ML20140C564
| ML20140C564 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/11/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20140C552 | List: |
| References | |
| 50-245-97-01, 50-245-97-1, 50-336-97-01, 50-336-97-1, 50-423-97-01, 50-423-97-1, NUDOCS 9704170127 | |
| Download: ML20140C564 (50) | |
See also: IR 05000245/1997001
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket Nos.:
50-245
50-336
50-423
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Report Nos.:
97-01
97-01
97-01
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License Nos.:
DPR-65
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Licensee:
Northeast Nuclear Energy Company
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P. O. Box 128
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Waterford, CT 06385
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Facility:
Millstone Nuclear Power Station, Units 1,2, and 3
Inspection at:
Waterford, CT
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Dates:
January 1,1997 - March 10,1997
inspectors:
T. A. Easlick, Senior Resident inspector Unit 1
D. P. Beaulieu, Senior Resident inspector, Unit 2
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A. C. Cerne, Senior Resident inspector, Unit 3
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A. L. Burritt, Resident inspector, Unit 1
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R. J. Arrighi, Resident inspector, Unit 3
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L. L. Scholl, Reactor Engineer, Region l
N. J. Blumberg, Project Engineer, Region i
R. J. Urban, Project Engineer, Region I
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J. T. Furia, Senior Radiation Specialist, Region I, DRS
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J. E. Carrasco, Reactor Engineer, Region I, DRS
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Approved by:
Jacque P. Durr, Chief
Inspection Branch
Special Projects Office, NRR
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9704170127 970411
ADOCK 05000245
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TABLE OF CONTENTS
EX EC UTIVE S U M M AR Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii
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U1.1 Operations
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U101
Cond uct of O perations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
U102
Operational Status of. Facilities and Equipment . . . . . . . . . .
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U103
Operations Procedures and Documentation
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U105
Operator Training Qualification . . . . . . . . . . . . . . . . . . . . . . . . . 3
U108
Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . 5
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U 1.ll M ainte na n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
U1 M2
Maintenance and Material Condition of Facilities and
Equipment
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U 1.lli Enginee ring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
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U1 E1
Conduct of Engineering
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U1 E2
Engineering Support of Facilities and Equipment . . . . . . . . . . . . . 9
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U1 E8
Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . 11
S2.1 Operations
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U2 01
Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
U2 O2
Operational Status of Facilities and Equipment . . . . . . . . . . . . . 12
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U2 08
Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . 14
U 2.ll M ainte n a n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
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U2 M8
Miscellaneous Maintenance issues
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U 2.lli Engine e ring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
U2 E8
Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . 17
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U3.1 Operations
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U3 01
Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
U~s 03
Operations Procedures and Documentation
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U3 07
Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . 26
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U 3.11 M ain te n a n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27
U3 M1
Conduct of Maintenance
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U3 M8
Miscellaneous Maintenance issues
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U 3.Ill Engine e ring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31
U3 E8
Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . 31
IV Plant Support
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R1
Radiological Protection and Chemistry Controls
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R8
Miscellaneous Radiological Protection and Chemistry issues
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V. M anage ment M eetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39
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Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39
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EXECUTIVE SUMMARY L
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Millstone Nuclear Power Station'
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Combined inspection 245/97-01; 336/97-01; 423/97-01
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Operations
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Numerous inaccurate Personal Qualification Statements (Form 398) were identified
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at all four Connecticut plants following NRC questions on two recent adverse
condition reports. Approximately two thirds of the Personal Qualification
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Statements submitted for recent license applicants were inaccurate. These
applications resulted in the conduct of NRC license examinations and the issuance -
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of licenses. In a significant number of cases, the licenses were issued without the.
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candidates fully completing the licensee's training and qualification program, and in
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a few cases the reactivity manipulations, specifically required by 10 CFR 55, were
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also not complete. This issue is unresolved for'each Millstone unit pending the
completion of the licensee investigation, resolution of allidentified deficiencies and
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implementation of programmatic corrective actions. (Section U1.05.1)
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The degraded conditions found in the Unit 1 spent fuel pool are representative of a
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icng standing disregard for foreign material exclusion (FME) during the conduct of
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refuelmg fioor activities. Past low standards for FME control allowed the
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accumulation of a large amount of debris, which could potentially have a significant
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-impact on the fuel assemblies stored in the pool. Once the recovery organization
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became aware of the extent of the problem, by reviewing video tapes, the
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inspectors noted a good response, including clear direction as to what needed to be
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done in the short term. Based on this information, the acceptability of the degraded
conditions in the spent fuel pool will be unresolved pending NRC review of the
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issues. (Section U1.02.1)
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At Unit 1, the licensee failed to evaluate and address a violation concerning the
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failure to prevent work which had the potential for draining the reactor vessel during
refueling. Further, the licensee failed to provide a comprehensive closure package -
of a quality consistent =with the process committed to in December 1996. The
' manner in which this issue was addressed provides evidence of the licensees ability
to perform effective reviews and to implement appropriate corrective' actions. As a
result of the inspector's concerns with the quality of the completion packages, the
licensee withdrew the NRC completion package schedule. A revised schedule was
still in developmcat at the end of the inspection period. (Section U1.08.1)
Although the Unit 2 backlog of 798 adverse condition reports (ACRs) that are
greater than 120 days old indicates that timeliness for completing corrective actions
remains a concern, the reduction in the backlog of older ACRs from 940 to 798
since the last inspection period is a positive trend which reflects the licensee's
increased level of effort in this area. Timeliness and effectiveness of corrective
actions are areas '... which the licensee must demonstrate sustained improved
performance. (Section U2.02.1)-
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A Unit 2 licensee event report discussed that while recovering from a loss of a
direct current (dc) bus, an operator failed to enter an action statement when three
channels of wide range nuclear instrumentation were rendered inoperable. This was
characterized as a non-cited violation. The primary concern associated with this
event was the fact that there was minimal procedural guidance provided to
operators to recover the loss of a de bus. The licensee is in the process of
preparing 12 abnormal operating procedures for recovering various de buses and
distribution panels. This concern is being tracked by an unresolved item. (Section
U2.02.2)
In addition to the physical plant design controls, a longstanding NRC concern at Unit
2 is that operating procedures do not reflect the Final Safety Analysis Report
(FSAR), and an NRC open item has existed since 1993 to address this concern.
This inspection report closes the old open item npqt because adequate corrective
actions have been taken, but because this concern is being addressed and tracked
by more recent items. The issue includes an evaluation of the procedure change
process, as well as the design control process, to ensure future operation is
conducted in accordance with the FSAR. (Section U2.08.1)
The procedure upgrade program has been effective in standardizing procedure
formats across the site. Because of the number of individuals involved in the
procedure upgrades and the long period of time to complete the task, the quality of
procedures vary substantially. As an adjunct to the Demand For Information
process [10 CFR 50.54(f)], which incorporates a verification of the design and
licensing bases, procedure accuracy will be verified. (Section U3 03)
The licensee's root cause investigation and corrective action plan for control of high
energy line break (HELB) doors were determined to be good. However, the
requirement to label the required HELB doors with a minimum number of turns to
ensure proper latching should have been included in the adverse condition report
corrective action plan if it was deemed necessary to prevent recurrence. (Section
U3.01.2)
Good' contingency planning and appropriate consideration of the applicable standard
and regulations were in evidence for both planned operational evolutions and
emergent shutdown conditions. Where necessary to improve shutdown risk
margins, temporary modifications or special system lineup were considered and well
controlled. The licensee development of a standardized approach for disseminating
operations policy for interpreting the language and action statement applicability of
the Unit 3 technical specifications appeared warranted. The examples noted during
this period will be reviewed further as an inspector follow item. (Section U3.01.1)
Maintenance
At Unit 1, the preparation and conduct of work associated with the entry and video
survey of the reactor water cleanup (RWCU) demineralizer room were well
controlled. The inspector noted that good radiological practices were used.
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The material condition in the room was acceptable and no deficiencies were
identified. (Section U1.M2.1)
Maintenance and surveillance activities were performed professionally and
thoroughly. All observed maintenance activities were performed with the work
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package or surveillance procedure present at the job site and personnel were noted
to be closely following the procedures. Review of the surveillance procedures
revealed that the requirements of the applicable technical specifications were
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appropriately incorporated into the implementing procedure. (Section U3.M1.1)
The licensee developed a Fix-It-Now (FIN) multi-discipline work team approach to
augment the way maintenance is performed at the unit. This process is in addition
to the normal work control process. The FIN work process is being implemented in
a conservative manner. Any work required on protected train equipment was not
being assigned to the FIN team. All monitored work activities performed by the FIN
team were performed in accordance with the unit and station procedures. FIN team
members appeared to be well qualified. (Section U3.M1.1)
Plant inspection-tourt revealed improvements in the Unit 3 areas of housekeeping,
material conditions, and work controls. Field observations raised no new
unresolved safety issues, but did highlight the need for additional management
attention to a previously identified concern regarding the control of temporary
equipment with the potential to adversely impact safety-related components. This
" Seismic II/l" issue will be tracked as an inspector follow item and will receive
further evaluation as a "significant item" in the NRC Restart Assessment Plan.
(U3.M8.1 )
Engineering
A review was performed at Unit 1 of the licensee's progress in resolution of the
Unresolved Safety issue (USI) A-46 outliers documented in the Licensee Event
Report (LER)96-003, Rev. 2. These deficiencies involved inadequate anchorage of
the emergency diesel generator day tank and the turbine building secondary closed
cooling water air coolers. The regulatory requirements for reportability were met
and the corrective action prescribed in the LER were adequate in general, based on
the detailed walkdown of the A-46 modifications, the inspector concluded that the
licensee performed a substantial number of field modifications to accommodate the
seismic loading on mechanical and electrical equipment identified in the USl A-46
scope and documented in LER 96-003; this LER is closed. The followup of the
licensee's commitments to resolve the A-46 program outliers prior to startup for
cycle 16 operation and the assessment of the implication of this event on the
operation of Unit 1 will be unresolved pending further NRC review. (Section
U 1.E1.1 )
The corrective actions taken by the Unit 1 Cnmponent Engineering
Services / Nondestructive Test Engineering (CES/NTE) concerning the use of test
equipment was acceptable. The corrective actions appeared to be broad-based. A
majority of the short term corrective actions were complete, and the long term
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corrective actions were being tracked for closure., The significance of the UT
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instruments being past their calibration due dates was minimal because they were
. subsequently found to be within tolerance. (Section U1.E2.1)
The NRC is concerned about the operation of the new containment isolation check
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valve 1-CU-29 at Unit 1, which is operating with lower than expected flow rates
during this extended shutdown. While the use of non-intrusive check valve testing
has verified that the check valve is backseated, this is a short term indication. The
long term effects of the low flow operation have yet to be determined. This issue is
unresolved pending the NRC review of the licensee's final determination of the
operability of the valve prior to plant startup. (Section U1.E2.2)
The inspectors found that overlap test reviews performed in 1993 were not
adequate. The licensee failed to identify this deficiency and improperly concluded
that the 1993 reviews accomplished the actions requested by NRC Generic Letter 96-01, " Testing Of Safety-Related Logic Circuits." (Section U3.E8.1)
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Plant Support
The licensee has demonstrated a significant increase in management attention
towards work control and maintaining occupational exposures as low as is
reasonably achievable. However, two of the licensee's activities were determined
not to be in compliance with NRC regulations. The Unit 1 violation involves a long-
standing situation (since initial plant start-up), concerning an unmonitored release
pathway in the ventilation system for the radwaste storage building. The violation
at Unit 2 involves a failure to adhere to the licensee's radiation protection program
concerning proper use of electronic dosimeters. (Section R1)
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Report Details
Summarv of Plant Status
Unit 1 remained in an extended outage for the duration of the inspection period. The
licensee continues to implement configuration management program activities, engineering
reviews, and docketed correspondence assessments to verify compliance with the
established design and licensing basis of the unit. The successful completion of these
activities is required by NRC order prior to restart of the unit. During this period, the
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licensee implemented a major revision to the corrective action procedure. The goal was to
simplify the process, and at the same time make it more responsive towards restart and
needed organizational improvements. Under the new process, " condition reports" have
replaced " adverse condition reports" to capture both the regulatory defined adverse
condition, as well as other conditions that do not meet managements expectations.
The licensee has recently made two changes to the organizational structure at Unit 1. A
project management group was established to facilitate the implementation of plant
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modifications prior to startup. Additionally, a restart manager was selected to oversee
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completion of the Operational Readiness Plan, which will be used to identify and control
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the actions necessary to achieve and maintain improved performance. The restart manager
will also be responsible for the review of corrective action completion packages, which will
provide objective evidence of corrective action completion. Section U1.08.1 & U1.E8.1 of
this report provides an assessment of the licensee's progress in the area of completion
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package development. The effectiveness of these changes, as well as the new corrective
action program, will be assessed as part of future NRC inspections.
U1.1 Operations
U101
Conduct of Operations
01.1 General Comments (71707)
Using Inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing
plant operations. The inspectors reviewed operability determinations, availability
determinations, and witnessed the conduct of management review team discussions
regarding the disposition and closure of condition reports (CRs). During a routine tour of
the Unit 1 intake structure, the inspector found the material condition of systems and
components to be adequate. The licensee initiated a material condition improvement
project in the Spring of 1996. The inspector observed some material improvement work
on-going. Specific events and noteworthy observations are detailed in the sections below.
U102
Operational Status of Facilities and Equipment
O 2.1 Soent Fuel Pool Cleanliness
a.
