ML20155D457
| ML20155D457 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 10/28/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20155D387 | List: |
| References | |
| 50-277-98-08, 50-277-98-8, 50-278-98-08, 50-278-98-8, NUDOCS 9811030223 | |
| Download: ML20155D457 (40) | |
See also: IR 05000277/1998008
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- ' U.'S, NUCLEAR REGULATORY COMMISSION
REGION I
License Nos. DPR-44
Report Nos. 98-08
98-08
Docket Nos. 50-277
50-278
. Licensee: PECO Energy Company
Correspondence Control Desk .
P.O. Box 195
Wayne, PA 19087-0195
Facility: Peach Bottom Atomic Power Station Units 2 and 3
Inspection Period: August 11,1998 through September 21,1998
Inspectors: A. McMurtray, Senior Resident inspector
M. Buckley, Resident inspector
B. Welling, Resident Inspector
B. Maier, Senior Reactor Engineer
S. Dennis, Operations Engineer
R. Nimitz, Senior Radiation Specialist
J. Carrasco, Reactor Engineer
L. Peluso, Radiation Specialist
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Approved by: Clifford J. Anderson, Chief
Projects Branch 4
Division of Reactor Projects
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9811030223 981028
PDR i
O ADOCK 05000277 '
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EXECUTIVE SUMMARY
Peach Bottom Atomic Power Station
NRC Inspection Report 50-277/98-08,50-278/98-08
This inspection report included aspects of licensee operations; surveillances and
rnaintenance; engineering and technical support; and plant support areas.
Operations:
> * The operators in the control room demonstrated very good communication
practices in their extensive use of three part communications. The operators
also demonstrated very good questioning attitudes in their pursuit of the scope of
a breaker problem and their review of procedures.
Peer checking and self checking were usually employed effectively. One error '
was noted in which the improper unit's procedure was initially used to substitute
computer variables for heat balance calculations but later corrected. Logs I
generally were kept accurately, but an erroneous plant status entry went
undetected through a shift turnover, indicating a cursory review of that entry.
(Section 01.1)
- Two Unit 3 reactor water cleanup (RWCU) system events occurred during this
inspection due to poor system configuration control. These events resulted in an
entry into emergency operating procedures due to a steam leak on the non-
regenerative heat exchanger and an automatic engineered safety feature (ESF)
isolation. The causes were less than adequate turnovers between senior reactor
operators and non-licensed operators, incomplete post-maintenance testing
instructions, and an inadequate RWCU startup procedure.
Station personnel failed to properly maintain the RWCU startup procedure, l
resulting in a violation of Technical Specification 5.4.1, " Procedures." Although
station personnel had previously developed some initiatives to reduce plant
configuration control problems, they had not made sufficient progress
implementing them to preclude these events. (Section 02.1)
- Operators did not verify that a torus-to-drywell vacuum breaker was closed
within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> of the discovery of an unreliable indication, as required by
technical specifications. This event was caused by the failure to adhere to
equipment operator rounds and log review practices by operations personnel.
This non-repetitive, licensee-identified, and corrected violation is being treated as
a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement
Policy. (Section 02.2)
- On August 22,1998, during performance of the Unit 3 turbine building rounds
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an equipment operator inadvertently shutdown the 3C drywell chiller. Since the
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Executive Summary (cont'd)
chiller was quickly restarted, the temperature and pressure increases in the
drywell were small and posed a small safety risk to the plant. l
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An engineering evaluation for a similar event that occurred on March 25,1997,
was not effective to preclude the August 22,1998 event. (Section O2.3) I
a On August 21,1998, unit 3 operators commenced a down power maneuver due
, to loss of cooling to the main transformer. The reduced load prevented a loss of
the main transformer and plant transient when the deluge system activated.
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The initial response by the operators for the loss of cooling to the mam
transformer and subsequent deluge activation was good. Due to the power I
reduction, the flow in the recirculation system loops became mismatched in l
excess of the Technical Specification limit. This cc,ndition was identified and
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corrected within the time allowable by Technical Specifications. Failure of ,
operations personnel to fully understand the effect of xenon on recirculation flow l
and closely monitor recirculation flow contributed to this involuntary entry into a !
Technical Specification Action and Limiting Condition for Operation. (Section 1
04.1)
- The Senior Reactor Operator Limited to Fuel Handling (LSRO) program was good
overall. The LSRO program guidelines and examinations were comprehensive
and well maintained by the program coordinator and LSRO license maintenance
was well documented. The areas of exam security, remediation, operator
feedback, and medical records were acceptable. (Section 05.1) ;
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Maintenance: i
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e Control room deficiencies were controlled and adequately prioritized so that !
critical Main Control room deficiencies were corrected in a timely manner.
However, some weaknesses, of minor safety significance, were noted with the {
clarity and implementation of the requirements in OM P-10.3, Revision 3, j
" Equipment Status List / Tagging of Deficiencies." (Section M2.1) j
e Weaknesses in maintenance planning and work practices led to a significant
water leak on the station fire main on August 23,1998. Water from the leak
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entered the safety related emergency service water /high pressure service water l
pump house via underground electrical conduits and degraded penetration seals.
The engineering evaluation, that the penetration sealleakage was within design
assumptions for a design basis flooding event, and pump operability was not
affected, was adequate. (Section M4.1)
Enaineerina:
e Construction activities on the east retaining wall of the independent spent fuel
storage installation were acceptable. Engineering personnel resolved
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Executive Summary (cont'd)
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construction deficiencies regarding as-built keyways and soil compaction in an
- effective manner, thus the ana / zed as-found condition of the east wall was
acceptable. The concrete mix delivery, testing, and pouring activities for the
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east retalning wall were acceptable. (Section E1.1)
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o Engineering personnel took prompt and effective corrective actions following
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their identification of the potential to bypass the pressure suppression function ,
of the torus during simultaneous purging of the torus and drywell as a result _
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of postulated failures, in accordance with the NRC Enforcement Policy,
Section Vll.B.3, Violations involving Old Design issues, the NRC is exercising
- enforcement discretion and not citing this violation. (Section E2.1) ,
Plant Sucoort: 1
o PECO implemented, an effective radioactive waste processing, handling, storage,
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and radioactive material transportation program. Wastes were properly classified '
and packaged. (Section R1.1)
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, e Overall, radioactive waste and material processing and storage areas were
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! properly posted and controlled, and exhibited very good material condition. I
(Section R2)
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e Individuals responsible for radioactive waste processing activities exhibited a j
good knowledge level of reculatory requirements'and program procedures.
(Section R4)
.e- PECO provided generally good training of personnelinvolved in radioactive waste
activities. However, one individual had not received the a-priori specified training
for mechanics involved with radioactive weste activities and there was no
defined training program for new radwaste personnel brought into the radwaste
group and involved in radwaste shipping activities. PECO took action on these
matters. (Section R5)
e PECO implemented an appropriately staffed and defined organization for
radioactive waste processing, handling storage, and shipping. -(Section R6)
e PECO performed audits of appropriate depth and scope of radwaste processing,
handling, storage, and transportation activities, including training and
qualification of personnel. Corrective actions were initiated for identified
concerns. (Section R7) i
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e PECO provided generally good radiological controls oversight of incoming fuel
shipments. However, a violation of radiation protection procedures associated !
. with source checking of an alpha contamination counting instrument was
identified by the NRC and was promptly corrected by PECO. (Section R8.3) l
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TABLE OF CONTENTS
EX EC UTIV E S U M M A RY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii
TA BL E O F CO N T ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v
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Summary of Plant Status ............................................1
