ML20212H369
| ML20212H369 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 04/15/1998 |
| From: | Doerflein L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20212H360 | List: |
| References | |
| 50-317-98-01, 50-317-98-1, 50-318-98-01, 50-318-98-1, NUDOCS 9804210362 | |
| Download: ML20212H369 (76) | |
See also: IR 05000317/1998001
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U.S. NUCLEAR REGULATORY COMMISSION '
REGION 1
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License Nos. DPR-53/DPR-69
SNM-2505
Report Nos. 50-317/97-08:50-310/97-08
Licensee: Baltimore Gas and Electric Company
Post Office Box 1475
Baltimore, Meryland 21203
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Facility: Calvert Cliffs Nuclear Power Plant
l Units 1 and 2; and
l Independern Gpent Fuel Storage Installation
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Location: Lusby, Maryland
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l Dates: February 8,1998 to March 14,1998
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Inspectors: Fred L. Bower Ill, Resident inspector
l Henry K. Lathrop, Resident inspector
l Ronald Nimitz, Senior Radiation Specialist
l Suresh Chaudhary, Senior Reactor Engineer
l Timothy J. Kobetz, Project Manager, Spent Fuel Licensing
l Section, Spent Fuel Project Office (SFPO), Office of Nuclear
l Material Safety and Safeguards (NMSS)
Henry W. Lee, Senior Structural Engineer, Spent Fuel Technical
Review Section, SFPO, NM S
wu d tow _
Ii!98
Approved by: Lawrence T. Doerflein, Chief Date
Projects Branch 1
Division of Reactor Projects
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9804210362 980415
PDR ADOCK 05000317
G PDR
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,. Enclosure 1 2
angles by bending vice shearing would cause the canister guides sleeves to deform
and impinge on the installed spent fuel assemblies. In discussing drop accidents.
Section 8.2.5.2 of the USAR states that the maximum spacer disk deflection was
computec' to be 0.077 inches and that the gap between the guide sleeve and the
fuel assembly exceeds 0.05 inches
This is a Severity Level IV violation (Supplement I-Reactor Operations).
Pursuant to the provisions of 10 CFR 2.201, Baltimore Gas & Electric Company is hereby
required to submit a written statement or explanation to the U.S. Nuclear Regulatory
Commission, ATTN: Document Control Desk, Washington, D.C. 20555 with a copy to the
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Regional Administrator, Region I, and a copy to the NRC Resident inspecW at the facility j
that is the subject of this Notice, within 30 days of the date of the let:ei uansmitting this q
Notice of Violation (Notice). This reply should be clearly marked as a " Reply to a Notice of
Violation" and should include: (1) the reason for the violation, or, if contested, the basis for
disputing the violation or severity level, (2) the corrective steps that have been taken and
the results achieved, (3) the corrective steps that will be taken to avoid further violations,
and (4) the date when full compliance will be achieved. Your response may reference or
include previous docketed correspondence, if the correspondence adequately addresses the
required response, if an adequate reply is not received within the time specified in this
Notice, an order or a Demand for information may be issued as to why the license should
not be modified, suspended, or revoked, or why such other action as may be proper should
not be taken. Wh6re good cause is shown, consideration will be given to extending the
response time.
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If you coi. test this enforcement action, you should also provide a copy of your response to
the Director, Office of Enforcement, United States Nuclear Regulatory Commission,
Washington, DC 20555 0001. !
Because your response will be placed in the NRC PDR, to the extent possible, it should
not incluJe any personal privacy, proprietary, or safeguards information so that it can be
placed in the PDR without redaction. If personal privacy or proprietary information is
necessary to provide an acceptable response, then please provide a bred.eted copy of ,
your response that identifies the informatien that should be proteca., and a sedacted I
copy of your response that deletes such information. If you request withholding of
such material, you muit specifically identify the portions of your response that you seek to
I have withheld and provide in detail the bases for your claim of withholding (e.g., explain
I why the disclosure of information will create an unwarranted invasion of personal privacy
l or provide the information required by 10 CFR 2.790(b) to support a request for i
withholding confidential commercial or financial information). If safeguards information is ;
necessary to provide an acceptMie response, please provide the level of protection l
described in 10 CFR 73.21.
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Dated at King of Prussia, Pennsylvania 1
this 15th day of April,1998.
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. EXECUTIVF. SUMMARY
Calvert Cliffs Nuclear Povier Plant, Units 1 and 2, and independent Spent Fuel Installation
l Inspection Report Nos. 50-317/98-01 and 50-318/98-01
l This integrated inspection report includes aspects of BGE operations, maintenance,
l engineering and plant support. The report covers a five week period of resident inspection j
and the results of specialist inspections in independent spent fuel storage, radioactive
l waste, and engineering.
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Plant Operations
Plant operations were conducted safely with a proper focus on continued nuclear safety.
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In general, the conduct of plant operations was professional and safety-conscious.
l Operation's shift turnover briefings were effectively conducted. BGE's efforts to reduce
the number of control room deficiencies and the number of deficiencies requiring
compensatory operator action have been positive.
Safety tours conducted by the General Supervisor Nuclear Plant Operations and operator
performance observations conducted by shift supervisors were two examples of
management oversight initiatives implemented by Operations management. The
inspectors concluded that Operations management implemented aggressive efforts to
reduce valve, switch, and breaker mispositionings during the last twelve months which
have been successful as demonstrated by the reduced number of mispositionings.
Maintenance
Prompt and effective replacement of a power supply in the control element drive coil power j
programmer cabinets precluded an unnecessary transient on the plant. System engineering
provided effective support to mainter'ance during the power supply replacement.
The inspectors concluded that the Worker Risk Assessment Process (WRAP) has been an
l effective initiative to aid in continuous improvement of industrial safety practices for
maintenance personnel. Maintenance management has recently expanded this program in
an effort to improve radiation safety practices for maintenance persennel.
Overall, the observed maintenance was conducted safely and in accordance with BGE
approved procedures and controls. Workers were knowledgeable and performed work
effectively. Good supervisory oversight of maintenance was observed during this period. l
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The observed surveillances were conducted safely and effectively demonstrated system
operability. Operators demonstrated good use of self-checking techniques. The pre test
briefings performed by operations personnel were excellent in scope, content, and level of
' detail.
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Executive Summary (cont'd) .
Enaineerina l
BGE established an adequate process to track design issues, received from Transnuclear l
West, associated the NUTECH Horizontal Modular Storage System. BGE developed and
followed a comprehensive process to ensure all design and quality assurance issues which
might have affected its dry shielded canister were resolved.
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BGE's application for a license for use of the NUHOMS system at the Calvert Cliffs
independent spent fuel storage installation, under a site specific license, did not provide
complete and accurate information regarding the behavior of the dry shielded canister
during a vertical top drop accident. The inspectors identified this as a violation of
10 CFR 72.11 (VIO 50-317&318/98-01-01). l
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BGE did not adequately resolve all of the design issues and, more significantly, did not '
identified an unreviewed safety question involving an equipment malfunction of a different ,
type than any evaluated previously in the Updated Safety Analysis Report. The inspectors
concluded that this was a violation of 10 CFR 72.48(a)(2)(ii)(VIO 50-317&318/98-01-02).
The inspectors further concluded that the safety significance of these findings was
minimized due to the low probability of, and minimal consequences associated with, a
vertical top end drop accident.
Overall, the continuing training provided to selected engineering and technical support staff
personnel provided an excellent overview of the development, results, and applications of
the Calvert Cliffs Probabilistic Risk Assessment. The probabilistic risk assessment training
was appropriate in scope and detail and the stated learning objectives were effectively
met.
Plant Support
Overall, BGE implemented an adequate radioactive waste processing, handling, storage,
and transportuJon program. Good efforts continued to reduce quantities of radioactive
waste, includi..g liquid radioactive waste, released from the facility.
Personnel involved in radioactive waste and material shipping received appropriate training
and were knowledgeable of applicable regulatory requirements. Regulatory documents
(e.g, certificates of compliance and disposal facility licenses) were maintained current and
were effectively implemented. BGE effectively updated its UFSAR to describe its present
waste processing, handling, and storage activities.
