ML20133G988
ML20133G988 | |
Person / Time | |
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Site: | Hatch |
Issue date: | 01/06/1997 |
From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
To: | |
Shared Package | |
ML20133G985 | List: |
References | |
50-321-96-14, 50-366-96-14, NUDOCS 9701170012 | |
Download: ML20133G988 (35) | |
See also: IR 05000321/1996014
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'U.S. NUCLEAR REGULATORY COMMISSION
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REGION II
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i Docket Nos: 50-321, 50-366
1 License Nos: DPR-57 and.NPF-5
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] Report No: 50-321/96-14. 50-366/96-14 i
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j Licensee: Georgia Power Company (GPC) l
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- Facility
- E. I. Hatch Units 1 & 2
Location: P. O. Box 439
i Baxley, Georgia 31513
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- Dates
- October 27 - December 7, 1996
! Inspectors: J. Moorman Senior Resident Inspector (Acting)
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E. Christnot. Resident Inspector
- J. Canady. Resident Inspector
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G. Kuzo. Senior Radiation Specialist ,
(Sections R1. R3. R5. R7. R8) !
j W. Kleinsorge. Reactor Inspector !
- J. Coley, Reactor Inspector
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, C. Rapp. Reactor Engineer
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i Approved by: P. H. Skinner. Chief. Projects Branch 2
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Division of Reactor Projects
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9701170012 970'106
PDR ADOCK 05000321
G PDR Enclosure 2
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EXECUTIVE SUMMARY
Plant Hatch. Units 1 and 2 -
NRC Inspection Report 50-321/96-14. 50-366/96-14
This integrated inspection included aspects of licensee operations,
engineering, maintenance, and plant support. The report covers a 6-week
period of resident ins)ection; in addition, it includes the results of
announced inspections )y a regional senior radiation specialist and two
regional reactor inspectors. An in-office review of open issues relating
to the Service Water System Operational Performance Inspection was also
conducted '
Ooerations
e The operator respon.se to the unexpected trip of the Unit 2 reactor
feed pump turbine was good. (Section 01.4)
e The inspectors concluded that the licensee was appropriately
monitoring the increased unidentified leakage rate observed in the
Unit 1 Drywell. (Section 01.5)
e Problem annunciators were resolved in accordance with the
appropriate procedures. The procedural reviews and evaluations
were conducted by appropriate operations personnel in a timely
manner. The supply of jumpers appeared to be adequately
maintained. (Section 01.6)
e The inspectors concluded that a review of the licensee's discovery
that a channel functional test of the reactor mode switch in the ;
shutdown control rod block function had not been performed that j
included all required TS surveillances. The inspectors also -
concluded that the missed test was of minor safety significance.
(Section 03.1)
e The licensee's activities in the area of self assessment
demonstrated a safety conscious attitude and aggressive
involvement by site management. (Section 07.1)
e Although the licensee's transfer program (Technical Specification
Improvement Program) was not a formal. proceduralized program, the
licensee appropriately transferred actions and surveillance from
the Technical Specifications to the Improved Technical
Specifications procedures, manuals or programs. A few exceptions
were noted. (Section 08.1)
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! Maintenance
i e Maintenance activities observed by the inspectors were thorough.
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i e The work activities observed during the Unit 2. Loop B. Core Spray
1 outage were conducted in a safe and timely manner. Housekeeping
- and implementation of Foreign Material Exclusion (FME) protection
i was excellent for the work activity performed on the Core Spray
l Test Valve. (Section M1.2)
e The licensee's preparations for cold weather were considered
j adequate. The inspectors noted a potential single failure
vulnerability in the freeze protection for the 1B Diesel
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Generator. (Section M2.1)
! e The licensee discovered that the instantaneous overcurrent trip
! setting for tne normal supply breaker to Motor Control Center
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1R24-5026 was improperly set. This is an example of a failure to
implement configuration control requirements. (Section M2.2)
e During observation of the Unit 1 High Pressure Coolant Injection
_ (HPCI) monthly surveillance test, the inspectors identified two
i potential deficiencies in the test procedure. These deficiencies
i are identified as Inspector Followup Item (IFI) 50-321,
366/96-14-04: Potential Deficiencies in the HPCI Surveillance
- Procedure. (Section M3.1)
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Enaineerina
I e A recently installed nitrogen system relief valve lifted during a
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tank filling operation. This is an example of Violation (VIO)
50-321. 366/96-14-03: Failure to Implement Configuration Control
Requirements. (Section El.1)
e The licensee successfully tested the Unit 1 Main Steam Isolation
Valves while at power to demonstrate that they did not contain a
defective part identified by a recent Part 21 notice.
(Section E2.1)
e The engineering group's discovery of a deficiency involving the
normal supply breaker to a 1B diesel generator motor control
center was excellent. The temporary modification to correct the
deficiency was installed in accordance with an approved process.
(Section E2.2)
Enclosure 2
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i e The inspectors concluded that administrative procedures clearly
- assign the system engineers the responsibility for trending
i repetitive equipment problems. Root cause analysis are performed
based on the category of events or at the discretion of ;
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management. (Section E4.1).
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e A loss of power to the Reactor Protection System Bus 1B was caused
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by contractor personnel error. Contractor personnel had miswired
- part of the current transformer circuit during a cable reroute.
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Engineering failed to find the error when the circuit was checked.
(Section E8.2)
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o Personnel conducting troubleshooting attempted to de-energize the
l 1G Torus-to-Drywell Vacuum' Breaker, by opening a control panel
- link, which inadvertently de-energized the 1H due to an error in
the electrical connection drawing. This is an example of a
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failure to implement configuration control requirements.
(Section E8.4)
Plant Suocort
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l e Radiological controls for high and very high radiation areas were
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maintained in accordance with TS requirements. Area postings and
- labels for containers of radioactive material were appropriate.
- Improvements were noted for general housekeeping and cleanliness
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relative to observations made during a previous inspection. ,
(Section R1.2) l
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e Excluding two packaging Quality Assurance (0A) issues. I
transportation and packaging activities for radwaste and material
shipments met 10 CFR 71.5 and 49 CFR 100-179 requirements. The
revised Department of Transportation (DOT) guidance was
implemented, as applicable. Radwaste characterization was in l
accordance with 10 CFR 61.55. Two concerns were identified for i
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radwaste packaging quality control documentation. (Section RI.3) i
e Effluent release documentation and radiological environmental
monitoring program results were prepared in accordance with Off-
site Dose Calculation Manual (0DCM) requirements. The release
data and environmental monitoring results verified offsite
releases and resultant doses were a small fraction of the
allowable limits. (Section R3)
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o Hazardous material (hazmat) training provided to personnel
handling radioactive materials was conducted at the appropriate
frequency, and included recent changes to D0T regulations. The
training provided was effective as demonstrated by the knowledge
and proficiency of interviewed staff members. Identified issues
observed included training data base upgrades and timeliness in
completion of hazmat examination. (Section RS)
Enclosure 2
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i e Audits for the radwaste and effluent processing programs were .
! thorough and comprehensive, and met TS. 10 CFR Part 20 and '
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10 CFR Part 71 requirements. .The low-level waste characterization
- study was identified as a lic '1see initiative to quantify and .
{ prioritize radwaste program activities. (Section R7.1)
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.e The licensee continues to experience delays entering the Emergency
l . Plan regarding toxic gases. During the nitrogen release event,
i they did not enter the Emergency Plan within a time frame
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considered reasonable. This is identified as a weakness.
(Section P4.1)
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! Enclosure 2
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Beoort Details
Summary of Plant Status
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Unit 1 began the report period at 100% rated thermal power (RTP) and
continued until November 3. when power was lowered to 70% to allow stroke
testing of main steam isolation valves. Power was returned to 100% the l
next day and was maintained at this condition for the remainder of the
report period except for routine testing activities. :
Unit 2 began the report period at 100% RTP. The unit operated at that
power level until November 17 when power was reduced to 60% for repair ;
work, rod sequence exchange, and scram time testing. The unit was l
returned to 100% on November 18 and operated at this power level
throughout the remainder of the report period except for routine testing
activities.
I. Doerationji l
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01 Conduct of Operations
01.1 General Observations (71707)
Using Inspection Procedure 71707, the inspectors conducted
frequent reviews of ongoing plant operations. In general, the l
conduct of operations was professional and safety-conscious:
specific events and observations are detailed in the sections
below. l
01.2 Use of Overtime
The inspectors reviewed the use of overtime and did not identify
any deficiencies. Appropriate measures were taken for the change i
from daylight savings time to standard time.
01.3 System Lineuos
The inspectors conducted a check of the major flow paths and
components of the Core Spray System for Units 1 and 2. The major
valves in the flow paths of the subsystems were in the correct
position for the operating condition. A limited walkdown of the ,
system was also performed to verify that the major manually
operated valves that are required to be locked by design drawings
were locked and in the correct position. The inspectors also
walked down the electrical lineup of the 1E 4160V/600V AC
electrical boards for Units 1 and 2. No discrepancies were found
during the walkdowns.
Enclosure 2
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01.4 Reactor Feed Pumo Turbine Trio
On November 17. during the power reduction for control rod .. ,
sequence exchange, the 2A Reactor Feed Pump Turbine (RFPT) tripped
due to. low bearing lube oil pressure. The low oil pressure
condition was caused by a clogged lube oil filter. No significant ;
plant transient occurred since reactor power was at 67% RTP. The '
operators responded properly, the standby filter was placed in
service, and the RFPT was returned to operation. Personnel
attempted to replace the bad filter element but were unable to ,
isolate it. At the end of the report period licensee personnel i
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were_ reviewing the specific circumstances involving the clogging
of the filter, the relatively rapid decrease in oil pressure, and
the inability to isolate the filter. The licensee is monitoring
the filter differential pressure on an increased frequency.
01.5 Unit 1 Drywell Unidentified Leakaae
The inspectors observed and reviewed the increasing Unit 1 drywell
unidentified leakage rate. The inspectors attended management
meetings at which the increasing leakage rate was discussed. The
inspectors review of the daily operator logs indicated that on
September 24. the leakage rate was 0.1 gpm and at the end of the
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report period the rate was approximately 1.4 gpm. There were no
indications or activities that have been identified which would
have caused this change. The highest change in leakage rate,
observed by the inspectors, appeared to be approximately 0.5 gpm
per month. At this rate, the TS limit for unidentified leakage. 5
gpm, will be reached during June.1997. The licensee was
appropriately monitoring the increased leakage rate and was
evaluation actions to correct the problem.
