewed, and interviews were occasionally conducted with
security force members. Tours of the central and secondary alarm
station were routinely conducted.
No violations, deviations, unresolved or open items were identified.
8.
Safety Assessment / Quality Verification (35502, 40500)
The effectiveness of management controls, verification and oversight
activities, in the conduct of jobs observed during this inspection,
was evaluated.
The inspector frequently attended management and supervisory meetings
involving plant status and plans and focusing on proper coordination
among Departments.
The results of licensee auditing and corrective action programs were
routinely monitored by attendance at Corrective Action Review Board
(CARB) meetings and by review of Deviation Reports, Event Reports,
Radiological Incident Reports, and security incident reports. As
applicable, corrective action program documents were forwarded to
NRC Region III technical specialists for information and possible
followup evaluation.
No violations, deviations, unresolved or open items were identified.
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9.
Reportable Events (92700, 93720)
The inspector reviewed the following Licensee Event Reports (LERs) by
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means of direct observation, discussions with licensee personnel, and
review of records. The review addressed compliance to reporting
requirements and, as applicable, that immediate corrective action
and appropriate action to prevent recurrence had been accomplished.
(Closed) LER 255/84001:
Loss of offsite power and communications.
Appropriate modifications, training, and procedure changes were made,
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which resolved the myriad of issues involved in the event.
Licensee
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Reports D-PAL-84-003, D-PAL-84-007, D-PAL-87-004, NAPO Report P85-36,
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and PRC minutes 85-37 were reviewed.
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(Closed) LER 255/840022-02:
Containment temperature exceeded the
analyzed value. The licensee revised their surveillance procedures
to require a plant shutdown if the containment temperature exceeds the
currently analyzed 145 degrees Fahrenheit. A Technical Specification
limit is planned to be submitted as part of the Restructured Technical
Specifications. The licensee is continuing to evaluate methodology for
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determining air temperature and the adequacy of installed
instrumentation.
(Closed) LER 255/86004-02:
Main Steam Relief valves failed to meet
the "as found" acceptance criteria. A Technical Specification change
was submitted and approved relaxing the tolerance requirements from
+/- 1 percent to +/- 3 percent, which bounded the tolerances identified
during this testing.
(Closed) LER 255/86017-01:
Primary Coolant System (PCS) leakage was
greater than one gpm requiring a plant shutdown.
Two sources were found.
Relief Valve RV-2006 which protects the letdown system piping had not
reseated and was discharging the PCS to the quench tank.
Disassembly
of the valve determined that the bellows were distorted.
Evaluation
of the valve determined that it was suitable for its service and it was
repaired and reinstalled. The secondary cause of the leakrate, was the
failure of three valves in the reactor head vent system to reseat.
These
valves, PRV-1067, PRV-1068, PRV-1072, were found to contain metal shavings
which were left in the system during fabrication. This conclusion was
reached'after the shavings were analyzed by Battelle (Columbus) to
determine their source.
The valves were replaced, the head vent system
was flushed and corrective action was taken to improve cleanliness
controls during modification work (refer to D-PAL-86-154).
No violations, deviations, unresolved or open items were identified.
10.
Engineered Safeguards System Walkdown (71710)
The inspector performed a walkdown of the Chemical Spray Additive System
and verified that each accessible valve in the flow path was in its
required position and operable, that power was aligned for components
that actuate on an initiation signal, that essential instrumentation
was operable; and that no conditions existed which could adversely
affect system operation.
The licensee was provided the identification
of a valve which was missing a label, and questions about changes to
the valve lineup for their resolution.
No violations, deviations, unresolved or open items were identified.
11. Allegation Followup (99024)
(Closed) Allegation (AMS No. RIII-88-A-0163): The following discussion
relates to an allegation concerning the radiation protection program,
which was evaluated during this inspection. The evaluation consisted of
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a review of licensee records and NRC Inspection Reports, and discussions
with licensee personnel.
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Concern No. 1: A dose of 1800 mrem assigned to an individual for steam
generator work in February 1978 was too low.
Discussion:
The alleged stated that during each of three separate
entries (" jumps") into a steam generators, his dosimeter went off
scale. He stated that he was assigned a dose of 1800 mrem total for
the three jumps, but he contends that this assigned dose was too low.