BGE was not able to demonstrate that it was conforming to applicable NRC Branch
Technical Positions in the area of waste concentration averaging and determination of
scaling factors for hard to detect radionuclides (e.g., transuranics). BGE suspended
shipments of radioactive waste and radioactive material pending resolutic.n. BGE had
previously suspended shipment of chemical volume and control system wastes due to
potential concerns of exceeding Class C waste limits for C-14.
A quality assurance audit of radioactive waste activities was not well structured or defined,
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. TABLE OF CONTENTS
EX EC UTIV E S U M M A RY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii
TABLE O F CO NTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv
Summary of Plant Status ............................................1
1. O p e r a t i o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
01 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
01.1 General Comments (71707) ...........................1
08 Miscellaneous Operations issues (92901) . . . . . . . . . . . . . . . . . . . . . . . 2
08.1 (Closed) VIO 50-317&318/96-06-01: Safety System Misalignments
During Saltwater and Serv;ce Water System Maintenance ...... 2 J
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11. M a i n t e n a n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
M 1.1 G ene ral Comm e nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
M1.2 Routine Surveillance Observations . . . . . . . . . . . . . . . . . . . . . . . 4
M8 Miscellaneous Maintenance issues (92700) .....................5
M8.1 (Closed) LER bO-317/97-008:Two Reactor Protective Channels Out
of Service During Test ...............................5
111. E n g i n e e r i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
E2 Engineering Support of Facilities and Equipment ..................6
E2.1 Resolution of Design issues Associated with the NUHOMS Dry Cask
Sto ra g e Sy st e m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
E5 Engineering Staff Training and Qualification ....................12
E5.1 Probabilistic Risk Assessment (PRA) Continuing Training ...... 12
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E8 Miscellaneoun Engineering issues (92903) ................13
E8.1 (Closed) Unresolved item 50-317&318/97-04-04: Effectiveness of
Corrective Actions for Emergency Lighting Deficiencies . . . . . . . 13
E8.2 (Closed) Violation 50-317&318/96-09-01: Defacto Modification of
Auxiliary Feed Water Pump Base . . . . . . . . . . . . . . . . . . . . . . . 13
l'/. Pl a n t S u p p o rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 3
R1 Radiation Protection and Chemistry Controls (RP&C) . . . . . . . . . . . . . . 13
R1.1 Radioactive Waste Processing, Handling, Storage, and Shipping . 13
R2 Status of RP&C Facilities and Equipment ................. 16
R3 RP&C Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . 16
R4 Staff Knowledge and Performance in RP&C ....................17
R5 Staff Training and Qualification in RP&C . . . . . . . . . . . . . . . . . . . . . . 18 l
R6 RP&C Organization and Administration . . . . . . . . . . . . . . . . . . . . . . . . 18 '
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Table of Contents (cont'd)
R7 Quality Assurance in RP&C Activities . . . . . . . . . . . . . . . . . . . . . . . . . 18
R8 Miscellaneous RP&C issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
R8.1 (Update) Unresolved item 50-317&318/96-04-03: Verification of
Updated Final Safety Analysis (UFSAR) Commitments . . . . . . . . 20
R8.2 (Closed) Violation 50-317&318/96-07-02: Loss of integrity of
Radioactive Materials Shipment . . . . . . . . . . . . . . . . . . . . . . . . 20
R8.3 (Updated) Violation 50-317&318/97-06-04: Lack of Procedure to
S u rve y La u nd ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 0
R8.4 Dose Assessments for May 5,1997, Unit 2 Reactor Cavity Event
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V. M a nag eme nt Meeting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
X1 Exit M e eting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
X2 Review of UFSAR Commitments . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
ATTACHMENTS
Attachment 1: Partial List of Persons Contacted
Inspection Procedures Used
items Opened, Closed and Discussed
List of Acronymns Used
Attachment 2: CCNPP NUHOMS-24 DSC Evaluation
Attachment 3: BGE DSC NRC Structural issue Overview
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Report Details
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Summarv of Plant Status
Unit 2 had a small power reduction on March 6,1998, to support scheduled
maintenance, otherwise, both units operated at full power throughout the inspection
period.
1. Operation 3
01 Conduct of Operations
01.1 General Comments (71707)
Plant operations were conducted safely with a proper focus on continued nuclear j
safety. Using Inspection Procedure 71707," Plant Operations," the inspectors l
conducted frequent reviews of ongoing plant operations. In general, the conduct of
plant operations was professional and safety-conscious. The inspectors observed a
se! acted sample of shift turnover briefings that were attended by operations, safety
tagging, radiation control, chemistry, and fire and safety personnel. Plant status,
equipment and operational problems, and required compensatory actions were
discussed in appropriate scope and detail using the Shift Turnover Information i
Sheet. The inspectors concluded that Operation's shift turnover briefings were l
effectively conducted in accordance with administrative procedure NO-1-207,
" Nuclear Operations Shift Turnover."
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Operators effectively used trend recorders to monitor anomalous parameters and l
evaluate potential problems. Control room deficiencies were promptly identified and j
issue reports were written to initiate corrective actions. Recently the operations a
department initiated a coordinated effort with engineering and maintenance to
reduce the number of control room deficiencies and the number of deficiencies j
requiring compensatory operator action. Additionally, the established goals for
these deficiencies were lowered in January 1998. The inspectors noted that since
this deficiency reduction effort was initiated, the number of deficiencies has been ,
on a generally downward trend. BGE was meeting the goal for the backlogs of s
deficiencies that can be repaired online. Although the trend has been positive, the l
number of deficiencies has not yet met the newly lowered and more aggressive j
goals for total number of deficiencies. The inspectors concluded that BGE's efforts
to reduce the number of control room deficiencies and the number of deficiencies
requiring compensatory operator action have been positive. ,
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On February 25, the inspectors observed the General Supervisor Nuclear Plant l
Operations (GS-NPO) and the on-shift operations Shift Supervisor perform a safety
tour of the plant. Discussions with operations personnelindicated that these safety
inspections are conducted weekly. On March 8, the inspectors observed the on-
shift Operations Shift Supervisor conduct a performance observation of a non- l
licensed plant operator. Discussions with operations personnelindicated that these <
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types of management observations are conducted monthly. Safety tours
conducted by the GS-NPO and operator performance observations conducted by l
shift supervisors were two examples of management oversight initiatives
implemented by Operations management.
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During this period the inspectors also reviewed the effectiveness of an Operations
initiative to reduce valve, switch, and breaker mispositionings. These initiatives
included an increased emphasis on BGE's "Stop-Think Act-Review"(STAR)
processes, peer checks and placekeeping techniques. Over the last year, there has
been a steady downward trend in the number of mispositionings. The six month
rolling average has been reduced from an average of 7 mispositioning per month in
February 1997 to an average of less than 1.5 mispositionings per month in February
1998. Additionally, the six month rolling average of failures to properly track
locked valve deviations has also been reduced from an averbge of two in February
1997 to almost zero in February 1998. The inspectors concluded that the
operations department initiated aggressive efforts to reduce valve, switch, and
breaker mispositionings during the last twelve months. These efforts have been
successful as demonstrated by the reduced number of mispositionings.
08 Miscellaneous Operations issues (92901)
08.1 (Closed) VIO 50-317&318/96-06-01: Safety System Misalignments During
Saltwater and Service Water System Maintenance
The Notice of Violation identified BGE's failure to follow procedures during the
return to service of the 11 saltwater header following maintenance. Specifically,
when the 11 saltwater header was returned to service after maintenance, the
associated emergency core cooling system (ECCS) room cooler fan switch was in
the STOP position versus the required AUTO position.
BGE determined that the cause of the event was human error due to the lack of
self verification and the poor human factors aspects of the handswitch design.