Inspection Scope (71707)
NRC inspection report 245/96-08, dated December 3,1996, discussed the continued
identification of discrepant conditions in the spent fuel pool, indicating a need to accelerate
the evaluation portion of the spent fuel pool cleanup / recovery plan. At that time, the
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-inspectors concluded that all discrepant conditions warrant identification and evaluation in
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the short term to ensure the collective impact of these issues were addressed. On January
10,1997, the inspectors reviewed a video tape surveillance of the spent fuel pool
conducted on January 3, using an under water camera. The video tapes were reviewed to
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evaluate the conditions in the spent fuel pool including the fuel storage racks and stored
fuel bundles.
b.
Observations and Findinas
As discussed in NRC report 245/96-08, the earlier video tapes identified improperly seated
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fuel bundles. In light of the recent video surveys, the licensee determined that the
improperly seated bundles were caused by one of three conditions: 1) There are 55 fuel
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bundles elevated as a result of their channel fasteners being caught on the spent fue!
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racks; 2) 14 fuel bundles are elevated due to unknown reasons, although it is suspected
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that debris is in the fuel rack preventing proper seating; 3) One additional fuel bundle was
resting on a 1/4 inch metal tube (suspected to be a boron tube from a control rod blade
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segment) that is lying on the floor liner and bends upward into the bottom of the fuel
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storage cell. An evaluation was performed to address all relevant issues including: the
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effect of a bundle drop on the fuel bundle itself, the fuel rack, and spent fuel pool liner;
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seismic response of the fuel racks; the criticality margin; fuel assembly cooling; and water
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shielding. The licensee concluded that the storage racks that contained elevated fuel
assemblies were operable, but were not full qualified, since the fuel assemblies were not
fully seated. In response to this concern, procedural controls were put in place to ensure
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that no fuel assemblies are transferred within the spent fuel pool until all fuel is fully
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seated.
The video tapes also revealed a significant amount of debris on the fuel bundles, fuel
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racks, and the floor of the fuel pool. The debris included rope, cable, boron tubes, a broom
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head, filter hoses, nuts, and unidentifiable objects. In addition, the bottom of the fuel pool
was covered by a layer of sediment. Additionally, the viden ^) view identified that the
velocity limiter portion cf '.our control rod blade assemblies vere stored vertically on top of
one another without support. The velocity limiter sections were located on the spent fuel
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pool floor in the space between a spent fuel rack and the control rod blade storage rack.
An engineering evaluation of the velocity limiter storage configuration concluded that the
structural integrity of the spent fuel rack, the control rod blade storage rack, and the pool
liner would be maintained in the event of an impact caused by the velocity limiters falling
over.
A dent was identified in the spent fuel pool floor liner from an impact of an unknown
object. The dent was approximately 4 inches in diameter and was relatively uniform and
smooth, with no obvious nicks or gouges. A final operability determination and safety
evaluation concluded that there were no operability or safety issues, and that the dent did
not challenge the leak tightness or structural integrity of the fuel pool.
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A previously known condition concerning the storage of a damaged, irradiated fuel
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assemble stored in a " damaged fuel container," was roviewed to consider the collective
impact of this issue and the other fuel pool discrepan:les. The fuel assemble was
damaged in 1974, placed in a storage container in 1976, and moved to its current location
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in 1989. The licensee performed a safety evaluation, as part of an operability
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determination, that addressed the storage configuration of the damaged fuel container and
its location in the control rod storage rack. Similar to the unseated bundles, criticality,
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seismic response, water shielding, and decay heat removal were evaluated. Based on that
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evaluation, the licensee concluded that the storage configuration is safe and does not
constitute an unreviewed safety question. In addition, a procedural restriction was placed
in procedure EN 1067, Supplemental Procedure for Inventory and Control of Special
Nuclear Material, to prevent storing fuelin locations adjacent to the damaged fuel
assembly. This was required since the criticality margin for storage of fuel assemblies in
adjacent rack locations has not been fully evaluated and full qualification has not been
verified,
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Conclusions
The inspectors concluded that the degraded conditions in the spent fuel pool are
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representative of a long standing lack of concern for fore:gn material exclusion (FME)
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during the conduct of refueling floor activities. Past low standards for FME control allowed
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the accumulation of a large amount of debris, which could potentially have a significant
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impact en the fuel assemblies stored in the pool. Once the recovery organization became
aware of the extent of the problem, by reviewing the video tapes, the inspectors noted a
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good resoonse, including clear direction as to what needed to be done in the short term.
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The appropriate operability determinations and safety evaluations were prepared, and
adverse condition reports were initiated to document the findings.
In a letter to the NRC dated February 21,1997, the licensee documented the current
conditions in the spent fuel pool and their future plans for correcting the adverse
, conditions. Prior to core reload all unseated fuel assemblies will be properly seated and all
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new and reload fuel bundles will be visually inspected from below to check for foreign -
material before placement in the core. The licensee committed to cleaning up the pool,
including removal of debris and various used components, prior to Refueling Outage 16.
Based on this information, the acceptability of the degraded conditions in the spent fuel
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pool will be unresolved (URI 245/97-01-01) pending NRC review of the issues and the
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completion of the licensee's root cause analysis.
U103
Operations Procedures and Documentation
03.1 Operations Procedures
During May and June,1996, two Unit 1 Operations department staff engineers performed
a self assessment of the Procedure Upgrade Program (PUP) for Unit 1 Operations
procedures. The Unit 1 operations self-assessment contained a significant number of
negative findings concerning the Unit 1 PUP process for the Operations Department and for
the quality of the upgraded procedures produced. The inspector discussed this assessment
with one of the staff engineers and the Unit 1 Operations Manager. Although the
assessment applied to Unit 1 only, the Unit 1 Operations Manager stated that he would
share the self assessment results with other Unit 1 Departments and with the operations
managers of the other Millstone units. The licensee's resolution of the problems identified
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in the Unit 1 self assessment are unresolved (URI 97-01-02) pending the NRC's review of
the associated corrective actions.
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U105
Operator Training Qualification
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05.1 Inaccuracies in Personal Qualification Statements Certifications
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a.
Inspection Scope
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Two adverse condition reports (ACRs) were initiated to address operator license training -
related deficiencies. The ACRs document the failure of license candidates to complete all
classes, on the job training (OJT), and on shift watch standing time, along with the failure
to comply with procedures resulting in weaknesses in the systematic approach to training.
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The issues were identified as a result of preliminary findings and insights gained from an
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independent root cause investigation to address poor candidate performance during a
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recent Millstone Unit 1 initial license examination. The inspectors reviewed the short term
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actions taken in response to these two ACRs. The reviews focused on the accuracy of
Personal Qualification Statements (Form 398) submitted to the NRC staff as an application
for an operators license. The Form 398 contains assertions by the applicant, the training
coordinator and senior management representative on site, that among other things, the
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applicant completed the licensee's requirements to be a licensed operator.
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b.
Observations and Findinas
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During the review subsequent to the initiation of the ACRs, the licensee identified
numerous discrepancies which resulted in inaccurate Personal Qualification Statements
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(Form 398). In some cases, the errors resulted in candidates not meeting the licensee's.
minimum program requirements prior to signing of the 398 forms; however, the necessary
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training was completed prior to the license examination. In other cases, the candidates
were issued licenses without the program requirements being met. The discrepancies
include failure to complete the required on shift watchstanding time, OJT, and the required
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number of reactivity manipulations, in addition, several candidates failed to meet the
program prerequisites such as technical degree or additional experience requirements.
At Millstone Unit 1, the four most recent license classes were reviewed by the licensee, in
the two most recent classes,12 of 13 candidates submitted inaccurate 398 forms. In the
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two prior classes, only 1 of 9 candidate's 398 form was inaccurate. Most of the
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discrepancies involved the failure to complete required under-instruction watches, but also
included the failure to complete the required OJT. In the worst case, the candidate
completed little more than 3 of the required 13 weeks of OJT specified by the training
program description.
At Millstone Unit 2, a review of the two most recent license classes revealed 14 out of 16
candidates submitted inaccurate 398 forms. These discrepancies generally consisted of
insufficient hours of under-instruction watchstanding, but also included one case of
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insufficient reactivity manipulations and two cases in which OJT records appear to be lost.
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At Millstone Unit 3, the review of the most recent license class revealed 3 of 10 of the
candidates submitted inaccurate 398 forms. These discrepancies included one missed
under-instruction watch, one case of insufficient reactivity manipulations and the failure to
meet the program prerequisites, and one case in which OJT requirements were
accomplished after the assertion that all training program requirements were completed on
the 398 form.
At Connecticut Yankee, the review of the most recent license class revealed 10 out 12
candidates submitted inaccurate 398 forms. These discrepancies included insufficient
hours of under-instruction watchstanding, insufficient reactivity manipulations in two
cases, and program prerequisites not met in two cases. Additionally, OJT records were
lost or signed after the 398 was completed.
On March 3,1997, the licensee issued a letter to the NRC staff to discussing these issues.
Subsequently, the NRC staff issued a confirmatory action letter. The reviews are being
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expanded on Millstone Unit 3 and Connecticut Yankee. Millstone Unit 2 is still evaluating
if review scope expansion is warranted and Millstone Unit 1 is preparing a position that
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additional expansion is not necessary. The licensee has removed numerous individuals
from watchstanding duties and requested the withdrawal of two licenses. However, in the
case of Millstone Unit 2, some licensed operators removed frorn watchstanding duties have
been restored to an active status following the completion of missed requirements. The
licensee believes the majority of the discrepancies can be attributed to unclear
expectations on program requirements, the failure to maintain the programs currer t using
the systems approach to training, and poor record keeping practices. All of the fctors are
~!
the result of inadequate management oversight.
c.
Conclusion
The licensee identified numerous inaccurate Personal Qualification Statements (Form 398)
following NRC questions on two recent ACRs. Approximately two thirds of the Personal
Qualification Statements submitted for recent license applicants were inaccurate. These
applications resulted in the conduct of NRC license examinations and the issuance of
licensees. In a significant number of cases, the licenses were issued without the
candidates completing the licensee's training and qualification program, and in a few cases
the reactivity manipulations, specifically required by 10 CFR 55, were also not complete.
This issue is unresolved (URI 245,336,423/97-01-03) pending the completion of the
licensee's investigation, resolution of allidentified deficiencies and implementation of
programmatic corrective actions.
U108
Miscellaneous Operations issues (92700)
08.1 (Undate) Violation 50-245/95-42-01: Failure to Prevent Work Which Had the
Potential for Drainina the Reactor Vessel Durina Fuel Movements
This violation concerned the failure to prevent work which had the potential for draining
the reactor vessel while fuel removal was in progress. In addition, the licensee does not
have a formal process to ensure all applicable technical specifications are properly
implemented during refueling. Further, based on the inspector's review of the licensing
,
.
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6
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bases for the current Technical Specification 3.5.F 7, it did not appear that the conditions
j
.
. initially established and reviewed by the NRC were appropriately maintained during
s"
subsequent amendments.
The licensee developed a process for preparation of corrective action completion packages
and a schedule for providing them to the NRC, as a result of a previous NRC request. The
inspector reviewed the first corrective action completion package prepared to address the
three issues discussed above. The documentation package contained a root cause analysis
-(RCA), license event report, and a violation response, which were developed to address the
3
issues. However, these documents were not consistent with each other and generally did
not address the cited violation. The identified causes and many of the corrective actions,
address maintenance and planning issues that led to the unplanned draining of a small
amount of reactor water during maintenance on a recirculation discharge valve. The RCA
i
appears to have been performed prior to the licensee's acknowledgment of the technical
'
,
specification compliance issue. The majority of the corrective actions specified, involve
I
improvements to the shutdown risk program; however, the inspector determined that these
actions would not preclude a recurrence of the technical specification non-compliance.
'
The violation response discussed the development of mode change checklists and
j-
enhanced logs; however, these actions, which may address technical specification
'
compliance, were not implemented by the end of the inspection period.
The licensee did not address the adequacy of Technical Specification 3.5.F.7, nor verify
that the conditions initially established and reviewed by the NRC were appropriately
maintained during subsequent amendments. The license event report was submitted 3
months after the event without a detailed reason for the delay and no corrective actions
i
4
'
were specified for the late reporting.
2
This item will remain open pending resolution of this item. The licensee failed to
0
appropriately evaluate and address these issues for more than a year since the event.
!
Further, the licensee failed to provide a closure package consistent with the process
committed to in December 1996. The manner in which this issue was addressed provides
evidence of the licensees ability to perform effective reviews and to implement appropriate
corrective actions. As a result of the inspector's concerns with the quality of the
completion packages, the licensee withdrew the NRC completion package schedule. A
..
revised schedule was stillin development at the end of the inspection period.
U1.ll Maintenance
U1 M2
Maintenance and Material Condition of Facilities and Equipment
i
i
M2.1 RWCU Demineralizer Room Material Condition
a.
insoection Scope (71750)
'
The inspector observed activities associated with the entry into, and video survey of, the
'
reactor water cleanup (RWCU) demineralizer room. The purpose of the entry was to
-i
i
determine the material condition of the infrequently accessed room. A remote controlled
robot was used, which supported both a video camera and radiation detection equipment.
.
.
7
b.
