ML20135E554
| ML20135E554 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 02/27/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20135E550 | List: |
| References | |
| 50-317-96-10, 50-318-96-10, NUDOCS 9703070136 | |
| Download: ML20135E554 (25) | |
See also: IR 05000317/1996010
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
License Nos. DPR 53/DPR 69
Report Nos. 50-317/96-10;50-318/96-10
Licensee: Baltimore Gas and Electric Company -
Post Office Box 1475
Baltimore, Maryland 21203
Facility: Calvert Cliffs Nuclear Power Plant, Units 1 and 2
Location: Lusby, Maryland
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Dates: November 31,1996 through January 11,1997 '
inspectors: J. Scott Stewart, Senior Resident inspector
H. Kirke Lathrop, Resident inspector
Fred L. Bower lil, Resident inspector i
Tim Kobetz, Senior Engineer, Spent Fuel Project Office, NRR i
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Approved by: Lawrence T. Doerflein, Chief i
Projects Branch 1
Division of Reactor Projects
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PDR ADOCK 05000317 i
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EXECUTIVE SUMMARY 4
Calvert Cliffs Nuclear Power Plant, Units 1 and 2
Inspection Report Nos. 50-317/96-10 and 50 318/96-10
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This integrated inspection report includes aspects of BGE operations, maintenance, I
l engineering, and plant support. The report covers a seven week period of resident ;
inspection and includes the results of an announced inspection by a spent fuel project i
specialists.
Plant Operations
e The inspectors identified that during fuel handling in the spent fuel pool, had a fuel .
handling event occurred, some of the radioactive material released may not have i
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passed through charcoal adsorbers prior to release to the environment. Also, the i
inspectors found that BGE activities during the fuel handling were deficient in that
pre-evolution briefings had not been conducted with control room personnel and a t
controlled copy of the fuel handling procedure was not in the control room.
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I * The inspectors found that operations personnel exhibited poor work practice and did l
l not have a questioning attitude during venting of a chemical and volume control
system filter. Control of the evolution using the safety tagging procedure was
ineffective and contributed to a plant auxiliary operator mispositioning a vent valve
l to an on-line purification ion exchanger resulting in a lowering of volume control
tank level.
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l Maintenance
e During fuel handling operations in the spent fuel pool, maintenance was conducted .
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in the auxiliary building that was poorly planned in that an unmonitored path
between the fuel handling area and the environment was intermittently created and
compensatory actions to ensure that radioactive material would be contained were
not prescribed.
e The inspectors concluded that emergency diesel generator realignment maintenance
l activities were very well planned and effectively implemented. Advance planning
i for the maintenance was extensive and included walkdowns of the job with a
l vendor technical representative, prefabrication of speciallifting and alignment tools, '
l dry runs on a spare EDG, a detailed risk assessment, and good coordinGion
between maintenance and engineering.
Enaineerina
e The inspectors determined that BGE engineers conducted a thorough and rigorous
examination of a piping defect and evaluated potential safety consequences of
continued operation. The technical content of the engineering evaluation was
excellent. The Plant Operating Safety Review Committee demonstrated a strong
safety perspective and questioning attitude in their review of the potential nuclear
safety consequences of the leak and the engineering assumptions used to justify
operability.
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- The inspectors deterrnined a BGE procedure did not provide adequate guidance to ;
ensure that a dry shielded canister would not be over pressurized during unloading i
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operations. BGE engineering did not develop supporting documentation to !
determine the required re-flood rate, and select and test equipment required for
canister re-flood operations. .
- Due to recent industry events and as a voluntary initiative, BGE developed a project
. plan to perform a review of the Updated Final Safety Analysis Report to assure that
the report accurately reflected the current plant design and operating practices.
- The inspectors found that BGE corrective actions to ensure that the as-built versus
as-designed configurations for electrical separation barriers were inadequate. The
specific weakness was the challenge to electrical separation resulting from damaged
or missing marinite separation barriers. The inspector also found that some design
documents did not reflect the as-built configurations.
Plant Suooort i
- During the conduct of operator rounds, the inspectors considered the actions of an
auxiliary plant operator to be a very good demonstration of sound ALARA and
radiation controls practices, j
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l TABLE OF CONTENTS
EX ECUTIV E SU M M A RY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii
TAB LE O F CO NTE NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv
! Summ ary of Plant Status . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
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l 1. O p e ratio n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 ;
j 01 Conduct of Oper ations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
01.1 General Comments (71707) . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
01.2 Spent Fuel Handling Operations ........................ 1
01.3 Valve Mispositioning in the Chemical and Volume Control ,
System.......................................... 4
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07 Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
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07.1 (Closed) LER 5 0-318/9 6004 . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 J
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11. M a in t e n a nc e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
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M1 Conduct of Maintenance .................................. 6 ,
l M1.1 Routine Maintenance Observations . . . . . . . . . . . . . . . . . . . . . . 6
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M1.2 2A Emergency Diesel Generator Realignment . . . . . . . . . . . . . . . 6
M1.3 Routine Surveillance Observations . . . . . . . . . . . . . . . . . . . . . . . 7
l M1.4 Reactor Trip Circuit Breakers .. ....................... 7
l 111. Engineering ................................................... 8
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El Conduct of Engineering . ................................. 8
E1.1 General Comments ................................. 8
E1.2 Degradation of the Safety injection Pump Recirculation Piping . . . 8
E3 Engineering Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . 9
, E3.1 Deficient Procedure for Unloading a NUHOMS Cask . . . . . . . . . . 9
l E7 Quality Assurance in Engineering Activities . . . . . . . . . . . . . . . . . . . . 11
l E7.1 Update Final Safety Analysis Report (UFSAR) Review Project .. 11
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E7.2 (Closed) Unresolved item 50-317&318/96-08-01 Cable
i Separation Issues ................................. 12
l E8 Miscellaneous Engineering issues (92902) . . . . . . . . . . . . . . . . . . . . . 15
l E8.1 (Closed) Unresolved item 50-317&318/93-25-01 . . . . . . . . . . . 15
l E8.2 (Closed) Unresolved item 50-317 and 318/96-04-01 Inoperable
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LPSI pump circuit breaker due to bent trip-paddle problem. . . . . 16
I V. Pl a n t S u p p o rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
R4 Staff Knowledge and Performance in Radiation Protection and
C h e mi st ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
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V. M an a geme nt M e eting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
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X1 Exit Me eting Sum m ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
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Table of Contents (cont'd) ,
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ATTACHMENT
Attachment 1 - Partial List of Persons Contacted
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Inspection Procedures Used !
