ML20133C493
| ML20133C493 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 12/09/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20133C475 | List: |
| References | |
| 50-440-96-06, 50-440-96-6, NUDOCS 9701070199 | |
| Download: ML20133C493 (22) | |
See also: IR 05000440/1996006
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U. S. NUCLEAR REGULATORY COMMISSION
REGION lli
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Docket No: 50-440
i License No: NPF-58
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Report No: 50-44')/96006 ,
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Licensee: Centerior Service Company
Facility: Perry Nuclear Power Plant ;
Location: P. O. Box 97, A200 '
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Perry, OH 44081
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Dates: July 27 - September 14,1996
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Inspectors: D. Kosloff, Senior Resident inspector
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R. Twigg, Resident inspector
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C. O'Keefe, Resident inspector (Fermi)
. K. Zellers, Resident inspector (Davis-Besse)
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, Approved by: J. M. Jacobson, Chief, Projects Branch 4 ,
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Division of Reactor Projects
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9701070199 961209
PDR ADOCK 05000440
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,. EXECUTIVE SUMMARY
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- Perry Nuclear Power Plant, Unit 1
. NRC Inspection Report 50-440/96006
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This inspection included aspects of licensee operations, engineering, maintenance, and
j plant support. The report covers a 7-week period of resident inspection.
, Doerations
j e. Conduct o' operations continued to be professional and safety-conscious. The
j operators continued to consistently repeat back and acknowledge oral
j communications related to plant operations (Section 01.1).
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i e Reporting of abnormal conditions improved. Operator response to abnormal plant
conditions continued to be prompt and appropriate. Communications with
- engineering improved. Identification of the suppression pool level increase
i demonstrated an appropriate questioning attitude (Section 02.1).
e A power reduction to identify a main condenser tube leak was well controlled and
the failure to locate the leak was recognized as an opportunity to improve
performance (Section 01.2).
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- The licensee continued to use a variety of self-assessment techniques to identify
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issues that required corrective actions. The licensee recognized weaknesses in its
corrective action process and continue <1 to pursue improvements in that process
{ (Section 07.1).
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1 Maintenance
- . e The licensee continued to aggressively pursue corrective actions for previously
I identified weaknesses in work planning and timely execution of work. This led to
improved performance (Section M1.1).
e An unresolved item was identified related to the Division 3 EDG air-start system
which required NRR assistance for future resolution (Section M2).
e inspectors identified a procedure violation with two examples. They involved the
improper storage of M&TE and scaffolding components (Section M4.2).
Enaineerina
e A personnel error allowed improper use of backfill for protection of Emergency
Service Water piping. This was a non-cited violation (Section E4.1).
e Previously identified personnel errors caused a modification to alter the operation of
the RCIC system in a way that had not been conveyed to the operators. This was
a non-cited violation (Section E5.1).
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l e Personnel error caused six sets of electrical relay contacts to be omitted from a
j surveillance test. The contacts were subsequently tested satisfactorily however,
j the earlier failure to test the contacts was a violation of Technical Specifications
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Plant Suonort
- e A hot short concern related to motor operated valve operation was reviewed
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Resort Details
Summarv of Plant Statug i
. The plant operated at full power throughout the inspection period except for short power
reductions for testing, control rod realignments, and an attempt to identify a condenser
tube leak.
I. Operations
01 Conduct of Operations
01.1 General Comments (71707)
Using Inspection Procedure 71707, the inspectors conducted frequent reviews of
ongoing plant operations. In general, the conduct of operations continued to be
professional and safety-conscious. The operators continued to consistently repeat
bad and acknowledge oral communications related to plant operations.
01.2 Resoonse to h'.i Condenser Leak
a. Insoection Scone (71500, 71707. 92901)
On September 2 at about 2:30 p.m., with the plant at full power, chemistry
personnel observed high conductivity readings in the hotwell. The operators
concluded that there was a condenser tube leak of circuhting (lake) water into the
hotwell and reduced plant power to 80 percent to maintain plant water chemistry
within limits. The inspectors observed control room operations at various times
during the associated power changes and leak isolation process.
b. Observations and Findinas
The next day the licensee decided to reduce power to 60 percent in en
unsuccessful attempt to locate and repair the leak. The licensee maintained power
at about 60 percent from September 3 until September 4 while attempting to locate
the leak. However, some time after the pewer reduction started, the leak had
resealed. Power was increased to 100 percent by September 4 at 5:15 p.m.
Because of the failure to locate the leak, the licensee decided to develop a plan to
complete leak location prerequisites more promptly to increase the probability of
locating a leak once power was reduced.
c. Conclusions
The inspectors concluded that the evolution was well controlled and the failure to
locate the leak was recognized as an opportunity to improve future performance.
No further leakage had been identified by the end of the inspection period.
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02 Operational Status of Facilities and Equipment
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02.1 Plant EnuinmentEnficiencies
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a. Insosction Scone (71500. 71707. 92901)
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l The inspectors performed routine inspections of the plant and observed a few minor 1
i equipment deficiencies that had not been previously identified by the licensee. The
j - inspectors observed operator actions in response to two self-identifying conditions.
l b. Observations and Findinas ,
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- The inspectors observed that licensee personnel continued to identify potential
- equipment deficiencies with def;ciency tags and potential issue forms (PlF).
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- On August 23 at about 12:20 p.m. the control room operators observed several
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annunciators illuminate for the rod control and information system, the reactor core
isolation cooling system, and the low pressure core spray system. The operators
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notified the inspectors who responded to the control room. The inspectc<s verified
- that the licensee had taken remedial actions and identified the proximate cause of
l the annunciated conditions, a broken lug for a fuse that interrupted power to a set
, of electronic trip units for various components. Although the trip units failed to
i their downscale positions, thereby causing incorrect logic signals to be provided to
j some safety equipment, none of the safety equipment was actually degraded. The
j inspectors observed that the operators promptly took appropriate compensatory
- actions and were aware of additional manual actions they would have to take for
F various plant transients. Also, the inspectors observed appropriate management
1- oversight and support from other organizations. The inspectors monitored repair
i activities and equipment testing, and verified that the operators completed prcper
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restoration of all equipment functions by about 5:48 a.m. the next day.