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01.6 Review of Annunciator Control Book and Jumoer Storace l
The inspectors reviewed the Annunciator Control Logs and conducted ;
- a limited inventory of the jumper storage cabinets for Units 1
and 2. This review and limited inventory was conducted in
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conjunction with a review of procedures 30AC-0PS-009-05. Control
Room Instrumentation. Revision 4. and DI-0PS-31-0596N. General i
Guidelines for Use of Jumpers and Links. Revision 0.
The inspectors verified that the active control room prob'lem
annunciators were properly identified and labeled. Documentation
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was present in the Annunciator Control Books indicating that the
j Ar.nunciator Control Sheets were reviewed by Operations Supervision
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on a monthly basis. The documentation also indicated that the
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Manager of Engineering Support had been notified of problem
annunciators or evaluations older than three months and the Plant
! General Manager had been notified of those that were older than
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Enclosure 2
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six months. The appropriate compensatory actions were documented
where applicable.
With the assistance of the SS, the inspectors conducted a limited
inventory of.the storage cabinet that contained the jumpers used
by the operators. The jumpers within the cabinets were neatly
arranged and clearly labeled. No discrepancies were identified.
c. Conclusions on Conduct of Ooerations
The inspectors concluded that the control room operator response
to the unexpected tri? of the Unit 2 reactor feed pump turbine was
good. The shift supervisor mair.tained an awareness of the power
to flow map throughout the event.
The problem annunciators reviewed by the inspectors were resolved
in accordance with procedures. The documentation indicated that
procedural reviews and evaluations were conducted by the
appropriate operations personnel in a timely manner. The supply i
of jumpers in the jumper storage cabinet appened to be adequately
maintained.
03 Operations Procedures and Documentation
03.1 Surveillance reauirements not included in olant orocedures
a. Insoection Scoce (92901)
On October 22, 1996. licensee personnel determined that the
periodic channel function test of the Reactor-Mode-Switch-in-
Shutdown, control rod block function, was not in any surveillance
procedure and had not been performed. The functional test is
required by Unit 1 and Unit 2 TS Surveillance Requirements (SRs)
3.3.2.1.6. Perform Channel Function Test.
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b. Observations and Findinos i
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The inspectors observed that SR 3.0.1 requires. in part, that SRs
shall be met during the applicable modes or other specified
conditions. Table 3.3.2.1-1 lists the SRs with the applicable
modes or other specified conditions for the Rod Block Monitor. Rod
Worth Minimizer and Reactor Mode Switch-Shutdown Position. SR 1
3.3.2.1.6 is listed for the mode switch. The SR is not required
to be performed until one hour after the mode switch is in the
shutdown position. The given frequency is 18 months. The
inspectors also observed from information received from the i
licensee that SR 3.10.6.3 associated with LC0 3.10.6. Multiple
Control Rod Withdrawal - Refueling, was also not in any plant
procedure. This SR is only required to be met during fuel
loading; it requires the verification that assemblies are being
loaded in a spiral sequence: and has a frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The inspectors determined from reviews and discussions with
licensee personnel that SR 3.3.2.1.6 should have been performed on
Enclosure 2
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both units during their last respective refueling outages, fall
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1995 for Unit 1 and spring 1996 for Unit 2. Also, during the
- outages, plant conditions did not require a spiral fuel loading
i sequence and LCO 3.10.6 was not invoked by either unit. When the
! deficiencies were discovered, neither unit was in a mode or ;
condition which required that the LCOs be invoked.
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- Not performing the SRs during this period and/or not including the
SRs in applicable surveillance procedures was determined by the
j NRC resident staff to have had minimal safety significance.
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The inspectors reviewed procedures 34SV-C71-003-1S/2S " Reactor
Mode Switch in Shutdown Functional Test." Revision 4 for Unit 1
1 and Revision 3 for Unit 2. Revised on November 28. 1996, both
- procedures contain the steps to perform the SRs.
c. Conclusions
l With the revision of the prc 7dures and the plans to perform the
SRs at the next opportunity, the licensee identified failure to
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include these SRs in applicable surveillance procedures is
identified as a Non-Cited Violation (NCV) 50-321, 366/96-14-01:
I Failure to include Surveillance Requirements in Applicable >
Procedures. consistent with Section IV of the NRC Enforcement
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Policy.
l 07 Quality Assurance in Operat.ons
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07.1 Licensee Self-Assessment Activities
a. Insoection Scooe (40500)
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l The inspectors reviewed and observed various self-assessment
activities.
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b. Observations and Findinas i
During the inspection period. the inspectors reviewed licensee
! self-assessment activities. Inspectors regularly attended the
Morning Management Meeting and t1e Plan of the Day r.eeting.
On October 27. at the Morning Management Meeting. . tams identified
, as Operations Open Issues were discussed. Among toe issues
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discussed were the following:
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the 2A RFPT duplex oil filter replacement
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the 1B RHRSW Heat Exchanger tube leak
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the deep well water pump power supply
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the 4160V switchgear circuit breakers
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the hydrogen recombiner controller.
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Each issue was discussed to determine current status and possible ;
solutions. Focus on reactor safety during the meetings was good.
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The inspectors were informed that the Operations Open Issues would
be discussed each Friday.
Enclosure 2
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During the Plan of the Day (POD) meeting, the licensee discussed
ongoing Required Action Lists for both units. Fire Action List.
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and the status of ongoing approved corrective / preventive
maintenance work items. Specific items, depending on the day of
the week, were discussed as follows: Monday - deficient :
annunciators and current caution tags: Tuesday - instrument out-
of-service reports: Wednesday - clearances currently in force and
i red circled readings (items identified as out of specifications
I during rounds): Thursday - temporary modifications currently in
plant systems: and Friday - forced outage work list.
The inspectors found that the items were discussed in detail:
specific individuals / groups were tasked with the responsibility
for each ittm: deficiencies in performance identified by
management we e discussed: deadlines and updates were established:
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and management expectations were outlined.
c. Conclusions
The inspectors concluded that licensee self assessment activities
conducted at routine meetings was effective in providing solutions
to short term problems as well as identifying potential
intermediate and long term problems.
08 Miscellaneous Operations Issues (92901) ,
08.1 Imoroved Standard Technical Soecification Imolementation Audits
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(Temocrary Instruction 2515/130)
a. ScoDe
The licensee transferred actions and surveillances from their
" custom" TSs to the Improved Standard Technical Specifications
(ITS) and other documents. This process was identified by the
licensee as Technical Specification Improvement Program (TSIP).
The results of TSIP were evaluated by the Office of Nuclear
Reactor Regulations (NRR) and documented in Safety Evaluation
Report (SER) related to Facility Operating License Amendment No.
195 for Unit 1 and No. 135 for Unit 2. To evaluate the
effectiveness of the process, the inspectors selected ten original
TSs for each unit and compared them with the new ITS. the TS
Bases, the SER. the FSAR. the Technical Requirements Manual (TRM)
and procedures, to verify that these requirements had been
transferred / relocated appropriately. The TS. ITS. TS Bases.
FSAR. TRM and implementing procedures reviewed are indicated in
Table 1.
The inspectors examined the licensee's implementation of the
controls of modifications to relocated requirements by evaluating
the implementation of those controls for the TS listed in Table 1.
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Enclosure 2
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Table 1 Transferred TS Examined.
Old TS New TS or Relocation Con- Implementing
Other Manual trol Procedures
2-3/4.1.3.1 2-3.1.3 10 CFR 50.92 34SV-C11-003-2S
2-3/4.4.3.2 2-3.4.4 10 CFR 50.92- 34SV-SUV-019-2S
2-3/4.6.6.5 2-3.3.1.2 10 CFR 50.92 345V-C51-001-2S
3/4.5.2 2-3.5.1 10 CFR 50.92 345V-B21-003-25
2-3/4.4.6.2 2-3.4.10 10 CFR 50.92 34SV-019-2S
2-3- 2-3.6.1.3 10 CFR 50.92 42SV-SUV-044-2S
4.6.6.5.2
2-3/4.9.12 2-3.9 10 CFR 50.92 34FH-0PS-001-0S
2- 2-5.5.8 10 CFR 50.92 00AC-REG-001-0S
3/4.11.1.4 64CH-ADM-001-0S
2-3/4.8.2.1 2-3.8.7- 10 CFR 50.92 34SV-SUV-013-05-
2-3/4.7.3 2-3.5.3 10 CFR 50.92 345V-E51-002-2S
1-3/4.6.C 1-3.4.9.5.6.& 10 CFR 50.92 52GM-MME-004-1S
1-3/4.6.G FSAR 4.10.3 10 CFR 50.59 34SV-SUV-019-1S
Procedures 42SV-T46-003-1S
1-3/4.6.H.1 TS Bases & 10 CFR 50.59 & 345V-B21-004-15
Procedures TS 5.5.11 52GM-B21-005-0S
1-3/4.6.I 1-3.4.2 & 10 CFR 50.92 & 34SV-SUV-023-15
Procedure 10 CFR 50.59
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Enclosure 2
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Table 1 Transferred TS Examined.
l Old TS New TS or Relocation Con- Implementing
Other Manual trol Procedures
1-3.6.K TRM 3.4.2 10 CFR 50.59 Numerous ISI & IST
1-3.6.K.a-c Program
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Procedures.
These Procedures
Were Not Verified.
l 1-3/4.6.L TRM P3.7.1 10 CFR 50.59 52-SV-SUV-001-0S
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1-3/4.8.A.1 TRM 3.7.3 & 10 CFR 50.59 62RP-RAD-007-0S
Procedures 60AC-HPX-007-05
l 1-6.9 Unit 2 FSAR 10 CFR 50.59 00AC-REG-001-0S
13.6.4 42CC-ERP-014-05
1-6.10 Unit 2 FSAR 10 CFR 50.59 20ADM-002-0S
13.6.5.A&B
1.6.13 5.5.2.b 10 CFR 50.92 52SV-E11-001-1S
To evaluate the licensee's corrective action program related to
the TSIP Process, the inspectors reviewed corporate and site
audits. Deficiency Cards (DCs). LERs and NRC inspection findings.