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Items Opened, Closed, and Discussed
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List of Acronyms Used
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Report Details
Summarv of Plant Status
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Unit 1 remained at full power during the inspection period. l
Unit 2 operated between 95 and 100 percent power during the inspection period. (See I
01.1 General Comments)
1. Operations
01 Conduct of Operations ' l
01.1 General Comments (71707)
Overall plant operations were conducted with a proper focus on continued nuclear
safety. A deficiency in the cooling system for the Unit 2 main transformer required
that reactor power be reduced 20 megawatts electric at 50 degrees fahrenheit (*F)
ambient temperature and 20 additional megawatts for each additional 10 degrees of
outdoor temperature. When temperature dropped, power could be restored. These
power swings were frequently conducted during the inspection period without
complication. BGE planned to repair the transformer during the upcoming Unit 2
refueling outage. As a result, Unit 2 operated between 95 and 100 percent power
throughout the inspection period.
01.2 Spent Fuel Handlina Ooerations
a. Insoection Scope
During a plant walkdown, the inspectors observed that air from the fuel handling
area was flowing into the auxiliary building while spent fuel was being moved in the
spent fuel pool. The circumstances of the observation were reviewed by the
inspectors.
b. Observations and Findinas
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On January 10, during a routine plant walkdown, the inspectors observed that !
spent reactor fuel was being moved in the spent fuel pool to prepare for the
upcoming Unit 2 refueling outage. Simultaneously, the inspectors found that a door i
from the spent fuel pool area into the auxiliary building stairwell was ajar, with !
indications that air was flowing out of the spent fuel handling area into the auxiliary l
building through the doorway.
The inspector noted that the basis for Technical Specification 3/4.9.12, " Spent Fuel j
Pool Ventilation System," stated that "The limitations on the spent fuel pool
' Topical headings such as 01, M1, etc., are used in accordance with the NRC
standardized reactor inspection report outline found in MC 0610. Individual reports are not
expected to address all outline topics. ,
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ventilation system ensure that all radioactive material from an irradiated fuel l
assembly will be filtered through the HEPA filters and charcoal adsorber prior to
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discharge to the atmosphere." The auxiliary building ventilation system did not j
- . include charcoal adsorbers. The ventilation limitations were repeated in the Calvert }
( Cliffs UFSAR Section 9.8, " Spent Fuel Pool Ventilation" and Section 14, " Accident l
Analysis". The inspector also noted that the UFSAR, Section 9.8, stated that the i
, - spent fuel pool ventilation system was capable of maintaining a negative pressure j
with respect to ambient and surrounding areas of the auxiliary building. l
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The inspectors questioned operations department supervisors as to the adequacy of l
the fuel handling area ventilation during the spent fuel movement. Plant operators j
secured the fuel movements until the open door was repaired and shut. On j
> additional questioning from the inspector, BGE initiated an investigation and *
- identified that work was being conducted oa the auxiliary building ventilation l
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system and that two o.* the three supply fans for the system were out of service. l
Then, a BGE engineering review was conducted which idontified that because o' i.he !
maintenance, the auxiliary building ventilation system was out of balance and j
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) auxiliary building. Since fuel handling was in progress when the fuel handling l
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ventilation system was out of balance, on January 10, BGE made a report to the :
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NRC in accordance with 10 CFR 50.72.b.1.ii(b), for a condition outside the design !
basis of the plant.
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i The inspectors were later informed that maintenance was in progress to replace the I
11 and 12 auxiliary building supply fan discharge gravity dampers. The supply fans
, nad been stopped when work started on January 6. The work was completed on .
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i January 10. Fuel handling operations were conducted on January 8,9, and 10.
During the maintenance,' supply fan dampers were sequentially removed from the
system and replaced. The work order specified that any plant condition was
adequate to support the maintenance and the system engineer stated that fuel
handling operations were not considered when planning the work
The work conducted on the ventilation supply resulted in a breach of the ducting
when the supply dampers were removed. The breach provided a path to the
outside environment that was not monitored for radioactive release. The inspectors
were informed that during the fuel movements, fuel handling and auxiliary building
exhaust ventilation ran continually and air flow remained from the outside into the
building through the breach and back to the environment through mo.nitored flow
paths.
The inspectors also found that the fuel movements were conducted by two
contractor personnel using fuel handling procedure FH-340. Both contractors had
completed fuel handling qualifications administered by BGE. The fuel handling area
ventilation system exhaust was aligned for the fuel movement with a charcoal
adsorber in service. A prepared set of fuel movements were specified and
completion of the moves was documented on the appropriate form. The fuel
movements involved two procedures; FH-340, " Component Movement in the
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Auxiliary Building," and Operating instruction OI-25A, " Spent Fuel Handling i
Machine." '
Two procedure compliance discrepancies were identified by the inspectors. FH-340 ,
step 2.1.B, stated that the controlled copy of the procedure shall be maintained by '
the control room when core components were being moved in the spent fuel pool. *
The inspectors found that no copy of the procedure was in the control room and the
controlled copy was maintained by the fuel management group in the engineering
department.
FH-340, Attachment FH-340-1, " Spent Fuel Pool Component Moves," stated that
the operators and the control room supervisor will be briefed as part of the pre-
evolution brief. The briefing instruction specified that communications during fuel
moves, ventilation lineups, and actions to be taken if a radiation monitor alarms or if
a fuel handling incident occurred would be included in the briefing. The inspectors
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found that the briefing did not include any licensed or control room personnel, but !
instead was conducted by the fuel management group with the contractors. The
control room operators were aware that fuel handling was planned and had placed
charcoal adsorbers in service per Operating Instruction 22D and were informed by ;
the contractors when fuel handling had started and stopped. However, control
room personnel had not reviewed the fuel handling precautions and procedures and
had not briefed or prepared for actions in event of a fuel handling problem. No
evacuation routes for personnel in the auxiliary building had been planned or
discussed with applicable work groups. The work group assigned the auxiliary i
building ventilation work were working in an area only accessible through the spent J
fuel handling area and had not been informed that fuel moves were in progress or
informed of their responsibilities in event of a fuel handling incident.