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On August 31 at about 8:00 a.m. the oncoming operations shift observed that
j suppression pool level had increased about 1 inch in the previous 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. They
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continued to monitor and record suppression pool level while they searched for the
- source of water. They could find no direct indications of the inleakage, but did note
i that the level increase stopped about the same time as 82 control rod hydraulic
l control units had their pressures reduced from about 1800 pounds per square inch
! (psi) to about 1750 psi. The system pressure had increased as a result of placing a
j rebuilt control rod drive hydraulic pump in service. The operators reported their
- observations to their management and to engineering. The licensee's investigation
j determined that the increased pressure had not caused the level increase and at the
j end of the inspection period the source of the water had not been identified.
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c. Conclusions
j General reporting of abnormal conditions improved and operator response to
. abnormal plant conditions continued to be prompt and appropriate.
j Communications between engineering and operations improved. The identification
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of the suppression pool level increase demonstrated an appropriate questioning
attitude even though the suspected cause was not validated. i
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07 Quality Assurance in Operations !
07.1 Licensee Self-Assessment Activities (40500)
a. Inspection Scope
The inspectors reviewed the following self-assessment activities that addrestM i
multiple functional areas, as well as operations
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- Licensee routine managers' meetings l
- Licensee management meeting to discuss deficient fill around underground !
emergency service water piping
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Potential issue forms (PlF) i
b. Observations and Findinas
The meetings were attended by appropriate personnel and there was substantive
discussion of specific issues and general methods of improving the corrective action
process. More than 300 PlFs were written by a variety of personnel who I
represented a wide cross section of plant organizations.
c. Conclusions
The licensee continued to use a variety of self-assessment techniques to identify
issues that required corrective actions. The licensee recognized weaknesses in its
corrective action process and continued to pursue improvements in that process.
08 Minellaneous Operations issues (92720,92901,92902,92903)
QL1 (Closed) Violation 50-440/94011-01: " Interruption of Shutdown
Cooling." This event was caused mainly by personnel error (inattention to detail)
during the preparation, review, and approval of a work order. The inspector
reviewed the corrective actions which included improvements to the design change l
program, development of a Work Managernent Section, revision of the fuse policy, !
and personnel training. The inspectors concluded that the corrective actions were
appropriate.
08,2 (Closed) LER 50-440/94-017-00: " Inadequate work order results in RHR B
Shutdown Cooling system isolation." This LER was for the same event discussed
in item 08.1 above, and is closed on that basis. l
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! QL3 (Closed) IFl (50-440/96005-09): On July 8,1996, while testing a program revision j
l for the simulator computer, a scenario file was inadvertently started. The scenario ;
file affected computer screen indications in the control room such as control rod l
positions and core thermal power averages. The operators determined the 1
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information to be inaccurate based on control panel indications and plant
performance. Through discussions with plant operators and simulator operators,
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the inspectors concluded that the simulator computer's capability to alter control
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room indications was limited ts acn-ccfsty related indecations. Corrective actions
j included adoitional computer graphics to indecats to the user that the simulator
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computer was monitoring control room indications. ;
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j .QBA (Closed) LER 50-440/93-014-00: "RACS Power Supply Failure Causes Technical
4- Specification 3.0.3 Entry." This LER was closed in IR 9tlOO4, but was erroneously
j described as LER 50-440/95-014-00. :
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l II. Maintenance ;
} M1 Conduct of Maintenance l
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! M1. ? General Comments
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- a. Insoection Scone (62703, 61726. 92902)
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The inspectors observed all or porties of the following work and surveillance
testing activities during Division 1 and Division 3 online system outages:
- R85-13OO6 Lube, megger, and perform general inspection of motor control
center for the emergency service water pump, Division ill <
- 95-4613, E21COOO1 Low Pressure Core Spray pump relay replacement ;
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- S85-10209, SVI-E22-T12OO High pressure core spray (HPCS) pump-
discharge pressure high channel functional test
- 95-1565 Division lil Diesel, recalibrate oil temperature switch
- 96-1478 Division lli Diesel, retorque inspection covers
The inspectors observed all or portions of the following additional surveillance tests:
- SVI-B21-TO212-A ATWS-RPT Reactor Vessel Pressure High Division i
Channel A Ft,actional Test for B21-N403A Analog trip Module
- SVI-E31-TOO86B NUMAC LDM Calibration for E31-N700B
- SVI-821-TO369A SRV Pressure Actuation Channel A Functional Test i
for 821-N668A
b. Observations and Findings _
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The inspectors found that the observed work activities were well coordinated with
appropriate supervision present. The craft assigned to perform a " bucket check"
for the Division 3 Emergency Servica Water Pumn electrical breaker were l
knowledgeable of potential deficiencies to look fur, such as cracked or damaged !
fuse clips. The craft were electrically safety conscious, having verified the tag out j
and metered the breaker to ensure it was doenergized. Procedures for the work
activities were present and were used. Appropriate documentation, such as
operation's approval to conduct the work, were inc!uded in the work packages.
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Responsible system engineers were observed to be engaged with the work
activities. This included temporary assignments as online divisional system outage
coordinators. A major activity, replacemer't of an air regulator on one of two l
subsystems for the air start system on the Division 3 Emergency Diesel Generator
(EDG), is discussed in Section M2.
The Division 1 system outage was completed about 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> later than scheduled.
The licensee evaluated its performance during this outage with a thorough formal
post-outage critique. The Division 3 outage, two weeks later, was completed on
schedule. A formal post-outage critique was held fc! that outage as well. Prior to
each outage some planned work items had to be deleted because not all preoutage i
requirements could be completed. !