The inspectors reviewed licensee self-assessment audits. The
scope of the review included: verification of requirement
relocation: adequacy of procedures, programs, and manuals
supporting relocations: evaluation of implementation of controls;
and conversion verifications.
b. Observations And Findinas
Except as noted below, procedures, manuals and, programs
ap3ropriately implemented requirements transferred to ITS and
otler documents as authorized by NRR.
e Although Revision 11 of Procedure 64CH-ADM-001-0S Chemistry
' Program, relocates the requirements of the ITS Page 59 of 82
still references footnote "a" to the TS. This footnote was
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deleted in the transfer process and is not appropriate to the
ITS. The licensee indicated that the discrepancy would be
corrected with the next substantive change to the procedure.
Enclosure 2
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e Procedure 42SV-SUV-044-25. Stand-By Gas Treatment Excess Flow
Isolation Damper Surveillance. Revision 1. dated 12/5/89
references SR 4.6.6.5.2.a. the pre-TSIP identification. This
surveillance was transferred to ITS without change and is now i
identified as SR 3.6.1.3.13. The licensee indicated that the !
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discrepancy would be corrected with the next substantive change
to the procedure.
e In Section 3/4.6.L of the old TS. a note at the bottom of page
3.6-10a for Unit 1 and page 3/4.7-11 for Unit 2 was deleted
when the requirements were relocated to the TRM. This deletion
was not justified in an applicable 50.59. The licensee issued
a DC C09604999 to correct this discrepancy. The note was ,
administrative in nature and of minor safety significance. :
c. Conclusions
Although the licensee's transfer program was not formal and
proceduralized. the licensee appropriately transferred actions and
surveillances from the TS to the ITS', procedures manuals or
programs. A few exceptions were noted. An NCV involving
surveillances that were not transferred is documented in
Section 03.1. ,
08.2 (Closed) VIO 50-321/96-10-01: Inadequate Procedure for Verifying
Plant Service Water Pump Discharge Valve Positions.
This violation was identified when it was observed that the pump
discharge valve positions were not being verified in accordance
with the Improved TS. which were implemented in July. 1995.
The inspectors reviewed the response, dated September 23. 1996,
and observed that an inadequate procedure was stated as the reason
for the violation. The response indicated the following: the
valve positions not being checked as required for both units were
checked and found to be in their correct positions; procedures
34SV-SUV-012-15 and 25. Plant Service Water and Residual Heat
Removal Service Water Valve Position Verification, were revised to
include the required valve positions: and other procedures used to
check valve positions were reviewed and no additional problems
were found. The inspectors reviewed procedure 345V-SUV-012-15.
Revision 3. Plant Service Water and Residual Heat Removal Service
Water Valve Position Verification, and procedure 34SV-SUV-012-25.
Plant Service Water Residual Heat Removal and Standby Service
Water Subsystem Valve Position Verification. Revision 13. The
inspectors observed the changes made to the procedures and that
the effective date for both procedures was August 14. 1996. Based
on the licensee actions and inspector reviews, this violation is
closed.
Enclosure 2
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08.3 (Closed) LER 50-321/96-04: Inadequate Procedure Results in l
Reactor Pressure Increase and Automatic Reactor Shutdown. l
This problem was discussed in IR 50-321.366/96-04. No new issues -l
were revealed by the LER. 1
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II. Maintenance
M1 Conduct of Maintenance
M1.1 General Comments
a. Insoection Scooe (62703)
The inspectors observed all or portions of the following work
activities:
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MWO 1-96-2889: 1C EDG End Turn Windings Inspection
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MWO 1-96-2806: Lube Oil Replacement in 1C EDG Bearing
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MWO 1-96-3011: Repair of 1C EDG Room Automatic Louvers
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MWO 2-96-1442: Support Maintenance Engineering for VOTES
Testing
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MWO 2-96-1868: Limitorque Operator Grease Inspection
b. Observations and Findinos
The inspectors found that the work performed during these
activities was thorough. All work observed was performed with the
work packages present and being actively used. The craft
personnel were knowledgeable of their assigned tasks. The
inspectors observed that supervisors and engineers monitored
specific work activities and routinely gave directions to craft
personnel. Quality control personnel were present during the
performance of Maintenance Work Order (MWO) 1-96-2889.
During the performance of MWO 1-96-2806. the inspectors observed
that the lube oil containers were clearly marked and labeled as
required.
The as-left static and dynamic Valve Operation Test and Evaluation
System (VOTES) testing was acceptable for 2E21-F0318. Core Spray
Minimum Flow Valve (MWO 2-96-1442). The dynamic VOTES testing
showed a large margin available for design operations. A small
amount of grease was added to the stem and drive sleeve of valve
2E21-F0158. Appropriate post maintenance tests were performed.
c. Conclusions on Conduct of Maintenance
Observed maintenance activities were generally completed
thoroughly. No deficiencies were identified by the inspectors.
Enclosure 2
_. - - -. . - - . . .--- - -. - - . - -
l .
L
.
.
10
M1.2 Unit 2 Core Soray Looo B Maintenance Outaae
a. Insoection Scoce (62707)
The inspectors observed portions of the activities associated with
l the Unit 2 Core Spray Loop B outage. In addition to the
l observations the inspectors reviewed selected maintenance work
packages and procedures. Maintenance. engineering and operations
personnel involved with the work activities were interviewed.
b. Observations and Findinas
On November 17, the licensee began an online maintenance outage ;
for selected components of the Unit 2 Loop B Core Spray System.
The scope of the work included the following for the selected
components:
-
Performance of 18 month grease inspection of limitorque
operators
. - - - . . - .
-
Cleaning and lubrication of valve stems and re-torquing of
valve packing as required
l
-
Changing oil in the pump motor and meggering the pump motor I
wiring !
-
Installation of a new motor on the operator for Core Spray
Outboard Injection valve. 2E21-F004B
-
Replacement of thermal overload heater elements for the Core
Spray Test valve. 2E21-F015B. and Torus Suction valve.
The inspectors observed removal of the spring pack on the
limitorque operator for the Core Spray Test valve. The 18-month
grease inspection was performed in conjunction with this activity.
The inspectors also observed the static VOTES testing on the
Minimum Flow valve. 2E21-F031. The inspectors determined from a
discussion with operations personnel and an independent review of
Technical Specification (TS) 3.5.1 that the applicable Required
Action Statement (RAS) was entered for the out of service Core
Spray loop.
The inspectors discussed the issue of clearance and tags used to
control the work activities with maintenance personnel.
Maintenance personnel indicated that most of the valve work was
performed by the use of personalized Danger Tags. The inspectors
i verified that the appropriate clearance tags were hung at the
1
Motor Control Center (MCC) and in the Control Room for the valves
being worked. The inspectors reviewed procedure 30AC-OPS-001-05.
Control of Equipment Clearances and Tags. Revision 15. No
discrepancies were found between the implementation of the Danger
Enclosure 2
.
. ,
,
11
Tags and the procedure. The inspectors also reviewed the
Clearance Index and Audit Sheet and found no discrepancies.
The inspectors reviewed work packages associated with the V0TES
testing of the Core Spray Minimum Flow valve: the 18-month grease
inspection on the valve operator for the Core Spray Test valve,
and the motor replacement for the Core Spray Outboard Injection
valve. 2E21-F004B.
The motor operator on the 2E21-F004B valve was replaced as part of
the licensee's on-going activities to replace shafts that are
susceptible to cracking. Similar problems were also identified in
Inspection Reports 50-321, 366/95-17.
Surveillance procedure 34SV-E21-002-25. Core Spray Valve
Operability. Revision 7 was satisfactorily completed on
November 19 and the system was returned to service.
c. Conclusions
The inspectors concluded that the work activities observed were
conducted in a safe and timely manner. Maintenance personnel
performing the tasks were conscientious. Housekeeping was
excellent. Excellent implementation of Foreign Material Exclusion
(FME) protection was observed for the work activity performed on
the Core Spray Test valve.
M2 Maintenance and Material Condition of Facilities and Equipment
M2.1 Cold Weather PreDarations
a. Inscection Scooe (71714)
The inspectors observed activities, performed walkdowns, and
reviewed documents and procedures associated with cold weather
preparations.
b. Observations and Findinas
Among the areas observed and reviewed were the following:
-
Review of procedures used to calibrate and test equipment
associated with heat tracing, space heaters and thermostats
-
System walkdowns to observe heat tracing, space heaters and
insulation installed on susceptible systems. Walkdowns were
also performed to observe the material condition of automatic
and manual louvers
-
Review of instructions and checklists used to implement
responses to actual cold weather conditions
-
Review of previous corrective maintenance and deficiencies
identified during the last cold weather period.
Enclosure 2
.. .. . . . _ . . - - - - - - ~ _ - - - - _ _ - - - - - _ - .
- -
<
.
.
l
.
!
12
,
The inspectors reviewed maintenance procedure 52PM-MEL-005-05, l
- Cold Weather Checks. Revision 9. and Operations Department
Instruction DI-0PS-36-0989N. Cold Weatler Checks. Revision 9. The !
procedure and instruction provide for testing and repair of
'
equi) ment associated with cold weather protection as well as a l
chectlist to ensure exposed equipment was adequately protected i
during cold weather conditions. !
,
l
The inspectors performed walkdowns of the Emergency Diesel
4
Generator (EDG) building, intake structure fire pump building,
service water valve ait, switchyard deluge buildings fire water
storage tanks, and a)ove ground piping systems. These areas
- contain systems that are important to safety and/or could cause a
'
plant transient. During the walkdowns the inspectors observed the
following:
) -
Several heat trace indicating light lens were missing in the
l fire pump house storage tank, intake structure valve pit. and
EDG building areas
-
Insulation damage on the piping for the fire ptep suction,
mi.ssing insulation on a level switch on a fire pump storage
. tank, and insufficient insulation on the travelling screen wash
system
'
.