Observations and Findinas
The health physics (HP) department was well prepared for this activity since preparations
'
and staging were completed the day before. This allowed potential problems to be
identified and corrected prior to the start of work. In particular, it was identified in
advance that the robot would need to be lifted into the room through the access in the
wall, and preparations were made to account for this. Positive control over personnel
access, was observed with only people that were needed for the activity permitted in the
area. The workers' awareness of radiological hazards was evident. HP supervision and
the system engineer provided oversight of this activity. The video survey indicated that
the room was in good condition and the structural integrity of the piping and three
'
domineralizer tanks was intact. There was no indication of any system leakage and
radiation levels in the general area were normal.
c.
Conclusions
Based on the above review, the inspector determined that the preparation and conduct of
work associated with the entry and video survey of the RWCU demineralizer room was
well controlled. The inspector noted that good radiological practices were used. The
material condition in the room was acceptable and no deficiencies were identified.
U1.Ill Enaineerina
U1 E1
Conduct of Engineering
E1.1
Unresolved Safety issue USl A-46 " Seismic Qualification of Eauipment in Operatina
Plants."
a.
Insoection Scope (37550)
The scope of this inspection was to review the licensee's progress in resolving the outliers
identified during the implementation of the Unresolved Safety issue (USI) A-46 " Seismic
Qualification of Equipment in Operating Plants."
b.
Observations
The inspector reviewed the Licensee Event Report (LER)96-003, Rev. 2 that documented
deficiencies involving inadequate anchorage of the emergency diesel generator (EDG) day
tank and the turbine building secondary closed cooling water (TBSCCW) air coolers.
Backaround
in December 1980, The NRC staff initiated an Unresolved Safety issue, (USI) A-46,
" Seismic Qualification of Equipment in Operating Plants," related to seismic adequacy of
mechanical and electrical equipment in older nuclear plants. After technical research by
the Seismic Qualification Utility Group (SQUG) and the NRC regarding this issue, the NRC
.,__ _
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7
b.
Observations and Findinas
i
The health physics (HP) department was well prepared for this activity since preparations
and staging were completed the day before. This allowed potential problems to be
identified and corrected prior to the start of work. In particular, it was identified in
"
advance that the robot would need to be lifted into the room through the access in the
wall, and preparations were made to account for this. Positive control over personnel-
,
access, was observed with only people that were needed for the activity permitted in the
j
]
area. The workers' awareness of radiological hazards was evident. HP supervision and
'
the system engineer provided oversight of this activity. The video survey indicated that
l
the room was in good condition and the structural integrity of the piping and three
,
l
demineralizer tanks was intact. There was no indication of any system leakage and
l
radiation levels in the general area were normal.
c.
Conclusions
t.
Based on the above review, the inspector determined that the preparation and conduct of
j
work associated with the entry and video survey of the RWCU demineralizer room was
well controlled. -The inspector noted that good radiological practices were used. - The
material condition in the room was acceptable and no deficiencies were identified.
,
I
U1.Ill Enaineerina
'
I
U1 El
Conduct of Engineering
E1.1
Unresolved Safety issue USI A-46 " Seismic Qualification of Eauioment in Ooeratina
,
Plants."
I
a.
Inspection Scope (37550)
l
The scope of this inspection was to review the licensee's progress in resolving the outliers
identified during the implementation of the Unresolved Safety issue (USl) A-46 " Seismic
i
Qualification of Equipment in Operating Plants."
b.
Observations
The inspector reviewed the Licensee Event Report (LER)96-003, Rev. 2 that documented
deficiencies involving inadequate anchorage of the emergency diesel generator (EDG) day
tank and the turbine building secondary closed cooling water (TBSCCW) air coolers,
i
Beckaround
in December 1980, The NRC staff initiated an Unresolved Safety issue, (USI) A-46,
" Seismic Qualification of Equipment in Operating Plants," related to seismic adequacy of
mechanical and electrical equipment in older nuclear plants. After technical research by
the Seismic Qualification Utility. Group (SQUG) and the NRC regarding this issue, the NRC
.
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1
8
4
>
published a detailed approach for resolving USl A-46, in Generic Letter 87-02, "Verif!:ation
of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, USl A-
-
4 6. "
The Generic Letter procedure set forth an approach for verifying seismic adequacy of
i
equipment using earthquake experience data supplemented by test results and analyses, as
,
necessary. Licensees subject to USl A-46 were encouraged to participate in the generic
'
'
program to accomplish seismic verification of equipment. As a result, SQUG developed the
" Generic Implementation Procedure (GIP) for seismic verification of Nuclear Plant
equipment."
l
USl A-46 Proaram at MS1
At Millstone Unit 1, the USl A-46 program was conducted to address the concerns
expressed in GL 87-02 regarding the seismic adequacy of safety related electrical and
d
mechanical equipment. The resolution of the seismic adequacy issue appeared to be
conducted in accordance with the SOUG approach, using the generic implementation plan
(GlP) as approved by the NRC in Supplemental Safety Evaluation Report (SSER) No.2.
I
Supplement 1 to GL 87-02, transmitted May 22,1992 includes SSER-2 which reviews the
l
GIP, requires the licensee to identify within 120 days a schedule for implementation and
'
any anticipated deviation from the GIP methods. The inspector verified that the licensee
met this requirement by reviewing the licensee's letter dated September 21,1992. In this
letter the licensee identified a schedule for MS1, which stated that the submittal of the
final report will take place six months after refueling outage 15, which is stillin progress.
The inspector noted that the implementation of the USI A-46 review program has resulted
'in the identification of outlier conditions which challenged the operability of the plant
,
components. These outliers were reported in an LER. The LER described the proposed
corrective action, which includes resolution of all outlier conditions prior to start up from
Refueling Outage (RFO) 15.
The LER selected for this inspection addresses operability concerns involving inadequate
anchorage of the EDG day tank and the TBSCCW air coolers. These operability concerns
were properly documented as LER,96-003, Rev. 2. The inspector reviewed the LER 96-
003, to ensure that regulatory requirements for reportability were met. The licensee has
properly identified this design deficiency as a USl A-46 program outlier, and has properly
characterized it as being reportable in accordance with 10 CFR 50.72 and 10 CFR 50.73.
The inspector found the LER's event described in a chronological sequence and the
prescribed corrective action appeared to be appropriate.
Conclusion
in terms of reportability, the proper characterizations were given to the outliers of the USl
A-46 program. The licensee prepared the LERs documenting these outliers in accordance
with established regulatory requirements. Based on these inspection results, LER 50-
.
245/96-003 is closed. The followup of the licensee's commitments to resolve the A-46
program outliers prior to startup for cycle 16 operation and the assessment of the
,
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9
,
implication of this event on the operation of Unit 1.will be unresolved pending further.NRC.
.
. review (URI 50-245/97-01-04)
Review of the Corrective Action for the EDG Day Tank
,
Since the EDG system is an emergency ac power system and the EDG day tank is a safety
,
related component, the inspector focused his review on the licensee's corrective action
- package. In the package, the Design Change Notice (DCN) designed and installed the
seismic restraint that consists of a box frame around the tank supported by the block
wall's structural reinforcements.
f
The design modification package was found to be complete, and the frame was designed
l
in accordance with the American Institute of Steel Construction (AISC) Manual for Steel
Construction, 9th Edition. However, key design parameters in the calculation were not
properly referenced making it difficult for an independent auditor to determine whether or
l
l
not these parameters are correct. These key design parameters questioned by the
inspector included acceleration values, friction values between the day tank and the
concrete base pad; the calculated reaction of the block wall; and the shear capacity of the
block wall. All the inspector's questions and observations were properly resolved by thea
.
licensee.
d
1
Conclusion
.
The inspector determined that,the modification package was complete and the frame
properly designed.
Walkdown of the Modifications
The inspector and the licensee design engineer walked down the modifications for the EDG
day tank and the TBSCCW air coolers, with the following details.
With regard to the EDG day tank, the general area was inspected and the framing to
distribute the tank's load appeared to be structurally sound. The tank was inspected and it
was noted that the impact on the block wall (T-27B) was minimal and limited to shear in
the plane of the block wall. The inspector also noted that these block walls had been
previously upgraded in response to NRC Bulletin 80-11. Outside the EDG day tank along
the hallways the inspector noted that proper seismic bracing and anchorage was evident
on the following:
Several modifications to vital electrical equipment (switchgear, load centers and
motor control centers) were installed.
Modification to vital batteries consisted of shim material which was installed to
address A-46 outlier conditions.
Modifications to prevent seismic interaction between lighting fixtures and vital
equipment were installed.
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i
10
)
. At the turbine building, elevation 14'-6" the inspector noted that the TBSCCW air
. coolers identified in the USl A-46 program scope were modified to accommodate
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seismic bracing and anchorage. in the EDG room, more air coolers and other
i
components (air-start tanks, motor control center, and control panel) were
j
seismically anchored or braced to address USl A-46 outliers.
Conclusion
Based on the detailed walkdown of the A-46 modifications, the inspector concluded that
the licensee has performed a substantial number of field modifications to accommodate the
seismic loading on mechanical and electrical equipment identified in the USl A 46 scope.
U1 E2
Engineering Support of Facilities and Equipment
'
E2.1
Adverse Condition Report (ACR) Review
i
a.
Inspection Scope (37550)
i
1
The inspector reviewed an ACR issued on August 16,1996, to assess whether appropriate
)
'
corrective actions were identified and implemented to prevent recurrence of the adverse
condition. The ACR (M1-96-0427) reviewed by the inspector concerned Component
l
Engineering Services / Nondestructive Test Engineering (CES/NTE) ultrasonic test (UT)
instruments that had exceeded their calibration due dates.
]
b.
Observations and Findinos
Six CES/NTE UT instruments were found to have exceeded their yearly calibration due
dates, with four of them having been possibly used during examinations at Unit 1. As
"
' stated on the ACR,'the person who previously handled the CES/NTE material and test
equipment (M&TE) program no longer workea for the company, and none of the job
functions were replaced or reassigned. As a result, there was no ownership of the
CES/NTE M&TE program.
The six instruments in question were sent offsite to a vendor for calibration. All six were
found to be within tolerance when they were received. Additionally, before and after each
examination was performed, the instruments were calibrated in accordance with procedure
NU-UT-1, using a step wedge or calibration blocx The missed yearly calibrations are
j
performed to verify instrument operability only, and do not represent a quality related
i
calibration. In other words, the calibrations done during examinations might be sufficient
to preclude sending these instruments offsite for yearly calibrations. Since these yearly
calibrations are not required by the ASME Code, the licensee will determine whether to
suspend them.
The inspector reviewed Quality Assurance Audit Package A-60607, " Measuring & Test
i
Equipment," an audit that was conducted from August 26,1996 through September 18,
1996. This audit evaluated the key elements and processes of the M&TE program and
determined that the program at Millstone and Connecticut Yankee was ineffective in
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11
fulfilling its mission and did not fully comply with 10 CFR 50 Appendix B criteria. in
- < .
response to this audit, ACR M1-96-0614 was written to address the adverse condition.'
1
i
i
To address the ACRs and audit report, CES/NTE developed a corrective action plan and
dedicated an individual to implement the plan and take ownership of the CES/NTE M&TE
program. Additionally, the licensee ensured that all CES/NTE quality related equipment
currently being used at Millstone and Connecticut Yankee was in calibration.
c.
Conclusion
The inspector concluded that the corrective actions taken by CES/NTE associated with the
ACRs and the QA audit was acceptable. The corrective actions appeared to be broad-
based. A majority of the short term corrective actions were complete, and the long term
corrective actions were being tracked for closure. The safety significance of the UT
j
instruments being past their calibration due dates was minimal because they were
subsequently found to be within tolerance.
E2.2 Containment isolation Check Valve.1-CU-29
a.
Inspection Scope (37551)
'
The inspector reviewed adverse condition report (ACR) 96-0539, which documents an
issue concerning the design specifications of the new containment isolation check valve 1-
1
CU-29. The valve was replaced with a smaller size valve for better flow characteristics,
and to allow testing and maintenance during this outage.
]
b.
Observations and Findinas
i
!
The replacement valve was a specially designed 6" check valve with a 4.5" disc. The disc
'
size was selected based on two flow conditions other than normal operations: 1) shutdown
flow conditions occurring approximately 10% of the time (70 days per operating cycle)
with a minimum system flow of 300 gpm; 2) and startup flow condition, occurring about
1
0.2% of the time with a system flow rate of 100-200 gpm. Due to the current extended
outage, it is not known if the shutdown flow rate will affect the valve. In fact, the RWCU
system is unable to produce the 300 gpm used in selecting the disc size due to the flow
restriction created by the pressure control valve 1-CU-10. During the followup of this
issue, the inspector noted that the actual system flow rate is approximately 190-200 gpm
using the auxiliary cleanup pump, it is not known what effects, if any, will be created by
i
operating the valve at these lower flow rates for this extended period.
In a effort to determine if the valve disc was fluttering or banging into the backseat due to
the low flow conditions, the licensee employed Liberty Technologies to perform non-
intrusive check valve testing. Liberty Technologies performed both acoustic and magnetic
testing on CU-29. Their report stated that the test results provide positive indication of full
opening during the flow initiation test. Acoustic data recorded during steady state flow
indicated no anomalous behavior (such as excessive trim wear or rattling). The report also
'
noted that due to large valve body signal strength from the magnetic sensors during the
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12
valve opening and steady state tests were insufficient to provide useful results. However, .
they determined the acoustic data was a sufficient basis for the conclusions documented.
-
A member of the engineering staff informed the inspector that ACR 96-0539 was currently
open and under review. Engineering needs to determine a method for ascertaining the
actual impact of the lower flow operation on the valve. Localleak rate testing is being
considered as a possible method to determine if any valve degradation has occurred,
c.