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Surveillance procedures were appropriate for testing conditions and correctly
l identified parameters needed to verify proper equipment performance. Test ;
personnel followed the procedures during the observed testing and equipment
problems were promptly identified and resolved. '
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! c. Conclusions
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! All activities observed were conducted in a professional manner, tagouts were
appropriately completed, and engineering and management involvement was
apparent. The licensee continued to aggressively pursue corrective actions for
! previously identified weaknesses in work planning and timely execution of work.
l This led to improved perfc~ nance in the divisional system outages.
M2 Maintenance and Material Condtion of Facuttles and Equipment
a. insoection Scone (61726,92902)
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The inspectors evaluated the licensee's corrective actions for an observed increase
in Division 3 Emergency Diesel Generator (EDG) surveillance test start times.
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b. Observations and Findings
in May 1996 the Division 3 EDG start time increased from an average of about 9
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seconds to about 9.7 seconds and remained constant for each monthly start until i
, August. The technical specification (TS) start time limit was 10 seconds. Systems !
l Engineering concluded that either an EDG governor or air-start system problem had i
l- developed. Testing equipment monitoring the air-start system during the August
j premaintenance EDG start revealed a malfunctioning air regulator in one of the air-
l start subsystems thus rendering that subsystem inoperable. The regulator was
replaced and the EDG start time returned to the expected range. The RSE was
effective in locating and correcting the degraded regulator. However, the start time
records indicated that the regulator had been malfunctioning since May,1996, and
called into question the timeliness of the identification. i
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On &ly 14,1996, the licensee began using improved TS (ITS) based on NUREG-
1434, " improved BWR-6 Technical Specifications." The ITS included a new
specification (3.8.3) for the air-start system that was not part of the previous TS.
The ITS bases for 3.8.3 stated that " Division 1,2, and 3 have two independent air
start subsystems per DG. For Division 1 and 2 DGs, gia air start subsystem for an
engine is required for OPERABILITY of each DG" and the licensee tested the EDG 1
and 2 air-start subsystems independently. Bases 3.8.3 continued, "For the Division
3 DG, bgg air start subsystems are required for OPERABILITY" and the licensee
tested the EDG 3 subsystems together. The inspectors were concerned that the
plain meaning of the bases statement appeared to be that both air-start subsystems
for EDG 3 were required to be operable for EDG 3 to be operable. Therefore, once
the ITS became effective, with one air-start subsystem degraded and inoperable,
the inspectors considered the Division 3 EDG inoperable until the defective regulator
was replaced.
The licensee stated that the ITS bases statement merely incorporated information
that hM been in licensee Technical Specification Position Statement (TSPS) No.
063. .icensee considered the EDG to be operable as long as it started in 10
seconds. Therefom, the licensee did not conclude that the degraded condition i
made the EDG incperable. l
A review of the UFSAR revealed apparent inconsistencies. Section 9.5.6.4 air-start
system description for EDGs 1 and 2 was identical to the UFSAR Section 9.5.9.3.4
description of the EDG 3 air-start system. However, the air-start system design for
EDG 3 was significantly different from the air-start systems for EDGs 1 and 2.
Additionally, section 9.6.5 of the May 1982 NUREG-0887 plant Safety Evaluation
Report, May 1982, described the air-start systems in a manner that was unclear to
the inspectors in light of the actual design of the systems.
c. Conclusions
The inspectors' understanding of the meaning of the TS basis statement on Division
l 3 EDG air-start systems differed from the licensee's. The inspectors requested NRR
to provide the proper interpretation of the TS basis statement. Until the inspectors >
receive a.. answer, the past operability of the Division 3 EDG will be considered an 1
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unresolved item URI (50-440/96006-01(DRP)).
M4 Maintenance Staff Knowledge and Performance
M4.1 Control of Measurina and Test Eauioment Control (M&TE) .
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a. Insoection Scoos (62703. 92902)
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During a general plant inspection the inspectors identified a FLUKE multimeter that i
was unattended.
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b. Observations and Findmgs
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Because the multimeter was used for testing of safety-related electrical
components, it was calibrated and controlled under the licensee's procedure for
Control of Measuring and Test Equipment, PAP-1201, Rev. 7, effective December
l 19,1995. The procedure was intended to achieve and maintain a high degree of
l confidence in the accuracy of safety-related measurements. Section 6.3.1, step
3.c of PAP-1201 required that M&TE be stored in a secure area when not in use or
in the controlled issue area. The inspectors found the multimeter on top of an air
- distribution header for plant serv 6e air, which was not a secure area. Section 6.3
l also required documented control of M&TE when issued to individuals for use. The
licensee determined that the multimeter was documented as being in the controlled
issue area.
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M4.2 Scaffold Storage
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a. Insnection Scone (62703,92902)
l On September 5,1996, the inspectors accompanied the plant manager on an
l inspection of the plant. The inspectors observed some of the same areas of the l
! plant on September 10,1996.
b. Observations and Findings
l On September 5,1996, the inspectors observed one scaffold component stored
outside of the designated scaffold storage area on the 599 foot elevation of the
l intermediate building. The area boundaries were defined by Field Clarification
! Request (FCR) 13745, approved May 22,1990. The boundaries were clearly
- marked on the floor. This observation was shared with the plant manager. On
September 10,1996, the inspectors again observed scaffold components (4 pieces)
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at the same unauthorized location, which was within 1 meter of a safety related
valve.
c. Conclusions on Maintenance Staff Knowledae and Performance
l The safety consequences and the potential safety consequences of the improper
storage of M&TE and scaffolding components were minor. However, the regulatory
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consequences were of more than minor significance because the licensee had
identified repeated problems in both areas and had taken corrective actions which
were not fully effective. Therefore, the improper storage of M&TE and scaffolding
components were examples of a violation of 10 CFR Part 50, Appendix B, Criterion
V, which requires that activities affecting quality be prescribed by documented
instructions and procedures and be accomplished in accordance with those
instructions and procedures (50-440/96006-02a&b(DRP)).