-
The automatic louvers in the EDG building appeared to close.
but may not be shut off completely. The manual louvers in the
! fire pump house appeared to have deteriorated and may not be
- capable of being manually closed completely
,
-
The sensing lines for the fire main jockey pump and the
- electric fire pump are located near a manually operated louver ;
- and appeared to be susceptible _ to freezing '
I -
A section of the fire pump house which contains fire main
- piping does not have a space heater l
l
'
-
The inspectors questioned the installation of the freeze
protection system concerning single failure vulnerability
-
The inspectors observed that the freeze protection for the
service water system for the IB EDG appears to be supplied by
- Unit 1 only although the EDG serves both units.
4
- The inspector reviewed 8 MW0s and 3 Deficiency Cards (DCs) from
'
the previous cold weather period. The inspectors observed from
4 this review that the MW0s and DCs concerned heat trace problems.
! The inspectors also observed that once identified, these items '
were promptly corrected.
The observations made by the inspectors and the concerns were
discussed with licensee personnel. The licensee stated that
freeze protection systems had been evaluated in the 1979 time
Enclosure 2
l. j
l
l' l
'
l 13
frame but that the evalu tion did not include single failure l
vulnerability. '
1
c. Conclusions l
The inspectors concluded that the cold weather pre)aration program )
contains procedures for performing equipment opera]ility checks,
performing checks during actual cold weather conditions, and
'
correcting deficiencies when identified. The inspectors also
concluded that a loss of power from Unit 1 to the freeze
protection for the service water cooling piping to the 1B EDG ,
could impact the EDG's operability to support Unit 2 operation. l
At the end of the report period, no determination had been made
concerning single failure vulnerability of the EDG service water
freeze protection. This is identified as an Inspector Followup
Item (IFI) 321.366/96-14-02: Potential Single Failure
Vulnerability in the Freeze Protection System.
M2.2 Ad.iustment of MCC Breaker Trio Setooint for 1B EDG
a. Insoection Scope (92902)
1
The licensee determined that the trip setpoint for the normal l
supply breaker to MCC 1R24-5026. from the 1B EDG. was not set I
properly (Section E2.2). The inspectors reviewed circumstances
associated with the improper setting.
b. Observations and Findinas
The inspectors reviewed administrative procedures and documents
involving configuration control and maintenance. Among the
procedures reviewed were the following-
1
-
E.I. Hatch Nuclear Plant Configuration Management Program
Reference Manual
1
-
Procedure 10AC-MGR-015-05. Configuration Management Program. l
'
Revision 0
1
-
Procedure 40AC-ENG-003-05. Design Control. Revision 8 !
-
Procedure 50AC-MNT-001-05. Maintenance Program. Revision 24
The inspectors also reviewed maintenance procedure
52PM-R24-002-0S. Air Circuit Breaker. Type LA-600. Maintenance.
Revisions 0 and 1. and MWO 1-88-7355. From this review. the
inspectors determined that the MWO directed that the trip setpoint
for the MCC breaker be changed from the as found setpoint of 3X to
the required setpoint of 12X. as determined by corporate
engineering. The MWO also directed that the change be made in
accordance with LA-600. breaker maintenance procedure.
The inspectors found that the MWO clearly stated what activities
were to be performed and that the applicable procedure was
Enclosure 2
.
.
14
referenced. The inspectors also found that on November 6. 1996,
the licensee determined that the setpoint had been adjusted to 8X
instead of 12X. as directed by the MWO.
c. Conclusions
The inspectors concluded that the Configuration Management Program
Manual in conjunction with procedures 10AC-MGR-015-05.
40AC-ENG-003-05, and 50AC-MNT-001-0S established and implemented
the applicable requirements of 10 CFR 50, Appendix B.
Criteria III. Design Control.
The inspectors concluded that not adjusting the trip setting on
the MCC normal supply breaker in accordance with engineering l
requirements is an example of a failure to implement the '
configuration control process. This is identified as an example
of Violation 50-321. 366/96-14-03: Failure to Implement
Configuration Coritrol Requirements. l
i
M3 Maintenance Procedures and Documentation !
l
'
M3.1 Surveillance Observations
a. Insoection Scooe (61726)
The inspectors observed all or portions of the following Unit 1
and Unit 2 surveillance activities:
- 34SV-E42-002-1S: HPCI Pump Operability
- 345V-E21-002-2S: Core Spray Valve Operability
b. Observations and Findinas
The inspectors observed during the performance of the Unit 1 High
Pressure Coolant Injection (HPCI) Pump Operability monthly
surveillance, two potential procedure deficiencies. The first
involved a high radiation alarm in the HPCI room and the second
the torus bulk average temperature.
The inspectors were informed during the pre-;iob briefing that the
hydrogen flow for the water chemistry systca would not be reduced.
Hydrogen injection significantly raises radiation levels of the
steam leaving the reactor. Shortly after the start of the HPCI
turbine, the HPCI room experienced a high radiation condition and
the area high radiation alarm was received. The Health Physicists
(HP) Technician at the scene reported that the steam supply line
to the turbine was reading one Rad on contact. The inspectors
reviewed the surveillance procedure and found that the procedure
did not discuss the alarm as an expected alarm during routine HPCI
surveillance testing. This appeared to be contrary to the concept
of ALARA and is identified as a potential procedure deficiency.
The inspectors observed that during performance of the .
surveillance, safety related recorder 1T48-R647. labeled TORUS I
!
Enclosure 2 I
!
.. - .-.
4
15
AVERAGE BULK WTR TEMP. indicated a temperature of 108 degrees
Fahrenheit ( F). The inspectors also observed that the Safety
Parameter Display System (SPDS) indicated a temperature of 93 "F.
Review of the HPCI surveillance procedure found that operators
were directed to use SPDS indication to determine torus
temperature, but the safety-related temperature recorder was not
addressed. The procedure stated that the average Suppression Pool
water temperature shall not exceed 105 *F. It also stated that
testing must be stopped if the 105 *F temperature was exceeded.
c. Conclusions
For the surveillance. all data was within the required range and
the equipment met the surveillance procedure criteria. The
performance of the operators and crews conducting the surveillance
was generally professional. A potential procedure deficiency was
identified during the Unit 1 HPCI surveillance and is identified
as Inspector Followup Item, 50-321. 366/96-14-04: Potential
Deficiencies in the HPCI Surveillance Procedure.
III. Enaineerina
El Conduct of Engineering )
On-site engineering activities were reviewed to determine their
effectiveness in preventing identifying, and resolving safety
issues. events. and problems.
El.1 Inadvertent Lift of Nitroaen System Relief Valve and NOVE
a. Insoection Scoce (37551) (92.902).0
While attempting to fill the Um t 2 nitrogen storage tank, a
relief valve on the fill line lifted. The discharae from the
relief valve was into the tank room, partially filling the room
with nitrogen. This resulted in the declaration of a Notification i
of Unusual Event (NOUE) due to release of toxic gas in accordance !
with the licensee's Emergency Plan. The inspectors reviewed ,
operator logs. event reports and design documents. The inspectors !
interviewed operators and engineers involved in the event.
b. Observations and Findinas
The inspectors reviewed circumstances associdted with the event. !
On October 28. 1996. plant operators were conducting a routine
refill of the Unit 2 nitrogen storage tank in accordance with ;
34S0-T48-002-25. Containment Atmospheric Control and Dilution
systems. Shortly after commencing the fill, a relief valve on the
plant side of the fill line lifted, discharging nitrogen into the
tank room. The inspectors determined that the relief valve lifted
within the tolerance of it's set pressure of 145 psig.
The inspectors determined that the relief valve was installed to
correct a discrepancy with the system drawing noted by the system
Enclosure 2
- .- - .-. - --- -- - - . _ . - -
'
I
.
d
16
engineer during a system walkdown. During the walkdown, the
! engineer discovered that a ball valve was installed where the
'
system drawing indicated there should be a relief valve. The ball
valve was determined to have been installed outside of the Design
'
Change Request (DCR) process at some point prior to June 1987.
Since the drawing showed a relief valve, the engineer initiated a
Maintenance Work Order (MWO) to remove the ball valve and install
d
a relief valve. The design change process was not required to be
"
implemented in this case since the work involved restoration of
the original system configuration and not a design change. As a
. result, much of the independent review involved in the
- installation process was not available to detect a problem.
When procuring the relief valve and determining its' lift
setpoint. the system engineer referenced the vendor manual for the
nitrogen system. The manual listed two values for the relief
valve lift setpoint. 150 psig on a drawing of the valve and 170
psig in a table listing all relief valves in the system. The
licensee later determined that the correct value was 170 psig. An
installed relief valve was set at 145 psig. It was during the
next fill of the tank that the relief valve lifted, resulting in
4 the NOUE.
The inspectors also observed that the system drawings did not
include the vent piping that had been installed by a previous
Design Change Notice (DCN) to allow nitrogen to vent outside of
the enclosure during relief valve lifts and normal system purges.
!
This vent piping was subsequently removed during the last
installation of the relief valve.
c. Conclusions
This failure to adequately control plant configuration is a
3
violation of the requirements of 10 CFR 50. Appendix B. Criterion
4 III. Design Control. This is an example of Violation 50-321,
366/96-14-03: Failure to Implement Configuration Control
Requirements.
.
E2 Engineering Support of Facilities and Equipment
.
E2.1 Main Steam Line Isolation Valve Soecial Test
a. Insoection Scooe (92903)
The inspectors observed a special test of the Unit 1 Main Steam
.
Line Isolation Valves (MSIV) on November 3.1996. The special
testing was performed in response to a 10 CFR 21 Potential Defect
Notification issued on October 28. 1996, by the valve vendor.
Automatic Valve.
b. Observations and Findinas
On November 3. the inspectors observed testing of the outboard and
inboard MSIVs on Unit 1. Unit power was reduced to approximately
70% rated thermal power (RTP) to conduct this test. The
Enclosure 2
- .. .- - . ..- - - - . - -- -. . . - - . - _ _ - - .
.
1 f
.
,
17
inspectors observed that management was present during the entire
time of MSIV testing.
-
The inspectors reviewed Section 4.6.5 of the Final Safety Analysis
.
Report (FSAR) and determined that the MSIVs may be tested and
exercised individually to the fully closed position after reducing
reactor power to less than 75%.