10 CFR 50, Appendix B, Criterion V stated, " Activities affecting quality shall be
prescribed by documented instructions, procedures, or drawings, of a type
appropriate to the circumstances and shall be accomplished in accordance with
these instructions, procedures, and drawings". The failure of BGE to estab!ish
conditions that assure that had a fuel handling event occurred, all of the released
gases would be filtered through charcoal adsorbers and the failure to follow the
briefing and procedure control guidance in procedure FH-340, were in the
aggregate, a violation of NRC requirements. (VIO 50-317&318/96010-01) l
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When informed of the inspector findings, BGE management initiated a review of
spent fuel pool operations, including fuel handling evolutions and ventilation
adequacy in different operating modes. Also, a review of ventilation adequacy in
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other areas of the auxiliary building was initiated. BGE management also informed i
the inspectors that the management expectation was for procedure compliance in
plant operations and that this expectation was not met in the fuel handling
operations observed by the inspectors. ;
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c. Conclusions '
The inspectors found that BGE activities during fuel handling in the spent fuel pool
were deficient in that pre-evolution briefings had not been conducted with control
room personnel and a controlled copy of the fuel handling procudure was not in the ,
, control room. Also, due to a procedure inadequacy, had a fuel handling event !
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occurred, some of the radioactive material released may not have passed through
charcoal adsorbers prior to release to the environment.
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During fuel handling in the spent fuel pool, maintenance was conducted in the l
auxiliary building that was poorly planned in that an unmonitored path between the
fuel handling area and the environment was intermittently created and
compensatory actions to ensure that radioactive material would be contained were
not prescribed.
BGE management responded promptly to the inspector findings by declaring the fuel i
pool ventilation system out of service until the various modes of operation could be
l evaluated. BGE management also initiated a review of fuel nandling operations and !
auxiliary building ventilation.
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01.3 Valvo Micoositionina in the Chemical and Volume Control System '
a. Insoection Scope
The inspectors rev'ewed a valve mispositioning occurrence at Calvert Cliffs Unit 2.
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b. Obs3rvations and Findinas
On January 15, a radiation protection technician requested control room operators i
open a vent valve for the 22 purification filter in the chemical and volume control
system. Prior to the request, the filter had clogged, a work order had been
prepared, and a tagout had been issued to isolate the filter. Control room operators
told the inspector that the technician made the request so that a radiation protection l
survey could be completed.
An auxiliary building operator was instructed by control room personnel to open the
purification filter vent valve, 2-CVC-122. Instead, the operator opened 22-
purification ion exchanger vent valve,2-CVC-140. The operat.or did not read the
valve label resulting in the wrong valve being operated.
Since the ion exchanger was in service at the time, partially depressurized reactor
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coolant passed from the ion exchanger through the vent valve ari into the waste
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tank level dropped approximately one inch (60 gallons) when a control operator
suspected a problem and requested the auxiliary building operator to shut the vent
valve. On returning to the valve, the auxiliary building operator observed that the
wrong valve had been manipulated and informed control room personnel.
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in response, control room personnel requested radiation controls personnel to
conduct surveys of the waste gas header to determine if radiation level changes had
l occurred in the auxiliary building and to verify the extent of contamination of the
i header by reactor coolant. No unusual radiation levels were detected. The
l operating crew reported that no unusual radiation alarms had occurred during the
event and a review was conducted which verified that there had been no
measurable release of radioactive material through the plant vent. Subsequently, a
drain valve on the waste gas header was opened to drain the coolant from the
i normally dry waste gas header, but no liquid passed through the drain valve.
Further BGE investigation identified that the 60 gallons of coolant had entered the
11 miscellaneous waste receiver tank through an open vent line on the tank. The
l operations department determined the occurrence to be a significant event near I
l miss and an issue report was written. As followup action, BGE management j
( reviewed safe plant operations and the need to complete self-verification prior to l
l operating plant equipment with all operating personnel, including the operator who
l had mispositioned the ion exchanger vent valve.
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l The inspectors became aware of the mispositioning in discussions with control room
personnel during a control room walkdown. The inspectors reviewed the event and
found that a safety tagout and work package to support replacing the filter had
l been issued on January 15. Although not tagged, a note on the tagout stated that
l the vent valve would be operated by the work group after a concrete shield block
l was removed by plant mechanics to access the valve. Otherwise the valve was
I inaccessible.
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When the control room was contacted to open the vent valve, a safety tagging
technician was contacted to authorize opening the valve, and permission was given.
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Control of the shield plug was apparently not raised in this discussion and direction
was given to the auxiliary operator without mentioning that shield plug removal was
necessary. Control room personnel did not review either the safety tagout nor the
work package prior to giving direction to open the valve and no written procedure
l directing the venting was available in the control room. On questioning, BGE ,
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management stated that the evolution was controlled by the Calvert Cliffs safety 1
l tagging procedure.
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The inspectors were informed that the involved auxiliary building operator had not
l previously positioned valves in either the purification or the purification filter
! systems and was unsure of the proper valve location. The inspectors found that
control room personnel did not have a questioning attitude when the decision to
vent the filter was made because the removal of the shield plug to access the valve
i was not considered and no written procedure step was sought directing the action.
T,ie mechanical work group vented the filter on January 15 and completed the work
. on January 16.
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c. Conclusions
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The inspectors found that operations personnel exhibited poor work practice and did
not have a questioning attitude during a venting evolution. Control of the evolution
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using the safety tagging procedure was ineffective and contributed to a plant
operator mispositioning a vent valve to an on-line purification ion exchanger.