1. Operations .....................................................1
- 01 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
01.1 Sustained Control Room Observation . . . . . . . . . . . . . . . . . . . . . 1
02 Operational Status of Facilities and Equipment ................... 3
O2.1 Reactor Water Cleanup System Configuration Control Events
(Unit 3) and (Closed) LER 50-278/3-98-004 . . . . . . . . ....... 3
02.2 Torus /Drywell Vacuum Breaker Loss of Seated Indication (Unit 2) . 6 '
02.3 Inadvertent Shutdown of the 3C Drywell Chiller ............. 7
03- Operations Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . 9
03.1 Review of Normal Plant Startup Procedure . . . . . . . . . . . . . . . . . 9 !
03.2 Unexpected Start of the Motor Driven Fire Pump During Testing . . 9
04 Operator Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . . . 10
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04.1 Operator Perforrnance During Loss of Cooling to the 3C Main
Transformer .....................................10
05 Operator Training and Qualification ..........................12
05.1 Limited Senior Reactor Operator (LSRO) Requalification Program . 12
11. M a in t e n a n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 14
M1.1 General Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
M2 Maintenance and Material Condition of Facilities and Equipment . . . . . . 14
M2.1 Main Control Room Deficiencies .......................14
M3 Maintenance Procedures and Documentation ...................15
M3.1 Fix It Now Team Planning and Documentation . . . . . . . . . . . . . . 15
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M4 - Maintenance Staff Knowledge and Performance . . . . . . . . . . . . . . . . . 16
M4.1 Fire M ain Le a k . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 '
111. E ng i n e e ri ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
,- El Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
E1.1 Installation of the ISFSI East Retaining Wall ............... 17
E2 Engineering Support of Facilities and Equipment .......... ...... 19
E2.1 (Closed) URI 50-277(278)/97-06-03and (Closed) LER 2-97-007
Potential for Bypass of Pressure Suppression Pool . . . . . . . . . . . 19
E2.2 Access and Alarm Failures to Protected Area and Vital Areas
Doors Due to Security Multiplexer Failure . . . . . . . . . . . . . . . . . 20
y E2.3 Core Spray, Residual Heat Removal and High Pressure Service Water
Motor Operated Valve Thermal Overload Wire Discrepancies . . . 21
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Table of Contents (cont'd)
IV. Plant Support ................................................23
R1 Radiation Protection and Chemistry Controls (RP&C) . . . . . . . . . . . . . . 23
R1.1 Radioactive Waste Processing, Handling, Storage, and Shipping . 23
R2 Status of RP&C Facilities and Equipment ......................24
R3 RP&C Procedures and Documentation ........................25
R4 Staff Knowledge and Performance in RP&C ....................25
R5 Staff Training and Qualification in RP&C . . . . . . . . . . . . . . . . . . . . . . 26
R6 RP&C Organization and Administration . . . . . . . . . . . . . . . . . . . . . . . . 27
R7 Quality Assurance in RP&C Activities . . . . . . . . . . . . . . . . . . . . . . . . . 27
R8 Miscellaneous RP&C /sctivities .............................28
R8.1 (Closed) Violation (VIO) 50-277(278)/97-03-02 Failure to Assure
that the Turbine Building Atmosphere was Processed Through the
Turbine Building Gaseous Waste Treatment System . . . . . . . . . . 28 ,
R8.2 (Closed) VIO 50-277(278)/97-04-03 Violation of Locked High
Radiation Area Key Control . . . . . . . . . . . . . . . . . . . . . . . . . . . 28
R8.3 Inspection of incoming Fuel Shipments . . . . . . . . . . . . . . . . . . . 28
R8.4 Security Oversight of Radwaste Activities ................ 30
V. Management Meetings ..........................................30
X1 Exit Meeting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30
X2 Review of Updated Final Safety Analysis Report (UFSAR) Commitments . 31
ATTACHMENTS
Attachment 1 - List of Acronyms Used
< - Inspection Procedures Used
-Items Opened, Closed, and Discussed
Attachment 2 - Maintenance Observations
- Surveillance Observations
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Report Details
Summary of Plant Status
PECO operated both units safely over the period of this report.
Unit 2 began this inspection period at 73% power, in end-of-cycle coastdown. Unit 2
power was at 60% at the end of the inspection period.
Unit 3 began this inspection period at 100% power. On August 14, unit load was reduced
to 84% due to a loss of service water to a main generator hydrogen cooler. Unit load was
reduced to 67% on August 21 due to degraded cooling of the 3C main transformer. Unit
power remained at 100% for the rest of the period.
1. Operations
01 Conduct of Operations
01.1 Sustained Control Room Observation
a. Insoection Scope (71715)
The inspectors conducted augmented observations of control room and other in-
plant activities from September 7 through September 11,1998. Some of the
activities the inspectors observed included:
- Diesel generator surveillance run for operability determination
- Investigation documentation of a problem associated with racking out a
4 kilovolt diesel output breaker
e Biocide injection in service water systems
- Pre-evolution brief of 3B recirculation pump scoop tube lockup and clamping
for motor refurbishment
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- Several position turnovers and beginning of shift briefings _
- Generation of troubleshooting, repair and test procedures for control room !
indication circuit j
- Performance of special test procedure for hydrogen injection l
' Topical headings such as 01, M8, etc., are used in accordance with the NRC standardized reactor inspection report
outline. Individual reports are not expected to address all outline topics.
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b. Observations and Findinas
The operators conducted their activities in acccedance with the operations manual
procedures. They used three part communications routinely in their exchange of
information. Annunciator response was prompt, and alarm response procedures
were consulted for any annunciator alarms not anticipated. Anticipated alarms were
announced to the crew by the cognizant operator. !
Peer checking was used where called for by the operations manual. Peer checking
was used for the diesel generator surveillance run and the recirculation pump scoop
tube lockup. Operators applying blocking tags on components performed self
checking to ensure that the correct components were tagged. The control room ,
supervisors showed appropriate oversight of infrequent tasks and test procedures, i
One exception to the otherwise excellent quality verification of control room j
activities was in the use of the wrong unit's routine test procedure for inserting a i
substitute value for recirculation loop flow rate in the plant computer heat balance I
equation. The inspector noted that a Unit 2 procedure initially was used on Unit 3 I
because it was printed on white paper, the color used for common unit procedures.
The operators quickly located the correct procedure prior to the actual performance
of the task referenced in the procedure, and the shift manager counseled the control
room team on the need to verify procedural accuracy when performing tasks. l
Control room logs were kept current. Technical specification action logs were
accurate. Entries were made in this log for potential as well as actual technical
specification action entries dealing with equipment unava'. lability. Two minor
exceptions were the insertion of erroneous unit status data into the unified
computer log at two shift turnovers. The inspectors noted these discrepancies, and
also noted that the first log error had not been noted by the control room team in
the log review conducted at the most recent shift turnover.
The operations teams, as led by the shift managers and their supervisory staffs, all
showed an excellent questioning attitude to abnormalindications and occurrences.
When a loose cable cover damaged a terminal strip during a breaker change out, the
operations team investigated several similar breakers for the same condition. One
control room supervisor noted a step in a surveillance procedure that was not
signed off even though there was no requirement that he conduct such a complete
review. He initiated action to correct the omission.
c. Conclusions
The operators in the contrni room demonstrated very good communication practices
in their extensive use of three part communications. The operators also
demonstrated very good questioning attitudes in their pursuit of the scope of a
breaker problem and their review of procedures.
Peer checking and self checking were usually employed effectively. One error was
noted in which the improper unit's procedure was initially used to substitute
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computer variables for heat balance calculations but later corrected. Logs generally
were kept accurately, but an erroneous plant status entry went undetected through
a shift turnover, indicating a cursory review of that entry.
02 Operational Status of Facilities and Equipment
O2.1 ' Reactor Water Cleanuo System Confiouration Control Events (Unit 3) and (Closedl
LER 50-278/3-98-004
a. Insoection Scooe (71707)
The inspectors reviewed two reactor water cleanup (RWCU) system configuration l
control events that resulted in an unplanned entry into emergency operating
procedures and an automatic engineered safety feature (ESP) isolation.
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b. Observations and Findinas
Non-Recenerative Heat Exchanaer Vent Valves left Op_en
On August 19,1998, while restoring the RWCU system to service following
maintenance, operators received a reactor building area temperature alarm in the 'B'
RWCU non-regenerative heat exchanger room. As required by procedure, operators ;
promptly entered emergency operating procedures for secondary containment '
control. An equipment operator responding to the alarm heard a steam leak from I
the 'B' non-regenerative heat exchanger room. The sound diminished when the
RWCU inboard and outboard isolction valves were shut.
Operators entered the 'B' non-regenerative heat exchanger room and found the heat
exchanger vent valves partially open, inctead of closed, as required. Upon further !
investigation, operations personnel identified that these valves were left out of
position due to poor configuration control of the system while preparing for
maintenance activities. 4
Operations personnel investigated this issue and determined that there were two
primary causes for these valves left out of position while preparing for maintenance
activities:
- Shift turnover information regarding the RWCU system was less than
adequate. The turnovers between senior reactor operators and between non-
licensed equipment operators did not address the detailed status of the
system.
- Shift supervision made incorrect assumptions with regard to the affected
trains when continuing with a RWCU system procedure. -.
The poor turnovers occurred during the performance of a cooldown and
depressurization procedure, which prepared the system for mahtenance. When
preparing the system for maintenance on August 16,1998, the operations shift
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partially opened the vent valves for both the 'A' and 'B' non-regenerative heat
exchangers based on the work control supervisor's interpretation of the intent l
of the depressurization and cooldown procedure. The relieving operations shift
crew continued with the procedure to prepare the system for maintenance. The
relieving crew believed, based on a precaution step in the depressurization and
cooldown procedure, that only the 'A' non-regenerative heat exchanger vent valves
had been opened in earlier steps. As a result, when the cooldown and
depressurization procedure directed the vent valves to be closed, the operators or:/ .
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closed the 'A' non-regenerative heat exchanger vent valves and did not close the l
, 'B' non-regenerative heat exchanger vent valves. The 'B' non-regenerative heat
exchanger vent valves remained partially cpen, when they should have been closed,
until the valves were found leaking steam, while returning the system to service, on :
August 19,1998. The system restoration procedure, SO 12.1.A 3, Revision 19, 1
"RWCU System Startup for Normal Operation or Reactor Vessel Level Control," did
not contain instructions to verify that the non-regenerative heat exchanger vent
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valves were closed during system restoration. '
The review by the inspectors of this event revealed the following.
The work control supervisors stated that their turnover was cursory with !
respect to the RWCU system configuration. They indicated that they relied ,
on the equipnient operators using the depressurization and cooldown l
procedure to conduct a detailed turnover.
- The equipment operators' turnover effectively communicated the steps of the
depressurization and cooldown procedure that were completed. However, ,
the tumover did not address the train (s) of the system that were affected by I
these steps. l
- The depressurization and cooldown procedure was infrequently performed.
One work control supervisor stated that this was the first time that he had
used the procedure.
- The system restoration procedure did not identify the mispositioned valves
before the event occurred.
The inspectors determined that this event was an important operations performance
issue. . This system configuration control event occurred despite ongoing operations
focus and initiatives on configuration control and system restoration. This event
was similar to others discussed in NRC Inspection Reports 50-277(278)/98-01,
98 02, and 98-06,in which shift supervision made improper assumptions regarding
system configuration or operation.
.Joadeouate RWCU Restoration Followina Post-Maintenance Testina
On August 20,1998, an automatic isolation of the RWCU system occurred due to a
high flow condition. Operators were in the process of returning the system to
service and wers opening the inlet valve to the 'B' RWCU demineralizer when an
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inrush of water into the 'B' demineralizer was heard. Following the isolation,
operators took appropriate actions to verify the isolation was complete and to check
the integrity of the system. Since this was an automatic engineered safety feature
(ESF) actuation, operations supervision made a four hour notification to the NRC per
The cause of the event was an incorrect system lineup following post-maintenance
testing, causing the 'B' demineralizer not to be properly filled and pressurized. After
investigation, operations personnel found that the 'B' demineralizer inlet valve had
been shut for post-maintenance testing and was not returned to the open position.