'
This special test was performed to determine if the licensee had
4 any of the defective solenoid valves associated with the MSIVs
i that were described in the 10 CFR 21 (Part 21) Notification of
Potential Defect. The notification identified that plungers in
the defective solenoid valves had the incorrect length due to a
! machining operation error at the time of manufacture. Valves with
- the defective plunger were found to have substantially longer !
closure times at normal operating temperature.
'
<
The inspectors reviewed Special Purpose Procedure
34SP-103196-CM-1-15. Main Steam Line Isolation Valve Special Test.
The operators had this procedure in their possession during the
performance of the MSIV testing.
.
I
The 1821-F028A and 1821-F028C MSIVs experienced limit switch
- problems during the test. The closed indicating light came on 1
-
when the respective valve received the close signal but the open i
-
indicating light did not extinguish. Operations personnel I
- determined that the valves had closed based upon observation that
- steam flow indication in the respective steam lines went to zero
while the flow indication in the remaining three steam lines
increased. Maintenance personnel entered the steam chase area to
make adjustments to the problem limit switch on each valve
i following the testing that resulted in dual light indications.
'
Testing of each valve subsequent to the limit switch adjustment ;
demonstrated proper valve indication and acceptable stroke times.
'
i
The 1821-F022D inboard MSIV experienced limit switch problems 1
during testing similar to that which occurred during testing of
the F028A and F028C outboard MSIVs. The licensee observed the
time it took the steam flow indicator to decrease to zero once the
-
MSIV was given a closed signal. This timing 3rovided a reasonable
judgement as to how fast the valve closed. T1e licensee concluded
that the F0220 MSIV closed within an acceptable time.
<
The licensee entered Technical Specification (TS) Required Action
Statement (RAS) 3.3.3.1 for the inoperable instrumentation
associated with the MSIV. The inspectors determined that the
- licensee's actions were appropriate.
c. Conclusions
Testing of the Unit 1 MSIVs demonstrated that the valves did not
contain the defective part described in the Part 21 notice. Power
was reduced to within the limits specified in the FSAR.
Maintenance coordination and support was timely. TS actions taken
Enclosure 2
, _ . ..
.
.
18
were appropriate. Management's presence enhanced the
effectiveness of coordination with offsite personnel and
facilitated the decision making process.
1
E2.2 Motor Control Center (MCC) Feeder Breaker Trio Device
c. Insoection Scoce (92902) l
l
The licensee determined tha; the trip setting for the normal
supply feeder breaker to the EDG 1B MCC. 1R24-S026. was incorrect
for the application (Section M2.2). This was identified during an
engineering walkdown.
b. Observations and Findinos
The inspectors reviewed information supplied by engineering
personnel and determined that the correct breaker for this
application would have a trip unit with a short and long time
delay overcurrent trip and no instantaneous trip. The breaker
that was found installed had a long time delay overcurrent trip
and an instantaneous overcurrent trip. The existing alternate
supply breaker had correct trip features.
The inspectors observed that two temporary modifications (TMs)
were initiated. One. 1-96-40 (TM40). was to install a breaker
with the acceptable trip features. The onsite engineering l
personnel identified a spare breaker in the warehouse with
applicable trip features. The other. 1-96-41 (TM41). was to
implement the following changes: Move the largest non-safety
load, the Technical Support Center (TSC) power feed, to another
MCC: disable the trips on the normal supply breaker; and adjust
the upstream 4KV overcurrent protective relays.
'
The inspector reviewed both TM40. which technically justified the
use of the spare breaker, and TM41. The review also included the
10 CFR 50.59 review. The testing activities for TM40 were
controlled by MWO 1-96-4267 and procedure 52PM-R24-002-05. Air
Circuit Breaker. Type LA-600. Maintenance. Revision 1. The
inspectors reviewed the MWO and procedure and did not identify any
deficiencies.
The inspectors observed the following circuit breaker tests: the
initial successful long time pick-up test corresponding to a 600
amp load; the unsuccessful short time pick-up test corresponding
- to a 1500 amp load; and the unsuccessful long time pick-up retest.
l Site personnel could not determine why the short time test and
retest were not successful. The licensee had the breaker shipped
off-site to be tested. The licensee was subsequently informed
that the breaker could not be tested. The license canceled TM 40
and implemented TM 41.
( The inspectors reviewed the implemented TM 41 and observed that
I the TSC power feed was rerouted from 1B EDG MCC 1R24-S026 to the
1B EDG MCC 1R24-SO48. The inspectors also observed that the
Enclosure 2
.. _ _ . _ . . _ . . . _ _ . - . _ _ _ _ _ . _ _ _ _ _ _ _ _ . . _ _ _ - _ _
.
-
,
,
, 19
l
j. restoration activities stated that the normal supply breaker would
l be replaced with a new breaker having the proper trip settings
l with the applicable trip features per a design change.
!
l c. Conclusions
4
i The inspectors concluded that the TMs were initiated. controlled
l and TM41 was implemented with engineering oversight Plant Review
l Board (PRB) approval, anc ooerattonal testing. The engineering
groups discovery of this deficiency was excellent. Until the
normal supply breaker to 1R24-5026 is replaced. this is identified
as IFI 50-321/96-14-05: Restoration of 1B EDG Motor Control
Center.
E4 Engineering Staff Knowledge and Performance
E4.1 Trendina Recetitive Eauioment Problems and Root Cause-
'
a. Insoection Scoce (92903)
The inspectors reviewed the licensee program for trending of
repetitive equipment problems and the analysis for root cause
program.
b. Observations and Findinas
The inspectors reviewed procedures and instructions that involved
trending and root cause. Among these were the following:
-
AG-ENG-04-0288N: Plant Performance Indicator Program
-
DI-ENG-41-1089N: Use of System Engineers
-
DI-ENG-45-1290N: Conduct of Engineering
-
AG-MGR-27-0687N: Root Cause Analysis
-
10AC-MGR-012-05: Event Review Program
The inspectors observed from the reviews that: procedure AG-ENG-
04-0288N. Revision 0, applies, in part, to all performance
indicator data submitted to the Plant Hatch Performance Trending
Program: instruction DI-ENG-41-1089N. Revision 4. applies, in
part to engineers assigned to the Systems Engineering Sections
and to other engineers assigned systems responsibilities:
instruction DI-ENG-45-1290N. Revision 0 applies. in part. to all
engineers, nuclear specialists and other personnel in the
Engineering Support Department; procedure AG-MGR 27-0687N.
Revision 3 provides a guideline in performing a methodical root
cause analysis and applies to any individual performing an
analysis; and procedure 10AC-MGR-012-05. Revision 3 applies to
category 1 and 2 events or any other event deemed by plant
management.
The inspectors observed Section 4.3. Duties and Responsibilities,
of DI-ENG-1089N Sub-Section 4.3.3. under Priority I, which states
the following:
Enclosure 2
__ _
- - .. - . - - _.
_ .- -. -_ - _. -- -
.
,
20
" Evaluate system performance to improve equipment reliability
and efficiency. This will normally include the trending of
selected system parameters to detect deteriorating system
performance and computation of unavailable figures for
, equipment. Significant changes or developing trends will be
! evaluated and reported to Engineering Supervision, along with
l any suggested corrective actions. A history of system
l
'
performance will be maintained so that the need for, and the
success of, corrective actions can be demonstrated."
The inspectors observed that DI-ENG-45-1290N discussed the conduct
,
of engineering in terms of goals, performance of engineering
l activities, and professional accountabilities.
l
The inspectors found from these reviews that the responsibility
for trending of repetitive equipment problems is assigned to the
system engineers. The inspectors also found that the type of
analysis for root cause is determined by event categories, which
are classified into four levels. The event classification is based
on the relative probability and the relative consequences.
Category 1. 2 and 3 events require a type of root cause analysis
and category 4 events are at management discretion.
c. Conclusions
The inspectors concluded that administrative procedures clearly
assign the system engineers the responsibility for trending
repetitive equipment problems. Root cause analysis are performed
based on the category of events or at the discretion of
management. j
E8 Miscellaneous Engineering Issues (92700) (92903)
E8.1 (Closed) VIO 50-321.366/96-07-02: Failure to Conduct Testing
Following Molded Case Circuit Breaker Instantaneous Trip Setpoint
Changes.
This violation was identified when. as a result of an inspector
identified issue, the licensee started adjusting the trip
setpoints on 577 molded case circuit breakers (MCCB). The
inspectors observed that post adjustment testing was not being
performed.
The inspectors reviewed the response, dated August 6.1996, and
observed that the use of an inaccurate value for motor locked
rotor amps for adjusting the setpoint was stated as the reason for
the violation. The response indicated that equipment was not
cycled after setpoint changes due to personnel incorrectly
cencluding that verifying the setpoint represented adequate post
maintenance acceptance criteria. The response stated that
personnel responsible for the error in the calculation and for the
assignment of the inappropriate functional test have been made
aware of their errors and the consequences of those errors.
Enclosure 2
. . . .. -. -- - - --- - . _ - . .. - . -
.
4
.
21
The inspectors documented in irs 50-321, 366/96-07, 96-10. and
96-13 observations of licensee personnel changing setpoints and
performing post change testing in accordance with instructions.
The inspectors also observed operations involvement and
engineering oversight. Based on the actions of the licensee and
the observations by the inspectors, this violation is closed.
E8.2 (Closed) LER 50-321/96-05: Loss of Power to Reactor Protection
System Bus Results in Actuation of Engineered Safety Features.
This LER discussed the loss of power to the Reactor Protection
System (RPS) 1B on April 14. 1996, as a result of the opening of
the supply breaker to the 4160VAC/600VAC Station Service
Transformer 10.
The event was caused by personnel error. Contractor personnel had
miswired 3 art of the current transformer circuit during a cable i
reroute t1at was implemented in Design Change Request (DCR)94-044. . Licensee personnel had failed to identify the wiring
error when the circuit was checked (" red lined") following the
completion of the DCR work.
As corrective actions, the licensee implemented a Maintenance Work
Order (MWO) 1-96-1393 on April 15, 1996, to correct the wiring
error. The responsible personnel were counseled. The inspectors
reviewed MWO 1-96-1393, which indicated that the error in the l
transformer circuit was and corrected.