07 Quality Assurance in Operations
07.1 (Closed) LER 50-318/96004: Missed Surveillance Due to Less than Adequate
Technical Review of Surveillance Test Procedure. BGE identified that a Unit 2
surveillance test procedure for verification of containment closure did not verify a
closed containment as specified by Calvert Cliffs Technical Specification 3.9.4.b.
Specifically, a steam generator sample drain valve was not verified shut when the
steam generator was open to the containment atmosphere during core alterations.
The discrepancy was identified by BGE during a review of the test procedure. The
inspectors reviewed the LER and verified completion of the long term corrective
actions including a technical review of containment closure for fuel movement and i
that the applicable procedure had been appropriately revised. The issue was !
considered a Non-Cited Violation, consistent with Section Vll.B.1 of NUREG 1600, 1
NRC Enforcement Policy. The LER is closed,
ll. Maintenance
M1 Conduct of Maintenance
M 1.1 Routine Maintenance Observations
Using Inspection Procedure 62707, the inspectors observed the conduct of
maintenance and surveillance testing on systems and components important to
safety. The inspectors also reviewed selected maintenance activities to assure that
the work was performed safely and in accordance with proper procedures. The
inspectors noted that an appropriate level of supervisory attention was given to the
work depending on its priority and difficulty. Maintenance activities reviewed
included:
MO2199601746 21 Charging Pump Suction & Discharge Valve Replacement
MO2199604378 23 Saltwater Pump Volute Cleaning Due to Low Flow
MO2199304705 Replace 21 AFW Pump Turbine Stop Valve Position Switch
M01199603843 EQ Replacement of Solenoid Valve on 12 Component Cooling
Heat Exchanger
M1.2 2A Emeraency Diesel Generator Realianment
The inspectors reviewed and observed selected portions of scheduled corrective
maintenance conducted to relieve crankshaft strain on the 2A emergency diesel
generator (EDG) during plant operation. NRC inspection report 50-317 & 318/96-06
documented the inspectors' previous review of maintenance act;vities that identified
crankshaft strain of -0.00275 inches. At that time, BGE was attempting to meet an
acceptance criteria of +0.001 to -0.001 inches. In November 1996, the EDG
vendor provided BGE information that the crankshaft strain acceptance criterion had
been revised to 0.000 to + 0.001 inches.
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Relief of the strain was accomplished by shimming the generator to obtain an
improved alignment to the engine. The final crankshaft strain of -0.00025 inches
was determined to be acceptable by BGE and the manufacturer. However, BGE ;
would monitor for changes in the strain during periodic surveillances. Also, an issue
report has been entered into the corrective action system to further evaluate the
discrepancy between the as-left crankshaft strain and the November vendor letter.
The inspectors concluded that the 2A EDG realignment maintenance activities were . I
very well planned and effectively implemented. Advance planning for the
maintenance was extensive and included walkdowns of the job with a vendor '
technical representative, prefabrication of special lifting and alignment tools, dry = ,
runs on a spare EDG; a detailed risk assessment, and good coordination between '
maintenance and engineering personnel. Although the issue of the conflict between
the as-left strain and the November 1996 vendor letter required an engineering ,
review prior to returning the EDG to service, the issue was resolved and the engine j
was returned to service without challenging the allowed outage time. The *
maintenance work order was effectively implemented with strong support provided l
by vendor technical representatives and system engineering personnel. .!
M1.3 Routine Surveillance Observations -
The inspectors witnessed / reviewed selected surveillance tests to determine whether
approved procedures were in use, details were adequate, test instrumentation was ,
properly calibrated and used, technical specifications were satisfied, testing was
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performed by qualified personnel, and test results satisfied acceptance criteria or
were properly dispositioned. ;
The surveillance testing was performed safely and in accordance with proper l
procedures. The inspectors noted that an appropriate level of supervisory attention
was given to the testing depending on its sensitivity and difficulty. Surveillance
testing activities that were reviewed are listed below:
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STP M-200-2 Reactor Trip Circuit Breaker Functional Test i
S.7 v-70-2 Monthly Test of "A" Train Containment Cooling Units, lodine
Removal Units, and Penetration Room Exhaust Filter l
STP O-65B-2 21 Service Water Subsystem Valve Quarterly Operability Test :
I M1.4 Reactor Trio Circuit Breakers 2 ;
! The inspectors informed the licensee of a condition identified at other nuclear ,
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facilities that involved the testing of the undervoltage and shunt trip electricallogic
paths for the Reactor Trip Circuit Breakers (TCBs). Either condition would cause the l
TCBs to trip, however, the test that was historically performed did not test each of
the trip devices independently. The inspectors and BGE personnel reviewed j
Technical Specification 4.3.1.1.1, and applicable diagrams and surveillance test
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procedures and concluded that the same conditions did not exist at Calvert Cliffs.
The Calvert Cliffs technical specification did not have a specific surveillance
- . requirement (similar to the other nuclear facility) to verify the independent
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operability of the undervoltage and shunt trips. However, the testing performed by
BGE was complete. The reactor protective system matrix functional test performed !
quarterly verified the logic matrices and the matrix relays. This procedure included .
verification of the change of state of installed indicating lights provided in both the ?
shunt and undervoltage trip paths. The reactor TCB functional test was performed
monthly to independently verify the response time of both the undervoltage and
shunt trip devices and verify operat!~ of the TCBs. The inspectors concluded that :
the TCB testing performed by BGE was appropriate. ;
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E1 Conduct of Engineering (37550) .
E1.1 General Comments :
On December 27,1996, during the process of reviewing design specifications in
preparation for purchasing new Dry Shielded Canisters (DSCs) for the Independent
Spent Fuel Storage Installation (ISFSI), BGE identified conflicting information
concerning the weight of fuel assemblies at Calvert Cliffs. A 1992 fuels vendor
letter identified the bounding maximum weight as 1300 pounds versus a 1995 letter 1
that identified the bounding weight as 1327 pounds. ISFSI Technical Specification l
3/4.1.7 specified that the maximum assembly mass including control components i
shall not exceed 1300 pounds. BGE entered this discrepancy into their corrective
action process. Additionally, BGE cancelled two scheduled DSC loadings and
postponed all future loadings until the issue was resolved. BGE personnel informed
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the inspectors that the safety significance.of this issue was low due to the design
of the canisters. The safety significance was based on a review of the system;
however, BGE had not quantified the available margin for their dry fuel storage
activities.