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E2 Enginsedng Support of Facilities and Egidpnient
i E2.1 Review of Undated Final Safety Analysis Ranort (UFSAR) Commitments
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) a. Insoection Scone (37001. 92903)
While performing the inspections discussed in this report, the inspectors reviewed
I applicable portions of the UFSAR that related to the areas inspected. The
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inspectors also reviewed situations that the licensee had identified as potential
- inconsistencies between wording of the UFSAR and actual plant practices, ,
- procedures, and parameters. !
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! b. Qhagrvations and Findinas
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- E2.1.1 UFSAR Table 3.5-7, " Safety-Related Systems / Components Located
l Outside Seismic Category I Structures," listed seven
! systems / components with brief descriptions of now they were
protected from external missiles. The table stated that " Emergency l
i Service Water Piping" was protected by being covered with !
l " compacted ear fill" instead of " compacted earth fill" as other listed
} systems / components were. The inspectors concluded that this was a j
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minor typographical error which does not merit further followup. i
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j E2.1.2 UFSAR Figure 9.2-13, " Condensate Transfer and Storage System" !
! was a copy of plant <trawing 302-102. The licensee determined i
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during a UFSAR update review that drawing 302-102 was changed in
l June 1993 by Drawing Change Notice 4199 without a required
i UFSAR change request and safety evaluation. The licensee
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completed a safety evaluation which was reviewed by the onsite
! review committee on September 5,1996. This issue will be
i raconsidered in the future and is an unresolved item URI (50-
440/96006-03).
E2.1.3 UFSAR Section 9.5.9.2.4, " Inspection and Testing Requirements" for
l the Division 3 EDG describes the air-start system incorrectly. See
j Section M2b of this report. This description appeared to be part of
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- the original issue of the UFSAR. This issue will be reconsidered in the
! future and is an unresolved item URI (50-440/96006-04).
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l E2.1.4 UFSAR Section 17.2.1.3.5.1, " Company Nuclear Review
j Board (CNRB)" describes the compodion and functions of the CNRB.
j The licensee determined that the CNRB had delegated some of its
j review functions to subcommittees. That delegation was not
l specifically described in the UFSAR and the licensee was evaluating
whether its actual practices met the intent of the UFSAR description.
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l This issue will be reconsidered in the future and is an unresolved item
URl (50-440/96006-05).
c. Conclusions
The inspectors will evaluate the findings and determine whether a violation of NRC
requirements has occurred.
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! E4 Staff Knowledge and Performance
E4.1 Serv ce Water (SW) Modification Imnacts Ememer.cv SW Tornado Protection
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i An inspector responded to the site on August 3,1996, to evaluate the licensee's
immediate corrective actions following their discovery that improper backfill had
i been used for temporary protection of a small portion of emergency SW (ESW)
! piping from postulated tomado generated missiles. The inspector also observed a
licensee management meeting about the issue and inspected the area where the f
l ESW piping was improperly covered and other site areas where ESW piping was
propeny covered.
b. Observations and Findinas
l On August 2,1996, the licensee determined that uncompacted safety-related Class
l A engineered backfill (a course sand mixture) temporarily placed over about 3
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meters of underground safety-related ESW piping was required to be compacted.
About 1.2 meters of compacted fill was required by Table 3.5-7 of the UFSAR.
This requirement was in engineering documents that had been given to the
construction contractor performing the nonsafety-related SW work. The
engineering documents had been completed by an engineering contractor. The SW
! piping being installed by the contractor was being connected to an abandoned Unit
2 underground ESW pipe. Ttat pipe was parallel to a Unit 1 underground ESW pipe
buried at the same elevation. The contractor excavated an area about 3 metem
square down to the Unit 2 pipe, which was about 4 meters deep. They did not
uncover the Unit 1 ESW pipe but left only about 0.6 meters of compacted fill above
and to the side of the pipe. Whenever direct access to the pipe was not required
l the contractor partially filled the excavation by placing about 0.7 meters of
uncompacted fill over the compacted fill. The contractor also maintained the {
capability to place the uncompacted fill within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of declaration of a tomado
watch. The licensee determined that since there was no analysis that
! demonstrated that the uncompacted fill could protect the ESW pipe from tornado
generated missiles, the associated train of ESW would have been declared
l inoperable if a tornado watch were to have been declared. That was an interim
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measure while the licensee developed an independently verified engineering analysis
that identified an acceptable method of temporary backfill that could be practically
j removed for construction activities. On August 4,1996, the licensee concluded
l that, for temporary missile protection,1.2 meters of uncompacted crushed rock
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l was equivalent to 1.2 meters of compacted engineered backfill, and placed that
material in the excavation. No tornado watch was declared during the inspection
period. .
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c. Conclusions j
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! The personnel error in translating the engineering information into an actual work j
l practice that allowed the improper use of backfill had no actual safety consequence l
l because these was no tornado watch while the violation existed. The potential
l safety consequence was minor because the inspectors concluded, after discussion
with Region lli menagement and staff, that the size and shape of the excavation
minimized the probability that a tornado generated missile could have damaged the
ESW pipe. However, the regulatory significance is more than minor because, in
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slightly different circumstances, a similar communications error could have more
significant consequences. The communications error appeared to be the result of
the way the backfill information had been presented in the engineering documents
prepared by the engineering contractor. The licensee's planned corrective actions
included a request for the engineering contractor to evaluate other engineering work
for similar problems and its programs for creating and reviewing its engineering
documents. The licensee also planned to train its engineering staff on the
communications error and its potential consequences. Procedure PAP-0305,
" Safety Evaluations," Rev. 7, effective September 1,1993, step 3.61. required
responsible / sponsoring managers to " Ensure required 10CFR50.59 Applicability
Checks and Safety Evaluations, ... are performed for necessary items prepared
within their section." The communications error that allowed the uncompacted fill
to be used (an inadvertent temporary change to the facility as described in the
l UFSAR) prevented step 3.61. from being fulfilled. Failure to fulfill step 3.61. is a
l
violation of TS 6.8.1 which required written instructions to be implemented for the
activities recommended in Appendix A of Regulatory Guide 1.33, Rev.1, February
1978. This licensee identified and corrected violation is being treated as a Non- .
Ci;ad Violation NCV (50-440/96006-06(DRP)) consistent with Section Vll.8.1 of '
the NRC Enforcement Policy (60 FR 34380, June 30,1995).