The failure to properly install DCR 94-044 in accordance with the
Hatch Nuclear Plant Configuration Control Program is a violation
of the requirements of 10 CFR 50. Appendix B. Criterion III. l
Design Control. This failure is identified as an example of VIO
50-321. 366/96-14-03: Failure to Implement Configuration Control
Requirements.
i
Based upon the inspectors * review. the licensee's corrective 1
actions, and the identification of this issue as an example of VIO
50-321, 366/96-14-03. this LER is closed.
E8.3 (Closed) LER 50-321/96-01: Clogged Servo Valve Strainers Result
in Turbine Control Valve Closure and an Automatic Reactor Scram.
This problem was discussed in IR 50-321, 366/95-27. No new issues
were revealed by the LER.
E8.4 (Closed) IFI 50-321/96-13-06: Additional Review of Wiring
Deficiency on Electrical Drawings.
This item was identified when licensee personnel discovered that
the wiring in control panels did not match the applicable
drawings. These wiring discrepancies did not result in the
licensee declaring any safety equipment inoperable. The IFI
documented a concern about the process for addressing wiring
discrepancies. The inspectors were informed that the Deficiency
Enclosure 2
_. _ _ _ . __._ ~ _ . . __ __ - _ _ _ _ . _ _ ._ -_ _. .-
.
l-
22
l Card (DC) system is the process used when wiring discrepancies are
l identi fied.
Additional reviews and observations by the inspectors and
information from the licensee indicated that an apparent error was
l
made during engineering specification implementation involving the
Torus-to-Drywell Vacuum Breakers. This resulted in personnel
inadvertently de-energizing the 1H Torus-to-Drywell Vacuum Breaker
indication instead of the 1G. The troubleshooting personnel were
using electrical connection drawing 17801, which indicated that
link JJ44 was in series with the 1G vacuum breaker. When the link
was opened the 1H vacuum breaker de-energized instead of the'1G.
Personnel restored the link and issued a DC. The error is
identified as an example of VIO 50-321. 366/96-14-03: Failure to
Implement Configuration Control Requirements. Based on the
issuance of the violation, this -IFI is closed.
,
E8.5 (CLOSED) IFI 50-321.366/94-01-01: Plant Service Water System Flow
Model Verification.
The licensee had conducted Plant Service Water (PSW) system flow
testing to obtain additional data for benchmarking of the PSW
system flow model. As a result of this testing, the licensee
identified and corrected fouling of the Unit 1 High Pressure
Coolant Injection cooler piping and a mispositioned throttle valve
l for Unit 1 Low Pressure Coolant Injection inverter room cooler.
Based on the results of this testing, the PSW system flow model
was updated. The inspector concluded the licensees actions were
sufficient.
E8.6 (CLOSED) URI 50-321.366/94-01-02: High Energy Line Break
Protection for Service Water Systems within the Containment
The licensee had conducted a study of service water system piping
inside containment and determined that twelve locations required
further analysis. These twelve locations were analyzed and found
to have sufficient margin for long term operation. The inspector
,
concluded the licensee had appropriately analyzed the locations in
l
question.
IV Plant Sucoort
.
R1 Radiological Protection and Chemistry Controls
R1.1 Observation of Routine Radiolooical Controls
l a. Insoection Scooe (71750)
General Health Physics (HP) activities were observed during the
report period. This included locked high radiation area doors,
proper radiological posting, and personnel frisking upon exiting
the Radiologically Controlled Area (RCA). The inspectors made
frequent tours of the RCA and discussed radiological controls with
Enclosure 2
-
-- .- .- .. ._. . -. - - . - . - -. .
.
.
.,
'
23
HP technicians and HP management. No deficiencies were
identi fied.
R1.2 Radiolooical Controls
a. Insoection Scoce (84750. 86750)
Radiological controls associated with radioactive waste (radwaste)
processing equipment and stora.ge areas were reviewed and
evaluated. In particular. housekeeping and cleanliness, area
postings. radwaste container labels, and controls for high and
very high radiation areas were reviewed for adequacy.
The inspectors made frequent tours of the RCAs. and reviewed and
discussed specific procedural guidance and selected survey results
with HP staff and supervisors,
b. Observations and Findinas
Locked high and very high radiation area controls were verified to
be implemented in accordance with Technical Specification (TS)
requirements. Posting of radwaste storage areas were proper and
in accordance with TS or 10 CFR 20 Subpart J requirements.
Overall, containers holding radwaste, materials or contaminated
equipment were labeled in accordance with 10 CFR 20.1904
requirements. Excluding some isolated examples of dirty floors,
e.g., the Unit 1 (U1) Waste Sludge Tank area, and a potential leak
in the U1 Laundry Drain Tank room. cleanliness and housekeeping
within the RCA and outside radwaste processing and storage areas
were considered to be acceptable.
c. Conclusions
Radiological controls for high and very high radiation areas were
maintained in accordance with TS requirements. Area postings and
labels for containers of radioactive material were appropriate.
Improvements were noted for general housekeeping and cleanliness
relative to observations made during a previous inspection
conducted July 29 through August 2. 1996, and documented in
Inspection Report 50-321, 366/96-10 dated August 30, 1996.
R1.3 Radioactive Waste and Material Transoortation Activities
a. Insoection Scoce (86750. TI2515/133)
The inspectors reviewed Radiological Control (RC) program
activities associated with the packaging and shipping of
radioactive material and waste to either vendor processing
facilities or directly to a licensed burial facility. The review
included evaluation and verification of radwaste classification
activities, and the subsequent packaging and shipping for the
radwaste shipments listed below. The inspection activities also
verified and evaluated implementation of revised 49 CFR Parts
100-179 and 10 CFR Part 71 regulations.
Enclosure 2
_ - - - . - . _ _ _ _ -. _
.-. . -. - - _ - - -
'
I
,
24
'
Records for radwaste and material shipments made between January 1
through November 12, 1996 were reviewed and discussed. In
particular, all documentation associated with the following l
- shipments were reviewed and discussed with licensee '
representatives,
o Shipment No. 96-1020. Radioactive material. Low Specific j
Activity (LSA) n.o.s.: 7 UN2912. Fissile Excepted /RO. May 23. !
1996.
-
o Shipment No. 96-1029. Radioactive material, n.o.s. 7 UN2982.
Fissile Excepted /RO August 7. 1996.
-
o Shipment No. 96-1020. Radioactive material. Low Specific i
Activity (LSA) n.o.s.: 7 UN2912. Fissile Excepted /RO. May 23. '
1996.
1
The following procedures were reviewed and discussed to determine
adequacy in relation to 10 CFR Part 20, 10 CFR Part 61.
10 CFR Part 71 the revised 49 CFR Parts 100-179 and 10 CFR Part 71
] regulations.
o Radiation Protection (RP) procedure. 62RP-RAD-011-05. Shipment
of Radioactive Material. Revision (Rev.) 8. effective April 1.
-
1996.
o 62RP-RAD-042-05. Solid Radwaste Scaling Factor Determination
and Implementation. Rev 3. effective March 26. 1996.
b. Observations
In general, licensee procedural guidance met applicable regulatory
requirements and recent revisions to 49 CFR Parts 100-179 and
10 CFR Part 71 regulations were implemented as required. No
concerns were identified for determination of radwaste scaling
factors. For the Type B radwaste shipments made, the inspectors
verified that licensee was a registered user of the shipping casks
used, and that the a3plicable Certificates of Compliance (C0C)
were maintained at t1e facility and used to develop licensee
procedures for packaging and shipping activities.
In general, shipping paper documentation was completed and
maintained as specified. One issue discussed with licensee
representatives for potential improvement was to enhance
visibility of the emergency phone number on the shipping paper
documentation. In addition, the following two concerns were
identified regarding shipping document Quality Control activities.
o For the Type B shipment documentation reviewed the inspectors
noted that not all of the routine determinations specified in
10 CFR 71.87 were documented as required by 10 CFR 71.91. In
particular the inspectors noted that records did not document
that any structural part of the package which could be used to
lift or tie down the package during transport is rendered
Enclosure 2
__
_ - - _ .- _. .. . . . - -. .- -- -
. I
4
.
d
25
' inoperable for that purpose. From discussion with responsible I
licensee representatives, the inspectors determined that all i
structures were rendered inoperable for the shipments made but i
that the determinations were not documented. Prior to the l
November 15. 1996, licensee representatives initiated a change l
to require the appropriate documentation on the Cask Users
4
Checkoff Sheet which is implemented in accordance with
o During review of C0Cs and associated documentation for package
type USA /5805/B( ) used for an August 7. 1996 Type B shipment
- of irradiated hardware to a licensed burial facility, the
. inspectors noted pages missing in copies of the referenced
documents maintained by Radwaste personnel. The inspectors i
were informed that the noted documentation was received i
'
directly by the radwaste staff from the vendor. Subsequent
review of document control records failed to locate all of the
required documentation. From discussions with applicable
, licensee representatives and review of Administrative Control
Procedure (AP) 20AC-ADM-003-05. " Vendor Manual Review and .
Control." Rev. 4. effective January 22. 1996, the inspectors
verified that program controls were in place to meet the intent
of 10 CFR 70.113 for the packaging Quality Assurance (0A)
program. However, by the end of the onsite inspection.
'
licensee representatives had not determined if the vendor
documents used to make the subject shipment in the
USA /5805/B( ) cask were received, processed and maintained in
accordance with the applicable administrative procedure.
t c. Conclusions
4
Excluding two packaging 0A issues, transportation and packaging
activities for radwaste and material shipments met 10 CFR 71.5 and
49 CFR 100-179 requirements. The revised Department of
Transportation (DOT) guidance was implemented, as applicable.
Radwaste characterization was in accordance with 10 CFR 61.55.
Two concerns were identified for radwaste packaging quality
control documentation. Consistent with Section IV of the
.
Enforcement Policy, one issue corrected prior to the end of the
onsite inspection was identified as NCV 50-321. 366/96-14-06:
Failure to Maintain Records Required by 10 CFR Part 71.91 for
Routine Determinations Specified by 10 CFR 71.87 for a Type B
Shipment. Pending licensee record search, the second issue was
identified as Unresolved Item (URI) 50-321, 366/96-14-07:
Determine If Processing and Control of COC and Associated
Documents for Package Type USA /5805/B( ) Vendor Documents Were in
Accordance with AC Procedure 20AC-ADM-003-05. " Vendor Manual
i Review and Control ."