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Corrective actions included changes to the administrative controls for positioning
the switch and improvements in the design of the switch indication layout. BGE ,
changed the operating instructions (Ols) for the saltwater system to add instructions !
for performing an independent verification of system restoration. BGE also reviewe'd
the Ols for other risk significant systems and added second barriers to prevent i
misoperation of the salt water, and chemical and volume control systems. To
address the human factors concerns with the ECCS fan cooler switch, BGE moved
the white power available light so that it was not directly under the AUTO switch
position. The switch labeling was also improved to better coincide the switch
position and the switch position pointer. Operations personnel also walked down
the control board to identify additional potential layout and human factors problems.
These issues were referred to design engineering and are in the backlog of items
requiring review and disposition.
The inspectors have not identified recurring examples BGE's failure to follow
procedures during the restoration of the saltwater systems following maintenance
since the implementation of these corrective actions. Further, as noted in report
section 01.1 above, the inspectors have observed an improving trend in the number
of valve, switch, and breaker mispositionings over the last twelve months.
Therefore, this item is closed.
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11. Maintenance
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M1 Conduct of Maintenance
M1.1 General Comments
a. Insoection Scoce (62707)
The inspectors reviewed maintenance activities and focused on the st.rus of work l
that involved systems and components important to safety. Component failures or
system problems that affected systems included in the BGE maintenance rule
program were assessed to determine if the maintenance was effective. Also, the l
inspectors airectly observed all or portions of the following work activities. ;
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MO2199700966 21/22 SRWHX Rmv. Lead Paint, Asbestos and Reinsulate !
MO2199705831 Remove Fixed Contamination in Unit 2 SRWHX Room l
MOO 199800219 Troubleshooting OC EDG Hi Exhaust Temperature Sensor l
MO2199703562 Clean 22 Component Cooling HX 1
MO2199703671 22 ECCS Cooler Anode Replacement
MO1199800417 Unit 1 Control Element Drive Cabinet #5-12VDC Power Supply
b. Observations and Findinas
The inspectors found that the selected maintenance activities were performed safely
and in accordance with approved procedures and work order packages.
Technicians were experienced and knowledgeable of the assigned duties. Pre-job
briefings were effective in ensuring that the work was conducted in accordance
with BGE work protocols and plans. When applicable, appropriate foreign material
exclusion controls were practiced. The inspectors noted that an appropriate level of
supervisory attention was given to the work.
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During rounds of the Unit 1 cable spreading room on February 10, operators l
identified a burning smell coming from section 5 of the control element drive coil l
power programmer cabinets. The operators suspected that the ventilation fan was
degrading. An issue report was written and was classified as a priority 2
maintenance item. Subsequently, day-shift electricians were assigned to investigate l
the problem. After initial troubleshooting, the electricians suspected a degrading
power supply in the cabinet was a more likely cause of the burning smell than a
failing ventilation fan. With system engineering support, the technicians used
thermography and voltage meters to identify the specific degraded power supply. l
BGE determined that the control element assemblies (CEAs) were trippable.
However, BGE concluded that an attempt to move any of the CEAs associated with
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the degraded power supply could result in a failure of the CEAs to move or dropped
CEAs. BGE entered a six hour action statement for limiting condition for operation
(LCO) 3.1.3.1, Moveable Control Assemblies CEA Position.
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BGE obtained and bench tested a replacement power supply. A maintenance orc'er
was planned and prepared to support the replacement of the power supply. System
engineering provided support during the power supply replacement. The
replacement was well coordinated with Operations. Replacement of the power
supply was well controlled. Appropriate safety precautions and independent
verification of lifted and landed leads were used. Oversight of the job was provided
by the first line supervisor and the General Supervisor - Electrical Maintenance.
Prompt and effective replacement of the 12 VDC power supply in section 5 of the
control element drive coil power programmer cabinets precluded an unnecessary
transient on the plant.
On February 24, the inspectors observed the General Supervisor Mechanical
Maintenance performing Worker Risk Assessment Process (WRAP) observations.
The inspectors discussed the WRAP process with BGE personnel. The WRAP
process was implemented to identify at-risk personal safety behaviors and reduce
industrial safety accidents. The behaviors monitored were in the categories of
personal protective equipment, environment, body mechanics, tools and equipment,
and other, such as, communications. The use of the WRAP program has been
effective in improving industrial safety for maintenance in 1997, approximately
3100 observations were performed. Ninety-five percent of the observations were
classified as safe and five percent of the observations were identified as at-risk.
The program was recently expanded to monitor radiation safety behaviors including
contamination control, wearing of dosimetry and as-low-as-reasonably-achievable
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c. Conclusions
The observed maintenance was conducted safely and in accordance with BGE
approved procedures and controls. Workers were knowledgeable and performed
work effectively. Prompt and effective replacement of the 12 VDC power supply in
section 5 of the control element drive coil power programmer cabinets precluded an
unnecessary transient on the plant. System engineering provided effective support
to maintenance during the 12 VDC power supply replacement. Good supervisory
oversight of maintenance was observed during this period. The inspectors also
I concluded that the Worker Risk Assessment Process (WRAP) has been an effective
l initiative to aid in continuous improvement of industrial safety practices for
l maintenance personnel. Maintenance management has recently expanded this
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program in an effort to improve radiation safety practices for maintenance
personnel.
M1.2 Routine Surveillance Observations
a. Insocction Scoce (61726)
The inspectors observed all or portions of the following selected surveillance tests:
STP-M-2OO-2 Reactor Trip Circuit Breaker Functional Test
STP-O-56C-1 ESFAS Equipment Response Time - Modes 1 & 2
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STP-O-56D-1 ESFAS Equipment Response Time - Modes 1 & 2 {
STP O 56C-2 ESFAS Equipment Response Time - Modes 1 & 2
l STP-043D-2 ESFAS Equipment Response Time - Modes 1 & 2
l STP-O-5-2 AFW System Monthly Surveillance Test
STP-0 731-1 HPSI Pump & Check Valve Quarterly Test i
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! b. Findinos and Observations
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The observed surveillance testing was performed safely and in accordance with
approved procedures. Pre-test briefings included test objectives, assigned
l responsibilities, means of communication, test control details, contingency actions, l
and use of event free tools. The inspectors noted that an appropriate level of '
supervisory attention was given to the testing, including direct observation of test
steps and independent verification of important steps and calculations. The
inspectors found that approved procedures were in use, details were adequate,
l technical specifications were satisfied, testing was performed by qualified ,
l personnel, and test results satisfied acceptance criteria. l
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c. Conclusions
! The observed surveillances were conducted safely and effectively demonstrated i
l system operability. Operators demonstrated good use of self-checking techniques.
l The pre-test briefings performed by operations personnel were excellent in scope, i
l content, and level of detail.
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i M8 Miscellaneous Maintenance issues (92700)
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l M8.1 (Closed) LER 50 317/97-008:Two Reactor Protective Channels Out of Service
i During Test
The Licensee Event Report discussed a " Channel Inoperable" alarm that was
l received on Unit 1 Channel D Wide Range Nuclear Instrumentation (WRNI) while
l Channel C was out-of-service for surveillance testing. BGE identified this as a
condition prohibited by Technical Specification (TS) 3.3.1.1 because two of four
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channels were inoperable versus the minimum of three operable channels specified
i by TS table 3.3.1. BGE entered TS 3.0.3, terminated the calibration procedure, and
began to restore Channel C to an operable status. The Channel D alarm cleared
j itself within approximately 11 minutes. Channel C was restored to an operable
, status within 50 minutes. The initial troubleshooting of Channel D focused on the
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power supply and was inconclusive, but BGE committed to perform a root cause
analysis.
Subsequent troubleshooting discovered a failed isolator in Channel D. Based on
discussions with vendor design engineering personnel for the WRNI system, BGE
concluded that the isolator was the most probable cause for receiving the " Channel
Inoperable" alarm. This particular type of isolator has experienced an elevated
number of failures due to excessive heat build-up. The vendor has designed a
replacement isolator with less heat build-up, and an expected longer life. BGE has
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not experienced an increased failure rate of these isolators that would indicate a
generic end of-life concern. BGE has made other modifications to the WRNI
drawers that has reduced the heat build-up in the drawers. Therefore, BGE plans to
replace these isolators as they fail.