E5 Enaineerina Staff Trainina and Qualification
E5.1 Modification Alters Safety-Related System Response
a. insoection Scone (37551. 92903)
The inspectors reviewed the licensee's evaluation of Unresolved item 96005-03
l which was opened as the result of a self-identified concern with a modification to
the Leakage Control System completed during the recent refueling outage (RF05).
The modification had been controlled by Design Change Package (DCP) 87-0725.
L The inspectors reviewed the human performance enhancement systems (HPES)
evaluation report that was completed as part of the associated PlF investigation.
The inspectors also discussed the results of the investigation with licensee
personnel.
l 13
1
l
l
t
. , _ , , ,c - . -,-
- . - . _ . . - .. - _ - . - -- - - . -- -._-- - - _---.....- - -.. -
..
,
- *
I
? l
5 !
j b. Observations and Findinas l
! !
During a review of simulator operations the hcensee observed that the Reactor Core
j isolation Cooling (RCIC) system isolated during a simulated loss of electrical power
to the General Electric (GE) Nuclear Measurement Analysis and Control (NUMAC) !
j steam leak detection monitor equipment. Prior to installation of the NUMAC
l equipment, RCIC did not isolate on a loss of power. Licensee personnel
{ appropriately documented their nbservation with PIF 96-2568. The licensee l
'
promptly investigated the obser.Jd change and found that power monitoring relays,
- which prevented RCIC isolation in the previous design, had been removed when the
} NUMAC equipment was installed. The licensee verified that RCIC remained
l operable based on its continuing capability to perform its Technical Specification
- required safety functions and because RCIC had no design basis safety function ;
l relative to the UFSAR accident analyses. However, the changes introduced by DCP l
l 87-0725 were of concern because RCIC, a safety-related system, would no longer
have performed in the manner that operators had been trained to expect under l
l certain conditions. '
!
l The licensee promptly recognized the potential impact of such unintended changes
and began an investigation to identify the extent of the condition and the causes.
i
) Although most of the engineering work for DCP 87-0725 had been completed in
{ 1992 and 1993, the licensee selected 13 recent DCPs (including 87-0725) for
j review for similar problems. Additional personnel errors were identified in the
! design or implementation of three of the DCPs. The inspectors reviewed the errors
j and concluded that they were not programmatic. One error was not an engineering l
{ error, but an error in the validation of an emergency operating procedure. Another
j involved the failure to identify an unusual heat transfer process. This item, DCP
j 94-0027 for emergency closed cooling system temperature control valves, was ,
- reviewed in IR 96008. The third error involved a communication failure related to l
i
'
the implementation of temporary compensatory actions during excavations for an
SW DCP. That item is discussed in section E4.1 above.
i
I The HPES evaluation identified four personnel errors that caused the design error
j and allowed it to remain undetected until after the modification had been placed in
service. The initiating error was a failure by GE to recognize the isolation function
i of the power monitoring relays, even though they had committed to provide the
, licensee with specific information about the relays. In checking the GE design,
i licensee engineers then made two independent errors by not identifying that GE had
j changed the design by deleting the power monitoring relays. These errors involved
- a failure to follow up on the earlier question to GE; the advantage of a good
j questioning attitude had been lost. The fourth error was a failure to provide a post-
l maintenance test that would have identified a failure mode that had been
{ overlooked during preparation of the DCP. The licensee was planning to further
i evaluate that error in conjunction with its evaluation of DCP 94-0027 for the ;
- temperature control valves. l
i
i
l !
14
i
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.
.
- , , -
. .. _ __ _ ._ _ _ _._._._ _. _ ___ __ __ _._ _ __ _ __ _
- .
,,
)
i
e
c. Conclusions
!
l The licensee's investigation was thorough, including the root cause evaluation and
the review of the extent of the condition. The inspectors concluded that the
licensee's programmatic corrective actions in 1995 for other design control
,
problems combined with additional corrective actions planned as a result of this
- newly discovered error should prevent recurrence. Nuclear Engineering Instruction
l NEl-0357, Rev. 9, effective November 20,1992, step 6.2.5, required that, for a
- DCP, the " Design Engineer / Project Coordinator ... obtain necessary design
4 review / verifications ... for the portions changed." The two personnel errors that
,
allowed the design error (an inadvertent change to the design) to remain undetected
, prevented step 6.2.5 from being fulfilled. Failure to fulfill step 6.2.5 is a violation
! of TS 6.8.1 which required written instructions to be implemented for the activities
l recommended in Appendix A of Regulatory Guide 1.33, Rev.1, February 1978.