Enclosure 2
.
'
i
26
R3 Radiation Protection and Control Procedures and Documentation i
l
a. Insoection Scooe (84750. 86750)
)
The inspectors reviewed selected effluent release and radiological
environmental monitoring program data for January. 1995 through
October 30, 1996. Reviewed reports included the Plant Radioactive
Effluent Releases. January 1.1995 through December 31, 1995, and
Radiological Environmental Operating Report for 1995.
Also, selected radiation control /radwaste performance indicators
were reviewed and discussed with licensee representatives.
b. Observations and Findinas
'
For 1995 no abnormal effluent releases were identified. For 1995
and 1996 year-to-date data, dose estimates from effluents were
small percentages of the Offsite Dose Calculation Manual (ODCM)
limits. No significant trends or changes in radiological
environmental monitoring program sample radiological analyses were
identi fied.
c. Conclusions .
!
Effluent release documentation and radiological environmental
monitoring program results were prepared in accordance with ODCM
requirements. The release data and environmental monitoring
results verified offsite releases and resultant doses were a small
fraction of the allowable limits.
R5 Staff Training and Qualifications in Radiation Protection and
Chemistry
a. Insoection Scoce (86750. TI 2515/133)
The training provided to designated staff to meet the requirements
of 49 CFR Part 172 Subpart H. was reviewed and discussed with
licensee representatives. Further, training details provided to
staff regarding implementation of recent DOT changes to
49 CFR Parts 100-179 were evaluated.
From interviews and discussion with applicable Health Physics (HP)
staff members regarding shipping documents and procedures, the
inspectors evaluated the training effectiveness regarding
implementation of 49 CFR Parts 100-179 requirements.
b. Observations and Findinas
Review of training records verified that responsible Health
Physics staff members directly involved in handling and packaging
of radioactive materials were receiving hazardous material
(hazmat) training within the required frequencies. From review of
current training documents. the inspectors verified that recent
DOT changes to shipping and packaging requirements were covered in
Enclosure 2
_ _ _
--. . . . . - . . . , _ _ .. - .. -. - _
..
.
,
'
27
the course material. From discussion of shi) ping procedures and
shipping papers, the inspectors determined tlat responsible
licensee representatives were knowledgeable of the recent DOT
,
changes.
From review of licensee's hazmat training base documents, the
inspectors identified two training program issues. The first
issue involved the lack of including all staff who had completed
the appropriate training, e.g. a first-line radwaste supervisor,
was not listed in the Hazmat training data base. In addition, the
inspectors noted that a time limit was not established for
completing and returning the training take-home exams provided by
a vendor. Several time periods exceeding one month between course
and satisfactory completion of take-home exams were identified.
Training representatives noted that the identified issues would be
addressed.
- c. Conclusions
Hazmat training provided to 3ersonnel handling radioactive
materials was conducted at t1e appropriate frequency. and included
recent changes to DOT regulations. The training provided was
effective, as demonstrated by the knowledge and proficiency of
interviewed staff members. Identified issues observed included
training data base upgrades and timeliness in completion of hazmat
examination.
R7 Quality Assurance in Radiation Protection and Chemistry Activities
R7.1 Licensee Self-Assessment Activities
, a. Insoection Scooe (84750. 86750)
! During the inspection period, the following audit reports and
associated checklists regarding Radiation Control; and Radwaste
processing packaging and transportation program activities
recuired by TS 10 CFR Part 20. and 10 CFR Part 71 were reviewed
anc discussed with licensee representatives.
o Hatch Safety Audit and Engineering Review (SAER). Audit of the
Radioactive Waste Shipping Program. Report Number
(No.) 94-RW-1 dated November 10. 1994.
o Hatch SAER Audit of the Radioactive Waste Program. Report No.
95-RW-1. Dated October 5. 1995.
,
In addition. initial results of a Low Level Waste Characterization
Study, conducted by the Electric Power Research Institute, dated
May 1996. were reviewed and discussed.
b. Observations and Findinas
The audits addressed Process Control Program (PCP), waste
processing, chemistry, radiological controls, radwaste and
Enclosure 2
_ _.
.
.
28
transportation program guidance. implementation and documentation.
Both compliance-based and performance-based strengths, issues,
weaknesses and recommendations were documented. The audits
included review and followup of previously identified items. In
particular, the inspectors reviewed and discussed licensee
followup actions regarding labeling issues identified. No similar
issues were identified during the current review of radiation
control activities.
The low-level waste characterization study results provided
quantitative data regarding sources of radwaste generation and
cost-effective reduction efforts. The project evaluated existing
generation, minimization, processing, and disposal programs and
provided potential mechanisms for reducing costs. Licensee
management informed the inspectors that final results will be used
to focus on radwaste program initiatives.
c. Conclusions
Audits for the radwaste and effluent processing programs were
thorough and comprehensive and met TS. 10 CFR Part 20 and
10 CFR Part 71 requirements. The low-level waste characterization
study was identified as a licensee initiative to quantify and
'
prioritize radwaste program activities.
R8 Miscellaneous RP&C Issues l
a. Insoection Scooe (84750. 86750)
The status of selected radiation control and radwaste performance
indicators were reviewed and discussed with licensee
representatives.
b. Observations and Findinas l
Since 1993, annual dose expenditure per unit outage continued to
decrease and dose expenditures were met. As of November 12. 1996,
dose expenditure was approximately 422 person-rem relative to the
575 person-rem goal.
For 1995 and 1996 year-to-date (YTD) approximately 2.382 and
121.919 curies (Ci) of radwaste were sent to a licensed burial
facility for disposal. The majority of material buried consisted
of resins in 1995 and irradiated hardware in 1996. Waste disposal
volumes for 1995 and 1996 were 10.047 and 9.338 cubic feet (ft')
respectively.
No reduction in personnel contamination events. 177 reported in
1995 and 174 reported YTD in 1996. was observed.
c. Conclusions
No significant declining performance trends were observed for the
performance indicators reviewed.
Enclosure 2
_ _. _ ._ _ _ _ _ . . _ _ _ . _ __ _ _ _ . _ - - . _ . -
- __ _ _
,
- i
l
l'
.
'
29
P4 Staff Knowledge And Performance In EP
P4.1 NOUE Associated with the Inadvertent Lift of Nitrocen System
Relief Valve
a. Insoection Scoce (71750)
On October 28, 1996, while attempting to fill the Unit 2 nitrogen
storage tank, a relief valve on the fill line lifted (See
Section E1.1). The valve discharged into the tank room, partially
filling the room with nitrogen. This resulted in the declaration
of a Notification of Unusual Event (NOUE) due to release of toxic
, gas in accordance with the licensee's EP. The inspectors reviewed
l control room logs and interviewed operators and members of the
emergency planning staff.
'
- b. Observations and Findinas
On October 28, 1996, plant operators were conducting a routine
refill of the Unit 2 nitrogen storage tank in accordance with .
34S0-T48-002-25. " Containment Atmospheric Control and Dilution i
Systems." During the fill, a relief valve on the fill line
lifted, discharging nitrogen into the tank room. Partially
'
filling the room with nitrogen gas caused the conditions in the
room to meet the EP definition of " toxic gas release" This
resulted in the declaration of a NOUE. as defined by section 11.3. l
" Hazards to Plant Operation. Toxic Gas." of the licensee's EP.
The event occurred at approximately 11:00 A.M. The shift crew
began to take actions to secure the nitrogen release and re-
establish the appropriate atmosphere in the tank room. At
approximately 11:15 A.M. , the nitrogen release was secured. At
,
approximately 12:00 P.M. . the shift recognized that the conditions
i warranted entry into the EP. The NOUE was declared at 12:20 P.M. ,
and the initial notifications were made within the required time
frame. The NOUE was terminated at 1:04 P.M., when oxygen levels in
the room were determined to have returned to normal.
The inspectors reviewed the licensee's activities and procedures
associated with the event. In accordance with the requirements of
Section A of the EP, the licensee is responsible to " Recognize and
declare the existence of an emergency condition." Implicit in
this requirement is that a decision to enter the EP and classify
If the
.
! the emergency must be made in a timely manner.
'
classification is not made promptly, following the availability of
indications that an emergency condition exists, the goal of the
classification scheme is undermined and the intent of Emergency
l Preparedness regulations would not be met. During this event,
ample time and opportunity existed for the recognition and
declaration of the appropriate emergency classification. The
failure to recognize and declare the existence of an emergency in
a timely manner is a weakness of the implementation of this
portion of the EP.
Enclosure 2
I
l
l
!'
30
'
An event similar to this occurred on October 5, 1994. This event
involved the inadvertent release of carbon dioxide gas into the
control building which resulted in the declaration of a NOUE.
This event occurred at 9:55 A.M. , and the NOUE was not declared
until 11:13 A.M. Corrective action from this event included
training for operators on interpretation of the toxic gas
emergency action level (EAL) and an evaluation as to whether the
EAL was appropriately titled. The training for operators was
conducted in the first 1995 requalification training cycle.
c. Conclusion
While actions to implement the EP after declaration of the
emergency were good, the delay in recognition and declaration of
'
the emergency represents a weakness. This is especially
noteworthy since there had been corrective actions taken as the
result of a recent similar event.
S2 Status of Security Facilities and Eauioment
The inspectors toured the protected area and observed that the
perimeter fence was intact and not compromised by erosion nor l
disrepair. The fence fabric was secured and barbed wire was '
angled as required by the licensee's Plant Security Program (PSP).
Isolation zones were maintained on both sides of the barrier and
were free of objects which could shield or conceal an individual.
The inspectors observed that persoMel and packages entering the ;
protected area were searched either by special purpose detectors 4
or by a physical patdown for firearms, explosives and contraband.
Badge issuance was observed, as was the processing and escorting
of visitors. Vehicles were searched, escorted and secured as
described in the PSP. ,
The inspectors concluded that the areas of the PSP inspected met
the PSP requirements.