They noted that a maintenance operator had shut both the demineralizer inlet and
outlet valves in order to satisfy an interlock and perform post-maintenance testing
(PMT) on solenoid valve SV-3-36B-030B,'B' demineralizer plenum vent valve. The
inlet valve should have been opened following PMT to provide the appropriate
configuration for restoration of the system.
The inspectors noNd that the PMT instructions in the work order did not address
positioning of the demineralizer inlet and outlet valves. Also, the documentation of
the completed PMT did not discuss any operation of the inlet and outlet valves.
The inspectors reviewed the procedure used for restoration of the RWCU system,
SO 12.1.A-3. The procedure stated that use of the RWCU system check-off list
was optional, "as directed by shift management." In this instance, shift
management determined that completion or partial completion of a check-off list
was not necessary. The inspectors also noted that none of the steps in this
procedure verified that the 'B' demineralizer inlet valve was in the proper position j
for system restoration. The inspectors concluded that SO 12.1.A 3 did not provide ;
adequate instructions for verification of the 'B' domineralizer inlet valve position. !
Operations adequately addressed short-term corrective actions for this issue.
The inspectors noted that SO 12.1.A-3 was revised following this event, on
September 12,1998, to include verification of the demineralizer inlet valve
positions. The inspectors performed an in-plant review of Licensee Event Report
(LER) 50-278/3-98-004,and identified no additional concerns,
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NRC Inspection Reports 50-277(278)/98-01,98-02 and 98-06 discussed a number ,
of instances of plant status / configuration control problems, some of which were the '
result of improper system restoration after maintenance or PMT. Violations for
plant status / configuration control problems were cited in NRC Inspection Reports
50-277(278)/98-01 and 98-06. As corrective actions for these issues, operations i
personnel developed the following initiatives:
- improve configuration control within a clearance boundary
- add system and equipment restoration details in work packages
- implement plant impact plans i
Peach Bottom Atomic Power Station Technical Specification (TS) 5.4.1 requires that )
written procedures be established, implemented, and maintained covering the l
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activities in Regulatory Guide 1.33, Appendix A, which includes procedures for
reactor cleanup system startup. The inspectors determined, based on both of these
events, '. hat PECO failed to fully maintain SO 12.1.A-3 with regard to verification of
j' system configuration prior to startup. (VIO 50-278/98-08-01)
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l- The inspectors determined that this violation was repetitive since the corrective
- actions for the previous violation cited in NRC Inspection Report 50-277(278)/98-01
j included making enhancements to station procedures. These enhancements were
j to preclude improper system restoration after maintenance and PMT. Also, the
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expectation was communicated to all station personnel of the importance of
notifying the control room of any system configuration changes made during the
, performance of maintenance and/or testing. Station personnel had not made
sufficient progress implementing these initiatives in order to preclude these events.
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c. Conclusions
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I Two Unit 3 reactor water cleanup (RWCU) system events occurred during this
i inspection due to poor system configuration control. These events resulted in an
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entry into emergency operating procedures due to a steam leak on the non-
i regenerative heat exchanger and an automatic engineered safety feature (ESF)
l isolation. The causes were less than adequate turnovers between senior reactor
operators and non-licensed operators, incomplete post-maintenance testing
i instructions, and an inadequate RWCU startup procedure.
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i Station personnel failed to properly maintain the RWCU startup procedure, resulting
! in a violation of Technical Specification 5.4.1, " Procedures." Although station
i personnel had previously developed some initiatives to reduce plant configuration
i control problems, they had not made sufficient progress implementing them to
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preclude these events.
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02.2 Torus /Drvwell Vacuum Breaker Loss of Seated Indication (Unit 2)
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a. Inspection Scope (71707)
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The inspectors reviewed operator rounds and logkeeping performance issues that
!- led to a technical specification violation associated with a torus /drywell vacuum ;
breaker,
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b. Observations and Findinas
- On August 24,1998, operators recorded on electronic rounds data that
torus /drywell vacuum breaker, AO-2-078-2504C,had lost its " seated" indication.
, On August 30,1998, operations personnel determined that the actions to verify
'
that the vacuum breakers were closed had not been performed, as required by
.
technical specifications.
1
- Peach Bottom Atomic Power Station Technical Specification 3.6.1.6 bases specify ,
- that if a torus /drywell vacuum breaker position indication is not reliable, then an l
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alternate method of ver'ifying that the vacuum breakers are closed shall be
performed within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. -This is necessary to ensure that a vacuum breaker is
not open, which would create the potential for overpressurization of the torus
during a loss of coolant accident. Operators failed to complete the required actions,
resulting in a violation of Technical Specification 3.6.1.6. This non-repetitive,
, licensee-identified and corrected violation is being treated as a Non-Cited Violation,
consistent with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 50-277/98-
08-02)
This event resulted, in part, from deficiencies in operations personnel
implementation of the rounds and log review processes:
- Operators did not always enter comments on their logs for out-of-
coecification readings, contrary to operations rounds guidance.
i
Operators did not always notify supervision of unsatisfactory readings or
conditions.
- Some electronic rounds data was not reviewed by shift management in a
timely manner as specified by operations guidance.
Operators also failed to recognize the potential safety impact of the loss of seated
indication for the vacuum breaker.
The inspectors reviewed completed corrective actions, which consisted of promptly
verifying the vacuum breaker position, revising the daily rounds data procedures,
>
and conducting briefings by operations management. Planned corrective actions
included an evaluation of equipment operator training on round sheet parameters,
inspection of the torus /drywell vacuum breaker, and revisions to electronic rounds
format. The inspectors determined that these completed and planned corrective
actions were adequate.
,
- c. Conclusions
Operators did not verify that a torus-to-drywell vacuum breaker was closed within
10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> of the discovery of an unreliable indication, as required by technical
specifications. This event was caused by the failure to adhere to equipment
operator rounds and log review practices by operations personnel. This non-
repetitive, licensee-identified, and corrected violation is being treated as a Non-Cited
Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy.
02.3 Inadvertent Shutdown of the 3C Drvwell Chiller
a. Insoection Scooe (71707 & 37551)
Inspectors reviewed the impact of an inadvertent shutdown of a Unit 3 drywell
chiller during reactor plant operation.
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b. Observations and Findinas I
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.On August 22,1998, during performance of the Unit 3 turbine building rounds an l
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equipment operator inadvertently shutdown the 3C drywell chiller. The control
room operators received the "Drywell Chiller Trouble Alarm" annunciator and took
actions in accordance with the alarm response card. The equipment operator l
ir. formed the control room operators as to the cause of the shutdown. Since the )
cause was known, no troubleshooting had to be completed before chiller restart.
Fourteen minutes after shutdown, the 3C drywell chiller was restarted by the
equipment operator. During the drywell chiller shutdown the drywell bulk average
temperature increased about 2 F and drywell pressure increase .05 psi.
The inadvertent shutdown of the 3C chiller on August 22 was documented in PEP
10008858. Part of the corrective action for this PEP included an engineering change ;
request (ECR) to evaluate the inadvertent shutdowns. This ECR resulted in an j
action request to fabricate and install a plastic guard to cover the ' Auto' and 'Stop' '
buttons on both Unit 2 and 3 chillers.
A similar event on the Unit 2 drywell chiller occurred on March 25,1997, resulting
in a temperature and pressure rise of the Unit 2 drywell. Operations personnel
initiated Performance Enhancement Program document (PEP) 10006793 to
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investigate the apparent cause of this event. According to this PEP, the human j
factoring of the microprocessor control panel could set-up any individual to i
inadvertently shutdown a drywell chiller. Although the original corrective action for I
this issue recommended providing a barrier on the control panel, engineering
personnel decided that a physical barrier was not needed. Engineering personnel
based this decision on this being an isolated event, and concluded that additional
training of equipment operators would prevent this event from recurring.
'
The inspectors concluded that the engineering evaluation of the March 25,1997
inadvertent shutdown of a drywell chiller, was not effective since a similar event
occurred on August 22,1998.
c. Conclusions
On August 22,1998,during performance of the Unit 3 turbine building rounds an
equipment operator inadvertently shutdown the 3C drywell chiller. Since the chiller
, was quickly restarted, the temperature and pressure increases in the drywell were
small and posed a small safety risk to the plant.
An engineering evaluation for a similar event that occurred on March 25,1997, was
not effective to preclude the August 22,1998 event.
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03 Operations Procedurea and Documentation
1
03.1 Review of Normal Plant Startuo Procedure i
a. Insoection Scoce (71707) l
1
The inspectors reviewed procedure GP-2, " Normal Plant Startup," Revision 88, ,
following a reactor scram event at another nuclear power station during a startup l
evolution, when operators continuously withdrew a control rod after reaching
criticality and before reaching the point of adding heat.
b. Observations and Findinas
Peach Bottom procedure GP-2, " Normal Plant Startup," contained written guidance
4
for control rod withdrawal modes following criticality. The inspectors noted that
one of the procedural steps specifically allowed single notch or notch override
(continuous withdrawal) following criticality and prior to the point of adding heat.
However, a caution statement contradicted this step by stating that only notch
mode was allowed until " nuclear heat begins to increase reactor water
temperature."
'
The inspectors discussed this discrepancy with operators and learned that, in
practice, operators used only single notch mode, consistent with the caution l
statement. Operations personnel considered the procedural step to be inconsistent
with operating practices and promptly revised GP-2.
c. Conclusions
The GP-2, " Normal Plant Startup," procedure and operating practices provided
>
adequate assurance that continuous rod withdrawal following criticality and prior to
the point of adding heat would not occur at Peach Bottom. However, inspectors
identified inconsistencies in the procedure, and operations personnel determined i
that a revision was necessary to ensure the procedure reflected operating practices.