This licensee-identified and corrected violation is being treated as a Non-Cited
i
Violation NCV (50-440/96006-07(DRP)) consistent with Section Vll.8.1 of the NRC
j Enforceinent Policy (60 FR 34380, June 30,1995).
j E5.2 Isolation Contacts Not Tested
i a. Insoection Scone (37551, 61726. 92903)
i
'
On August 20,1996, licensee personnel identified six sets of electrical relay
i contacts that had not been surveillance tested since the installation of the GE
! NUMAC modification in RF05. The inspectors reviewed the licensee's evaluation of
j this error and observed portions of the surveillance testing for the six relays.
i
j b. Observations and Findinas
!
l Technical Specification (TS) Surveillance Requirement (SR) 3.3.6.1.4 required that a
j channel calibration be performed on the primary containment and drywell isolation
- instrumentation every 18 months. TS SR 3.3.6.1.5 required that a logic system
- functional test be performed on the primary containment and drywell isolation
! instrumentation every 18 months. TS Table 3.3.6.1-1 identified the functions that
I required testing for SR 3.3.6.1.4 and SR 3.3.6.1.5. The following relay contacts
j were required to be tested to verify functions listed in TS Table 3.3.d.1-1: l
l
- 1E31-N702A-K5 contacts 9/5, Function 1.3.f l
} 1E31-N7028-K5 contacts 9/5, Function 1.3.f i
i 1E31-N702A-K6 contacts 9/5, Function 1.3.a
.
1E31-N7028-K6 contacts 9/5, Function 1.3.a i
j 1E31-N702A-K7 contacts 9/5, Function 1.2.b
j 1E31-N7028-K7 contacts 9/5, Function 1.2.b
!
-
The above TS SRs became effective on July 14,1996. The relay contacts listed
i - above, part of the leak detection system, had not been tested followir>g a
i modification to the system completed in February and March of 1996, and had not ,
i been tested in accordance with the old TS SR prior to the effective date of the I
!
i
j 15
__ _. _ _ _ _ .._ _ _ _ _ _ _ __ _ ___ _ ____ _.___ _ _ _ _
'
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.
, above TS SR. Therefore, once the above TS SR became effective, they were
l immediately required to be tested.
l The inspectors verified that all the untested contacts were associated with
i
temperature indication instrumentatien logic for isolation of piping following
postulated pipe ruptures, if temperatures were to exceed a preset limit in certain
,
rooms, the temperature indication signals from the affected temperature indicators
l would cause the logic relays to change state. As a result multiple sets of contacts
'
associated with the relays would then change state. The relays had been properly
tested and all sets of contacts on the relays had been properly tested except for
one set on each of six relays. It was possibic for one set of contacts to fail even
though the associated relay and other sets of ' contacts worked properly. The
untested contacts were required to be operable by TS Limiting Condition for
Operation (LCO) 3.3.6.1 as detailed in TS Table 3.3.6.1-1.
Upon discovering the untested sets of contacts the licensee complied with TS SR 3.0.3 which allowed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to perform a surveillance after discovery that a
surveillance had not been performed within its required frequency. The inspectors
verified that the surveillance testing of te untested sets of contacts was
successfully performed within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> J discovery. The testing confirmed that the
untested contacts had actually remali ad operable since they had been installed.
c. Conclusions
Once the untested contacts were discovered, the licensee took prompt and
appropriate actions to confirm their operability. However, the earlier failure to test
the contacts immediately upon the ITS becoming effective was a Violation (50-
440/96006-08(DRP)) of TS SR 3.3.6.1.4 and SR 3.3.6.1.5.
E8 Miscellaneous Enginsedng lasues (92720, 92903)
E8.1 (Closed) Unresolved item 50-440/96005-03: During a review of simulator
operations the licensee observed the t the Reactor Coro isolation Cooling (RCIC)
system isolated during a simulated loss of electrical power to the General Electric
(GE) Nuclear Measurement Analysis and Control (NUMAC) steam leak detection
monitor equipment. Prior to installation of the NUMAC equipment, RCIC did not
isolate on a loss of power accident. This item is discussed in section E5 and is
closed.
E8.2 (Closed) LER 50-440/93-015-00: " Unexpected Reactor Recirculation Pump Fast to i
Slow Speed Downshift." This LER was closed in IR 96004, but was erroneously
documented as LER 50-440/95-015-00.
!
!
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!
IV. Plant Sunnart !
F2 Status of Fire Protection Facilities and Equipment (71750)
F2.1 Review of Licensee Compensatory Actions for " Hot Shorts" issue
!
On September 10,1996, the NRC, including a NRR lead fire protection reviewer,
met with the licensee at the plant to discuss compensatory measures that would be l
, taken in response to the licensee's determination that safe shutdown motor l
'
operated valves (MOVs) could be susceptible to fire induced hot shorts. Based on 1
these discussions, reviews of guidance provided to plant operators, and walkdowns j
of affected portions of the plant, the NRC concluded that the compensatory actions l
were acceptable.
l
i
V. Manaaement Meetings
i X1 Exit Meeting Summary l
4
!
The inspectors presented the inspection results to members of licensee management after
'
the conclusion of the inspection on September 18,1996. The licensee acknowledged the !
'
findings presented.
l
The inspectors asked the licensee whether any materials examined during the inspection
should be considered proprietary. No proprietary information was identified.
a
,
'.
b
.