111. Enoineerina
E2 Engineering Support of Facilities and Equipment
E2.1 Resolution of Design issues Associated with the NUHOMS Dry Cask Storage
System .
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a. insoection Scooe (37001 and 60851)
Transnuclear West (TN West, formally VECTRA Technologies Inc.) initiated a design
review of the NUTECH Horizontal Modular Storage System (NUHOMS) dry shielded
canister (DSC). TN West initiated the review in response to concerns about the
methods previously used by VECTRA to disposition a DSC nonconformance issue.
The review was scheduled for completion in late March 1998. The TN West review
was focused on the Standardized NUHOMS system which may be used under a
general license. BGE has a site specific license to use a NUHOMS system, which is
similar but not identical, to the Standardized NUHOMS system. BGE planned to
load four casks before the start of the Unit 1 refueling outage in April 1998. In lieu
of waiting for TN West to complete its review, BGE elected to perform an
independent design review specific to its NUHOMS system.
The purpose of this inspection was to assess whether BGE adequately evaluated
concerns raised by TN West, regarding the design of the Standardized NUHOMS
system, for applicability to the site specific license for Calvert Cliffs.
b. Observations and Findinos
b.1 Evaluation Process used by BGE
By a letter dated December 5,1997, TN West notified BGE of several discrepancies
with the NUHOMS DSC structural calculations. TN West stated that an
independent review of the NUHOMS design calculations was in progress. The letter
noted that the major areas of potential noncompliance with the license were those
conditions related to the vertical end drop of the DSC. The letter also stated that
the system, although not in compliance with the license, had the ability to be safely
operated and handled without creating a substantial safety hazard.
Attached to the letter was a matrix of all issues that had been identified during an
independent evaluation of the Standardized NUHOMS DSC design. The matrix
briefly described the issue and its applicability to the NUHOMS design used at
Calvert Cliffs.
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7
In response to the findings of the independent evaluation, BGE contracted Hopper
and Associates to perform an evaluation (BGE Calculation CA04132, Revision 000)
of the structuralintegrity of the DSC and internals. The evaluation was only
applicable to nine DSCs fabricated for Calvert Cliffs but not yet loaded with spent
fuel. A vertical top end drop was evaluated. Each DSC component was
investigated for compliance with the Certificate of Compliance, the Updated Safety ,
Analysis Report (USAR) and the Safety Evaluation Report (SER). The calculations l
were reviewed and approved by BGE on February 16,1998.
In addition to its own review, BGE contracted with TN West to perform an
independent review of Hopper and Associates' evaluation. TN West divided its
review into two tasks. The first task assessed the completeness of the Hopper and
Associates' evaluation against the commitments from the generallicense
independent assessment. The second task assessed the technical adequacy of i
each issue. On February 17,1998, TN West notified BGE that the first task had l
been completed. On March 2,1998, TN West completed the second task. TN
West identified 38 issues during its second task review. One issue, the buckling of
the support rod, remained under review by BGE at the conclusion of the inspection. )
TN West further concluded that the new calculations completely revised the original i
BGE DSC design calculations.
The inspectors reviewed the above documentation and determined that BGE, with
tha assistance of TN West, had established an adequate process to track the
technicalissues received from TN West on December 5,1997, in addition, BGE
collaborated with TN West to review all corrective actions that resulted from a
Demand for Information, issued to VECTRA, for applicability to Calvert Cliffs' site
specific license, in both instances the inspectors determined that BGE had
developed and followed a comprehensive process to ensure all design and quality
assurance issues which might have effected its DSCs were resolved.
b.2 Enaineerina Evaluation (BGE Calculation CA04132. Revision 000) )
l
During the inspection, representatives for BGE and Hopper and Associates
presented the inspectors with an overview of the issues (see Attachments 2 and 3).
The concerns were bounded by a vertical top end drop accident of the DSC inside
of the transfer cask as described in Section 8.2.5 of the USAR. The original
calculations performed to support the USAR noted that during a vertical top end
drop accident the clip angles, which attached the guide sleeves to the bottom
support plate, would failin shear at an acceleration of 35 G and allow the guide
sleeves to fall through the spacer disc square holes. This would in turn prevent
additional stresses, due to the weight of the sleeves, to be transferred to other DSC
components. However, BGE determined through analysis and testing that the clips
actually failed by bending (not shearing) at approximately 43 G.
The inspectors later learned from TN West that the original calculations assumed
nominal dimensions of the clip angles in lieu of maximum tolerances variations
which could be up to 25 percent of the nominal values. When the larger design
values were used, the clips failed at a higher acceleration. As a result of the higher
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acceleration, more load was transferred to the bottom spacer plate and ultimately
the support rods.
Based on the results of BGE Calculation CA04132, Hopper and Associates
'
concluded, and BGE agreed, that following a potential vertical top end drop accident
the nine DSC assembliss met the applicable site specific licensing requirements and l
'
retrievability of the fuel assemblies would be maintained.
The inspectors performed an independent review of select portions of the
calculations and identified the following concerns with BGE Calculation CA04132,
Revision 000:
- The calculations did not evaluate the effects of bending moment on the
spacer disc-to-support rod welds. Calculations were only performed to
evaluate the shear stress of the welds. Since the clips do not fail at 35 Gs
as originally calculated, additional forces are transferred to the spacer disc-
to-support rod welds. Hopper and Associates stated that per the SAR the
DSC was designed to American Society of Mechanical Engineers Boiler and
Pressure Vessel (ASME B&PV) Code Section lil, Division 1, NB, Class I, and
as long as codu welds were applied the stresses did not need to be
evaluated. However, the inspectors determined from DSC drawings that the
welds were not code welds and that all weld stresses should have been
evaluated. TN West later also identified this issue during its technical
review. In response, the calculation was revised to include an evaluation of
the bending moment in the spacer disc-to-support rod welds. The inspectors
reviewed the revised calculation and identified examples of nonconservative
assumptions. One example was that the calculation stated that "Using a
rotation value for a pinned rod condition is overly conservative. It assumes
that the rods offer no rotational resistance to the spacer disk. In reality,
even if the rod forms a plastic hinge, a resisting moment equal to the plastic
moment is available." To account for this, the moment was reduced to
M' = 212-97 = 115 KIP-IN. The inspectors determined that this assumption
l did not have an adequate technical basis. Therefore, the assumption may be
i nonconservative because the support rod plastic moment should still be
l restricted by the spacer disc-to-support rod welds.
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- The calculation concluded that, at worst case, the clip angles would not
i shear but would bend and impinge on the guide sleeves and pinch the fuel
assemblies.10 CFR 72.122(l) requires that the fuel assemblies must be
retrievable. Therefore, the calculation demonstrated that while the fuel
assemblies were pinched inside of the guide sleeve, it would require an
additional lifting force of not more that 273 lbs. to break the clip angles and
remove the assemblies. This amount was within the capacity of the spent
fuel pool handling crane and plant procedures. However, the calculation
used nominal, in lieu of worst case clip angle dimensions to calculate the
additional removal force. The inspectors performed the calculation using the
maximum design dimensions and calculated that a force of 335 lbs would
be required. Not using maximum angle clip tolerances was of significant
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concern to the inspectors because it was that type of omission in the original
calculations that under estimated the breaking force of the clip angles.
- The calculation did not discuss effects of spacer plate deformation on the
guide sleeves and any resulting effects on clip impingement of the fuel.
Section 8.2.5.2 of the USAR states that: " .the
.. maximum spacer disk
deflection was computed to be 0.077". The gap between the guide sleeve
and the fuel assembly exceeds 0.5". Therefore, the fuel retrievability from
the DSC will be ensured for the postulated 80" drop accident."
BGE Calculation CA04132, Revision 000, concludes that the maximum
bottom spacer disk lateral deflection could be as large as 0.24" during the
vertical top end drop accident. However, the inspectors determined that the
calculation does not address the effects of the revised deflection on the gap
between the guide sleeve and the fuel assembly in conjunction with the
bending of the clip of the clip angle.