03.2 Unexoected Start of the Motor Driven Fire Pumo Durina Testina
a. Insoection Scoce (71707)
'
The inspectors reviewed documentation for the maintenance on the H-1 fire hydrant
and discussed with operations personnel the circumstances that resulted in the
unc- acted starr of the motor driven fire pump,
b. Observations and Findinas
On August 23,1998, the motor driven fire pump unexpectedly started during post-
maintenance testing of the H-1 fire hydrant. During this testing, the fire system
pressure dropped low enough to cause an automatic start of the pump when the
hydrant isolation valve was opened. The inspectors determined that the fire system
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had not been fully filled and vented and that the system pressure dropped as the !
system filled when the hydrant isolation valve was opened. The inspectors noted l
that there was no Action Request or PEP written for this issue.
The inspectors reviewed the work order activity report and the routine test, (RT)-O-
37B-382-2," Fire Hydrant inspection and Flush (Miscellaneous)" and discussed this
issue with operations personnel. The inspectors noted that neither the work order
or the RT had instructions to fill and vent the system before starting the post-
maintenance testing or any precautions that the motor driven fire pump could start
during the opening of the hydrant isolation valve. In NRC Inspection Report 50-
277(278)/97-08,the inspectors noted a similar event when an unexpected
automatic start of the motor driven fire pump occurred during clearance restoration
of the fire system.
c. Conclusions
On August 23,1998, the motor driven fire pump unexpectedly started during post-
maintenance testing of the H-1 fire hydrant. Neither the work order or the routine
test procedure contained any documentation to inform operators that the motor
driven fire pump could start during the hydrant post maintenance testing nor did
these documents contain instructions to fill and vent the fire system after work was
performed.
Several unexpected equipment status changes, some involving safety related
components, have been documented in NRC inspection reports during the past year.
Even though this issue involved an unexpected change in the status of the motor
driven fire pump, it was not documented in any of the licensee's corrective action
systems so that it could be tracked and trended.
04 Operator Knowledge and Performance
04.1 Operator Performance Durina loss of Coolino to the 3C Main Transformer
a. Insoection Scooe: (71707)
The inspectors observed and reviewed equipment and control rocm operators
actions for the 3C Main Transformer loss of cooling occurrence.
b. Observations and Findinas
On August 21,1998, with Unit 3 operating at 100%, the control room received the
"3 TRANS TROUBLE" Alarm. The #6 oil pump had failed due to a burnt wire and
when the operator, following the alarm response card, switched the local control to
manual, all of the cooling fans and oil pumps tripped off.
The unit 3 operators commenced a down power maneuver due to loss of cooling to
the main transformer to reduce the heat load on and potentialloss of the main
transformer.
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Electricians repaired the burnt wire for the #6 oil pump and checked the other main
transformers for similar problems. While bringing the cooling system back into
service, the operators started each cooling fan and oil pump at periodic intervals.
Although operations personnel were aware that a high temperature rise could
actuate the transformer fire protection deluge system, the operators did not fully
ac. count for the temperature rise wh7n the #5 fan and oil pump were started. This
resulted in the deluge system immediately detecting a high temperature rise and
actuating as designed. The operators quickly isolated the system.
Operations personnel told the inspectors that a plant transient could have occurred,
if the deluge system had sprayed on the transformer with the unit at full power, due
to transformer oil temperature and pressure changes. With generator output
significantly reduced, the deluge activation did not cause a transformer transient
and the unit remained on line. The inspectors determined that actuation of the
deluge system was not expected. Although the deluge system could have been
bypassed to prevent activation, there was no guidance in the procedure for
restoring the cooling system for the main transformer, to deactivate or bypass the
deluge system while bringing the cooling system back to operation.
During the down power maneuver, the operators created a speed mismatch of 50
RPM between the recirculation pumps which resulted in a loop flow mismatch that
was recorded in the reactor operator's log as within the Technical Specification (TS) '
requirements. During a subsequent panel walkdown by the shift supervisors, the i
loop flow mismatch had increased to greater than TS limits. A one hour Technical
Specification Action (TSA) was entered, the 'B' recirculation pump speed was i
lowered so that loop flow mismatch was reduced, and the TSA was exited within '
27 minutes,
iI
The inspectors independently determined that the operators allowed a loop flow
mismatch during the down power maneuver based on reviews of operator logs and
recirculation flow data. The inspectors noted that the effect of xenon following the ;
downpower caused recirculation loop flows to change which resulted in the
'
increase in loop flow mismatch. The inspectors determined, based on discussions
with shift management that, although the operators thoroughly understood the
effects of xenon on power they did exhibit a lack of full understanding of the
effects of xenon on recirculation loop flow. Therefore the operators did not control
recirculation flow before the effect of xenon increased loop flow mismatch outside
the LCO range. By more closely monitoring the recirculation loop flow, the
operators could have prevented an involuntary entry into a Technical Specification
Action and Limiting Condition for Operation.
c. Conclusion
On August 21,1998, unit 3 operators commenced a down power maneuver due to
loss of cooling to the main transformer. The reduced load prevented a loss of the
main transformer and plant transient when the deluge system activated.
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The initial response by the operators for the loss of cooling to the main transformer
and subsequent deluge activation was good. Due to the power reduction, the flow ;
in the recirculation system loops became mismatched in excess of the Technical
'
Specification limit. This condition was identified and corrected within the time
allowable by Technical Specifications. Failure of operations personnel to fully
understand the effect of xenon on recirculation flow and closely monitor
recirculation flow contributed to this involuntary entry into a Technical Specification
' Action and Limiting Condition for Operation.
05 Operator Training and Qualification
05.1 Limited Senior Reactor Operator (LSRO) Reaualification Proaram
'
a. 1Dsoection Scope (71001)
.
The inspectors evaluated the dual site, Limerick / Peach Bottom, PECON LSRO
requalification training program to verify its compliance with 10 CFR 55. NRC
Inspection Procedure 71001, Licensed Operator Requalification Program Evaluation, !
and NUREG-1021 Interim Rev.8 - ES-702 were used for the evaluation.
The inspectors evaluated the following program areas:
- Program guIdel;nes -
- Operating and written examinations ?
- Exam security
i
- Management oversight -license activation and maintenance of records,
remediation, training, attendance, feedback system, and medical records
i
. PECON procedures and documents associated with the LSRO training program and
. its implementation were also reviewed. ,
The observation of the annual operating exam was not performed during this
inspection and will be performed during the LSRO training cycle in 1999. i
b. Observations and Findinas
' Proaram Guidelines '
The inspectors reviewed PECON procedures LSRO-9500,"LSRO Course Plan," and i
LSRO-OOOO, " Multi -Site Fuel Handling Director," and determined they acceptably i
- described a program which met 10 CFR 55 requirements and previous written !
l
commitments by PECON to the NRC. Additionally, the inspector reviewed the LSRO
. program subject index and selected LSRO classroom and practical job performance ;
lesson plans and found that their content was comprehensive and well maintained i
by the program coordinator.
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Qpjeratina and Written Examinations
The inspectors reviewed three written biennial examinations and two annual
operating exams and determined thes/ 9cceptably sampled the items specified in 10
CFR 55. The inspectors also found f. hat the exams adequately assessed knowledge
level in the area of abnormal and emergency procedures. Additionally, it was noted
that a large percontage of the questions in the exams were of the more challenging,
higher order, analytical type.
The inspectors reviewed job performance measures (JPMs) and found that they met
the qualitative guidelines of the inspection procedure and the PECON program. The
JPMs reviewed included those for normal, emergency, and abnormal conditions,
Exam Security
The inspectors reviewed the security measures taken by the facility for exam
development and administration, and determined that programmatic controls were
satisfactory, with no indications of exam compromise.
Activation and Maintenance of Operator Licenses 1
The inspectors reviewed the programmatic controls that PECON used for
maintaining an active license and for reactivating a license while meeting the
requirements of 10 CFR 55.53 and found them to be Greptable. The inspectors ,
reviewed various training attendance records, includng niissed training make-up
sessions or exams, and determined that controls for rnain:enance and reactivation
of operator licenses were good.
Remedial Trainina Proaram
The inspectors reviewed remediation records for two individuals who had failed the
biennial written exams. The inspectors found that the remediation packages l
developed by the training coordinator were appropriate for the weaknesses
{
demonstrated and were properly documented in accordance with PECON i
procedures.
Operator Feedback :
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The inspectors reviewed the feedback records for the past three years and found ;
that management review and disposition was timely.
Medical Records i
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The inspectors also reviewed all LSRO medical files to ensure that medical exams !
were being conducted biennially in accordance with 10 CFR 55.21 and determined !
that requirements were met.
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c. Conclusion
The inspectora concluded that the Senior Reactor Operator 1.imited to Fuel Handling
(LSRO) program was good overall. The LSRO program guidelines and examinations
were comprehensive and well maintained by the prograra coordinator and LSRO
license maintenance was well documented. The inspectors also determined that the
areas of exam security, remediation, operator feedback, and medical records were
acceptable.
11. Maintenance
M1 Conduct of Maintenance
M 1.1 General Observations
NRC Inspection Procedures 62707 and 61726 were used in the inspection of plant
maintenance and surveillance activities. The inspectors observed and reviewed i
selected portions of the maintenance and surveillance test activities listed in '
Attachment 2.
The work and testing performed during these activities was professional and
,
thorough. Technicians were experienced and knowledgeable of their assigned
tasks. The work and testing procedures were present at the job site and actively
used by the technicians and operators for activities observed. Good pre-job briefs
were observed prior to the performance of the surveillances observed. Applicable
procedures were present in the control room and at the job sites during surveillance
testing and were appropriately used.