E1.2 Dearadation of the Safety Inlection Pumo Recirculation Pioina
a. Insoection Scoce (93702)
The inspectors reviewed the issues involving the discovery of a through-wall leak in
the ASME Class 2 combined recirculation pipe for the Unit 1 emergency core ,
cooling system (ECCS) pumps. )
b. Observations and Findinos
On December 11, a maintenance worker noted moisture on grouting near a 4-inch
pipe in the Unit 1 component cooling water pump room. The stainless steel
schedule 10 pipe (4"-HC-23-1005) provided a recirculation flow path from the
safety injection and containment spray pumps back to the refueling water tank
(RWT) during testing and other times when the pumps were in operation but not )
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injecting into the reactor coolant system. The inspectors responded to the site after l
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being informed of the finding and observed BGE efforts to characterize and resolve
the issue.
BGE engineers determined that the moisture was coming from a section of the pipe
which could not be isolated from the RWT. Using several non-destructive
examination (NDE) techniques, BGE determined that the moisture resulted from a
small pinhole leak at a welded support joint, and that this condition had probably
existed since original construction. The leakage could not be readily quantified, but
the surface of the weld would appear moist about 15 minutes after being dried. ;
Additionally, a Code repair, as required by the plant's Technical Specifications (TS), j
could not be conducted when the ECCS pumps were required to be operable. The <
reactor would have to be shut down and cooled to below 200 F. I
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BGE engineers evaluated the leak and conducted a risk assessment. The results of l
the evaluation and potential corrective actions were presented to the plant i
operational safety review committee (POSRC) on December 14. The POSRC ,
concluded that risk to safe plant operation was minimal and no POSRC member had '
a safety concern. The conclusion was based, in part, on the engineering evaluation j
which stated that the indication was unlikely to propagate because of the very low )
stresses (compared to design allowable) on the pipe. A compensatory measure to
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evaluate the flaw for growth was specified to validate the engineering assumptions '
until the defect could be repaired. In a 5-4 vote, the POSRC recommended to the
plant general manager that an ASME Boiler and Pressure Vessel Code exemption be
requested from the NRC until p!.snt conditions were conducive for a Code repair.
The minority of POSRC voteis believed that continued plant operation with the i
existing defect was contrary to TS requirements. The plant general manager
accepted the majority POSRC recommendation and BGE submitted their exemption
request on December 19. The request was under review by the NRC when the
inspection period ended,
c. Conclusions )
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The inspectors determined that BGE engineers conducted a thorough and rigorous ,
examination of the piping integrity defect and evaluated potential safety
consequences of continued operation. The technical content of the engineering I
evaluation was excellent. The POSRC demonstrated a strong safety perspective
and questioning attitude in their review of the potential nuclear safety consequences
of the leak and engineering assumptions used to justify operability.
E3 Engineering Procedures and Documentation
E3.1 Deficient Procedure for Unloadino a NUHOMS Cask
a. Insoection Scone (60855)
A follow-up inspection was conducted to assess BGE corrective actions for two
NRC identified weaknesses in independent spent fuel storage activities.
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b. Observations and Findinas ;
in a previous inspection, the NRC identified two weaknesses regarding independent '
spent fuel storage activities. These issues were described in NRC Inspection Report
50-317&318/96-07. One weakness was that the procedure used to unload fuel
from a spent fuel cask did not control re-flooding the cask such that over- ,
pressurization of the cask would not occur. The second issue was that the
procedure did not contain a method to sample the cask for damaged fuel prior to
removing the dry shielded canister (DSC) shield plug. The inspector interviewed
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BGE staff and reviewed the following documents: !
- Calvert Cliffs Technical Procedure ISFSI-02, Revisions 1,2 and 3, ,
" Independent Spent Fuel Storage Installation (ISFSI) Unloading" .
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- Calvert Cliffs Independent Spent Fuel Storage installation, Materials License ,
No. SNM-2505, as amended July 21,1995, i
The inspector found that BGE made changes to ISFSI-02 to include a step to sample
the DSC atmosphere following removal of the outer cover plate. The change was l
appropriate and the inspectors had no further concerns in this area. !
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BGE also added a step, prior to re-flooding the cask to ensure the Nuclear I
Engineering Unit had completed a calculation to determine maximum flow rate of j
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water into the canister such that canister pressure remained below 10 psig. The
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l inspectors had two concerns with the procedure change. ISFSI-02 was updated to i
l add a step to " Ensure Nuclear Engineering Unit has completed calculations to
determine maximum flow rate of SFP water into DSC such that DSC pressure
remains below 10 PSIG". However, BGE did not have e procedure in place to
- support the nuclear engineering unit in performing the calculation and selecting ,
l appropriate equipment to re-flood the DSC.
Also, BGE had not performed a preoperational test of the proposed re-flood system
to ensure it would adequately control the re-flood rate of fuel pool water into the
DSC which is a safety related component. In addition, BGE had not performed a
bounding analysis to determine what flow rates would be required to ensure that
the cask would not be over-pressurized. Although the procedure was updated in an
attempt to address NRC concerns documented in inspection Report
50-317&318/96-07, BGE did not develop supporting documentation to determine
the required re-flood rate and select and test equipment required for DSC re-flood
operations. This failure to develop supporting documentation was a violation of
10 CFR 50, Appendix B, Criterion V, which required that activities affecting quality
shall be prescribed by documented procedures of a type appropriate to the
circumstances (VIO 50-317&318/97010-02). The inspector followup item related
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to previously identified NRC concerns was closed. (Closed
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IFl 50-317&318/96-07-01).