Conclusions
The NRC is concerned about the operation of CU-29 with lower than expected flow rates
during this extended shutdown. While the use of non-intrusive check valve testing has
verified that the check valve is backseated, this is a short term indication. The long term
effects of the lower flow operation have yet to be determined. This issue is unresolved
-(URI 245/97-01-05) pending the NRC review of the licensee's finial determination of the
operability of the valve prior to plant startup.
U1 E8
Miscellaneous Engineering issues
E8.1
Closecut Documentation Packaae Review
The inspector reviewed the contents of a corrective action documentation package for NRC
unresolved item 96-04-07 (safety relief valve electric lift modification). The inspector
noted that the package included a draft licensee response to an NRC request for additional
information associated with the license amendment. Per previous arrangements the
licensee documentation packages were to be complete and contain only approved
documents. The package was returned to the licensee and no inspection was performed at
this time.
.
.
13
Report Details
Summary of Unit 2 Status
Unit 2 entered the inspection period with the core off-loaded. The unit was initially shut
down on February 20,1996, to address containment sump screen concerns and has
remained shut down to address an NRC Demand for Information [10 CFR 50.54(f)] letter
requiring an assertion by the licensee that future operations are conducted in accordance
with the regulations, the license, and the Final Safety Analysis Report.
U2.1 Operations
U201
Conduct of Operations
01.1 General Comments (71707)
Using Inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing
plant operations to ensure that licensee's controls were effective in achieving continued
safe operation of the faciiity while shut down. The inspectors observed that proper control
room staffing was maintained, access to the control room was properly controlled, and
operator behavior was commensurate with the plant configuration and plant activities in
progress. In general, the conduct of operations was professional and safety-conscious.
Operations Management has recently placed greater attention on improving performance
associated with operator response to control room alarms with a focus on communications
and use of alarm response procedures. The NRC has noted the improvements in this area,
i
particularly regarding control room operators communicating to the unit supervisors what
alarms were received and ensuring a mutual understanding of why the alarm was received.
The licensee discovery of potential discrepancies in the personal qualification statements
(NRC Form 398) of certain Unit 2 licensed operators has been assessed for immediate
impact and determined to require further evaluation. This is considered an unresolved
,
issue as described in Section U1.05.1 of this inspection report.
The inspector toured the Unit 2 intake structure and found the material condition of
systems and components to be acceptable, in the Spring of 1996, the licensee instituted a
corrective action plan to address material deficiencies. A number of items have been
corrected, which has improved the material condition of the intake structure.
U2 O2
Operational Status of Facilities and Equipment
02.1 Adverse Condition Report Backloa
a.
Insocction Scope
The NRC evaluated the timeliness in which the licensee completed corrective actions
associated with Unit 2 adverse condition reports (ACRs).
.
.
14
b.
Observetions and Findinas
Timeliness for completion of corrective actions has been a longstanding concern at
'
Millstone. Having an ACR backlog in itself is not a reflection of poor performance because
as the threshold for writing ACRs decreases, the ACR backlog willincrease accordingly.
The concern is the number of ACRs that are not closed in a timely manner. To help
provide the NRC some sense of the licensee's progress in addressing the timeliness
concern, the licensee was asked to provide the number of ACRs having outstanding
corrective actions that are greater than 120 days old. Although the NRC does not consider
120 days a level of excellence nor is it acceptable when addressing immediate safety
concerns, it does provide some understanding of licensee management effectiveness in
addressing the corrective action timeliness issue.
Several months ago, the NRC raised a concern that the licensee's ACR database did not
allow them to determine the number of ACRs having outstanding corrective actions. The
1
licensee's previous understanding, as documented in NRC Inspection Report (IR) 50-
336/96-09, was that the ACR data entries had been corrected to provide reliable ACR
'
backlog numbers. However, additional licensee reviews of the ACR database indicate that
the number of ACRs greater than 120 days old as of December 31',1997, was 940 ACRs,
i
no_t 732 ACRs as stated in IR 50-336/96-09. The increase of 208 ACRS is based on a
licensee review of previously closed ACRs that they decided to reopen based on
incomplete closure documentation. At the end of the current inspection period (February
24,1997), there were 798 ACRs greater than 120 days old that have not been closed.
DEPARTMENT
ACRs OLDER
THAN 120 DAYS
Operations
56
Design Engineering
211
Technical Support (System Engineering)
254
Work Planning
28
Maintenance
55
_
l&C
42
Safety / Licensing
25
Other
127
TOTAL
798
c.
Conclusion
Although the backlog of 798 adverse condition reports (ACRs) that are greater than 120
days old indicates that timeliness for completing corrective actions continues to be a
,
>
15
concern, the reduction in this backlog of older ACRs from 940 to 798 since the last.
inspection period is a positive trend which reflects the licensee's increased level of effort in
this area. As discussed in NRC Inspection Report 50-336/96-04, timeliness and
effectiveness of corrective actions is an area in which the licensee must demonstrate
sustained improved performance.
U2 08
Miscellaneous Operations issues (92700)
08.1 (Closed) Violation 50-336/94-17-10: O_peration Outside System Desian Parameters
a.
inspection Scope
The sc-ope of this inspection included a review of Violation 50-336/94-17-10.
b.
Observations and Findinas
This violation involved the failure to correctly translate design basis temperature limits of
the service water (SW) and reactor building closed cooling water (RBCCW) 2ystems into
operating procedures. As a result, on May 24,1993, a reactor trip occurred when the SW
and RBCCW system temperature limits were exceeded during a main condenser thermal
j
backwashing evolution. This violation was previously reviewed in NRC Inspection Report
50-336/96-05 which concluded that the violation could not be closed because although
'
the specific procedures regarding thermal backweshing were adequately aabemd, the
corrective actions were too narrow in that they failed to address the possibility eat other
plant procedures did not insure operation was in accordance with the plant's oesign basis.
c.
Conclusion
Violation 50-336/94-17-10, which resulted from a 1993 unresolved item, reflects that the
failure to operate the plant in accordance with the design basis had been a longstanding
NRC concern. The licensee failure to address this type of concern eventually culminated in
their current extended shutdown and 10 CFR 50.54(f) effort which is intended to ensure
i
the plant is designed and operated in accordance with the licensing and design basis.
There are several outstanding violations including Escalated Enforcement items 50-336/96-
'
06-05 & 96-08-06 which also address plant operation that is inconsistent with the
licensing basis. Therefore, Violation 50-336/94-17-10 is being closed nql because
adequate corrective actions have been taken but because this concern is being addressed
and tracked by more recent items.
'
08.2 (Closed) LER 50-336/96-15, (Open) Unresolved item 50-336/96-01-04: Failure to
Enter Action Statement Reaardina the Number of Operable Nuclear instrument
Channels
a.
inspection Scooe
The scope of this inspection included a review of Licensee Event Report 50-336/96-15.
.
.
16
b.
Observations and Findinas
On March 12,1996, while the unit was shut down, the "B" train vital de bus was
inadvertently deenergized due to operator error. The resultant loss of various vital and non
vital power supplies overflowed the reactor building closed cooling water surge tank
through a failed open make-up valve, and challenged the operators to recover from this
complex event. The event was complicated by the fact that there was minimal procedural
guidance for operators to use to recover the bus. One train of shutdown cooling remained
in operation throughout the event and normal power was restored within four hours.
During the recovery, operators were required by procedure to deenergize bus VA-40, a
vital 120 Volt ac instrument panel which supplies channel "D" of the reactor protection
system (RPS). Operators were aware that a loss of RPS channel "D" would cause the
channel "D" wide range nuclear instrument to be inoperable. Prior to the loss of the "B"
train vital de bus, in an unrelated situation, channel "B" and "C" wide range nuclear
instruments had been declared inoperable. The action statement for Technical Specification (TS) 3.3.1.1 is applicable when less than two channels of wide range nuclear
instruments are operable. The action statement requires the immediate verification of
shutdown margin and at least once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter. The operators did not
~ recognize the need to erter the TS 3.3.1.1 limiting condition for operation when bus VA .
40 was deenergized.
LER 50-336/96-15 stated that a contributing cause for failing to enter the TS 3.3.1.1
action statement was that the shift was occupied with the restoration of the deenergized
dc bus and was considering the effects of deenergizing loads on plant operatico. The
licensee's corrective actions included training of operators to not only consider the effects
of their actions on plani operations, but they must also assess TS requirements, in
addition, operating procedures were changed to remind operators to determine if TSs are
affected when deenergizing an electrical bus.
c.
Conclusion
The failure of operators to enter the action statement for TS 3.3.1.1 when three channels
of wide range nuclear instrumentation were inoperable is considered a violation. This
licensee-identified and corrected violation is being treated as a Non-Cited Violation,
consistent with Section Vll.B.1 of the NRC Enforcement Policy. The primary concern
associated with tnis event was the fact that there was minimal procedural guidance
provided to operators to recover the loss of de bus. The licensee is in the process of
preparing 12 abnormal operating procedures for recovering various dc buses and
distribution panels. This concern is being tracked by Unresolved Item 50-336/96-01-04.
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U2.ll Maintenance
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U2 M8
Miscellaneous Maintenance issues (92903)
M8.1 (Closed) Unresolved item 50-336/96-04-09: Troubleshootina Controls
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a,
lnspection Scope
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in April 1996, the NRC resident inspectors identified programmatic concerns regarding the
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conduct of troubleshooting. The concerns centered around the practice of performing
" troubleshooting" under the guise of " investigating" to avoid implementing the
administrative requirements for the performance of troubleshooting that are contained in
procedure WC-1, " Work Control Process." This issue was unresolved pending licensee
changes to WC-1.
b.
Observations and Findinas
in July 1996, the licensee issued Attachment 5.2 to procedure WC-1. This attachment
provides guidelines to be used in conjunction with a work order when a formal
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troubleshooting plan is not required by Attachment 5 of WC-1.
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The inspector reviewed the instructions contained in WC-1 and reviewed several work
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orders that performed troubleshooting since the issuance of the change to WC-1.
c.
Conclusions
The inspector found that Attachment 5.2 of WC-1 contains appropriate directions to
ensure that all troubleshooting work is documented, supervision is consulted prior to
.
performing repair or replacement of components and retest requirements are determined
following the completion of the troubleshooting. No problems were identified during the
review of the troubleshooting work orders. This item is closed.
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M8.2 (Closed) Unresolved item 50-336/96-06-06: Hioh Pressure Safety Iniectiori Check
Valve Backflow Testino
a.
Insoection Scope (92903)
The inspectors identified that the licensee had unnecessarily relaxed the frequency of
backflow testing of high pressure safety injection system (HPSI) pump discharge check
valves. The licensee agreed to revise the surveillance procedure to perform quarterly
backflow testing. This issue was unresolved pending NRC review of the planned
procedure changes,
b.
Observations and Findinas
The licensee revised procedure SP 21136, " Safety injection and Containment Spray
System Valves Operational Readiness Test," to include quarterly backflow testing of the
HPSI pump discharge check valves.
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c.
Conclusions
'
The NRC concluded that the licensee had appropriately resolved the check valve testing
concern in Revision 10 of procedure SP 21136 and the associated data forms. This item is
closed.
U2.lll Enaineerina
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U2 E8
Miscellaneous Engineering issues
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E8.1 LClosed) Unresolved item 50-336/95-07-06: Condensate Storaae Tank Siohon Break
a.
Inspection Scope (92903)
On February 10,1995, the licensee discovered that the condensate storage tank (CST)
level had dropped to approximately 30% due to a heat exchanger tube leak. At the time of
the event the plant was defueled and no minimum tank volume was require by the plant.
technical specifications. During an investigation of the inadvertent loss of CST inventory
the licensee discovered that a siphon break (a 1/2 inch hole) in the tank recirculation piping
was missing. This issue was unresolved pending further review of how the plant design
change process missed the removal of the siphon break,
b.
Observations and Findinas
Plant information Report (PIR) 2-95-174 documented the licensee's investigation of the
missing siphon break. The licensee's review concluded that the section of piping, in which
the siphon break was located, may have been removed and replaced during a modification
performed in 1992 to install a CST nitrogen blanketing system. The modification included
the addition of a stiffener beam on the internal tank wall that required the recirculation loop
suction piping to be modified to provide clearance for the beam.
The interference problem between the stiffener and the recirculation piping was not
identified during the initial modification design. When the problem was identified a design
change notice (DCN) was processed to provide details on how to modify the piping.
However, the drawing provided with this DCN did not depict the existence of the siphon
break hole.
The work description in the work order that modified the piping was to " fabricate and
installinternal piping to the CST per drawing #25203-13006 Sheet 44". Although only
the section of piping at the elevation of the stiffener beam was affected, the work order
did not preclude the replacement of the section of piping that contained the siphon break
hole, nor did the drawing provide the details necessary to drill the hole in the event the
piping was replaced.
The cause of the event appears to be inadequate attention to detail during the preparation
of the DCN and the associated drawing since other drawings (the system piping and
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instrumentation drawing and the original piping isometric drawing) showed the existence of
a siphon break hole.
The licensee briefed the design engineering staff on this event and the need to pay
particular attention to less obvious design attributes such as siphon breakers when
developing design changes. Also, since the time of this event, the licensee has
implemented programmatic improvements to the design control process to improve the
effectiveness of the process.
The inspector noted that the plant technical specifications require the maintenance of a
minimum volume of 150,000 gallons of water in the CST when in modes 1,2 or 3 and the
volume is verified every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. These requirements reduce the possibility of a
significant, undetected loss of CST inventory when the plant is in operation.
c.