M2 Maintenance and Material Condition of Facilities and Equipment
M2.1 -Main Control Room Deficiencies
a. Insoection Scoce (61726 & 62707)
The inspectors reviewed the Equipment Status list, action requests designated as
Main Control Room (MCR) deficiencies, and the non-outage maintenance backlog to
assess the effectiveness of the licensee's corrective maintenance of MCR
deficiencies that impact the operators ability to maintain reliable and safe plant
operation.
b. Observations and Findinas
Main control room deficiencies (MCRDs) were identified and tracked by Action
Requests (ARs). The ARs dealing with MCRDs were noted as either control room
deficiencies or critical control room deficiencies. The inspectors noted that
currently there were 56 control room deficiencies listed on outstanding ARs.
Several of these deficiencies were corrected but the ARs remained open to provide
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for equipment monitoring. When critical control room deficiencies were identified
during this inspection, they were corrected in an expedited manner.
An equipment deficiency log and control rod drive (CRD) deficiency log were kept
by control room personnel to identify control room deficiencies. These logs made
up the equipment status list. The equipment status list was updated biweekly. The
operators reviewed and maintained both logs to track the status of control room
significant deficiencies.
The inspectors reviewed the equipment deficiency log. During this review, the
inspectors noted that the oldest outstanding significant MCRD was over a year old
on unit 2 and over six months old for unit 3 and that five and 13 significant MCRDs
were open for units 2 and 3, respectively. However, the inspectors determined that
all items on the significant MCRD log had minor safety impact and were scheduled
for work.
The inspectors noted that some of the requirements in OM-P-10.3, Revision 3,
" Equipment Status List / Tagging of Deficiencies" were vague. During tours of the
main control room, the inspectors observed that each MCRD had an equipment
deficiency tag, but that some tags were inconsistent with OM-P-10.3 or other work
control procedures. The inspectors determined that these inconsistencies were of
minor safety significance,
c. Conclusions
Control room deficiencies were controlled and adequately prioritized so that critical
Main Control room deficiencies were corrected in a timely manner. However, some
weaknesses, of minor safety significance, were noted with the clarity and
implementation of the requirements in OM-P-10.3, Revision 3, " Equipment Status
List / Tagging of Deficiencies."
M3 Maintenance Procedures and Documentation
M3.1 Fix It Now Team Plannina and Documentation
a. insoaction Scope (62707)
The inspectors reviewed approximately 15 completed work orders performed by the
Fix l.t Now (FIN) team.
b. Observations and Findinos
The FIN team work order documentation was usually consistent with FIN
. . administrative procedures. The documentation appropriately reflected such items as
work scope, parts, procedures / prints, and post-maintenance testing requirements.
In one instance, incomplete corrective maintenance documentation led to repetitive
problems on temporary emergency cooling tower replenishment pumps. These
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problems reduced the number of pumps available, but did not affect the operability
of the replenishment capability.
c. Conclusions . 1
>
Most Fix-It-Now (FIN) team work order documentation was consistent with FIN
administrative procedures, however; incomplete documentation led to repetitive ,
problems on temporary emergency cooling tower replenishment pumps. I
M4 Maintenance Staff Knowledge and Performance
M4.1 Fire Main Leak 1
a. Insoection Scope (37551. 62707 & 71707) l
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The inspectors reviewed an event in which a pipe coupling separated on the station
fire main during maintenance on a fire hydrant. Water from the leaking fire main
exceeded the capacity of the storm drains and then entered the Unit 2 safety-
related emergency service water /high pressure service water pump hause via
. underground electrical conduits.
I
b. Observations and Findinas 1
On August 23,1998, a leak on the station fire main occurred wher, a 6" pipe
separated from a coupling upstream of a block valve. Maintenanco workers were
working in the vicinity of the coupling. This event was caused by weaknesses in
both maintenance planning and work practices.
Planning issues: A lack of knowledge or understanding of the design of the
pipe coupling upstream of the' block valve contributed to this event. This
<
coupling was a slip-fit compression fitting that separated while maintenance
technicians were working on the downstream hydrant coupling. Normally,
the failed coupling was supported laterally by two tie rods clamped between
the hydrant and the piping upstream of the coupling, and by a thrust block at
the hydrant. Both the tie rods and the thrust block had been removed to <
permit replacement of the hydrant, thus allowing the coupling to separate.
Planners did not have detailed information on the design of the coupling or
the function of the tie rods. Planners did not direct maintenance personnel
to uncover the coupling, thus missing an opportunity to visually check the
coupling before the work was accomplished. Also, planners did not use two-
valve isolation for the work.
- Maintenance Practice issue: Maintenance personnel left the pipe coupling
.
covered with dirt, thus they were not aware of its configuration and the
potential hazard associated with removing the tie rods.
Water accumulated in the outside yard area in the vicinity of the fire hydrant, due to
exceeding the capacity of the storm drains in the area. Water seeped into electrical
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conduits via manholes that connected to underground cable duct banks. The water
then leaked into the emergency service water /high pressure service water pump
house structure through these cable duct banks and degraded penetration seals.
Sump pumps in the pump house prevented any water accumulation. PECO
engineers promptly evaluated the impact of the degraded penetration seals on pump
operability and determined that no operability concems existed. Engineering
personnel also extrapolated the pump house in-leakage rate to the design basis
flood level and postulated that in-leakage at flood levels would have been within the
capacity of the sump pumps. (The design basis flood scenario is a flood of about
twelve feet above the ground level outside the pump housa. The water level
outside the pump house actually reached about three inches). The Peach Bottom
flood analysis allowed for cables to exist in a wet environment, and some conduit
seepage was acceptable. The inspectors reviewed the engineering analysis and
operability determinations from this event and had no concerns.
c. Conclusions '
7
WeakneIes in maintenance planning and work practices led to a significant water
lesk on the station fire main on August 23,1998. Water from the leak entered the
i
safety related emergency service water /high pressure service water pump house via i
underground electrical conduits and degraded penetration seals. The engineering l
evaluation, that the penetration sealleakage was within design assumptions for a
design basis flooding event, and pump operability was not affected, was adequate.
Ill. Enaineerina
E1 Conduct of Engineering
E1.1 Installation of the ISFSI East Retainina Wall j
a. Insoection Scope (60851 & 60853)
The inspectors reviewed independent spent fuel storage installation (ISFSI) ;
engineering and construction activities affecting the concrete placement of the east i
retaining wall, including physicalinspection of the installation. The inspector also
reviewed field activities associated with the soil compaction testing and inspection j
of the subsurface of soil of the ISFSI east retaining wall. The inspector evaluated i
the site and reviewed construction records. !
1
b. Observations and Findinas i
Assessment of Construction Activities ]
!
To protect the ISFSI storage pad from an adjacent hill and undermining due to a
slope drop, retaining walls were constructed to the east and the west of the storage
pad. The design and construction of these retaining walls was performed in
accordance with 10 CFR 72 Subpart G " Quality Assurance" requirements. These i
retaining walls are designated as important to safety (ITS). j
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The inspectors reviewed the construction of the east retaining wall, in particular, the
- construction joints. Concrete for the east retaining wall was poured in sections,
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which interlocked through construction joints called keyways. The inspectors
reviewed an evaluation of construction discrepancies made on the dimensioning of
<
two 6'f the keyways. The design specified that each keyway be 8" wide by 3"
, deep. PECO personnel identified that the keyways on the south construction joint
'
of wall No. 02A and the north construction joint of wall No. 05 used incorrect
2
keyway dimensions (approximately 5.5" wide by 3" deep). Engineering personnel
documented this evaluation in Engineering Change Record (ECR) 98-02355. As a
! part of the corrective actions, PECO personnel stopped additional concrete pours
and analyzed the as-found condition.
The inspectors reviewed calculation NCR-98-02355,which assessed the as-found I
configuration (reduced size of keyway) of the construction joint at the east wall.
The inspectors found the calculation and conclusion acceptable. The assumptions !
in the calculation were conservative, and the approach used to calculate the shear
forces acting on the reduced keyway were acceptable. The calculation showed that
the as-found configuration for the reduced size of the keyway was within shear i
allowables established in the main design calculation of the east wall.
Soil Comoaction Activities
The inspectors reviewed Field Change Request (FCR) No. 98-OO811-10,which was
prepared to document and disposition the two field density tests which failed to
achieve the required degree of compact'on (95 percent). The FCR disposition in
both cases accepted the condition "as-is" based on engineering judgement. The
inspectors reviewed the engineering judgement determination and determined that
this was acceptable because the deviation was not substantial. The inspectors
verified that this was an isolated case and PECO personnel have ensured 95% )
compaction was achieved in all tested locations since this deviation was discovered. '
Concrete Mix Deliverv. Testino. arid Pourina Activities
The inspectors observed the concrete mix delivery, testing, and pouring activities
for the east retaining wall. The inspectors noted that the concrete pounng was l
being conducted in an acceptable manner. i
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Concrete was being shipped from Havre de Grace, Maryland, which was l
approximately one hour from ISFSI site. The inspectors noted that the specified l
maximum time of 90 minutes, from batching the concrete till p.,uring, was being
enforced to maintain quality and achieve the desired compressive strength. This
was evident when PECO personnel rejected two truck loads of concrete because j
the loads were not poured within 90 minutes.
'
The inspectors noted that the testing of the newly arrived concrete was properly 1
completed. Tests included slump, air entrainment, concrete and ambient air i
temperatures, and weight. Batch tickets were reviewed by PECO personnel and
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contained appropriate information, including data on the concrete mixture, time of
batching, and truck number.
c. Conclusion I
The inspectors concluded that the construction activities on the east retaining wall
of the independent spent fuel storage installation were acceptable. Engineering .
personnel had resolved construction deficiencies regarding as-built keyways and soil l
compaction in an effective manner, thus the analyzed as-found condition of the east
wall was acceptable. The concrete mix delivery, testing, and pouring activities for
the east retaining wall were acceptable. !