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The inspector also found that the pressure gauge used to monitor DSC pressure
during fill of the DSC was located down stream of the DSC and provided a
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nonconservative indication of the actual pressure of the cask. In addition, the
pressure gauge was specified by BGE to read 0 - 100 psi, minimum. Based on
industry practice, the inspector considered this scale to be too large to accurately
monitor a pressure of less than 10 psig as required by the BGE procedure. Industry
practices such as those specified in ASME Boiler and Pressure Vessel Code, Section
Ill, required that gages used in testing shall be graduated over a range not less than
1 1/2 times nor more than 4 times the test pressure.
The inspectors were also concerned that BGE had not adequately demonstrated re-
flooding of the cask during the original dry run of DSC activities which were
performed to address Condition 15 of License No. SNM-2505. Although the
condition did not specifically state that a demonstration of re-flooding the DSC
should be performed, the condition stated that the activities should not be limited to
only those listed. The inspectors determined that since re-flood of the cask was
required prior to retrieval of the fuelit would have been appropriate to perform a dry
run of the re-flood during the demonstration of Condition 15.g, " Removing the cask
lid and cutting open the DSC (length may be truncated) assuming fuel cladding
failure."
c. Conclusions
The inspectors found a significant weakness in the methodology used by BGE to
determine that independent spent fuel storage installation unloading could be
conducted safely. Specifically, BGE had not demonstrated the ability to re-flood the
DSC such that overpressurization would be prevented. Also, BGE had selected a
pressure gage that was nonconservative because the indicating scale did not
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conform to industry practice.
E7 Quality Assurance in Engineering Activities
E7.1 Update Final Safety Analvsis Report (UFSAR) Review Proiect
Due to recent industry events and as a voluntary initiative, BGE developed a project
plan with an overall goal to perform a review and ensure that the UFSAR accurately
reflects the current plant design and operating practices. This plan was intended to
be accomplished in at least two phases.
The first phase included review of approximately 50 selected UFSAR sections, using
the guidance in NRC Regulatory Guide, RG-1.70. " Contents of Final Safety Analysis
Reports (FSAR)". The review was to ensure that the UFSAR accurately and
adequately described the design and operation of the plant. BGE system
engineering, design engineering, and operations personnel performed these reviews.
BGE initially identified approximately 40 issues that were entered into the BGE
corrective action program for resolution. BGE also planned to complete a root cause
analysis to assess whether there were generic problems with the UFSAR change
processes or their implementation. Examples of the types of discrepancies included:
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minimum-flow isolation valves. One UFSAR section identified that these
valves automatically close on a recirculation actuation signal (RAS), whereas !
another portion identified that this automatic feature is normally locked out.
- Conflicting statements within the UFSAR concerning the 36 minute minimum
time to switch over to the recirculation mode. One UFSAR section identified i
that this minimum time was based on operating two HPSI pumps, in the i
injection phase, whereas another section indicated that 36 minutes was ,
based on the operation of three HPSI pumps.
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- The UFSAR stated that each of the twelve containment pressure transmitters
has an individual sensing point whereas, in the plant, there were three
transmitters common to each sensing point. :
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Also, in review of the spent fuel handling issue (See 01 F .he inspectors found that
UFSAR Section 9.8.2.3 " Spent Fuel Pool Ventilation," .ated that an air supply j
system consisting of two 50 percent capacity air handling units provides ventilation :
for the spent fuel pool area. However, the inspectors were informed by BGE that l
the supply fans had been removed from service and not operated for more than five '
years. Instead, air was supplied only from leakage into the area from adjoining
areas. The inspectors considered that the discrepancy could have been identified j
during the ongoing BGE UFSAR initiative. i
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BGE planned to develop a second phase that will expand the scope and depth of
the UFSAR reviews based on the findings of the initial reviews. BGE informed
the inspectors that they plan to make a submittal to NRC concerning scope and
schedule for completion of the project. No operability issues had been identified
during the reviews. Enforcement action regarding design issues identified
during the BGE review was Unresolved (URI 50-317&318/96010-03), pending
completion of the BGE initiative and NRC inspection of the completed review. The
unresolved item is consistent with the General Statement of Policy and Procedures
for NRC Enforcement Actions, NUREG 1600, as published in the Federal Register,
Volume 61, Number 203, Page 54464.
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E7.2 (Closed) Unresolved item 50-317&318/96-08-01 Cable Seoaration issues i
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a. inspection Scope
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The inspectors reviewed safety related cable separation issues associated with
URI 50-317&318/96008-01, including the status of electrical cable separation
barriers and related modifications performed in 1990; BGE's corrective actions for a
prior NRC violation; BGE actions to update the configuration control drawings for
separation barrier installation changes; and whether appropriate administrative
controls have been applied to the storage and retention of the applicable project
records.
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b. Findinas and Observations
NRC violation 50-317/89-27-05 involved discrepancies in the BGE control of
electrical cable separation barriers. NRC Inspection Report (IR) 96-08 identified an
apparent lack of ownership for the cable separation barriers and the associated
modifications performed to restore these barriers to licensing basis conditions.
Discussions with BGE personnel indicated that BGE had assigned engineers
responsibility for cable separation. System engineers were responsible for cable
separation for the individual systems including the need to conduct periodic walk
downs on portions of the barriers tmed on accessibility, maintenance history, and
potential for damage. A project ensmeer was assigned responsibility for closeout of
the 1990 barrier modification package (FCR 90-10).
The inspectors found that some of the original project records related to the
walkdown, engineering evaluation, and repair of separation barriers for the project
plan appeared to be quality records that were not stored in the records vault and
were not available for general use by engineering personnel. The inspectors
questioned whether the appropriate administrative controls had been applied to the
storage and retention of the applicable project records. Discussions with BGE
personnel identified that BGE would be reviewing these records as part of the
closecut of the FCR 90-10. BGE personnel indicated that copies of the inspection
records and engineering evaluations were filed as quality records with the ;
nonconformance reports (NCRs) and maintenance orders (MOs) generated during i
FCR 90-10. Therefore, filing the originals would duplicate existing records.
However, the inspectors questioned whether the BGE inspection reports for areas
found acceptable had been filed and maintained as quality records. This issue was ;
still under review by BGE and BGE personnel indicated that the index of the NCRs I
and MOs were not currently retained as quality records, but would probably be
added to the records system.