No documentation was provided by the alleger to support his contention
that his dose should have been higher. A review of licensee's dosimetry
records maintained for the individual indicated that the individual was
provided with three separate dosimetry devices to monitor his radiation
dose. According to the records, on January 23, 1978, the ir;dividual's
self-reading dosimeter (SRD) went off scale and his TLD badge read
670 mrem (the inspecto 5 could not establish the range of the dosimeter
worn by the allegerb
, January 24, 1978, the dosimeter again went
offscale and the TLD adp read 999 mrem; and on January 26, 1978, the
dosimeter again went M M ale and the TLD badge read 413 mrem, for a
total exposure by TLD badge of 2082 mrem. The record also indicated
that on January 30, 1978, the dosimeter read 0 and the TLD badge read
4 mrem, and on January 31, 1978, the dosimeter read 0 and the TLD badge
read 2 mrem, for additional total exposure by TLD of 6 mrem.
In total,
for the month of January 1978, the record lists the individual's final
exposure according to his TLD badges as 2088, while his monthly film
badge reading was 1800 mrem.
(There was no indication that the individual
received any radiation dose in February 1978.) At the time, Palisade's
practice was to use the film badge as the official record of worker
exposure, to satisfy the requirements of 10 CFR 20, and use the less
accurate TLD badge and the SRD as secondary or back-up dosimeters to
quickly provide exposure information. The TLDs and SRDs were read onsite
after each entry, whereas the film badges had to be sent to an independent
contractor offsite for a monthly readout. The 14% difference between the
TLD dose and the film badge dose is consistent with the accuracy of the
two methods and is not a health and safety concern nor a regulatory
concern.
An inspection conducted at Palisades by NRC radiation specialists in
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January 1978, reviewed the licensee's control over outage activities,
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specifically steam generator work (Inspection Report No. 50-255/78-02).
)
The inspectors noted that during the work, the licensee experienced a
shortage of low-range (0-200 mrem) and high-range (0-1 rem and 0-5 rem)
SRDs; however, the inspectors verified that workers were sent into areas
only after surveys had been performed, wore film and TLD badges, and were
limited on their stay times so that an over-exposure was unlikely.
From
a health and safety perspective, the limitation on stay time imposed by
the licensee was an acceptable substitute for the lack of SRDs.
Findings: The allegation is not substantiated. Although the alleger's
SRD went off scale during three steam generator jumps, the individual's
dose was adequately monitored by a film badge and a backup TLD dosimeter.
The 1800-mrem dose assigned to the alleger was based on the film badge,
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which was the licensee's officially designated dosimeter at the time.
'
The allegation is closed.
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.
Concern No. 2: The licensee did not evaluate an uptake received by the
alleger when his airline became disconnected during a steam generator
jump.
Discussion: The inspectors were not able to retrieve any licensee
records that specifically dealt with the alleged airline disconnect.
However, the inspectors did review records that indicated that the
alleger received a whole body count on February 1,1978, shortly after
his steam generator entries. Whole body counting is an acceptable method
for evaluating possible uptakes of radioactive material, including that
found in steam generator channel heads during maintenance. The results
of this whole body count indicated the presence of approximately four
nanocuries of Cs-137 in the individual, a value which indicates exposure
to airborne radioactivity of no more than about 0.1% of the NRC limit
during the individual's steam generator jumps. NRC review in 1978 of
the licensee's whole body counting system and results of many counts
conducted during the outage determined that the licensee whole body
counting system was adequate and no exposures to airborne radioactive
material in excess of regulatory limits occurred during the outage
<
(Inspection Reports No. 50-255/78-02; 50-255/78-11).
I
Findings: The allegation is not substantiated. The licensee conducted
a whole body count of the alleger shortly after the steam generator work.
Although this whole body count may not have been specifically given to
evaluate any uptake resulting from the alleged airline disconnect, it
i
served that purpose. The amount of radioactive material detected by
the whole oody count indicated that the individual was not exposed to
airborne radioactivity in excess of regulatory limits during his steam
generator work.
NRC review during the period of concern determined that
the licensee was providing adequate internal exposure control and
assessment. The allegation is closed.
No violations, deviations, unresolved or open items were identified.
12. Management Interview (30703)
The inspectors met with licensee representatives (denoted in Paragraph 1)
on April 12, 1989, to discuss the scope and findings of the inspection.
In addition, the inspector also discussed the likely informational content
of the inspection report with regard to documents or processes reviewed by
the inspector during the inspectidn.
The licensee did not identify any
such documents / processes as proprietary.
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