E2 Engineering Support of Facilities and Equipment
l
E2.1 (Closed) URI 50-277(278)/97-06-03and (Closed) LER 2-97-007 Potential for !
Bvoass of Pressure Suporession Pool l
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a. Inspection Scope (92903)
4
The inspectors reviewed licensee actions taken in response to the identification of I
the potential to bypass the pressure suppression function of the torus.
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b. Observations and Findinas
>
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A potential bypass flow path existed by which the drywell air space could
.
communicate with the torus air space through a six-inch containment purge nitrogen
supply piping. On October 21,1997, station personnel reported this issue pursuant
to 10 CFR 50.72 as a condition outside the design basis of the plant. Licensee
Event Report (LER) 50-277(278)/2-97-07 reported this issue on November 19,
1997.
Following the identification of this condition, engineering personnel drafted non-
conformance reports and a shift update notice for operations to notify all shift
personnel. Engineers also changed several procedures to prevent simultaneous ,
purging of the drywell and torus. Further, operations personnelissued a
administrative clearance to disable the drywG inboard purge supply valves on both
1 units to prevent simultaneous purging of J+ 6rnell and torus.
1
-
PECO corporate engineers completed an evmuation of other potential bypass paths
- in June 1998. They concluded that other potentialleakage paths were either not
credible, or were significantly smaller than the equivalent of a one-inch hole limited
by technical specifications. This evaluation also concluded that the interim
i disposition of disabling the drywellinboard purge supply valves was acceptable until l
a final disposition was approved.
Engineering personnel attributed the cause of the event to an original design
deficiency in that the design requirements for lines which connect the drywell
airspace to the torus airspace were not adequately specified. Single failure and
.
.
20
electricalindependence design criteria were not originally applied to the drywell and
torus inboard purge supply valves.
The inspectors reviewed engineering activities for this issue, discussed them with
selected engineers, and conducted an in-plant review of the LER. The inspectors
determined that this issue was an apparent violation of 10 CFR 50 Appendix B,
Criterion Ill, " Design Control." However, the inspectors noted that it was licensee-
identified as a result of review', of industry operating experience and General
Electric 10 CFR Part 21 notification No SC97-4 dated October 15,1997. In
addition, the inspectors concluded that station personnel took prompt and effective
interim corrective actions as described above, and this issue was not likely to be
identified through routine efforts, in accordance with the NRC Enforcement Policy,
Section Vll.B.3, Violations involving Old Design issues, the NRC is exercising
enforcement discretion and not citing this violation as noted in separate
correspondence issued on October 28,1998. (NCV 50-277(278)/98-08-03)
c. Conclusions
Engineering personnel took prompt and effective corrective actions following j
their identification of the potential to bypass the pressure suppression function
of the torus during simultaneous purging of the torus and drywell or as a result
of postulated failures. In accordance with the NRC Enforcement Policy,
Section Vll.B.3, Violations involving Old Design issues, the NRC is exercising l
enforcement discretion and not citing this violation. l
l
E2.2 Aacess and Alarm Failures to Protected Area and Vital Areas Doors Due to Security
Multiplexer Failure
a. Inspection Scone (37551 & 71750)
The inspectors observed the response of security personnel to several failures of the
protected area and vital area doors due to a security computer multiplexer failure.
The inspectors also discussed this issue with security management and the security
system manager.
b. Observations and Findinas
Twice on August 12,1998 and then again on August 19 and August 24, the #1
security computer multiplexer failed. This failure caused the protected area and
vital area doors to f ail closed and rendered the alarm functions on the doors
inoperable. Security personnel implemented compensatory actions for this failure
per site security plan operating procedures. The inspectors discussed the
compensatory actions with security management. The inspectors noted that the
response by the security force to this multiplexer failure was adequate and the
failure of the doors had no adverse impact on plant operations. Operations
personnel were provided keys to allow access through locked doors while the
security computer was down. Each of the computer multiplexer failure events were
noted in the security log.
.
.
21
The system manager for the security system reviewed these failures and believed
that they were due to a power surge in the system. No root cause was determined
for these failures by the end of the inspection period.
The inspectors discussed this issue with engineering personnel. The inspectors
noted that engineering was initially slow to investigate this issue because the
system manager was offsite in training. Full investigation or this issue did not occur
until after the fourth failure of the security system when the system manager
returned from training,
c. Conclusions
Security personnel responded adequately to four failures of the security computer
multiplexer ir August 1998. The failures caused vital and protected area doors to
fait closed without alarm functions.
Engineering personnel failed to fully investigate and support this issue until after the
fourth failure. This contributed to delays in determining the root cause of the
multiplexer failures.
4
E2.3 Core Sorav Residual Heat Removal and Hiah Pressure Service Water Motor
poerated Valve Thermal Overload Wire Discrepancies,
a. Inspection Scone (37551)
The inspectors reviewed the findings and actions by engineering personnel for
thermal overload wiring discrepancies on motor operated valves (MOVs). l
Differences were identified between the installed wiring and the drawings for the l
Core Spray (CS), Residual Heat Removal, and the High Pressure Service Water
valves on both units.
b. Observations and Find _ings
l
During the review of a multiple high impedance fault calculation, engineering
personnel noted that the thermal overloads for the unit 2 CS suction MOVs (MO-2-
14-007A,B,C,D) were in the control circuit. However, on the unit 3 CS suction
MOVs (MO-3-14-007A,B,C,D),the thermal overloads for the control circuits were
permanently bypassed. The unit 3 CS valves were operable since they have no
automatic safety function, were key locked open, and would only be closed to act
as primary containment isolation valves when the CS system was secured and/or
tested.
~ . - , - - - - - - . - . - - - - - - - . - - . . - . . - . -
.. - . - - .~._ - - . -
..
, 22
A walkdown of the system by maintenance planning, on August 6 through
'
August 19,1998, revealed that the actual wiring did not match the schematic
drawings. Although the schematics showed that the wiring for the MOVs on both
units were the same, the as-found did not match the schematic drawings for the
unit 3 CS suction MOVs. Station personnel found the thermal overloads in series
with the control power ground connection on the unit 3 MOVs. This could result in
the control power being open circuited on thermal ovccload operation resulting in a
loss of position indication. However, an alarm in the control room would still )
indicate thermal overload activation. Although this wiring configuration was not in
accordance with the schematic drawing, it still resulted in the de-onergizing of the
MOV on a thermal overload condition. l
Station personnel compared the wiring configurations of several other MOVs with !
the schematics and found that similar discrepancies existed for the residual heat
removal (RHR) system suction and cross connect valves and the high pressure
ervice water (HPSW) outlet valves on the RHR heat exchangers. Initial evaluations ;
'
indicated that these valves functioned the same as the unit 3 CS suction MOVs and
were operable based on the same criteria used to evaluate the CS MOVs.
The inspectors noted that engineering's evaluation found the thermal overload
protection changed when the alternate power supply from the remote shutdown I
panel was used due to the as-found wiring for one of the RHR suction and HPSW l
MOVs on each unit. Further evaluation of these RHR and HPSW valves was on- I
going and tracked though the licensee's corrective action request system.
Engineering personnel noted that the initial operability evaluations did not address
the thermal overload protection concern that occurred when the alternate power
supply was used.
The inspectors reviewed documentation from the discovery of the wiring
discrepancies, the wiring rchematics, and the operability evaluations for the core
,
spray suction valves. The inspectors noted during this review that these wiring
discrepancies had existed for a long time.
c. Conclusion
Although discrepancies existed between the as-found condition and the schematics
for thermal overload wiring on several core spray, residual heat removal, and high
pressure service water motor operated valves, none of these discrepancies resulted
in the valves being inoperable. Initially, the most significant problem due to this
issue was the loss of valve position indication during a thermal overload actuation. ,
However, further evaluations of thermal overload protection for these motor !
operated valves, during the use of the alternate power supply, was under review by
the engineering personnel and was documented in the licensee's corrective action
system. I
!
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23
1
IV Plant Support
R1 Radiation Protection and Chemistry Controls (RP&C)
R 1.1 Radioactive Waste Processina. Handlina, Storaae, and Shionina
l
a. Insoection Scoce (86750) !
The inspectors reviewed and discussed sources of radioactive waste at the station,
waste processing and volume reduction efforts, and storage of waste. The
i
inspectors evaluated the methodology for radioactive waste concentration averaging !
and the development of scaling factors used to estimate hard to detect l
radionuclides (e.g., Pu-239, Am 241). The inspectors reviewed waste classification I
practices and selectively reviewed radioactive waste shipping records for shipments (
of radioactive waste and other radioactive materials made since the previous !
inspection. The inspectors also performed a test of the emergency response
contact listed on the licensee's radioactive material shipping papers. In addition,
the inspectors observed various radwaste loading and shipping activities including
loading of a cask with a high integrity container of spent resins and loading and
shipment of packages of slightly contaminated soil, i
l
The review was against selected criteria contained in 10 CFR 20; 10 CFR 61;
10 CFR 71; 49 CFR 100-179; 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 Findinas
PECO continued to aggressively review and evaluate methods to reduce generated
radioactive waste volumes. PECO implemented actions to minimize dry activated
waste and process waste and was closely tracking and monitorin0 numerous
performance indicators relative to radioactive waste program performance including
plant leaks to reduce waste volumes. Of particular note was PECO's initiatives to
request industry audits of its solid and liquid waste programs to identify areas for
enhancen Mt and waste volume reductions.
PECO implemented appropriate scaling factors for use in determining curie content
of hard to detect radionuclides and was implementing applicable NRC Branch
Technical Position (BTPs) guidance regarding waste concentration averaging.
Waste was properly classified relative to 10 CFR 61 requirements.