- The inspectors determined that the calculation did not evaluate the effects of
spacer disc-to-support rod bending moment on buckling the support rod.
Hopper and Associates revised the calculation, however, the inspectors
concluded that the revision did not fully address the rod buckling issue. This
conclusion was also reached by TN West in its technical evaluation of the
revised calculation,
b.3 BGE DSC Guide Steeve Standard Clio Desian Test
in January 1998, BGE conJucted two tests. The first test was performed to l
evaluate the f ailure mode of the guide sleeve clips and to quantify the magnitude of I
the load necessary to separate or break the clips which are welded between the
guide sleeves and bottom spacer disc. The second test was performed to confirm !
the results of the first test and to quantify the force required to extract a mockup
fuel assembly in the guide sleeve for a pinched condition. Both tests were
conducted at Ranor, Inc., the first test was supervised by TN West and the second
test by BGE.
The inspectors reviewed the methodology and results of the first test and agreed
with the conclusions. The clip angles failed at approximately 41 Gs during the test
which closely corresponds to the calculated upper bound value of 43 Gs.
However, the inspectors disagreed with the methodology and results of the second
test. The inspectors determined that the test did not adequately demonstrate fuel
retrievability. Specifically, the test did not accurately model the deformation of the
bottom spacer plate and the associated affects on the guide sleeves for a vertical
top end drop accident.
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b.4 10 CFR 72.48 Safetv Evaluation
, .The inspectors reviewed 10 CFR 72.48 safety evaluation (SE) ES199601368
Supplement 002, Revision 0000, and developed the following concerns:
- The SE states: "This safety evaluation is prepared to clarify and correct the
licensing basis for the NUHOMS system in use at the Calvert Cliffs
Independent Spent Fuel Storage Installation (ISFSI) with respect to the
postulated transfer cask drop accident. The USAR will be changed to
properly account for the behavior of DSC components, and to correct the
stress values and deflections that might be expected in the unlikely event of
a cask vertical drop accident."
The inspectors concluded that this staterm,nt indicated that the license
conditions, as submitted to NRC for approval of a site specific license, were
incorrect. The inspectors reviewed the USAR and referenced documents and
confirmed that several conclusions concerning the behavior of the DSC
during a vertical top end drop did not agree with the evaluation performed by
Hopper and Associates. The SE discrepancies included that the angle clips
would fail at a higher acceleration than the 35 Gs referenced in the USAR,
that the clip angles could fail in a manner that would cause the guide sleeve
to impinge in the spent fuel assemblies and that the bottom spacer plate
would deform in excess of the .077 inches stated in the SAR. Since these
discrepa cies existed the inspectors determined that the SAR, originally
submitted by BGE to NRC, did not completely and accurately describe the
behavior of the DSC resulting from a vertical top end drop accident. 10 CFR
72.11 requires that license applications include complete and accurate
j information.
- The SE states: "The possibility of a malfunction of a different type than
previously evaluated in the SAR will not be created as a result of this
proposed activity."
However, the inspectors determined that an equipment malfunction of a
different type than previously evaluated did exist in that the USAR had not
assumed that the clip angles would wedge the guide sleeve and pinch the
fuel assembly. The following statement from the USAR indicates that
contact between the guide sleeve and fuel assembly was not expected:
...the maximum spacer disk deflection was computed to be 0.077 inches.
The gap between the guide sleeve and the fuel assembly exceeds 0.5
inches. Therefore, the fuel retrievabiltiy from the DSC will be ensured for the
postulated 80" drop accident.
BGE stated that because they believed fuel retrievability was maintained with
the use of existing plant equipment, albeit with a larger extraction force, that
impingement of fuel assemblies was not a problem. The inspectors again
disagreed with this statement. As discussed in report sections b.2 and b.3,
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the calculational and testing methods utilized to demonstrate retrievability of
the fuel assemblies did not take into account worst case scenarios.
Based on the above, the inspectors concluded that the possibility for an accident or
l malfunction of a different type than eveluated previously in the Safety Analysis
Report existed. Therefore, the inspectors determined that BGE had not identified
! that an unreviewed safety question (USO) existed as defined in 10 CFR
72.48(a)(2)(ii). The inspectors also determined that BGE should have applied to
NRC for a license amendment in accordance with 10 CFR 72.56.
BGE disagreed with this conclusion and on February 22,1998, initiated ar' issue
report to document the differing views. As a result of the issue report, BGE
performed an operability determination to justify loading additional casks until the
issue was resolved. BGE concluded that the DSCs were operable. The operability
determination was based on the results of structural calculations and the low
probability of the a top end vertical drop accident outside of the Auxiliary Building.
On March 4,1998, BGE initiated actions to load DSCs with spent fuel. The
inspectors reviewed the operability determination and as discussed above did not
agree with the adequacy of the structural calculations. However, the inspectors
determined that calculation deficiencies did not affect the conclusion of the
operability determination, that the DSCs evaluated were operable for loading spent
fuel. Further, the inspectors agreed with the assumption that the vertical top end
drop accident was a low probability event.
c. Conclusions
The inspectors concluded that BGE established a good process to track the DSC 3
design issues received from TN West on December 5,1997. BGE developed and t
followed a comprehensive process to ensure all design and quality assurance issues
which might have affected its DSC were resolved. However, the inspectors further
concluded that BGE had not adequately resolved all of the design issues.
The inspectors determined that BGE's original license submittal, for use of the
NUHOMS System at Calvert Cliffs under a site specific license, did not contain
complete and accurate information regarding the behavior of the DSC under all
accident scenarios. The inspectors concluded that this was a violation of 10 CFR
72.11 which requires in part that information provided to the Commission by an
applicant for a license shall be complete and accurate in all material respects
(VIO 50-317&318/98-01-01)
l
Of more significant concern, the inspectors determined that BGE had not identified a l
USQ involving a malfunction of a different type than any evaluated previously in the l
Updatec' Safety Analysis Report. Specifically, that the failure of the angle clips
could cause the fuel to be impinged by the guide sleeve. Therefore, BGE should l
have applied to NRC for a license amendment in accordance with 10 CFR 72.56. l
The inspectors concluded that this was a violation of 10 CFR 72.48(a)(2)(ii),(VIO
l
50 317&318/98-01-02). l
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The inspectors further concluded that this issue, while it constituted a USQ, had
limited safety significance due to the low probability of a vertical top end drop
accident, in addition, if such an event did occur, the fuel would remain subcritical
and the containment barrier of the DSC would be maintained.
E5 Engineering Staff Training and Qualification
E5.1 Probabilistic Risk Assessment (PRA) Continuing Training
a. insoection Scoce (37551)
On February 26,1998, the inspectors observed a two-hour continuing session
training provided to approximately 30 engineering and technical support (E&TS)
staff personnel on probabilistic risk assessment (PRA).
b. Observations and Findinos
The training was provided by the principal engineer (PE) for the unit that developed
and maintains the Calvert Cliffs Probabilistic Risk Assessment (CCPRA). This
training was offered as a continuing training option.and was expected to be given to
approximately 180 of the 300 E&TS staff members. Discussions with the PE
indicated that similar training has been provided to the offsite safety review
committee (OSSRC) and the plant operating safety review committee (POSRC). The
stated learning objectives of the training was to describe the basic elements of a
PRA, gain an appreciation of the construction of the CCPRA, understand the results
of the CCPRA, and to understand the current and future applications of the CCPRA.
The inspectors observed that the training provided a good overview of PRA
terminology and development methods. NRC Generic Letter 88-20, Individual Plant
Examination (IPE) for Severe Accident Vulnerabilities, and the supplements related
to Individual Plant Examination of External Events (IPEEE) for Severe Accident
Vulnerabilities and BGE's associated submittals to the NRC were also discussed.
The training identified that updating of the CCPRA model was ongoing. A good
discussion of the results of the IPE, and IPEEE was also provided. Significant risk
contributors and top sequences were also discussed. Applications of the CCPRA
were also discussed, including the use of risk data during the development of the
weekly maintenance schedules.
.c. Conclusions
The inspectors concluded that the PRA training was appropriate in scope and detail.