Conclusion
l
The inspector concluded that the licensee had taken appropriate actions to resolve this
issue. However, the inspector noted that the licensee evaluation could have been more
thorough in that the investigator did not review the associated work order until questioned
by the inspector. A review of the work order was necessary to determine if the event may
have been a result of poor work controls, which has been a problem in the past at this site.
This item is closed.
E8.2 LClosed) Unresolved item 50-336/95-11-03: 10 CFR 21 Reportability Review
a.
Inspection Scone (92903)
Following the licensee identification of several design problems that affected replacement
components of the engineered safeguards actuation system (ESAS) cabinets, the NRC
inspectors questioned if the findings had been reviewed for reportability under the
requirements of 10 CFR 21, " Reporting of Defects and Noncompliance." At the time of
the inspection in 1995, no formal review had been initiated and the issue was unresolved
pending further actions to be taken by NU, and NRC review of the licensee actions.
b.
Observations and Findinas
in June 1995, the licensee performed an assessment of the design problems experienced
with the ESAS system during the previous two refueling outages. The assessment
identified four significant design problems that were related to defects in the components
provided by the vendor. The four problems were assessed for reportability in engineering
evaluation M2-EV-97-0004, Revision 0, " Evaluation of URI 95-11-03, Reportability of
ESAS Design Deficiencies."
Three of the four issues had been reported to the NRC in Licensee Event Reports (LERs)
94-12-00, 95-18-00 and 95-21-00. The issue associated with LER 94-12-00 had also
been reported by the vendor ~(Eaton Corporation) in accordance 10 CFR 21. The fourth
issue was documented in ACR 00506 and the licensee had determined that the issue was
not reportable.
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The engineering evaluation notea that problems reported in an LER did not require
additional reporting in accordance with 10 CFR 21. The bases for an LER fulfilling the
{
licensee's reporting obligations under 10 CFR 21 is contained in 10 CFR 21.2(c).
'
The inspector also noted in discussions with the licensee that the components that were
'the subject of LERs95-018 and 95-021 were designs that were unique to Millstone Unit 2.
c.
Conclusion
The inspector reviewed the licensee evaluation, LERs and ACR and found that the licensee
appropriately evaluated and reported the failures to the NRC. This item is closed.
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Reoort Details
Summary of Unit 3 Status
Unit 3 remained in cold shutdown (mode 5) status throughout the inspection period. The
licensee continued its implementation of the Millstone Unit 3 Recovery Plan and the
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configuration management program activities in support of the milestones leading to the
readiness for the unit restart. In accordance with commitments made to the NRC with
regard to corrective action progress and documentation of the completed work items, the
licensee provided the first set of corrective action completion packages for NRC review.
To date, the presentation of such packages to the NRC has been timely,. relative to the
scheduled workload. This documentation has also provided evidence of progress in the
resolution of open NRC inspection items, as well as an indication of the licensee efforts to
demonstrate corrective action program effectiveness. NRC review of the available closure
)
packages will continue as an ongoing process, with the individual technical issues
,
discussed, as appropriate, in the following report sections and in future NRC inspection
reports.
On January 22,1997, the appointment of Mr. M. H. Brothers, then the Unit 3 Director, to
the position of Vice President-Millstone Unit 3, was announced, in this new position, Mr.
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Brothers fills the role of Recovery Officer for the unit. On January 28,1997, Mr. Brothers
announced the appointment of Mr. G. D. Hicks, who had been serving on the Carolina
Power & Light Recovery Team for Unit 3, to the position of Unit 3 Director, in an acting
capacity. Both of these managerial changes became effective on February 3,1997. The
inspector noted that Mr. Hicks' qualifications to assume the Unit 3 Director position had
been reviewed by both the licensing department and the Nuclear Safety & Oversight
organization. -The inspector also reviewed section 6.3.1 of the unit technical specifications
and American National Standard, ANSI N18.1-1971, regarding " Plant Managers" and -
?
d
identified no qualification concerns or other questions regarding these licensee
management changes.
U3.1 ODer8tions
U3 01
Conduct of Operations
01.1 General Comments (71707)
Using inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing
plant operations. During a walkdown of the Unit 3 intake structure, the inspector observed
large scale painting and materialimprovement work in progress. Various degrees of work
activity have been on-going in the intake structure since the licensee instituted a material
condition improvement project in the Spring of 1996. The inspector noted that adverse
condition report M3-97-0370, dated February 1,1997, was written to report that station
air and instrument air supply piping in the intake structure was in poor extemal condition.
The licensee is planning to replace or paint the piping. In general, the conduct of
operations was professional and safety-conscious; specific events and noteworthy
i
observations are detailed in the sections below.
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< Over the course of this inspection, the inspectors witnessed and/or reviewed a number of
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,. operational activities, and noted the following observations and assessments of operations
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performance:
Good contingency planning (e.g., preparation for field flashing the "B" emergency _
f
diesel generator during the Battery 2 outage) was in evidence.
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Appropriate regulatory provisions (e.g', a technical specification [TS] " bases"
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change to allow operation of the safety injection pumps in mode 5 to fill an
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accumulator) were considered and dispositioned to address emergent operational
conditions.
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Consideration of how the incore flux mapping results affect the quadrant power tilt
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ratio (OPTR) was found to be consistent with both the TS definition for QPTR and
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the Westinghouse position statement on " core tilt".
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Evaluation of the calibration provisions for the range and accuracy of digital
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instrumentation utilized in the performance of operational surveillances was
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determined to meet the NRC guidance discussed.in NUREG-1482.
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Reductions in the range of reactor coolant system operating temperatures (i.e., T-
avg) to maintain margins relative to nil ductility transition (NDT) considerations were
!'
implemented consistent with both TS 3.4.10 and ASME Code requirements.
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The implementation of temporary modifications (e.g., a bypass-jumper for cross-
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tying trains in the auxiliary feedwater system flow paths) was determined to be
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conservative in providing additional heat sinks (i.e., two steam generators) for
mshutdown risk considerations.
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The licensee discovery of potential discrepancies in the personal qualification-
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statements (NRC Form 398) of certain Unit 3 licensed operators has been assessed
for immediate impact and determined to require further evaluation. (NOTE: See
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unresolved item associated with similar Unit 1 activities - Section U1.05.1 of this
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inspection report)
4
Additionally, during control room inspections and reviews of TS limiting condition for
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operability (LCO) action statements, the inspector raised a question regarding the
!
applicability of actions for single system / train inoperability when more than one system or
train is determined to be inoperable. Examples where the need for interpretation of the
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LCO actions might be appropriate for multiple system or component unavailability were
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identified in TS 3.7.7 and 3.7.1.2. The inspector also noted recent correspondence (i.e., a
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memorandum dated January 17,1997) from the Unit 3 licensing staff providing a
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clarification of TS terminology, e.g., how to interpret "at least once per 7 days". The
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inspector determined that the licensee needed to further develop its approach to
4'
promulgating such interpretive guidance. While no TS violations or technical concerns
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were evident in the areas of TS compliance questioned by the inspector, a standardized
method for disseminating Unit 3 policy in the interpretation of TS language and LCO
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actions appears prudent. The inspector discussed this issue with cognizant licensee
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personnel and intends to review this matter further as an inspector followup item. (IFl
423/97-01-06)
01.2 Control of Hiah Enerav Line Break (HELB) Doors
a.
Insoection Scooe (71707)
On several occasions during this inspection period the licensee identified that HELB doors
were open or not fully latched. These conditions were reported in accordance with 10 CFR 50.72 as a condition that could have prevented a safety system from functioning as
required. The inspector reviewed the licensee's corrective actions to assess the
effectiveness of the licensee's root cause determination and whether appropriate corrective
actions were identified and implemented to prevent recurrence of the adverse condition.
b.
Observations and Findinas
,
On January 10,1997, the licensee reported that the HELB door to the "A" train 4160 volt
switchgear room was open. The door had been opened to facilitate battery 301 A-1
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replacement. Within the next few days, four additional occurrences of an open or
improperly latched HELB door were identified, including one by the resident inspector. All
the doors had a HELB sign affixed on both sides of the door.
As a result of these incidents, a level "B" adverse condition report (ACR) was gererated
i
and an event review team assembled. As an immediate corrective action, a work stand
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down was held within all departments and licensee management briefed employees on the
control of plant doors. During the next several days, several other ACRs regarding HELB
issues were generated; including doors not being appropriately labeled in the field or on
prints.
The licensee's root cause investigation detarmined that the cause of the events was a
failure to develop and implement a HELB door control program, which resulted in a lack of
understanding of the HELB requirements associated with the plant design basis. Corrective
actions included: develop a door control program, insert a door control training module in
general employee training, and simplify and label all HELB doors. Although not specifically
stated in the corrective action plan, the licensee indicated that door labels would specify
the number of turns required to latch the doors.
As part of the investigation, the licensee concluded that the HELB door issue was not a
reportable condition with the plant in modes 5 or 6. Final Safety Analysis Report, Section
3.6.1 states that a high energy system is a fluid system that operates during normal plant
operating conditions. Normal plant conditions are defined as startup, operation at power,
hot standby, or reactor cooldown to cold shutdown conditions. Therefore, those
conditions that were identified while in mode 5 were retracted. However, the licensee
concluded that the status of HELB doors should be considered in the shutdown risk
program since a rupture of the hot water heating line in the service building could result in
the loss of safety-related equipment located in the switchgear room, independent of the
plant mode.
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The inspector toured the plant and verified that HELB doors 'nere properly latched.
Applicable HELB doors were labeled to indicated the required number of turns to properly
latch the door. The inspector also verified that a training module was being developed for
inclusion in the general employee training. The other corrective actions are scheduled to
be complete prior to the unit entering mode 4.
c.
Conclusion
The licensee's root cause investigation and corrective action plan for control of HELB doors
were determined to be good. However, the requirement to label the required HELB doors
with a minimum number of turns to ensure prope- U.ching should have been included in
the ACR corrective action plan if it was deemed necessary to prevent recurrence.
U3 03
Operations Proceduies and Documentation
a.
Inspection Scope
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The purpose of this inspection was to determine the adequacy of the procedure upgrade
program (PUP) as it applies to Unit 3. The licensee started the PUP in 1992 to standardize
procedure format for all units on the site and to improve the technical adequacy of all
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procedures. The process was a third iteration of previous procedure improvement
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programs which were started in the late 1980's. This inspection was performed from
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August 1996 through February 24,1997.
The onsite inspection included interviews with the site PUP group, Unit 3 procedure
coordinators and procedure writers, station oversight group, station quality assurance, and
Unit 1 operations personnel. Documents reviewed included, but were not limited to,
document control (DC) procedures: DC-1, " Administration of Millstone Procedures and
Forms"; DC-2, " Developing and Revising Millstone Procedures and Forms"; DC-3,
" Verification and Validation of Millstone Procedures and Forms"; and a sample uf Unit 3
procedures which had already been upgraded. in addition, the inspector reviewed the
Millstone Unit 3 PUP self assessment of their Operations Department (conducted March-
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May,1996); Millstone Unit 1 PUP self assessment of their Operations Department
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(conducted May-June,1996); Station oversight audit conducted during September 1996;
and various procedure related adverse condition reports (ACRs).
Although the focus of this inspection was Unit 3, the Unit 1 self assessment was reviewed
and discussed with Unit 1 personnel who performed the assessment and the Unit 1
Operations Manager. The inspector considered its potential applicability to Unit 3.
b.
Observations and Findinas
,
At the start of this inspection, there was a station wide Procedure Upgrade Group to
provide overall control of the program. This group developed and maintaineci the station
DC procedures for control of the program, the overall status of upgraded procedures,
coordinators for each Millstone Unit, and the hiring of contractors, as necessary, to write
the procedures. The actual upgrade of procedures was the responsibility of each
department within each unit. Since the licensee's reorganization in October 1996, the
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25
station PUP group has been decentralized. The group now controls the station
administrative procedures including the PUP DC procedures. It has no control of the
production of upgraded procedures. Despite the changes in PUP control, the quality and
quantity of upgraded procedures has depended on the individual technical departments in
each unit.
A review of the program as it applies to Unit 3 noted the following:
The program has been effective in standardizing procedure formats. The document
control procedures are lengthy and cumbersome to use, but appear to be
comprehensive. The quality of the upgraded procedures appears to depend on the
producers of the procedures and the adequacy of the performance verification and
validation (V&V) process rather than on any apparent process deficiency.
1
The V&V process can be by table top review, procedure walkdowns, or by
i
procedure performance. All three methods are used, but the table top review by a
technical peer is the most common form of validation for Operations and
Maintenance Department procedures. Instrument & Control procedure technicians
stated that when possible their V&V process consisted of procedure walkdowns. -
In general, management involvement in the upgrade process seems to be minimal.
Some department managers have more involvement than others. There is heavy
reliance on the procedure coordinators for each department. For example, the Unit
,
3 operations manager authorized his procedure coordinator to act on his behalf for
procedure review and approval.
,
The inspector noted that during the five years that the PUP has been in place, there
had been no Quality Assurance (QA) audits of the program itself. Unit 3 Operations
Department had performed a self assessment of the PUP process in May 1996, but
this assessment was fairly limited. As a result of this NRC inspection, the licensee
performed two further assessments of the PUP program as it relates to Unit 3.