7
N
,
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PARTIAL LIST OF PERSONS CONTACTED
Licensee
D. C. Shelton, Senior Vice President
R. D. Brandt, General Manager Operations
N. L. Bonner, Engineering Director ,
L. W. Worley, Nuclear Services Director
W. R. Kanda, Nuclear Assurance Director
J. Messina, Operations Manager
l INSPECTION PROCEDURES USED
l lP 37551: Onsite Engineering
IP 40500: Effectiveness of Licensee Controls in Identifying, Resolving, and
Preventing Problems
IP 61726: Surveillance Observations
IP 62707: Maintenance Observation
IP 71500: Balance of Plant inspection
IP 71707: Plant Operations 1
IP 71714: Cold Weather Preparation
IP 71750: Plant Support Activities 1
IP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power
Reactor Facilities
i IP 92720: Corrective Action j
l IP 92901: Followup - Operations l
IP 92902: Followup - Maintenance
IP 92903: Followup - Engineering 1
IP 92904: Followup - Plant Support
'
IP 37001: 10 CFR 50.59 Safety Evaluation Program j
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
50-440/96006-01 URI Diesel start testing
50-440/96006-02 VIO M&TE and scaffold control
50-440/96006-03 URI UFSAR Figure 9.2-13, drawing change
50-440/96006-04 URI UFSAR Sec. 9.5.9.2.4, EDG air-start system description
50-440/96006-05 URI UFSAR Sec. 17.2.1.3.5.1, CNRB delegation of functions
50-440/96006-06 NCV Use of incorrect temporary backfill
50-440/96006-07 NCV GE NUMAC modification affects RCIC
50-440/96006-08 VIO GE NUMAC surveillance testing inadequate
l Closed
- 50-440/93014-00 LER RACS power supply failure causes TS 3.0.3 entry
l 50-440/93015-00 LER Unexpected recirculation pump downshift
'
50-440/94011-01 VIO interruption of shutdown cooling
50-440/94017-00 LER Interruption of shutdown cooling
50-440/96005-03 URI Modification affects RCIC operation
50-440/96005-09 IFl Simulator computer interaction with control room
18
_ _ _ . . _ - . _ _ _ _ . ___ _ .__ ._._ __ ______._.___._._ . ._ _ _ . .. .
'
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l 50-440/96006-06 NCV Use of incorrect tamporary backfill
l 50-440/96006-07 NCV GE NUMAC modification affects RCIC
Discussed
None
,
LIST OF ACRONYMS USED
i ATWS ANTICIPATED TRANSIENT WITHOUT SCRAM
- BWR BOILING WATER REACTOR
,
CFR CODE OF FEDERAL REGULATIONS j
j CNRB COMPANY NUCLEAR REVIEW BOARD
J
DCP DESIGN CHANGE PACKAGE
DIV DIVISION 4
! DG DIESEL GENERATOR I
i .DRP DIVISION OF REACTOR PROJECTS
i EDG EMERGENCY DIESEL GENERATOR
ESW EMERGENCY SERVICE WATER l
I FCR FIELD CLARIFICATION REPORT 1
4 FSAR FINAL SAFETY ANALYSIS REPORT l
GDC GENERAL DESIGN CRITERIA :
- GE GENERAL ELECTF!lC l
l HPCS HIGH PRESSURE CORE SPRAY
j HPES HUMAN PERFORMANCE ENHANCEMENT SYSTEM l
l IFl INSPECTION FOLLOW-UP ITEM )
2
ITS IMPROVED TECHNICAL SPECIFICATIONS
j LCO LIMITING CONDITIONS FOR OPERATIONS l
'
LER LICENSEE EVENT REPORT
, LDM LEAKAGE DETECTION MONITOR !
MOV MOTOR-OPERATED VALVE :
M&TE MEASURING & TEST EQUIPMENT
' NEl - NUCLEAR ENGINEERING INSTRUCTION
NRC NUCLEAR REGULATORY COMMISSION
. NRR OFFICE OF NUCLEAR REACTOR REGULATION i
l NUMAC NUCLEAR MEASUREMENT ANALYSIS AND CONTROL
l PAP PERRY ADMINISTRATIVE PROCEDURE l
i PDR PUBLIC DOCUMENT ROOM
' PlF POTENTIAL ISSUE FORM
,
PNPP PERRY NUCLEAR POWER PLANT
"
psi POUNDS PER SOUARE INCH
j RACS ROD ACTION CONTROL SUBSYSTEM
l RHR- RESIDUAL HEAT REMOVAL
i RI RESIDENT INSPECTOR
! RSE RESPONSIBLE SYSTEMS ENGINEER
$ SR SURVEILLANCE REQUIREMENT
i SRI SENIOR RESIDENT INSPECTOR
l; SW SERVICE WATER
TS TECHNICAL SPECIFICATION
i
19
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-.
_ _ - _ . _ . _ _ _ _ _ _ . _ _ . - _ . _ . . . _ . . _ _ _ _ . . _ . _ _ . . _ . _ . _ _ _ . _ _ . _ . .
'
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e
,
1
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} TSPS TECHNICAL SPECIFICATION POSITION STATEMENT
I
UFSAR UPDATED FINAL SAFETY ANALYSIS REPORT
URI UNRESOLVEDITEM
!
t
i
2
PARTIAL LIST OF DOCUMENTS REVIEWED DURING THE INSPECTION
4
j 95-1565 Division lil Diesel, recalibrate oil temperature switch
- 95-4613 E21COOO1 Low Pressure Core Spray pump relay replacement
a 96-1478 Division ill Diesel, retorque inspection covers
Charts on various control room chart recorders.
Computer printouts on various control room printers.
i Control Room Standing Orders, various dates
Control Room Daily Instructions, various dates
j Control Room Daily Instructions, Supplemental Reading, various dates
Control Room Annunciator Status Books, revisable format, various dates
I
DCP 95-0022
I Drawing D-302-102, Rev.CC, 6-4-96
i Excessive Radwaste Sump Inleakage Report - Dated 09/09/96
i GCl 0016, Scaffolding Erection, Modification or Dismantling Guidelines, Rev.1, 8-4-95
] Managers' Meeting Report - 07/29/96
l Managers' Meeting Report - 07/31/96
- Managers' Meeting Report - 08/02/96
l Managers' Meeting Report - 08/05/96
j Managers' Meeting Report - 08/07/96
Managers' Meeting Report - 08/09/96
. Managers' Meeting Report - 08/12/96
Managers' Meeting Report - 08/14/96
Managers' Meeting Report - 08/16/96 l
! Managers' Meeting Report - 08/19/96
l Managers' Meeting Report - 08/21/96
j Managers' Meeting Report - 08/23/96
Managers' Meeting Report - 08/26/96
Managers' Meeting Report - 08/28/96
Managers' Meeting Report - 08/30/96
Managers' Meeting Report - 09/04/96
Managers' Meeting Report - 09/06/96
Managers' Meeting Report - 09/09/96
Managers' Meeting Report - 09/11/96
Managers' Meeting Report - 09/13/96
MEMORANDUM - J. Kloosterman to IRT on PlF 96-2568,08/16/96 ;
SUBJECT: REQUESTED INPUT TO INCIDENT RESPONSE TEAM (IRT) FOR PIF 96-2568, i
Monthly ALARA Report, July 1996
Monthly ALARA Report, August 199S
Monthly Performance Report, Perry Nuclear Power Plant, July 1996 ,
Monthly Performance Report, Perry Nuclear Power Plant, August 1996
NEl-0357, Design Change Packages, Rev. 9, effective November 20,1992
NEl-0357, Design Change Packages, Rev.11, effective 2-1-95 I
20
l
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_ __ _ . . _ _ __.._ _ ___ _ _ _ __ _ _ _ ._ . _ _ ___ _ __
'
! !