The stated learning objectives were effectively met. Overall, the continuing training
provided to selected engineering and technical support staff personnel provided an
excellent overview of the development, results and applications of the CCPRA.
_
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E8 Miscellaneous Engineering issues (92903)
E8.1 (Closed) Unresolved item 50-317&318/97-04-04: Effectiveness of Corrective
Actions for Emergency Lighting Deficiencies
NRC Inspection Report (IR) 50 317&318/97-04 documented an NRC identified
concern related to the adequacy of corrective actions for emergancy lighting ,
deficiencies and the reliability of emergency lighting between surveillance tests.
'
An additional thorough review of this issue was documented in IR 50 317&318/97-
05. BGE was cited for a failure to review emergency lighting test results and take
corrective actions to prevent repetition of lighting failures, (VIO 317&318/97-05-
04). Therefore, the unresolved item is closed.
E8.2 (Closed) Violation 50-317&318/96-09-01: Defacto Modification of Auxiliary Feed
Water Pump Base
This violation pertained to a non-conformance regarding an unauthorized
modification to the auxiliary feedwater (AFW) pump turbine base guide blocks. The
inspectors reviewed the corrective and preventive actions implement by BGE to
preclude such occurrences. These actions were documented in BGE's response to
the violation, dated December 12,1996. The inspectors' review indicated that in
response to the issue, BGE performed a root cause analysis to determine the
underlying deficiencies leading to this violation, and proposed corrective actions to
prevent such occurrences in the future. The corrective action steps included: (1)
counseling of system engineers regarding the need to obtain proper design reviews
prior to making plant configuration changes, (2) training to all plant engineering
personnel to emphasize that system engineers could not authorize plant
configuration changes without formal approval from design engineering, (3)
awareness training to maintenarice personnel regarding this event to assure that
maintenance did not inadvertently implement unauthorized plant configuration *
changes, and (4) design-related guidance issued to plant engineering during the
- 1996 outage was reviewed to determine if other similar events had occurred. The -
U
inspectors verified the documentation of the above actions, and concluded that the
corrective steps implemented were adequate to resolve this violation. This violation
is closed. ;
IV Plant Support
R1 Radiation Protection and Chemistry Controls (RP&C)
R1.1 Radioactive Waste Processing, Handling, Storage, and Shipping
a. Insoection Scope (86750)
The inspectors reviewed and discussed sources of radioactive waste at the station,
the processing (as appropriate) of the waste, and volume reduction efforts for
waste. The inspectors evaluated the methodology for radioactive waste
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14
concentration averaging (i.e., the determination of the average radioactivity
concentration for shipment of materials mixed together) and the development of
scaling factors used to estimate hard to detect radionuclides (e.g., Pu-239,
Am-241). The inspectors also selectively reviewed radioactive waste shipping
records for shipments made since the previous inspection, including shipments of
low specific activity (LSA) material, spent resins, and filter cartridges.
The review was against criteria contained in 10 CFR 20; 10 CFR 61; 10 CFR 71;
49 CFR 100-199;the Barnwell, South Carolina disposal facility license; applicable
certificates of compliance for various NRC licensed shipping casks; the Updated
Final Safety Analysis Report; and applicable NRC Branch Technical Positions,
b. Observations and Find nas
Waste Volume Reduction
BGE developed and was implementing plans and programs to reduce generated
radioactive waste volumes to a minimum. BGE significantly reduced its generated
waste volumes over the past three years and was continuing to evaluate other
techniques and methods to further reduce waste volumes. BGE was developing
water management plans for the upcoming Unit 1 outage to minimize unnecessary
discharge of water as radioactive liquid effluent and to maximize clean-up to reduce
total curies released. These plans were similar to those used during the previous
outage and were considered a very good initiative.
Waste Classification / Concentration Aversaina/Use of Scalina Factors
BGE performed a review of the methodology for determination of the curie content
of spent resins shipped for disposal. BGE concluded that, due to use of potentially
overly conservative scaling factors for hard to detect radionuclides, the curie
content of radioactive waste shipped for disposal may be overestimated by a farNor
of 2.5. BGE also determined that curie content calculations for Fe-55 in resin
shipments made in 1997 may have been nonconservative. Subsequently, BGE
suspended shipments of spent resins and initiated a review of this matter.
Preliminary reviews by BGE indicated that, though the curie content of the waste
was overestimated, the waste was shipped in proper shipping casks and the waste
was properly classified for cisposal purposes. BGE was continuing to review this
matter.
The inspectors' review indicated it was not apparent that BGE was conforming to
NRC Branch Technical Positions (BTPs) regarding waste concentration averaging,
particularly for cartridge filters. The inspectors reviewed the methodology for
concentration averaging and the determination of scaling factors used to quantifi/
the curie loading of hard to detect radionuclides on filter cartridges. The inspectors
noted that filters from different waste streams (e.g., the chemical volume control
system and the spent fuel pool cleanup system) were consolidated for shipping
purposes and that their radioactivity concentrations were averaged for determination
of average radioactive material curie content. However, the determination of
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
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average radioactivity concentrations for the filters did not appear to be in ,
conformance with the NRC Branch Technical Position (BTP). Specifically, samples I
of certain filters exhibited curie loading, per sample of cartridge filter, beyond the
factor of 10 parameter described in the guidance provided by the BTP. Further, i
radionuclides identified via laboratory analysis were not evaluated on a common unit
bases (e.g., microcuries per gram) as described by the guidance.
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The inspectors noted BGE had previously suspended shipments of cartridge filters l
from the chemical volume and controls system in light of similar concerns l
associated with determination of waste classification due to apparent elevated j
indications of Carbon-14 radioactivity. BGE subsequently determined that certain i
cartridge filters may exceed Class C waste limits and were unsuitable for disposal. I
These matters were under review by BGE. )
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While no violations were identified, BGE's waste classification and curie content j
determination methodology will be reviewed during a subsequent inspection and is
l
an inspector follow-up item (IFl 50-317&318/98-01-03). l
Radioactive Waste Shinoina
i
The radioactive waste / material shipping program was generally well implemented
relative to conformance with applicable Department of Transportation shipping l
regulations. Radioactive material shipping documentation was well maintained and
available for review. Individuals responsible for shipping activities were
knowledgeable of applicabie requirements. The inspectors noted that the curie
content of waste shipments may be conservative due to use of apparent overly
conservative scaling factors. This matter will be re-examined as part of 1
IFl 50-317&318/98-01-03, discussed above. l
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BGE was a registered user of the NRC licensed casks it used for shipping purposes )
and maintained up-to-date cask certificates of compliance and drawings. Shipments I
of radioactive materialin NRC licensed casks was performed in accordance with
certificate of compliance requirements. BGE maintained up-to-date dispcaal facility
licenses,
c. Conclusions
BGE's basic radioactive waste transportation program, including processing,
handling, storage, and transportation were adequate. However, technical
weaknesses were identified in the methodology used for waste concentration l
averaging and determination of scaling factors. Since these technical matters had
the potential to affect the accuracy of curie content for waste shipping and
classification purposes, BGE suspended shipment of radioactive waste and material l
pending resolution. No violations of NRC sequirements were identified. I
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R2 Status of RP&C Facilities and Equipment
a. Insoection Scoce (86750)
The inspectors toured and visually inspected various radwaste and radioactive
material storage areas including the Lake Davies area, the Materials Processing l
Facility (MPF), and the V/est Road cage area. I
b. Observations and Findinos
Areas were generally well maintained, properly posted and controlled. The
following observations were made.
- The Lake Davies area, a fenced in area, had unlocked boxes containing
radioactive material. BGE initiated a review of this matter and elected to lock
the unlocked boxes of material stored at the location.
- The sign providing special frisking instructions to personnel exiting the
Materials Processing Facility was faded and illegible. BGE initiated a review
of this matter.
- BGE routinely surveys lockers located near the main radiological control
access point as a secondary check for potential contamination. The lockers
located next to the main access point to the MPF, a radiological control area,
were not frisked. BGE initiated a review of this matter.