An assessment of the verification and validation process by Nuclear Oversight for
Unit 3 was performed in September 1996. This assessment noted some
weaknesses and problems in the V&V process. Another assessment was performed
on the PUP program for Units 1,2 and 3, November 13-16,1996, by the Region
One Procedere Working Group. This group is composed of persons from other NRC
Region i nuclear utilities. This assessment also noted strengths and weaknesses.
Both assessments noted deficiencies in the PUP program but did not conclude that
the program was seriously flawed.
On February 19,1997, the licensee forwarded to the inspector three recent QA
audits and fourteen quality control surveillance activities performed in the last
quarter of 1996 and the first quarter of 1997. These audits and surveillance
indicated that procedures are routinely reviewed as part of that activity and, where
appropriate, procedure deficiencies are identified. Also forwarded were recent
adverse condition reports identifying procedure problems.
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26
As part of the PUP, the licensee developed the procedure basis document. The
Intent'of this document was to duplicate the procedure and to add blocks at certain
points to indicate the source or basis for key technical information such as the Final
~ Safety Analysis Report (FSAR), vendor. manual, technical specification, regulatory
commitments, etc. The Region One Procedure Working Group report considered the
basis document a strength, inspector's conversations with many plant personnel
,
indicated that they had high expectations from this document.
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' The inspector reviewed numerous basis documents during the course of this
'
inspection. While the basis document concept generally appears to be a strength,
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the inspector noted numerous basis documents to be incomplete. The documents
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were essentially the procedure with one or two basis blocks added in a pro forma
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matter. Documents referenced in both the procedure and basis document were not
further identified in the basis document as to where the referenced material actually
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applied. This appears to be's weakness in the practical usage of basis documents.
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The licensee does not, however,' use the basis document as a procedure for plant
operations.
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^The inspector did a review of some Unit 3 operations procedures with a Unit 3 staff
7e
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engineer. Some minor technical (Hecrepancies were observed,' but were not
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considered by the inspector to be' safety significant. In some cases, the operations
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. engineer had difficulty in determining the source for specific information in certain
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procedures. For instance, the calculation for one instrument setpoint was only
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available in the desk drawer of an l&C engineer; and the source of the setpoint was
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not in the basis document. This was an example of one of several instances of
apparent configuration management control. However the configuration
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management problem has already been identified to the licensee and is being
addressed generically by the licensee.
As noted in previous NRC inspections, design basis discrepancies have been
identified at all three Millstone units. It is possible that some procedures may not
conform to the FSAR and other procedures may conform to FSAR conditions which
are contrary to the actual design basis. As a result of an NRC 10 CFR 50.54(f)
letter, the licensee is currently performing an extensive design basis review which in
turn will be independently verified. In a letter from the licensee dated July 22,
'1996, the licensee stated, in part, the following: "...As we believe was
communicated [to the NRC in a meeting conducted on April 30,1996], there was
no commitment to complete the PUP as a conditio 1 of restarting any of the
[ Millstone] units. As part of the Operational Readiness Plan for Millstone unit No. 3,
findings resulting from the 10 CFR 50.54(f) related work will be reviewed to
determine if any procedure modifications are required prior to restarting the unit.
This will be done independent of procedure upgrades completed via the PUP..."
The licensee's corrective action plan for any required " procedure modifications" will
be assessed in the future as part of the NRC restart assessment plan,
The inspector observed that Unit 3 Operations Department had established a 67
page handbook in order to implement the DC procedures for the PUP process. This
appeared to be an uncontrolled, unapproved and unofficial procedure. The licensee
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stated at the conclusion of the onsite inspection that this handbook would be made
into an officially controlled procedure. By telephone on February 26,1997, the
inspector was informed that the operations handbook had been deleted and only the
DC procedures were being used by the Operations Department as guidance for the
PUP process.
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From January 13 through January 30,1997, the licensee's Nuclear Oversight
Group conducted an audit in the areas of document control and the maintenance of
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quality records. This audit identified a site-wide breakdown in the control of
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procedures in that procedures in use had not been properly updated. The
operations handbook (now deleted) was an example of an uncontrolled procedure in
use. Because of its magnitude, the adverse condition report (ACR) generated by
this audit was initially recommended for classification as a Level "A" ACR.
c.
Conclusions
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The procedure upgrade program meets regulatory requirements and has been effective in
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standardizing procedure formats across the site. The technical adequacy of upgraded
procedures, except for a small sample of Unit 3 operations procedures, was not a subject
for this inspection. Because of the number of individuals involved in procedure upgrade
and long period of time to upgrade the procedures (5 years), the quality of procedures
vary. A number of ACRs reviewed indicate technical problems with some procedures
already upgraded. The licensee has committed as part of their 10 CFR 50.54(f) process
and their Configuration Management Prograin to ensure procedures will meet the applicable
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design bases. The licensee's basis documents, as implemented, did not appear to meet
the licensee's own expectations as an adequate foundation for future procedure reviewers
to have an adoquate technical basis for the information contained in each procedure.
U3 O'
Quality Assurance in Operations
07.1 General Comments (40500,92901)
The inspector reviewed station procedures, assessed planned program changes, and
discussed various quality assurance activities with representatives from the Nuclear Safety
& Oversight (NS&O) organization and Unit 3 licensing, engineering and operations
departments. The following topics were generally reviewed and evaluated during the
conduct of this inspection:
corrective action program changes (Revision 4 to the station procedure, RP-4,
addressing " Corrective Action", effective February 25,1997)
job rcAation between the Unit 3 line departments and the NS&O organization
engineering assurance recovery activities, conducted by the NS&O organization, to
include planning and preparation for an Integrated Assessment Plan imp!ementation
and review of the proposed changes to the Design Basis Document Package
upgrade program
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issuance of a " Millstone 3 Line/ Oversight Interface Agreement"
implementation of and subsequent release from a Quality and Assessment Services
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(QAS) " Hold" of all work on safety-related eqaipment requiring the use of "non-QA"
parts
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With the progress of the noted program revisions, recovery activities, and organizational
'
initiatives still ongoing, the impact and effectiveness of the changes have not yet provided
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measurable results. As of the end of this inspection period, the inspector observed
increased NS&O involvement in performance monitoring, interfacing analysis, and support
of the Unit 3 rnanagement and line staffs. Such involvement has included "real time"
evaluation and feedback on routine operational activities and nonroutine' events. NRC
assessment of NS&O effectiveness (including an expectation of demonstrable results of the
,
corrective action program improvements) and specific QAS activities (e.g., Hold 97-1,
Revision 1) will continue over the course of the next severalinspection periods; covenng
the ongoing recovery, open item closure, and work associated with the startup planning for
the unit.
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U3.Il Maintenance
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USM1
Conduct of Maintenance
M 1.1 General Comments
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a.
Insoection Scoce (62707/61726)
The inspector observed / reviewed all or portions of the following maintenance and
surveillance activities to verify proper calibration of test instruments, use of approved.
+
procedures, performance of work by qualified personnel, and conformance to technical
specification (TS) limiting conditions for operation.
M3-96-15203,
Battery 301 A-1 Removal
M3-96-25108,
Calibrate ITT Barton (SP 3481B01)
SP 3626.8,
" Control Building Air Conditioning Booster Pump,3SWP'P2A,
Operational Readiness Test"
The inspector found the work performed under these activities to be professional and
thorough. All activities observed were performed with the work package or surveillance
procedure present at the job site and personnel were noted to be closely following the
procedures. Review of the surveillance procedures revealed that the requirements of the
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applicable TS were appropriately incorporated into the implementing procedure,
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29
M1.2 Fix-It-Now (FIN) Conduct of Maintenance
a.
Inspection Scope (62707)
^
The licensee has implemented a number of changes to the work control process including
the development of a FIN multi-discipline work team approach to augment the way
maintenance is performed at the unit. This process is in addition to their normal work
control process. The inspector reviewed the FIN maintenance procedure and monitored
work that was performed under the FIN process to assess its implementation.
b.
Observations and Findinas
<
As part of the Unit 3 recovery process, the licensee developed a FIN work process to
reduce the time it takes to respond to operations department needs, and to perform work
more efficiently. In addition, a minor maintenance work process was developed to allow
,
qualified personnel to accomplish tasks in a more efficient manner. It was envisioned that
these work processes would contribute to reducing the corrective maintenance backlog.
These work processes were initiated in November 1996.-
Maintenance procedure MP 3705B, "Fix It Now Conduct of Maintenance," states that the
FIN work process shall not be implemented for work requiring: major plant modifications,
lengthy tag clearances, special radiological work permits that cannot be dispositioned by
the health physics (HP) team member, or repairs involving welding on specified plant
equipment. The minor maintenance procedure U3 WC 1.1, " Minor Maintenance Process
"
Controls," allows work to be performed without the generation of a work order if the work
i
is performed on non-QA equipment and the work doesn't impact plant operations or require
special work control needs.
,
i
The FIN team is comprised of mechanics, electricians, maintenance planners, and
'
instrument and control (l&C) technicians. In addition, there is a representative from the HP
and operations department, and an assigned first line supervisor (FLS). The inspector
reviewed the training records for each FIN team member and verified that they were trained
,
on the new procedure. In addition, team members appeared to be well qualified. At least
i
one member of the team was qualified for each of the matrices job tasks.
Each morning the FIN team members review all the trouble reports (TRs) that were
generated the previous day and determine which items can be performed by the team and
which need to go through the normal maintenance process. At the 6:45 a.m. morning
meeting, the FIN team FLS notifies the operations shift manager and the work planning
organization of the selected work items to ensure that these departments are cognizant of
all FIN planned maintenance activities.
The inspector attended the FIN and the 6:45 a.m. morning meeting. In addition to the off-
going and on-coming shift managers, individuals at the 6:45 a.m. morning meeting
included: the Unit Director, the operations, maintenance, l&C, and engineering department
managers, and representatives from work planning, chemistry, and HP. Each TR is
discussed then assigned to either the FIN or work planning department. Any TR that
potentially affects equipment operability is identified and an adverse condition report
.
l
.
30
generated. The inspector noted that all work on the protected train was assigned to the
work planning department.
There have been approximately one hundred TRs generated each week during the month of
January 1997. Of these, the FIN team completed approximately fifty-five percent. A
review of the TR backlog, those over one week old, revealed that the number has been
declining. The inspector monitored selected activities and reviewed those work items that
were performed by the FIN team for the month of January. A review of the work activities
revealed that work orders had been generated for all safety-related work activities in
accordance with procedure U3 WC 1.1. As a result of the high number of TRs being
generated and the minor maintenance work activities performed, the FIN team has been
unable to work off any corrective maintenance backlog items.
c.
Conclusions
The licensee is implementing the FIN work process in a conservative manner. Any work on
the protected train equipment is not being assigned to the FIN team. All monitored work
activities performed by the FIN team was performed in accordance with the unit and
station procedures. FIN team members appeared to be well qualified. No safety concerns
,
were identified from the specific activities observed.
U3 M8
Miscellaneous Maintenance issues
M8.1 Plant Insoection-Tours (62707. 92902)
The inspectors conducted inspection-tours of several areas of the plant during this
inspection period, observing work in progress and raising some questions regarding
completed field installations. As appropriate, discussions were held with workers, field
supervisors, and support personnel (e.g., health physics technicians). While most field
observations and questions were resolved prior to completion of the inspection-tours, the
following issues required followup, as documented below:
snubber removal on a residual heat removal (RHS) line in the engineered safety
features building; authorized by plant design change record (PDCR) MP3-90-003.
The inspector reviewed the applicable PDCR and design change notice DM3-P-154-
90, verifying proper re-analysis of the RHS piping system and control of the snubber
elimination list. Since design criteria discussed in ASME Code Case N-411 were
used in the pipe stress re-verification, the inspector reviewed the related discussion
of seismic design response spectra, provided on NRC Regulatory Guides 1.60 and
1.61, in the final safety analysis report (FSAR); and confirmed NRC approval for the
use of ASME Code Case N-411 at Unit 3.
white " frothing" of oil observed in the site glass for the speed increaser on the "B"
charging (CHS) pump, located in the auxiliary building.
The inspector discussed this observation with the responsible system engineer, who
confirmed that the subject " frothing" was likely due to the turbulence caused by the
.
.
31
meshing of gears in the CHS pump speed increaser. The inspector reviewed the
results of the most recent chemical analysis performed on this pump and identified
no adverse conditions or additional concerns.
treatment of reactor plant sampling (SSR) tubing runs and flexible hose connections
inside the containment building as ASME class 2 components, as discussed in the
Unit 3 FSAR.
.
The inspector reviewed the fabrication installation control drawings for the SSR
piping and common header connections from the containment penetrations to the
i
steam generator blowdown lir.es (i.e, a review of approximately 60 Stone &
Webster Engineering Corporation isometric drawings); and confirmed proper ASME
j
Code classification of the subject sample lines.
1
an unrestrained trolley assembly, located on a structural beam at the lower elevation
j
(-24'6") of the containment building, in proximity to some safety-related trisodium
1
phosphate baskets.
Inspector followup of the status of this assembly revealed that the trolley had been
installed as a temporary component during the first refueling outage in 1987, but
never removed. The licensee removed the unauthorized trolley and issued an
adverse condition report (ACR) M3-97-0563, documenting the concern that current
procedural controls for " incomplete work" were not being followed. Subsequently
(note: after the conclusion of this inspection report period), the licensee issued
another Condition Report (CR) M3-97-0850, documenting inadequate corrective
action implementation relative to a licensee event report, LER 3-96-003, involving
unauthorized temporary I-beams over safety-related equipment.