The inspectors also found that the 1990 modification (FCR 90-10) to address the
NRC violation for inadequate cable separation had not been closed since the effort 1
was completed in 1994. During further reviews with BGE personnel, the inspectors I
found that FCR 90-10 was stillin working status. This modification was among a
large number of modifications where the work had apparently been completed, but
j the modification had not been closed out. BGE personnel indicated that their quality I
l assurance department had identified that this was an issue requiring BGE l
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( management attention. One effort to address the backlog and manage the
! closecuts included the development of a relevant performance indicator. Review of
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the recently developed indicator revealed that there are approximately 375
i modifications that have had the work completed in the field, but the final closeout,
including updating the configuration control documentation, was not complete. On
questioning by the inspectors, BGE personnel could not identify the oldest
outstanding modification or the average age of the backlog of modifications
awaiting closecut.
BGE personnel identified that the drawings for Unit 2 had not been updated to
reflect the as-built conditions for changes made to the separation barriers by FCR
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90-10. BGE generated an issue report to adaress this issue. The inspectors
identified two instances where the as-built conditions and the drawings for the
Unit 1,45 foot switchgear room, did not agree. Discussions with BGE personnel
indicated that their reviews concluded that the Unit 1 drawings were updated during
the FCR 90-10 process. Additionally, marked-up drawings were developed for
Unit 2 during the FCR 90-10 process; however, the controlled drawings were not
updated and the marked-up drawings could not be located. As discussed below,
the inspectors concluded that BGE did not complete the corrective actions identified i
by their response to NRC violation 50-317/89-27-05. BGE told the inspectors of
plans to walkdown the Unit 2 drawings that were known to have not been updated.
These Unit 2 drawings included some Unit 1 and common areas. BGE planned to
determine the extent of the condition and specify corrective actions based on the
results of the walk downs and identified deficiencies. The inspector concluded that
the configuration control related to cable separation had been inadequate to ensure
that design documents reflected the as-built configurations. !
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The inspectors initially identified three examples where barriers did not meet the
electrical separation criteria in FSAR chapter 8.5, " Separation Criteria," and design I
document E-406, " Design and Construction Standards for Cable and Raceway".
During this inspection period, BGE and the inspectors identified eleven additional
cable separation related issues during system walkdowns. The deficiencies included
missing or cracked barrier material.
The inspectors reviewed BGE's response to NRC violation 50-317/89-27-05
documented in a March 9,1990 letter to the NRC, and reviewed the effectiveness
of the corrective actions. The inspectors concluded that two of these corrective
actions were not effectively implemented or were inadequate to preclude
recurrence. These issues were (1) failure to ensure that the design documents
reflected the as-built configurations; and (2) failure to ensure that the as-built versus
as-designed configurations continued to meet the criteria on a long-term basis.
The inspector also questioned whether BGE had established a clear understanding of l
the licensing basis and criteria for electrical separation. Specifically, the inspector
questioned the licensing basis and criteria for electrical separation of cables passing
through free air. For example, cables passing between cable trays, cables passing
from conduit to cable trays, or cables passing from trays and conduits to
penetrations. Currently, BGE design document E-406 showed that a separation j
barrier must be sealed at the penetration caused by a cable exiting a cable tray to ,
enter a conduit; however, E-406 did not require the protection and separation from l
redundant channels for this cable passing through free air. Discussions with BGE
personnel indicated that this issue had been identified and documented as an issue
in their correctiva action system. The inspectors found that the lic9nsing basis l
criteria for electrical separation relative to redundant cables passMg 1.hrough free air
was not clear.
As discussed above, the corrective action to ensure that the design documents
reflected the as-built configurations was found incomplete. BGE's corrective
actions to ensure that the as-built versus as-designed configurations continued to
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meet the criteria on a long-term basis were inadequate to preclude the challenge to
electrical separation resulting from damaged or missing marinite separation barriers.
The incomplete corrective action and the inadequacy of the corrective actions to
preclude recurrence were violations of 10 CFR 50, Appendix B, Criterion 16,
" Corrective Actions" (VIO 50-317&318/96010-04). Unresolved item (URI 50-
317&318/96-08-01) is closed.
c. Conclusions ,
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The inspectors concluded that the BGE corrective actions to ensure that the as-built )
versus as-designed configurations for electrical separation barriers were inadequate. i
The specific weakness was the challenge to electrical separation resulting from l
damaged or missing marinite separation barriers. The inspector also found that the j
design documents did not reflect the as-built configurations.
The inspectors concluded that following issues present potential challenges to
maintaining the licensing and design basis of electrical separation: (1) BGE's self-
identified backlog of modifications awaiting closeout; (2) the licensing basis criteria
for electrical separation relative to redundant cables passing through free air was
not clear; and (3) BGE's ongoing review to determine which cable separation barrier
modification project records should be filed and maintained as quality records.
E8 Miscellaneous Engineering Issues (92902)
E8.1 (Closed) Unresolved item 50-317&318/93-25-01: failure to promptly perform a
reportability evaluation. On August 5,1992, BGE determined that a fire in the
Unit 1 cable chase could potentially cause a loss of off-site power to both 4160 V
emergency busses, resulting in the loss of both trains of control room ventilation. A
BGE engineer wrote an issue report documenting the finding and provided the report
for supervisory review. Neither the supervisor or the initiator considered the issue
NRC reportable. The issue report was then reviewed by the issues assessment unit
which included members from both operations and nuclear regulatory matters
(NRM-licensing), with no reportability concern. However, the inspector determined
that as was BGE practice at the time, an NRM engineer had independently reviewed
the issue for reportability and had concluded that no report was required. This
conclusion was not formally documented until a more rigorous analysis was
performed several weeks later, with the same conclusion. The inspector reviewed
the formal analysis and concluded that BGE's evaluation and conclusions were
reasonable, and that the compensatory actions taken were appropriate. The
inspector concluded that BGE had reviewed and determined reportability of the
issue in accordance with BGE practices. However, documentation of the initial
review was weak. Since that time, BGE has upgraded their reportability process.