- A sign for the West Road cage was posted inside the fence and was not
readily observable. BGE initiated a review of this matter.
c. Conclusions
Overall, radioactive waste and material storage areas toured were properly posted
and controlled. Areas for improvement were noted.
R3 RP&C Procedures ana Documentation
a. Insoection Scope (86750)
The inspectors discussed changes in radioactive waste processing, handling, and
shipment procedures and programs since the previous inspection with licensee
personnel responsible for these areas.
b. Observations and Findinos
There were no major changes identified in the radioactive waste processing,
handling, storage, and shipment procedures and programs since the previous
inspection. New personnel coming into the radwaste group were provided training,
as appropriate, and prohibited from performing tasks for which they were not yet
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qualified. BGE was continuing a review of the radwaste shipping program. The
review was begun in the fall of 1997 and included efforts relative to the
determination of package curie content and waste classification in light of elevated
contamination identified in the Unit 2 reactor cavity during the 1997 refueling
outage following use of hydrogen peroxide for reactor coolant system cleanup
purposes.
BGE enhanced the program procedures for determining the curie content for strong-
tight packages (steel boxes). Procedures were also enhanced to require clear
survey documentation for radweste shipments. BGE also modified procedures to
enhance controls of for requiring documentation that all routine determinations
specified by 10 CFR 71.87 were completed. Procedures were modified to require
documentation that Certificate of Compliance requirements were verified prior to
shipping radwaste shipping casks, and to provide enhanced instructions to radwaste
shipment drivers.
c. Conclusions
No apparent program or personnel changes were noted that could adversely affect
the radioactive waste processing, handling, storage, and transportation program
were noted. Procedures and program implementation appeared effective.
R4 Staff Knowledge and Performance in RP&C
a. insoection Scope (86750)
The inspectors evaluated general staff knowledge of radioactive waste prucessing,
handling, storage, and shipping requirements during the inspection.
b. Observations and Findinos
The inspectors' discussions with personnel during the inspection indicated generally
a good knowledge level of regulatory requirements and program procedures. The
personnel were aware and knowledgeable of applicable regulatory requirements,
including procedural specifications, Department of Transportation rules and
regulations, and radiological survey and assessment methodologies. However, the
inspectors noted some apparent weaknesses in staff knowledge of NRC Branch
Technical Position guidance relative to waste concentration averaging. ,
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c. Conclusions
individuals responsible for radioactive waste processing activities exhibited a
generally a good knowledge level of regulatory requirements and program
procedures. The inspectors noted some apparent weaknesses in staff knowledge of
NRC Branch Technical Position guidance relative to waste concentration averaging.
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R5 Staff Training and Qualification in RP&C
a. insoection Scone (86750)
The inspectors selectively reviewed the training provided for personnel involved in
radioactive vmste generating, processing, and handling activities and personnel
handling mixed waste against criteria contained in IE Bulletin 7919 and 49 CFR
172, Subpart H. The inspectors reviewed selected training records and lesson plans
and d;scussed training with cognizant BGE personnel.
b. Observations and Findinas
BGE continued to provide appropriate training to Materials Processing Unit personnel
in accordance with IE Bulletin 79-19 and 49 CFR 172, Subpart H, guidance.
BGE provided training of station personnel on the requirements of IE Bulletin 7919
and 49 CFR 172, Subpart H, via structured peneral employee training.
c. Conclusions
The inspectors concluded that BGE was providing appropriate training as outlined in
IE Bulletin 7919 and 49 CFR 172, Subpart H.
R6 RP&C Organization and Administration
a. Inspection Scone (86750)
The inspectors reviewed the current radioactive waste processing organization,
including staffing, responsibilities and authorities. The inspectors evaluated BGE's
performance in this area by discussion with cognizant personnel and review of
applicable administrative and organizational records.
b. Observations and Findinos
The inspector's review indicated that there were no significant changes in the
organization or its responsibilities and authorities since the previous inspection in
this area. Responsibilities and authorities were appropriately defined.
c. Conclusions
GGE centinued to implement an appropriately staffed and defined organization
responsible for radioactive waste processing, handling storage, and shipping.
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R7 Quality Assurance in RP&C Activities
a. Insoection Scope (86750)
The inspectors reviewed BGE's audit of the radioactive materie's management
,
program against the criteria contained in the BGE's Quality Assurance Policy,
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Revision 49, and 10 CFR 71, Subpart H, Quality Assurance,
b. Observations and Findinas
BGE performed an audit of radioactive materials management in 1997 (Audit No.
97-09, dated October 23,1997). A technical expert supplemented the audit team.
No significant concerns were identified during the audit. The inspectors made the
following observations. l
- The audit report provided a summary of effectiveness and performance
ratings in " Business Function Elements." It was not apparent that the audit
covered all applicable topics outlined in 10 CFR 71.101 through 71.137.
- The inspectors requested the specific checklists used for the audit to
ascertain specific regulatory requirements audited as well as bases for
effectiveness and perfornance conclusions. It was not apparent that the
audit was conducted using written checklists as specified in 10 CFR 71.137.
The bases for performance ratings in some areas (e.g., training and
qualification of personnel) was unclear.
- The audit team leader had apparently not received any specific training in
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radioactive waste handling and shipping activities in approximately two
years. The inspectors noted that 10 CFR 71.137 requires that audits be
l performed by appropriately trained personnel.
- Based on the documents provided and the audit reviewed, it was not
apparent that BGE had developed and implemented a well structured and I
defined audit program for radioactive waste processing, handling, storage, ,
transportation. The inspectors was unable, based on the level of detail of i
the audit, and audit plan to clearly identify what specific regulatory l
requirements were reviewed and examined. i
i
BGE initiated a review of the above matters. This area requires further review to
,
ascertain adequacy with respect to regulatory requirements. (
l (IF150 317&318/98-01-04) i
1
c. Conclusions i
BGE performed an audit of radwaste processing, handling, storage, and
transportation, including training and qualification of personnel. However, the audit
did not appear to be well structured or include a detailed audit checklist. Although
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20
further review of this area will be required, no violations of NRC regulatory
requirements were identified.
R8 Miscellaneous RP&C lasues
R8.1 (Update) Unresolved item 50-317&318/96-04-03: Verification of Updated Final
Safety Analysis (UFSAR) Commitments
While performing the inspections discussed in this report, the inspectors reviewed
the applicable portions of the UFSAR that related to the areas inspected. In
particular, the inspectors evaluated waste storage matters. During a previous
inspection, severalinconsistencies associated with processing and storage of
radioactive waste and material at the Calvert Cliffs Nuclear Power Plant relative
to descriptions and commitments provided in Chapters 1,11, and 12 of BGE's
UFSAR were identified. The inspectors identified that the UFSAR was updated, via
Revision 20, to reflect current radioactive waste processing, handling and
transportation practices. BGE had initiated, but not completed, safety evaluations
for specific onsite radwaste and radioactive material storage locations not identified
in the UFSAR.
R8.2 (Closed) Violation 50-317&318/96-07-02: Loss of Integrity of Radioactive
Materials Shipment
On May 24,1996, a shipment of cs ntaminated scaffolding on a flatbed trailer was
shipped to the Chem-Nuclear Systems, Barnwell, South Carolina facility for
processing. Upon arrival at the facility, the container, a 20-foot long sea-land type
container was found to have a 5-6 inch hole in the floor of the container.
The inspectors verified that BGE implemented the corrective actions as detailed in
. Lorrespondence dated December 17,1996. The actions included revising
procedures to require the use of weight distributing plates to prevent concentrated
forces from punchy holes in shipping containers, revising driver instructions to
notify the shipper of emergency stops and transport conditions that may affect
package integrity, and revising radiation safety personnel training relative to
verification of package preparation.