The radioactive waste / material shipping program was generally well implemented.
Radioactive material shipping documentation was well maintained and available for
review. Individuals responsible for shipping activities were knowledgeable of
applicable requirements.
PECO was a registered user of the NRC licensed casks used for shipping purposes
and maintained up-to-date cask certificates of compliances associated drawings and
. - . . - . - _ - . - - - ---_.---- .- --
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. 24
- . disposal facility licenses. Shipments of radioactive materialin NRC licensed casks
,
were performed in accordance with certificato of compliance requirements.
! PECO implemented its emergency notification requirements for shipments by use of -
'
a vendor service listed on its shipping manifests as a point of contact for in-transit
shipping problems. When contacted, the vendor accurately described emergency
- response actions for the shipment (dewatered resin) in transit.
J
i
c. Conclusions
l-
4
PECO's radioactive waste transportation program, including processing, handling,
, -storage, and transportation was effective. Wastes were properly classified and
, packaged.
R2 Status of RP&C Facilities and Equipment
,
a. Inspection Scoce (86750)
,
)<
l
The inspectors toured various radwaste and radioactive material storage areas '
j including the radwaste building, the south radwaste storage location, and the low l
r level waste storage area.
.
b. Observations and Findirigs -
Areas were generally well maintained, properly posted, and controlled. No
abandoned areas containing unprocessed, stored, or spilled waste was detected.
There was a limited amount of waste stored in the radwaste processing and storage ,
'
areas. Waste storage and processing areas exhibited generally very good material
condition. PECO was actively cleaning and painting the facility.
- *
- The overhead pipes located in the 91'6" elevation of the floor drain pump room
exhibited some surface rusting. PECO had previously evaluated the pipes,
concluded the rust to be minor surface rust, and was monitoring the condition of
the pipes in the room.
The documented inventory of material contained in the south waste storage area
was not fully up-to-date. One package was marked as indicating it contained
material but the inventory list for the area indicated the container was empty.
PECO updated the list and took action to ensure the list was maintained current by
use of formal reporting of material transferred into the area,
c. Conclusions
Overall, radioactive waste and material processing and storage areas were properly
posted and controlled, and exhibited very good material condition.
.. . .- . - . - - - .-- --- - .-. - - - - . . - . . - - _
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.
4 '
25
R3 RP&C Procedures and Documentation
'
a. Inspection Scope (86750)
,
i
The inspectors discussed changes in radioactive waste processing, handling, and l
shipment procedures and programs since the previous inspection with personnel '
responsible for these areas.
i
b. Observations and Findinos I
There were no major changes identified in the radioactive waste processing,
handling, storage, and shipment procedures and programs since the previous
inspection.
There were very few staff or organization changbs in the radwaste management
4
group. New personnel coming into the group were provided training and prohibited
from performing tasks for which they were not yet qualified.
4
PECO was developing a 10 CFR 50.59 safety evaluation ta support injection of
9
noble metals into the reactor coolant in late 1998. PECO concluded that no adverse
radiological controls impact was associated with the injection including impact on l
radwaste processing, handling, or transportation programs.
PECO had established and implemented procedures for evaluation and survey of ,
,
materials not normally considered radioactive (e.g., sewage) to determine if the l
material should be considered contaminated and disposed of at licensed facilities.
- c. Qnclusions
There were no major changes identified in the radioactive waste processing,
- -
- handling, storage, or shipment procedures and programs since the previous
inspection. PECO implemented its program for monitoring of normally non-
radioactive material (e.g., sewage).
R4 Staff Knowledge and Performance in RP&C
'
4
a. Insoection Scope (86750)
The inspectors evaluated general staff knowledge of radioactive waste processing,
handling, storago, and shipping requirements during the inspection,
b. Observations and Findinos
PECO personnel responsible for radioactive waste processing, handling, storage and
radioactive material transportation exhibited a good knowledge level of regulatory
requirements and program procedures. The personnel were aware of, and
knowledgeable of applicable regulatory requirements, including procedural
specifications, DOT rules and regulations, and radiological surv;y and assessment
methodologies.
- - - - - . - - _-- --.-.___.-.__
-
s
y
.
.
26
c. Conclusions
Individuals responsible for radioactive waste processing activities exhibited a good
knowledge level of regulatory requirements and program procedures.
R5 Staff Training and Qualification in RP&C
a. Inspection Scone (86750)
The inspectors reviewed the training provided to personnel involved in radioactive
waste generating, processing, and handling activities against criteria contained in
NRC Bulletin 79-19 and 49 CFR 172, Subpart H. The inspectors reviewed training
records and lesson plans and discussed training with cognizant PECO personnel.
b. Observations and Andinas
Based on a review of job tasks at the station, PECO had previously established a ,
training matrix to ensure that appropriate training was provided to applicable
personnel. PECO met the training requirements for station personnel via general
employee training and special training modules developed to address applicable
requirements.
One individual (mechanic) involved with transfer of radioactive waste and loading of
a cask shipment on August 20,1998, had not received the training previously
prescribed for his position in the a-priori developed employee training matrix
described above. The mechanic transferred waste from the station to the low level
waste storage facility and routinely loaded waste into shipping casks for transport.
The individual was noted to be knowledgeable of his specific task requirements and j
. was provided continuous direct oversight by radioactive material shipping l'
coordinators.
PECO initiated an evaluation of the adequacy of the individual's training for his .
assigned tasks and believed he may have possessed appropriate training relative to
49 CFR 172, and NRC Bulletin 79-19, acquired in advanced radworker training and
waste minimization training. PECO initiated action to place this matter into its
corrective action program. PECO also initiated action to determine why the
i individual had not received the a-priori specified training hnd if the training was
necessary. PECO took action to ensure other mechanics involved in waste handling
i. had received appropriate training.
There was no apparent defined training program for new radwaste personnel
, involved with radwaste shipping activities. PECO was taking action to review, and
. adopt, as appropnate, a proposed common program to be implemented at its
j . nuclear stations. One individual had recently been transferred into.the radwaste
, group and a PECO radwaste manager provided job specific procedure training and j
L on-the-job training. The individual possessed previous radwaste oversight
- . experience. The training was considered adequate.
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- _ . _ - - . - - - . . . - . , .
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c. Conclusions
- PECO provided generally good training of personnel involved in radioactive waste
activities However, one individual had not received the a-priori specified training
for mechanics involved with radioactive waste activities and there was no defined
- training program for new radwaste personnel involved in radwaste shipping
[ activities. PECO took appropriate action on these matters.
i. .
R6 RP&C Organization and Administration
,
- a. Insoection Scoce (86750)
>
!
The inspectors reviewed the current radioactive waste processing organization, I
j including staffing, responsibilities and authorities. The inspectors evaluated PECO's
j performance in this area by discussion with cognizant personnel and review of .
I applicable administrative and organizational records.
>
l b. Observations and Findinas
! ,
.
i The review of the current radioactive waste processing organization indicated that !
there were no significant changes in the organization or its responsibilities and j
authorities since the previous inspection in this area. I
PECO radwaste management indicated that the current radwaste organization would
i be disbanded, in the near future, with its various subgroups incorporated into the
chemistry, radiation protection, and plant engineering groups, as appropriate, for
, purposes of enhancing efficiency and effectiveness. No immediate safety concerns
-
were identified relative to thic proposal. PECO was aware of the need to update
j organizational administrative documents, as appropriate, to reflect the new
-
organization, j
c. Conclusions
s
i. PECO continued to implement an appropriately staffed and defined organization for
i
radioactive waste processing, handling storage, and shipping.
1 :
l R7 Quality Assurance in RP&C Activities
t
,
a. Inspiq1 ion Scope (86750)
,
.The inspectors reviewed PECO's audits, assessments, and surveillances of its
i- radioactive waste handling, processing, and storage programs, as well as audits of
l the Process Control Program, against the criteria contained in its Quality Assurance
J. Program,10 CFR 20, and 10 CFR 71, Subpart H, Quality. Assurance.
,
p b. Observations and Findinas
i PECO performed various audits and surveillances of its radwaste processing,
j- -handling, storage,-and transportation programs including its process control
program. Training audits for applicable personnel were also conducted. The
<
4
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.
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, station's Nuclear Review Board recommended areas for additional review,
i > '
Oversight activities were performed using appropriate check lists and qualified
L personnel were used in lead audit capacities. Appropriate corrective action
{' - measures were init:ated for areas for enhancement. Audits were of appropriate ;
j depth and scope. !
!
c. Conclusions
4
2 '
, PECO performed audits of appropriate depth and scope of radwaste processing,
- ' handling', storage, and transportation activities, including training and qualification
i _- of personnel. Corrective actions were initiated for identified concerns.
R8 Miscellaneous RP&C Activities
fi
t R8.1 (Closed) Violation (VIO) 50-277(2781/97-03-02 Failure to Assure that the Turbine !
j' Buildina Atmosphere was Processed Throuah the Turbine Buildina Gaseous Waste
y Treatmcnt System
.
)
i During a review of the design modificetion involving the north wall of the Unit 3 -
turbine building, it was determined that the processing and monitoring of the turbine
building atmosphere was not adequately performed. In response to NOV
50-278/97-03-02, dated May 16,1997, PECO attributed the violation to a lack of j
detail in the work order and inadequate verbal communication between the work
planner and the Health Physics planner. To mitigate future errors the licensee
revised four procedures to clarify communication expectations and regulatory
i, Mance. The corrective actions were reviewed and found to be reasonable. The
violttion is closed.
R8.2 (Closed) VIO 50-277(278)/97-04-03 Violation of Locked Hiah Radiation Area Kev
Control
The corrective actions taken by PECO for this violation were previously described in
NRC Inspection Report No. 50-277(278)/97-04, dated July 24,1997. PECO
implemented the corrective actions described therein. A review of high radiation
area access controls during this inspection found that access doors to high radiation
areas were properly locked and proper administrative controls were implemented for
keys to these areas.