1
4
l PAP 020'/, Conduct of Operations, Rev. 9, effective 3-28-95
4
PAP O204, Housekeeping / Cleanliness Control Program, Rev. 8, effective 9-1-95
PAP-0305, Safety Evaluations, Rev. 7, effective September 1,1993
- PAP 0909, Scaffolding Erection / Teardown Requests and Scaffold Tracking Program
j PAP 1201, Control of Measuring and Test Equipment
! Perry Daily Report - 07/30/96
i Perry Daily Report - 08/01/96
.
Perry Daily Report - 08/06/96
!
'
Perry Daily Report - 08/08/96
Perry Daily Report - 08/13/96
i Perry Daily Report - 08/15/96 q
j Perry Daily Report - 08/20/96 l
l Perry Daily Report - 08/22/93 '
i Perry Daily Report - 08/27/96
Perry Daily Report - 08/29/96
Perry Daily Report - 09/03/96
Perry Daily Report - 09/05/96
Perry Daily Report - 09/10/96
Perry Daily Report - 09/12/'46 l
PERRY Lines, September E.,1996
PERRY NUCLEAR POWEP. PLANT - RADIATION PROTECTION SECTION
Fifth Refueling - ALARA/Hf ALTH PHYSICS - POST OUTAGE REPORT
January 21,1996 to April 10,1996
Plant Loc Vol. 31, Page No. 49 - 98 l
PNPP Man of the Day - 07/26/96
PNPP Plan of the Day - 07/29/96
PNPP Plan of the Day - 07/30/96
PNPP Plan of the Day - 07/31/96
PNPP Plan of the Day - 08/01/96
PNPP Man of the Day - 08/02/96 l
PNPP Plan of the Day - 08/05/96 '
PNPP Plan of the Day - 08/06/96
PNPP Plan of the Day - 08/07/96 1
PNPP Plan of the Day - 08/08/96 i
PNPP Plan of the Day - 08/09/96 l
PNPP Man of the Day - 08/12/96 j
PNPP Man of the Day - 08/13/96
PNPP Man of the Day - 08/14/96
PNPP Plan of the Day - 08/15/96
PNPP Plan of the Day - 08/16/96
PNPP Plan of the Day - 08/19/96 !
PNPP Man of the Day - 08/20/96 '
PNPP Man of the Day - 08/21/96
PNPP Man of the Day - 08/22/96
PNPP Plan of the Day - 08/23/96
PNPP Man of the Day - 08/26/96
PNPP Man of the Day - 08/27/S6
PNPP Man of the Day - 08/28/96
21
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a*
l
l
l
PNPP Plan of the Day - 08/29/96
PNPP Plan of the Day - 08/30/96
PNPP Plan of the Day - 09/03/96
PNPP Plan of the Day - 09/04/96
PNPP Plan of the Day - 09/05/96
PNPP Plan of the Day - 09/06/96
PNPP Plan of the Day - 09/09/96
PNPP Plan of the Day - 09/10/96
PNPP Plan of the Day - 09/11/96
PNPP Plan of the Day - 09/12/96
PNPP Plan of the Day - 09/13/96
j PNPP Potential lasue Form No. 96-2590 through 96-2936
! PNPP Potential lasue Form No. 95-1372
l R85-13OO6 Lube, megger, and parform generalinspection of motor control center for the
! emergency service water pump, uvision til j
l Radiation Protection Section Organization Chart (Effective Date: 8/5/96) i
l RADIOLOGICAL AWARENESS - September 4,1996
!
S85-10209, SVI-E22-T12OO, High pressure core spray pump discharge pressure high
!
channel functional test
Site Weekly Dose Summary On-Line Quarterly Schedule Week 03W4
09/02/96 Through 09/08/96
!
SV!-B21-TO212-A ATWS-RPT Resctor Vessel Pressure High Division i Channel A
Functional Test for B21-N403A Analog trip Module i
SVI-E31-TOO86B NUMAC LDM Calibration for E31-N7008
SVI-B21-TO369A SRV Pressure Actuation Channel A Functional Test for B21-N668A
Temporary Modification Tracking Report, September 1996,9-1-96
l Unit Log - Unit 1 - Vol. 88, Page No. 31 - 150
Updated Final Safety Analysis Report
Various Equipment Deficiency Tegs
,
Various Active LCO Log Sheets
l Various Operations Administrative Control Tags
Various Operations information Tags
Various Potential LCO Log Sheets
Various Fire Extinguisher Inspection Tags
Various Radiologically Restricted Area Radiation Surveys
Various Safety Tags
l Radiation Work Permit 96006
Weekly Effluent and Release Rate Data Report, about August 21
Weekly Effluent and Release Rate Data Report, about August 28
I Weekly Effluent and Release Rate Data Report, about September 4
l Weekly Effluent and Release Rate Data Report, about September 11
Work Process Performance indicators, July 2,1996
l
<
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