This item is closed.
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E8.2 (Closed) Unresolved item 50-317 and 318/96-04-01 Inocerable LPSI oumo circuit
breaker due to bent trio-caddle oroblem.
The item involved the BGE corrective actions in response to a circuit breaker failure
on June 14,1996. The failure was due to a bent trip-paddle in the General Electric
4160 Volt Magne-Blast circuit breaker. There were two issues in this unresolved
item. The first issue pertained to the bending of circuit breaker linkages by the
technicians during the preventive maintenance inspection. This issue was closed by
Section E6.4 of Inspection Report 50-317&318/96-07.
The second issue involved the BGE root cause analysis (RCA) for the breaker failure
and bent trip-paddle. The issue was updated in Section E1.1, part b, of NRC
Inspection Report (IR) 50-317 & 318/96-07.
The inspectors reviewed the completed root cause analysis. The analysis [
considered a General Electric evaluation that included information concerning testing
of the failed breaker. BGE concluded that the root cause for the bent trip paddle
was a weak trip shaft reset spring that would allow the trip-paddle to contact the
breaker frame with excessive force. General Electric concluded that the weak
spring was probably an isolated incident due manufacturing defect or installation
damage. BGE believed that it was also possible that the springs relaxed due to i
age-related degradation. In accordance with 10 CFR Part 21, BGE submitted a i
notification of these problems to the NRC by letter dated January 15,1997. The
long term corrective actions were to replace the springs during scheduled breaker
overhauls or to replace the 4160 volt breakers. The actions were to be completed
in the next two or three years. BGE expected that the ongoing modification of the
trip paddle would preclude further breaker failures. The inspectors concluded that i
BGE's review, analysis, and corrective actions for these issues were extensive and l
appropriate. The item is closed.
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IV. Plant Support
R4 Staff Knowledge and Performance in Radiation Protection and Chemistry
On January 3, the inspector observed are auxiliary building operator complete
routine log taking and operations duties. During the rounds, the inspector observed l
that the operator routinely used good radiological work practices that included
verification of radiation and contamination levels using radiological contols postings
prior to entry into radiation areas. Also, the operator verified the eC tence of low
radiation leveis in areas subject to high radiation by both checking local radiation
monitor readings and by completing spot radiation level checks using an alarmed
dosimeter. The operator also contacted radiological controls personnel prior to each
entry into a potentially contaminated area. The inspector considered the actions of
the operator to be a very good demonstration of sound ALARA and radiation
controls practices.
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V. Moneaement Meetinas
X1 Exit Meeting Summary :;
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During this inspection, periodic meetings were held with station management to T!
[ discuss inspection observations and findings. On January 27,1997, an exit ,
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meeting was held to summarize the conclusions of the inspecticn. BGF '
management in attendance acknowledged the findings presented. "
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ATTACHMENT 1
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PARTIAL LIST OF PERSONS CONTACTED
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P. Katz, Plant General Manager i
K. Cellers, Superintendent, Nuclear Maintenance i
K. Neitmann, Superintendent, Nuclear Operations *
P. Chabot, Manager, Nuclear Engineering
T. Camilleri, Director, Nuclear Regulatory Matters
8. Watson, General Supervisor, Radiation Safety
C. Earls, General Supervisor, Chemistry I'
T. Sydnor, General Supervisor, Plant Engineering
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INSPECTION PROCEDURES USED f
IP 62707: Maintenance Observation
IP 71707: Plant Operations
IP 93702: Prompt Onsite Response to Events at Operating Power Reactors
IP 61726: Surveillance Observations
IP 37550: Engineering
IP 37551: Onsite Engineering
IP 71750: Plant Support Activities
IP 83750: Occupational Exposure
IP 92700: Onsite Follow-up of Written Reports of Non-routine Events at Power
Reactor Facilities
IP 92002: Follow up - Engineering i
IP 82701: Operational Status of the Emergency Preparedness Program I
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Attachment 1 2
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
50-317&318\96010 01 VIO Failure of BGE to establish procedures that assure that
in a fuel handling event, released gases would be
filtered through charcoal adsorbers and the failure to
follow the briefing and procedure control procedures
50-317&318/97010-02 VIO Failure to develop documentation to support dry fuel
storage cask unloading
50-317&318/96010-03 URI Old design issues identified during the BGE UFSAR
review
50-317&318/96010-04 VIO Incomplete corrective actions for electrical separation
barriers
Closed
50-318/96004 LER Missed Surveillance Due to Less than Adequate Review
of Surveillance Test Procedure
50-317&318/93025-01 URI Failure to Promptly Perform Reportability Evaluation
50-317&318/96007-01 IFl Engineering for IFSF1-02, Cask Unloading
50-317&318/96008-01 URI Cable Separation issues
50-317&318/96004-01 URI Inoperable circuit breaker due to bent trip-paddle i
LIST OF ACRONYMS USED
ALARA As Low As Reasonably Achievable !
BGE Baltimore Gas & Electric
CFR Code of Federal Regulations l
ECCS Emergency Core Cooling System
EDG Emergency Diesel Generator
DSC Dry Shielded Canister
FCR Facility Change Request
HPSI High Pressure Safety injection
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IFl Inspector Followup Item
IR Inspection Report or issue Report (non-conformances)
ISFSI Independent Spent Fuel Storage Installation
LER Licensee Event Report
LPSI Low Pressure Safety injection
MO Maintenance Order
NCR Nonconformance Report or issue Report
l NDE Non-destructive Examination
NRC Nuclear Regulatory Commission
NRM BGE Nuclear Regulatory Matters Group
NUHOMS Nutech Horizontal Modular Storage
PDR Public Document Room 1
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Attachment 1 3 i
PSIG Pounds per Square Inch-Gauge
RAS Recirculation Actuation Signal :
RCA Root Cause Analysis "
RWT Refueling Water Tank
SFP Spent Fuel Pool :
TCB Trip Circuit Breakers '
TS Technical Specification l
! UFSAR Updated Final Safety Analysis Report i
URI Unresolved item !
VIO Violation ;
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