R8.3 Inspection of incomina Fuel Shioments
a. Insoection Scope (86750)
The inspectors reviewed radiological controls oversight of incoming fuel shipments.
b. Observations and Findinas
- PECO was receiving new fuel for the Unit 2 outage. Radiation protection (RP)
technicians performed radiological surveys of the incoming fuel shipping containers
-including using an alpha contamination smear counting system to check smears of
the incoming packages for alpha contamination.
p
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.
.
29
RP procedure HP-C-403, " Instrument Quality Checks," Revision 0, required in
Section 7.4.2, that if 3 or more consecutive values, were in the warning band,
notify the Instrument Physicist, who would evaluate the Control Chart and
determine the instruments physical condition and determine whether to place the
instrument out of service or continue use, if continued use was permitted, the
instrument Physicist was to denote this on the Control Chart and initial and date the
entry. The warning level was defined in the procedure as the range on the control
chart between + 2 sigma and + 3 sigma and between -2 sigma and - 3 sigma
values.
The inspectors reviewed the Control Chart source check data for the instrument on
August 18,1998. The inspectors determined that 3 consecutive instrument source
check values fell outside of two sigma during the period August 16-17,1998.
, The radiation protection technician who performed the source check did not act on
the matter and a second technician did not recognize the problem. The Control
Chart listed only one of several acceptance criteria. The instrument had been used
for counting of smears of incoming fuel shipments and the instrument's Control
Chart was not initialed to permit its continued use. This was identified by the
inspectors as a violation of Technical Specification 5.4.1 for failure to properly
, implement procedure HP-C-403. (VIO 50-277(278)/98-08-04)
PECO placed this matter in the PEP program, initiated an evaluation of the alpha
smear counting instrument, and determined that the instrument was functioning
properly and exhibited proper efficiency when source checked. PECO reviewed
beta-gamma smear survey results and did not identify any removable contamination
on incoming fuel shipments. PECO reviewed other in-field counting instruments and
did not identify any similar problems. PECO revised its instrument Control Charts to
include all procedure specified acceptance criteria for evaluation of source check
results. PECO concluded the individual who had performed the check was aware of
the procedure requirements but forgot to initiate a call to the Instrument Physicist. i
PECO coached the involved individual and discussed the event at all hands
meetings. !
l
c. Conclusion
PECO provided generally good radiological controls oversight of incoming fuel
shipments. However, a violation of radiation protection procedures associated with i
'
source checking of an alpha contamination counting instrument was identified by
the NRC and was promptly corrected by PECO.
t
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30
R8.4 Security Oversiaht of Radwaste Activities
a. Insoection Scoce (71750 & 86750)
' The inspectors reviewed PECO's loading and transfer of a high integrity container
into the process shield at the low level waste storage facility for transfer into the
protected area.
b. Observations and Findinas
PECO stored its vendor supplied high integrity containers, upon receipt, in a locked i
building outside of the protected area. The containers were subsequently loaded i
into a waste processing shield, under observation of security personnel, and
transferred into the protected area to the waste fill station by personnel authorized
Protected Area access. The large lids of the containers were sealed.
Although security personnel routinely provided oversight of the loading and transfer j
of the containers, there were no clearly described expectations regarding the degree
of security oversight to be provided for the activity (e.g., inspection of the bottom
of the transfer shield or opening and inspection of non-sealed small areas).
.
The acting Security Manager agreed that inspection of the containers could be I
enhanced and suspended transfer of the high integrity containers into the protected
area pending establishment of additional guidance for conducting an inspection of
the containers. The acting Security Manager stated that this additional guidance
would be added to the security training program to ensure that security personnel
met the revised expectations regarding review of the container loading and closure
operations.
c. Conclusions
PECO provided security oversight of high integrity containers transferred into the
Protected Area. However, clearly described expectations regarding the degree of
security oversight of this activity was not fully provided. PECO enhanced
inspection guidance and added the revised expectations to the security training
. program.
1
V. Manaaement Meetinas l
X1 Exit Meeting Summary
l
The inspectors presented the results of the inspection to members of the licensee
management on September 23,1998. The licensee acknowledged the findings I
presented. No proprietary information was identified by the licensee.
l
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._ - _ . _ . _ _ . - _ _ _ . _ _
e
.
31
X2 Review of Updated Final Safety Analysis Report (UFSAR) Commitments
, A discovery of a licensee operating their facility in a manner contrary to the
Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a
special focused review that compares plant practices, procedures and/or parameters I
to the UFSAR descriptions. While performing the inspections discussed in this
report, the inspectors reviewed the applicable portions of the UFSAR that related to
the areas inspected. The inspectors verified that the UFSAR wording was
consistent with the observed plant practices, procedures and /or parameters.
1
I
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3
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ATTACHMENT 1
LIST OF ACRONYMS USED -
~
'
'AO[ abnormal operating i
AR - action request
,
BTP Branch Technical Position
CM. Corrective Maintenance
CRD Control Rod Drive
'
DOT- Department of Transportation
ECR Engineering Change Request
ESF Engineered Safety Feature
FCR Field Change Request ,
FIN Fix-It-Now
GP general procedure ,
'
ISFSI independent spent fuel storage installation
ITS Improved Technical Specifications
- LCO. limiting condition for operation
j LER licensee event report
j. LOCA. loss of coolant accident
-
LSRO Limited Senior Reactor Operator
~
MCRD Main Control Room Deficiency
MOV _ motor operated valve
-NCV. non-cited violation
NOTICE notice of violation
i
PECO Peco Energy
'
[ PkCON Peco Nuclear
PEP performance enhancement program
PDR public document room ]
PMT Poct-Maintenance Testing
RO Reactor Operator *
RP radiation protection
RPM radiation protection manager
RT Routine Test
ST surveillance test
TS technical specification
TSA technical specification action
'UFSAR updated final safety analysis report - )
I
!
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__ __ _ .
,
S
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Attachment 1. 2
INSPECTION PROCEDURES USED
IP 37551 Onsite Engineering Observations
IP 60851 Design Control of IFSFI Components
IP 60853 On-Site Fabrication of Components and Construction of an IFSFl
IP 61726 Surveillance Observations
IP 62707 Maintenance Observations
IP 71001 Licensed Operator Requalification Program Evaluation
IP 71707 Plant Operations
IP 71715 Sustained Control Room and Plant Observation
IP 71750 Plant Support Observations
IP 84750 Radioactive Waste Treatment, and Effluent and Environmental Monitoring
IP 86750 Solid Radioactive Waste Management and Transportation of Radioactive
Materials
IP 92903 Followup - Engineering
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
50-278/98-08-01 VIO RWCU System Startup Procedure
Ooened/ Closed
50-277/98-08-02 NCV Torus /Drywell Vacuum Breaker Loss of Seated
Indication (Unit 2)
50-277/98-08-03 NCV Potential for Bypass of Pressure Suppression Pool
50-278/98-08-03 NCV Potential for Bypass of Pressure Suppression Pool
50-277/98-08 04 VIO Failure to Adhere to Radiation Protection Procedures for
Source Checking Instruments
50-278/98-08-04 VIO Failure to Adhere to Radiation Protection Procedures for
Source Checking Instruments
Qlosed
50-277/97-03-02 VIO Failure to Assure that the Turbine Building Atmosphere
was Processed Through the Turbine Building Gaseous
Waste Treatment System
50-278/97-03-02 VIO Failure to Assure that the Turbina Luilding Atmosphere
was Processed Through the Turt 'e Building Gaseous
Waste Treatment System
50-277/97-04-03 VIO Violation of Locked High Radiation Area Key Control
50-278/97-04-03 VIO Violation of Locked High Radiation Area Key Control
50-277/2-97-007 LER Potential for Bypass of Pressure Suppression Pool
50-278/2-97-007 LER Potential for Bypass of Pressure Suppression Pool
50-278/3-98-004 LER Reactor Water Cleanup System Automatic Isolation
50-277/97-06-03 URI Potential for Bypass of Pressure Suppression Pool
50-278/97-06-03 URI Potential for Bypass of Pressure Suppression Pool
. . -. . ~ . . - - - _ , . . . - - - - _.
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ATTACHMENT 2
Maintenance Observation _g: Observed On:
M-018-003 New Fuel Receipt and Inspection August 23
M-053-011 Cleaning and Inspection of Powell August 31 :
Series P-51000 Metal-Clad Switchgear l
C0183052 Refueling Water Pump B - Inspect / Repack September 10
Seals
C0182395 Recirculation Motor Generator Oil Cooler September 12 ,
Setpoint Change 1
1
M-056-001 480 Volt Motor Control Center September 16
Circuit Breaker Assembly and
Cubicle Terminal Maintenance
2
Surveillance Observations: Observed On:
TRT #98-025 Reactor Water Cleanup System 16A Valve August 24
'
Troubleshooting (Unit 2)
ST-O-052-704-2 E4 Diesel Generator 24 Hour August 26 i
Endurance Test
Sl2K-54-E32-XXFM Functional Test of E32 4KV Septembe'r 3
Undervoltage Relays
RT-O-40C-530-2(3) Drywell Temperature Monitoring September 16
ST-O-052-413-2 E3 Diesel Generator Fast Start September 16 !
and Full Load Test {
TRT #98-050 2B Loop of HPSW,2B Loop of RHR September 16
in S/D Cooling, Unit 2 ILRT Valve
RT-O-003 990-2 Control Rod Stroke Speed September 17
ST-O-10-306-3 B RHR Loop Pump Valve Flow September 17
and Unit Cooler Functional and !
Inservice Test
i
Sl3N-60B-RBM-AICS Calibration / Functional check of September 20
Rod Block Monitor "A"
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