R8.3 (Updated) Violation 50-317&318/97-06 04: Lack of Procedure to Survey Laundry
This matter was reviewed during NRC Inspution 50-317;318/97-07, dated
January 27,1998. The inspectors also reviewed this matter with respect to BGE's
January 5,1998, response. BGE implemented the corrective actions as outlined in
its letter. However, the inspectors noted that during sampling and monitoring of the
laundry, BGE identified a piece of laundry with elevated activity (apparent hot
particle) and removed that article from service. The inspectors performed an
estimate of shallow dose equivalent assuming the particle contacted the skin and
concluded that the particle could proouce a shallow dose equivalent of several
hundred millirad per hour. The inspectors determined that BGE had not initiated any
action to verify the quality of the affected laundered lot of protective clothing to
. 21 I
determine the potential or presence of other radioactive particles. BGE immediately
initiated action to remove the laundry from service and initiated a review with the
vendor. This matter remains open in that corrective actions did not address actions
to be taken on detection of elevated levels of radioactivity on returned protective
clothing.
R8.4 Dose Assessments for May 5,1997, Unit 2 Reactor Cavity Event 1
The inspectors met with cognizant personnel, including radiation protection ,
technicians, to better understand the sequence of events and the cdequacy of dose f
assessments for those personnel associated with the May 5,1997, Unit 2 reactor !
I
cavity airborne event. NRC IR 50 317&318/98-03 documented a previous review
of this event and the weaknesses with the in-plant evaluation of radiological
conditions. The inspectors noted that BGE's follow-up of the event und dose
assessments were on-going. Follow-up dose assessments indicated that no
ll
individuals had been identified as sustaining any significant apparent external or
internal exposure. The dose assessments appeared appropriate based on the
sequence of events described by involved radiation protection personnel. The
inspectors verified that the individuals performing the May 5,1997 flange cleaning j
activities, and wearing respirators, were qualified to wear respiratory protective
equipment. The inspectors will review the final dose assessment during a
subsequent inspection.
1
V. Manaaement Meetinas
X1 Exit Meeting Summary )
!
During this inspection, periodic meetings were held with station management to
discuss inspection observations and findings. On March 27,1998, an exit meeting i
was held to summarize the conclusions of the inspection. BGE management in
attendance acknowledged the findings presented. I
X2 Review of UFSAR Commitments
While performing the inspections discussed in this report, the inspectors reviewed
the applicable portions of the updated final safety analysis report (UFSAR) that
related to the areas inspected to verify that the UFSAR wording was consistent with
the observed plant practices, procedures and/or parameters. No concems were
identified. )
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ATTACHMENT 1
PARTIAL LIST OF PERSONS CONTACTED
B.G.E
C. Cruse, Vice President- Nuclear Energy Division
P. Katz, Plant General Manager
P. Chabot, Manager, Nuclear Engineering
K. Cellars, Superintendent, Nuclear Maintenance
K. Neitmann, Superintendent, Nuclear Operations
T. Pritchett, Acting Manager, Nuclear Engineering
S. Sanders, General Supervisor, Radiation Safety
T. Sydnor, General Supervisor, Plant Engineering
J. Lemons. Manager Nuclear Support Services Department
A. Edwards, Director Nuclear Security
C. Sly, Senior Engineer, NRM
J. Osborne, Nuclear Regulatory Analyst
M. Tonacci, General Supervisor- Chemistry
G. Tesfye, Senior Engineer, Nuclear Regulatory Matters
P. File, Principal Engineer, Nuclear Fuel Management
J. Remeniuk, Senior Civil Engineer l
M. Gahan, Principal Engineer, Civil Engineering
J. Lipold, General Manager, Technical Services Engineering
R. Gradle, Compliance Engineer
L. Smialek, Senior Plant Health Physicist (Interim Radiation Protection Manager)
W. Paulhardt, Radiation Safety Supervisor-Dosimetry
P. Jones, ALARA Supervisor
M. Rigsby, Supervisor-Radiation Technical Services
Hoooer and Associates
Philip Hasrouni l
David Hopper '
Transnuclear West
Usama Farradji, Project Manager
Walter Bak, Vice President, Engineering
N.B.C
A. Dromerick, Senior Project Manager, Calvert Cliffs, NRR
T. Hoeg, Reactor Engineer
E. Leeds, Chief, SFLS, SFPO, NMSS
E. McKenna, Project Manager, NRR
.
2
INSPECTION PROCEDURES USED
IP 61726: Surveillance Observations
IP 62707: Maintenance Observation
IP 71707: Plant Operations
IP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power Reactor
Facilities
IP 92901: Followup - Operations
IP 92902: Followup - Maintenance
IP 92903: Followup - Engineering
, IP 92904: Followup - Plant Support
IP 37001: 10 CFR 50.59 Safety Evaluation Program j
IP 60851:- Design Control of ISFSI Component 9
IP 83750: Occupational Exposure
ITEMS OPENED. CLOSED. AND DISCUSSED 1
i
Ooened
]i
50-317&318/98-01-01 VIO Failure to provide complete and accurate information in I
the ISFSI site specific license application
50 317&318/98 01-02 VIO Failure to identify an unreviewed safety question during
a 10 CFR72.48 evaluation of the DSC
50-317&318/98-01-03 IFl Review waste classification and curie content
]
determination methodology
50 317&318/98-01-04 IFl Review of structure and definition of audits of radwaste i
activities
Closed
50-317&318/96-06-01 VIO Safety system misalignments during saltwater and
service water system maintenance
50-317/97-008 LER Two reactor protective channels out of service during
test
50-317&318/97-04-04 URI Effectiveness of corrective actions for emergency
lighting deficiencies
50-317&318/96-09-01 VIO Defacto Modification of Auxiliary Feed Water Pump
Base
50-317&318/96-07-02 VIO Loss of Integrity of Radioactive Materials Shipment
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3 ,
Discussed
50-317&318/97-06-04 VIO Lack of Procedure for Survey of Laundry
50-317&318/96-04-02 URI Vcrification of Updated FSAR commitments
LIST OF ACRONYMS US1Q
ALARA As Low As Reasonably Achievable
ASME B&PV American Society of Mechanical Engineers, Boiler & Pressure Vessel
BTP Branch Technical Position
CCPRA Calvert Cliffs Probabilistic Risk Assessment
CDE Committed Dose Equivalent
CEA Control Element Assembly
CEDE Committed Effective Dose Equivalent
DAC Derived Air Concentration
DSC Dry Shielded Canister
EA' Enforcement Action
ECCS Emergency Core Cooling System
EDG Emergency Diesel Generator
eel Escalated Enforcement item
ESFAS Engtneered Safety Feature Actuation System
ET&S Engl.1eering & Technical Support
HP Health Physics
HPSI High Pressure Safety injection
HX heat Exchange
IFl Insrsection Follow-Up items
IPE Inuividual Plant Examination
IPEEE Individual Plant Examination of External Events
,
IR inspection Report
i ISFSI Independent Spent Fuel Storage Installation
LCO Limiting Condition for Operation
LER Licensee Event Report
MPF Materials Processing Facility
NUHOMS NUTECH Horizontal Modular Storage System
01 Operating Instructions
OSSRC Offsite Safety Review Committee
PDR Public Document Room
PE Principal Engineer
POSRC Plant Operation Safety Review Committee
i PRA Probabilistic Risk Assessment
l RP&C Radiation Protection & Chemistry
! RPM Radiation Protection Manager
RWP Radiation Work Permit
i SAR Safety Analysis Report
l SE Safety Evaluation
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.
4
SER Safety Evaluation Report I
STAR Stop-Think-Act-Review
SWP Special Work Permit
TLD Thermoluminescent Dosimeter I
TN West Transnuclear West
TS Technical Specifications
VDC Volts-Direct Current )
VEC1RA VECTRA Technologies Inc.
UFSAR Updated Safety Analysis Report
URI Unresolved item
USAR Updated Safety Analysis Review
USO Unreviewed Safety Question
VIO Violation
WRAP Worker Risk Assessment Process
WRNI Wide Range Nuclear Instrumentation
I
,
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ATTACHMENT 2
HOPPER AND ASSOCIATES ENGINEERS.
PROFESSIONAL ENGINEERING SUPPORT TO INDUSTRY
CCNPP NUHOMS - 24P
DSC EVALUATION
1
FEBRUARY 1998
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