CR M3-97-0850 also documented current licensee findings of heavy, unrestrained
tools and chain falls located in proximity to safety-related equipment. Based upon
the discovery of the unrestrained trolley assembly, as well as the more recent
licensee-identified issues of CR M3-97-0850, the inspector determined that
additional licensee management attention to such " seismic II/l" concerns would be
prudent. The Significant items List (SIL) enclosed with the NRC Restart Assessment
Plan for Millstone Unit 3 documents an item for " Resident Emphasis: Seismic II/l".
This issue, with emphasis upon the new problems documented above, will be
tracked as an inspector followup item (IFl 423/97-01-07) to evaluate both the
timeliness and adequacy of further corrective measures in this area.
Overall, the plant inspection-tours revealed improvements in Unit 3 areas of housekeeping,
material conditions, and work controls. With the exception of the problem with
unrestrained equipment, noted as an IFl above, the licensee provided adequate response to
the inspector questions and field observations and demonstrated continued progress in the
physical enhancements to the plant field conditions.
-
.
'
32
U3.lli Enaineerina
U3 E8
Miscellaneous Engineering issues
E8.1
IOnen) Unresolved item 50-423/96-01-08: Slave Relav/Overlan Test Deficiencies
a.
Inspection Scope (92903)
In 1993 the licensee identified slave relay and other testing deficiencies. As a result of
those findings, the unit director established an overlap testing task force to review the
adequacy of overlap testing for the reactor trip and engineered safeguard systems circuitry.
These reviews were completed in 1993 and the licensee later credited these reviews with
accomplishing the reviews requested in NRC Generic Letter 96-01, " Testing Of Safety-
Related Logic Circuits."
l
in 1996, the licensee performed a review of safety and non-safety related functions,' as '
described in the Final Safety Analysis Report (FSAR) and/or the Safety Evaluation Report
(SER) to determine if the functions were properly tested. The scope of that review
included systems that were considered to be accident mitigating or risk significant as
'
defined in the Maintenance Rule (10 CFR 50.65).
The inspector reviewed selected test procedures, elementary electrical and logic drawings,
Open item Reports (OIRs), ACRs and other documents associated with the testing review
efforts to assess the effectiveness of the reviews.
b.
Observations and Findinas
The licensee task force reviews performed in 1993 identified procedural deficiencies,
circuitry which required design changes, incorrect drawings and FSAR errors. Three LERs
were issued as a result of technical specification violations that occurred due to testing
deficiencies (LERs 3-93-005,-010,-017). In each case the affected circuit performed
satisfactorily when tested. The followup reviews of the FSAR and SER performed in 1996
identified a number of questions regarding testing adequacy and the licensee was
continuing to disposition the associated OIRs. The OIRs that had been dispositioned to
date had not identified logic testing deficiencies of the types discussed in GL 96-01.
The inspector performed an independent review of testing that was performed on reactor
protection and engineered safety features logic circuits to assess the adequacy of the test
procedures. These reviews included the steam generator low level reactor trip and
emergency feedwater pump start testing and portions of the emergency diesel generator
start, load shed and load sequencing testing. The inspector also reviewed several OIRs to
assess the significance of the issues and the adequacy of the licensee resolutions.
The inspector's review of the steam generator low level channel testing included the
following procedures:
SP 3444A01 (Rev. 04) - Steam Generator Water Level Channel Calibration
SP 3443A21 (Rev.10) - Protection Set Cabinet i Operational Test
_ _ . _ _ . _ _ _ _ _ _ . _ _ _ _ . _ _ . _ . _ _ _ _
. . _ _ .
. _ _ _ . . .
i
e
- .
4
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33
.
I
SP 3446B11_(Rev. 09) - Train A Solid State Protection System Operational Test
'
~ The review included testing of the circuitry from the steam generator level transmitters to
i
the output devices, the reactor trip breakers and the auxiliary feedwater pump controls.
The test procedures were thorough and no problems were identified, in addition to the
testing of the automatic reactor trip, the inspector also reviewed the testing of the manual
i
reactor trip push buttons that is performed in'accordance with test SP 3446F331, "SSPS
}
Refueling Tests". The inspector found_that the manual push buttons were properly tested
_
and ensured they would each independently open the reactor trip breakers on either the
- shunt trip or undervoltage trip mechanism.
The inspector also reviewed test procedures and selected drawings associated.with the
i
emergency diesel generator starting, load shedding and load sequencing functions. The -
j
following procedures were included in this review:
- SP 3646.A.1 (Rev.12) - Emergency. Diesel Generator A Operability Test- -
')
)'
SP 3646A.5. (Rev. 05) - Offsite Power Transfer Operability Test
-*
i
SP 3646A.8 (Rev.14).- Slave Relay Testing' Train A
j
[
SP 3646A.12 (Rev. 07) - Emergency Diesel Generator A Lockout Test -
-
SP 3646A.15 (Rev.11) - Train A Loss of Power Test
I
1
'
~
SP 3646A'.17 (Rev. 09)- Train A ESF With LOP Test
- .
SP 3646A.19 (Rev. 03) - SIS Transfer of DG From Test to Standby
- -
SP 3646A.21 (Rev. 05) - DG Auto Start on ESF Signal
j
SP 3448E51 (Rev. 01) - Diesel Sequencer Train' A Actuation Timer Test
j
SP 31447MA (Rev. 01)- MP3 Bus 34C Loss of Power Channel Calibration
'
Although this review did not include 100% of the circuitry, the inspector identified the
d
following testing issues:
During the loss of power testing, the EDG receives a start signal from relay 27Y2 in
the bus undervoltage logic circuit. A set of contacts from a different relay in the
]
undervoltage logic feeds a loss of power signal to the emergency generator loading
.
sequencer (EGLS) resulting in an additional EDG start signal. The existing test
]
procedures were not adequate to properly verify the operation of the parallel
signals.
,
The bus undervoltage relays provide trip signals to the emergency bus tie breaker
j
and the feeder breaker from the reserve station' service transformer (RSST). The tie
<
breaker trip circuit contains _three parallel trip circuits and the testing does not verify
j
each path. The RSST breaker trip was not included in any of the surveillance test
j
procedures.
1
1
The EDG starting control has two circuits that are designed such that a start signal
to either circuit will result in the opening of both air start solenoid valves even if one
or more circuit component failures may have occurred. The current testing'does not
independently test the redundancy designed into the circuits.
J
. - _ _ _ . _ . . _ _ . . .
_
. -
_ . _ _,
_ . _ . _ . _ . _ - . _
_ _ _ - . . _ _ _ . . _.
. .
.
34
j
Procedure SP 3646A17 contains notes that state:
-
'
"If possible, the SW pump not tested in lead during SP 3646A.15 should be lead."
t
!
"The CCP pump not tested in SP 3646A.15 should be tested here: the other must.
4
!
be in " PULL-TO-LOCK.""
- "The CHS pump not tested in SP 3646A.15 should be tested here; the other pump
'
must be in " PULL-TO-LOCK.""
,
These notes are intended to ensure that all of the equipment is tested either in SP
3646A15 or SP 3646A17.' These notes are worded as to provide
recommendations, rather than requirements, and as such may not ensure that the
1
standby and swing pumps get tested as required by technical specifications.
' ' "
~ Based on these findings th'e' inspector questioned the adequacy of the previous overlap ' -
+
-testing. reviews that were credited with ensuring that test procedures are adequate to
ensure the technical specification test requirements are met. As a result, the licensee
A:
developed an action plan to address this concern. . The' planned actions included: . _ ..e
A review for overlap test issues of several circuits by comparing electrical
- *
schematic drawings and logic drawings against plant surveillance test pro:edures to
'
ensure that all portions of the logic circuitry, including the parallel logic, interlocks,
bypasses and inhibit circuits, are adequately covered in the surveillance procedures
"
to fulfill the TS requirements. These reviews were to include the loss of power
schemes (undervoltage and degraded voltage), one ESF actuation system, and one
4
reactor trip instrumentation functional unit.
The revision of the surveillance testing for the loss of power initiation logic to
'
adequately verify parallel logic.
The performance of a root cause analysis to determine the cause of missed contacts
to determine if the failure was a generic issue that applied to the overall effort of
- the overlap task force. Based on the findings of the root cause analysis any
additional corrective actions would be determined.
- Subsequent licensee reviews identified additional testing deficiencies. The licensee
performed a self-assessment of the Unit 3 response to Generic Letter 96-01 and concluded
that the response did not meet the requirements of the letter. This was based on the
above specific findings where actuation' contacts were not tested, a lack of auditable
documentation and a difference in philosophy between the 1993 overlap task force and the
requirements of the Generic Letter. This issue was documented in ACR M3-97-0529 and
additional corrective actions were being developed by the licensee at the end of this
inspection period.
,
k-,-
- - + ,
c-
,s.w-
-
,-,e
- .- ,
, - ~
-
y-
c
y
- . _ . . . _ _ . - _
_ _ _ . _ _
,
!
..
35
c.
Conclusion
,
":
- ' ~ The inspector found that the licensee had expended significant resources in the past to~
i
review the logic testing and had identified and corrected numerous deficiencies. The test'
procedures reviewed were generally thorough. However, as discussed above, some test
-
deficiencies continued to exist. The failure to ensure all contacts are operable could result
,
in significant undetected problems. For example, if the emergency bus tie and feeder
1
breakers failed to trip on an undervoltage signal, the EDG output circuit breaker would be
prevented from closing to' energize the bus to power necessary safety equipment. The
inspector also noted that a more thorough self-assessment would have been appropriate
,
prior to the licensee submission of a response to GL 96-01.
]
.
.
This item remains open pending NRC review of additional licensee corrective actions, and
)
the assessment'of the significance of any additional findings and the results of additional
tes:ing that is performed.
i
.
E8.2 (Closed - Part of SIL ltem 67) ACR M3-96-0621: Potential For Overloadina Station
Blackout (SBO) Diesel
a.
Insoection Scope (92903)
This ACR identified a concern that the station blackout (SBO) diesel generator could be
overloaded if a safety injection or containment depressurization signal occurred while the
-
SBO diesel was supplying power to an emergency bus.
-
~
b.
Observations and Findinas
-
. In the event of a loss of all ac power,-the SBO dieselis manually started in accordance .a
Y
with procedure ECA O.0, " Loss of All'AC Power." Prior to energizing the bus from the
SBO diesel, the procedure directs the operators to place the control switches for large
)
loads in the pull-to-lock position. This action blocks the automatic start of the loads.
Following the energization of the bus, the operator then manually starts loads needed to '-
cope with the station blackout condition,
i
c.
Conclusions
- The inspector reviewed the associated ACR, procedure and elementary electrical drawings
)
and concluded that the licensee had appropriately reviewed and dispositioned this ACR.
This item is closed, (representing partial closure of SIL ltem 67).
o
.
l
'
36
IV Plant Support
Millstone Units 1,2, and 3
R1
Radiological Protection and Chemistry Controls
a.
Inspection Scope (83750)
i
The inspector reviewed the licensee's radiation protection programs established at each
unit and for the site. A review of specific work performed, the programs for maintaining
occupational radiation exposures as low as is reasonably achievable, and tours of the
various radiologically controlled areas (RCAs) were conducted by the inspector.
b.
Observations and Findinas
Unit 1
i
i
The inspector toured portions of the reactor and liquid radwaste buildings as part of the
j
inspection at Unit 1. The inspector noted a significant decrease in the number and size of
posted contaminated areas in the unit, which was described to the inspector as part of.the
unit's clean-up policy. The inspector noted that while some work was being performed in
the reactor building at the time of this inspection, significant radiological work, especially in
the drywell and on the refueling floor had yet to commence. The inspector also toured the
Xenon / krypton building, which houses some of the off-gas treatment systems and delay
tanks. This structure and its component systems had undergone a significant
refurbishment during 1996. Two areas within the structure were appropriately posted and
controlled as high radiation areas. The upper level of the structure housed two glycol
chiller systems, one of which was still under refurbishment.
For 1997, the licensee had established a goal of not more than 399 person-rem. This goal
'
is based on completing significant work and having the unit ready for restart of operations
during 1997. As described in a previous NRC Inspection Report (50-245/96-09), the unit
has significantly increased the number of personnel assigned to perform work planning and
ALARA functions. Seven persons within the Radiation Protection Department are now
assigned to ALARA, and two technicians are on loan as work week managers. Each of the
ALARA personnel have been assigned specific work packages and/or work areas for
planning purposes, and are responsibla for coordination with engineering and the work
groups to ensure proper ALARA controls are incorporated into the work packages.
Additionally, an ALARA Committee has been established, which includes all department
managers.
On January 15,1997, the licensec identified, through its Adverse Condition Reporting
(ACR) process (ACR # M1-97-0094) that fan HVE-14, which exhausts portions of the
Radwaste Storage Building, was potentially an unmonitored release pathway, as the fan
was not connected to the main plant stack, and no radiological effluent monitoring
equipment was located with this fan. The inspector discussed this ACR with a licensee
representative during this inspection, and reviewed a reportability evaluation performed bt
the site Engineering Department which analyzed the significance of this ACR. The
inspector determined that the reportability determination was invalid in that the evaluation
~
1
4
37
erroneously addressed a building that was different from the Radwaste Storage Building.
Subsequent to this finding, the licensee wrote another ACR (ACR # M1-97-0282) to
document this error, and subsequently determined that a notification to the NRC was
required, which was made on February 6,1997. Failure to monitor effluents released to
the environment from the Radwaste Storage Building to demonstrate compliance with
applicable regulatory limits, including 10 CFR 20.1301, is a violation of 10 CFR 20.1302.
(VIO 245/97-01-08)
Unit 2
The inspector toured various portions of the Unit 2 RCA, including the Auxiliary and
Containment Buildings, as part of this inspection. In general, all areas were determined to
be in compliance with NRC requirements for radiological postings and control of radioactive
material. On February 5,1997, the inspector observed the removal and subsequent
transfer of two highly radioactive pieces of debris previously found in the reactor vessel. A
metal nut and a tie wrap, each reading in excess of 10 Roentgens per hour on contact,
were removed from a storage bucket that was being kept in the refueling cavity,
transferred to a lead pig, and moved from the Containment Building, through the Auxiliary
Building and outside the unit to a designated storage area. The inspector observed the pre-
job briefing, which included a discussion of engineering controls for the minimization of
personnel exposure, and the conduct of all work until the shield pig with the objects was
removed from the unit. This activity also involved significant coordination between the
unit operations department, the health physics department, the site security organization
"
and the self-directed work group. Due to the careful planning process used, total exposure
for this work was less than 25 millirem.
For 1997, the unit established a goal of not more than 182 person-rem. Since the last
specialist inspection (50-336/96-09), the unit had flooded up the refueling cavity and
successfully off-loaded the reactor fuel to the spent fuel pool. Significant strides in
improving the unit ALARA program have also been made. ALARA coordinators have been
identified in each of the major work departments within the unit, and the Unit ALARA
coordinator was in the process of developing a training program for them. A unit ALARA
program procedure was also drafted, however, it was not issued at the time of this
inspection. Discussions with the ALARA coordinator, Health Physics Manager and Unit
Directr,r indicated the intent to establish an ALARA Committee, to include the major
depa:tment heads and their ALARA coordinators.
The inspector interviewed the health physics manager, and reviewed the documentation
associated with three ACRs (M2-97-0086, M2-97-0091 and M2-97-0142) written to
identify improper entries to the RCA which occurred during a ten day period in January
1997. In two of the instances, workers entered the RCA without having signed in on a
RWP, and without having on an electronic docimeter, as directed by the unit radiation
protection staff. In the third instance, a fire watch entered the unit with an electronic
dosimeter that had not been properly turned on. When discovered through self-checking,
the fire watch remained in the RCA with the non-functioning electronic dosimeter unti!
completion of the fire watch round. Procedure RPM 5.22 requires radiation workers to
comply with written instructions, including RWPs, from the radiation protection staff.
Although the safety significance of each of these events is individ ally low, as each worker
'
.
1
i
e
i
38
was wearing a thermoluminescent dosimeter (TLD) which is utilized to determine dose of
record, the number of events in such a short duration are of concern. Additionally, the fire
watch event may highlight a problem with the training given and the perception of the
workers performing this task relative to other plant requirements, such as radiological
j
safety. Both of these issues were discussed by the inspector with the unit Health Physics
Manager and Unit Director, and with the station Vice President - Work Services. Short-
)
term corrective actions taken by the licensee included posting a health physics technician
at the main RCA entrance to ensure that personnel entering the RCA were wearing a
functional electronic dosimeter. Long-term corrective actions were not identified, however,
at the time of this inspection. Failure to adhere to the licensee's radiation protection
program, specifically procedure RPM 5.22,is a violation of 10 CFR 20.1101. (VIO
336/97-01-09)
Unit 3
The inspector toured various portions of the Unit 3 RCA, including the containment and
auxiliary buildings. Since the last specialist inspection (50-423/96-07) a significant
reduction in the number of leaking valves was observed, based on the reduced number of
catch containments observed. The unit continues te have very low dose rates in most
i
,,
areas, and significant portions of the Containment rematri accessible as clean areas.
The inspector reviewed the licensee's ALARA program, including the 1997 ALARA goal of
not more than 170 person-rem. The unit focus in ALARA has been to improve the work
order process, to include having RWP and ALARA control information included in the work
-
orders, in addition, the licensee has begun to include detailed maps and pictures of areas
and systems to be worked in the work order package. This is the result of a significant
campaign completed in 1996 by the unit radiation protection staff to photograph over
30,000 components in the RCA.
Site Health Physics
The site health physics group is responsible for providing calibration and dosimetry
services, health physics engineering and health physics support to the units and to other
site-wide organizations. As part of this inspection, a review of certain activities was
i
coni:cted by the inspector.
l
The self-directed work group includes six health physics technicians whose primary focus
is to support the activities of the Waste Services Department. As previously noted in the
discussion on Unit 2, above, this group was involved in the transfer of two highly activated
pieces of material from the Unit 2 vessel to a storage location outside the unit. In addition,
the inspector also reviewed the performance of this group during a recently completed
waste transfer evolution.
As part of the Liquid Radwaste Remediation Project at Unit 1, waste concentrates spilled
inside the "A" Concentrator cubicle were removed in 1996. In addition to being
i
radioactive, this material also contained asbestos, and thus required specialized engineering
controls for handling, as required by the Occupational Safety and Health Administration
(OSHA). On January 24,1997, ten barrels of this material, each containing several
7
V
e
39
plastic-wrapped bags of the waste, were transferred to a processing liner, which ultimately
was to be solidified and buried as low-level radioactive waste. Due to OSHA requirements,
this transfer was conducted inside a tent-like structure erected around the liner by a team
'
of five specially trained contractors. Because of the OSHA requirements, the health
physics technicians from the self-directed work group could not enter the tent once work
began. Based upon interviews conducted by the inspector with all members of the
j
contractor work team, the contractor industrial hygienist, and the self directed work group
technicians, and a review of documentation associated with this activity (RWP, ALARA
review, pre-job briefing package, post-job review) the inspector determined that the work
i
was appropriately controlled in accordance with NRC regulations. While all five contractors
were contaminated on their person and/or clothing upon completion of the work, this was
'
not the result of a breakdown of radiological controls. None of the contaminations resulted
,
in a significant radiological exposure.
'
c.
Conclusions
Unit 1
Noticeable reductions in the amount of contaminated spaces within the unit were
observed. ALARA planning and staffing of the ALARA group have significantly improved,
however, the effectiveness of this cannot be determined until more radiologically
significant work commences. One violation of NRC requirement involving an unmonitored
release path was identified.
Unit 2
Good work planning and control was observed for the transfer of highly irradiated materials
j
from the reactor vessel. Significant changes in the planning and control of radiological ,
work is under development. A violation of NRC requirements involving poor radiological
worker practices was identified. While short-term corrective actions were implemented by
the licensee at the time of this inspection, long-term actions had not yet been identified.
Unit 3
Contamination control improvements, especially the reduction in the number and need for
catch containments, was observed during tours of the RCA. Incorporation of RWP and
ALARA information into the work orders was an improvement, although the effectiveness
of this will have to be evaluated once significant radiological work resumes.
Site Health Physics
Appropriate support to Unit 2 was observed during the transfer of highly irradiated
material. Appropriate work controls were implemented during the transfer of asbestos
contaminated concentrates wastes.
$
40
R8
Miscellaneous Radiological Protection and Chemistry Issues
A recent discovery of a licensee operating their facility in a manner contrary to the Updated
Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused
review that compares plant practices, procedures and/or parameters to the UFSAR
descriptions.
While performing the inspections discussed in this report, the inspector reviewed the
applicable portions of the UFSAR that related to the areas inspected. The inspector
verified that the UFSAR wording was consistent with the observed plant practices,
procedures and/or parameters, except in the area of the management organization and
responsibilities for radiation protection. Section 12.5 of the Unit 1 UFSAR and Section
11.2.3 of the Unit 2 UFSAR make reference to Section 12.5.1 of the Unit 3 UFSAR for a
full description of the health physics organization and reporting functions. This description
no longer is accurate due to the restructuring and unitization of the Radiation Protection
Program. The Work Services organization recognized the need to update the Unit 3 UFSAR
to reflect the management changes and identified it to the Site Licensing Director by
memorandum, dated November 29,1996.
V. Manaaement Meetinas
X1
Exit Meeting Summary
The inspectors presented the inspection results to members of licensee management at the
conclusion of the inspectian. The licensee acknowledged the findings presented.
X1.2 Final Safety Analysis Report Review
A recent discovery of a licensee operating their facility in a manner contrary to the final
safety analysis report (FSAR) description highlighted the need for additional verification
that licensees were complying with FSAR commitments. All reactor inspections will
provide additional attention to FSAR commitments and their incorporation into plant
practices, procedures and parameters.
While performing the inspections which are discussed in this report the inspectors
reviewed the applicable portions of the FSAR that related to the areas inspected.
inconsistencies were noted between the wording of the FSAR and the plant practices,
procedures and/or parameters observed by the inspectors, as documented in Sections
U3.E8.1 and R8.
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INSPECTION PROCEDURES USED
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IP 37550:
Engineering
IP 37551:
Onsite Engineering
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IP 40500:
Licensee Self-Assessments Related to Safety issues inspections
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IP 61726:
Surveillance Observations
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IP 62707:
Maintenance Observations
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IP 71707:
Plant Operations
IP 71750:
Plant Support Activities
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IP 83750:
Occupational Radiation Exposure
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IP 92700:
Onsite follow-up of Written reports of Nonroutine Events at Power Reactor .,
Facilities
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IP 92901:
Follow-up Operations
IP 92902:
Follow-up Maintenance
IP 92903:
Follow-up Engineering
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ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
URI 50-245/97 01-01
U 1.02.1
Spent Fuel Pool Cleanliness
I
URI 50-245/97-01-02
U 1.03.1
Operations Procedure Adequacy
1
URI 50 245,336,423/
U 1.05.1
Inaccurate Personal Qualification Statements
97-01-03
URI 50-245/97-01-04
U 1.E1.1
Resolution of A-46 Program Outliers
URI 50-245/97-01-05 -
U1.E2.2
Low Flow Operation of Containment isolation Check
i
Valve 1-CU-29
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U2.08.2
Failure to Enter TS Action for inoperable Nis
IFl 50-423/97-01-06
U3.01.1
Interpretation of TS Language and LCO Actions
IFl 50-423/97-01-07
U3.M 8.1
Seismic 11/1 Concerns
VIO 50-245/97-01-08
R1
Failure to Monitor Gaseous Effluents from the
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Radwaste Storage Building
VIO 50-336/97-01-09
R1
Entering RCA w/o Electronic Dosimeter or Signing
Closed
LER 50-245/96-03
U 1.E1.1
LER 50-336/96-15
U2.02.2
VIO 50-336/94-17-10
U2.08.1
Operation Outside Systtim Design Parameters
'
URI 50-336/96-04-09
U 2.M8.1
Troubleshooting Controls
URI 50-336/96-06-06
U2.M8.2 High Pressure Safety injection Check Valve Backflow
Testing
j
URI 50 336/95-07-06
U2.E8.1,
Condensate Storage Tank Siphon Break
-
-
URI 50-336/95-11-03
U2.E8.2
10 CFR 21 Reportability Review
Discussed
VIO 50-245/95-42-01
U 1.08.1
Failure to Prevent Work Which had the Potential for
Draining the Reactor Vessel
URI 50-336/96-01-04
U2.02.2
Loss of DC Bus Event
URI 50-423/96-01-08
U3.E8.1
Slave Relay / Overlap Test Deficiencies
Sianificant items List
i
Unit 3 SIL #67
Partial Closure
,
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43
LIST OF ACRONYMS USED
ACR(s)
adverse condition report (s)
American Institute of Steel Construction
as low as reasonably achievable
ANSI /ANS
American National Standards Institute /American Nuclear
American Society of Mechanical Engineers
reactor plant component cooling
CES/NTE
component engineering services / nondestructive test engineering
CFR
Code of Federal Regulations
CHS
charging system
CR(s)
condition report (s)
DCN
design change notice
diesel generator
emergency generator loading sequence
engineered safeguards actuation system
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engineered safety feature
Fix-It-Now
first line supervisor
GIP (s)
generic implementation procedure (s)
GL
Generic Letter
gpm
gallons per minute
high pressure safety injection
Independent Corrective Action Verification Program
IFl
inspector follow item
IR(s)
inspection Reports (s)
LCO
limiting condition for operation
LER(s)
licensee event report (s)
material & test equipment
nil ductility transition
NRC
Nuclear Regulatory Commission
Nuchar Reactor Regulation
Nut.,ar Safety Information Center
NS&O
nuclear safety and oversight
Nuclear Regulation
.OCA
Office of. Congressional Affairs
OIR(s)
open item report (s)
on the job training
. OSHA
Occupational Safety & Health Administration
PAO
Public Affairs Office
PDCR
plant design change record
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Public Document Room
PiR(s)
plant information report (s)
PUP
procedure upgrade program
quality assurance
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QAS
- Quality and Assessment Services
OPTR
quadrant power ti!! ratio
reactor building closed cooling water
radiologically controlled area
refueling outage
RI
Region I
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reaction protection system
reserve station service transformer
RWP(s)
radiation work permit (s)
station blackout
SER(s)
safety evaluation report (s)
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SPO
Special Projects Office
,
seismic qualification utility group
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SSER
supplementel safety evaluation report
SSR
reactor plant samples
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TBSCCW
turbine building secondary closed cooling water
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TLD(s)
thermoluminescent dosimeter (s)
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TR(s)
trouble report (s)
TS(s)
technical specification (s)
updated final safety analysis report
USl
unresolved safety issue
violation
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WC
work control
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