V. Manaaement Meetinas
X. Review of UFSAR Commitments
A recent discovery of a licensee operating its facility in a
manner contrary to the Updated Final Safety Analysis Report
(UFSAR) description highlighted the need for a special focused
review that compares plant practices, procedures and/or parameters
to the UFSAR description. 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.
X.1 Exit Meeting Summary
The inspectors presented the inspection results to members of the
licensee management at the conclusion of the inspection on
Enclosure 2
.
.
.
31
December 19. 1996. The licensee acknowledged the findings
presented. ,
!
.The inspectors asked the licensee whether any materials examined
during the inspection should be considered proprietary. No
proprietary information was identified.
PARTIAL LIST OF PERSONS CONTACTED
Licensee
Anderson J., Unit Superintendent
Arnold. B., Chemistry Supervisor
Bennett. D., Health Physics Superintendent
Betsill . J. , Operations Manager
. Coggin. C. . Engineering Support Manager
Coleman. V., Chemistry Supervisor
Curtis. S. , Operations Support Superintendent
Davis. D., Plant Administration Manager
Fornel. P., Performance Team Manager
Fraser. 0. Safety. Audit and Engineering Review Supervisor
Hammonds. J., Regulatory Compliance Supervisor
Kirkley, W., Health Physics and Chemistry Manager
Lewis, J., Training and Emergency Preparedness Manager
Moore. C.. Assistant General Manager - Plant Support
Reddick J., Health Physics Supervisor
Reddick, R., Site Emergency Preparedness Coordinator
Roberts. P., Outages and Planning Manager
Smith D., Chemistry Superintendent
Sumner. H., General Manager - Nuclear Plant
Thompson. J. , Nuclear Security Manager
Tipps. S.. Nuclear Safety and Compliance Manager
Wells.. P. , Assistant General Manager - Operations
,
!
!
!
Enclosure 2 i
j
- = - - . . .. . . - - . . - - .. . -- . .- -.- - .-
L- ;
,
32
l
l
INSPECTION PROCEDURES USED
IP 37551: Onsite Engineering !
Effectiveness of Licensee Controls in Identifying,
'
IP 40500:
Resolving, and Preventing Problems i
IP 61726: Surveillance Observations !
'
IP 62703: Maintenance Observations
IP 62707: Maintenance Observations
IP 71707: Plant Operations !
IP 71714: Cold Weather Preparations l
Plant Support Activities
,
l IP 71750: <
l IP 84750: Radioactive Waste Treatment, and Effluent and j
l Environmental Monitoring
I
IP 86750: Solid Radioactive Waste Management and Transportation
of Radioactive Materials 1
IP 92700: Onsite Follow-up of Written Reports of Non-routine
Events at Power Reactor Facilities
! IP 92901: Followup - Operations
l IP 92902: Followup - Maintenance / Surveillance
IP 92903: Followup - Followup Engineering
l 2515/130: Im roved Standard Technical Specification
Im lementation Audit
! 2515/133: Im lementation of Revised 49 CFR Parts 100-170 and
,
'
ITEMS OPENED. CLOSED AND DISCUSSED
Ooened
50-321.366/96-14-01 NCV Failure to Include Surveillance l
'
Requirements in Applicable
Procedures (Section 03.1).
50-321.366/96-14-02 IFI Potential Single Failure i
'
Vulnerability in the Freeze
Protection System (Section M2.1).
'
50-321.366/96-14-03 VIO Failure to Implement
Configuration Control
Requirements - Multiple Examples
(Sections M2.2. E1.1. E8.2, and
E8.4).
50-321.366/96-14-04 IFI Potential Deficiencies in the
HPCI Surveillance Procedure
(Section M3.1).
l 50-321/96-14-05 IFI Restoration of 1B EDG Motor
Control Center (Section E2.2).
Enclosure 2
.
.
33
50-321.366/96-14-06 NCV Failure to Maintain' Records
Required by 10 CFR Part 71.91 for
Routine Determinations Specified
by 10 CFR 71.87 for a Type B
Shipment (Section R1.3).
50-321.366/96-14-07 URI Determine If Processing and
Control of C0C and Associated
Documents for Package Type
USA /5805/B( ) Vendor Documents
Were in Accordance with AC
Procedure 20AC-ADM-003-05,
" Vendor Manual Review and
Control" (Section R1.3).
Closed
50-321,366/96-14-01 NCV Failure to Include Surveillance
Requirements in Applicable
Procedures (Section 03.1).
50-321/96-13-06 IFI Additional Review of Wiring
Deficiency on Electrical Drawings
(Section E8.4).
50-321/96-10-01 VIO Inadequate Procedure for
Verifying Plant Service Water
Pump Discharge Valve Positions
(Section 08.2).
50-321.366/96-07-02 VIO Failure to Conduct Testing
Following Molded Case Circuit
Breaker Instantaneous Trip
Setpoint Changes (Section E8.1).
50-321/96-05 LER Loss of Power to Reactor
Protection System Bus Results in
Actuation of Engineered Safety
Featurer (Section E8.2).
50-321/96-04 LER Inadequate Procedure Results in
Reactor Pressure Increase and
Automatic Reactor Shutdown
(Section 08.3).
50-321/96-01 LER Clogged Servo Valve Strainers
Results in Turbine Control Valve
Closure and an Automatic Reactor Scram (Section E8.3).
50-321.366/94-01-01 IFI Plant Service Water System Flow
Model Verification (Section
E8.5).
Enclosure 2
_ _.__._.___ _.__ _._ _ _ _.._ _._ _ _ _...___._.._ - _
= ,
, i
34
50-321.366/94-01-02 URI High Energy Line Break Protection
for Service Water Systems within ,
'
the Containment (Section E8.6),
LIST OF ACRONYMS USED .
!
l ALARA- As Low As Reasonably Achievable
AP -
Administrative Procedure
,
C1 -
Curie
!
CFR - Code of Federal Regulations
COC - Cartificate of Compliance
F - degrees Fahrenheit
DC - Deficiency Card ;
DCN -
Design Change Notice i
'
DCR -
Design Change Request
DOT -
Department of Transportation
EAL - Emergency Action Level
EDG - Emergency Diesel Generator
EP -
EPRI - Electric Power Research Institute
ft -
foot (feet)
FME - Foreign Material Exclusion
FSAR - Final Safety Analysis Report
GPC - Georgia Power Company
gpm - gallons per minute
Hazmat- Hazardous Material
HP - Health Physics
HPCI - High Pressure Coolant Injection
IFI -
Inspector Followup Item
IP -
Inspection Procedure
IR -
Inspection Report
ISI - Inservice Inspection
IST -
Inservice Testing
Improved Standard Technical Specifications ~!
ITS -
KV -
kilovolts ,
LC0 - Limiting Condition of Operation i
LER - Licensee Event Report
LPCI - Low Pressure Coolant Injection
MCC - Motor Control Center
MCCB - Molded Case Circuit Breaker
MSIV - Main Steam Isolation Valve
MWO - Maintenance Work Order
NCV - Non-cited Violation
NOUE - Notice of Unusual Event
NRC - Nuclear Regulatory Commission i
NRR - Nuclear Reactor Regulation
ODCM - Offsite Dose Calculation Manual
PCE -
Personnel Contamination Event
PDR - Public Document Room
PM -
Preventive Maintenance
POD - Plan of the Day
PRB - Plant Review Board
Enclosure 2
__ _
1
-
35
PSIG - Pounds Per Square Inch Gauge
PSP - Plant Security Program
PSW -
Plant Service Water
QA -
Quality Assurance
OC -
Quality Control
RAS - Recuired Action Statement
RC -
Raciation Control
FCA - Radiological Controlled Area
,
RCIC - Reactor Core Isolation Cooling
Rev - Revision j
RFPT - Reactor Feedwater Pump Turbine i
RG - Regulatory Guide
RHRSW- Residual Heat Removal Service Water ,
RPS -
RSDP - Remote Shutdown Panel i
RTP - Rated Thermal Power l
RWP - Radiation Work Permit !
SAER - Safety Audit and Engineering Review ,
SER -
Safety Evaluation Report i
SPDS - Safety Parameter Display System i
SR - Surveillance Requirement
SS - Shift Supervisor
TCV - Turbine Control Valve
TRM -
Technical Requirements Manual
TS - Technical Specifications ,
TSC - Technical Support Center !
TSIP - Technical Specification Improvement Program !
UFSAR- Updated Final Safety Analysis Report
URI - Unresolved Item ;
VAC -
Volts Alternating Current l
VIO -
Violation
V0TES- Valve Operation Test and Evaluation System ,
YTD - Year-to-Date l
Enclosure 2
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ATTACHMENT 1
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PARTIAL LIST OF PERSONS CONTACTED
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PECO Enerav
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i R. Boyce, Plant Manager
j .W. Sproat, Director Engineer
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G. Johnston, Director Maintenance
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R. Bickhart, Lead Assessor, Nuclear Quality Assurance
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G. Bird, Manager, Emergency Preparedness
- M. Karney, Manager, Security / Emergency Preparedness, Limerick
! D. LeQuia, Director, Site Support
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S. MacAinsh, Manager, Support Services
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C. Mengers, Manager, Limerick Quality Division
- G. Stewart, Engineer, Experience Assessment
P. Berry, Manager, Technical Support ]
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- W. Coyle, Manager Radwaste
D. DiCello, Radiation Protection Manager
M. Kaminski, Radwaste Engineer
NRC Reoresentatives
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R. Keimig, Chief, Emergency Preparedness and Safeguards Branch
N. Perry, Senior Resident inspector, Limerick
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F. Rinaldi, Limerick Project Manager
- J. Shea, Peach Bottom Project Manager
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D. Jaffee, Acting Limerick Project Manager
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- c. Conclusions ;
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- The inspector noted that management expected to complete reviews for the abandonment
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of selected radwaste equipment in December 1996. The updating of the UFSAR in
accordance with 10 CFR 50.71 (e) remains open pending additional NRC review.
R8.2 Plant Tour Observations
During the inspection, the inspector made various tours of the RCA. The inspector's
j review indicated an apparent need for improved housekeeping. During the tours, the
- inspector noted oil leaks, wrenches on the floor, metal wire clippings, and rags left in l
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The inspector noted plant personnel were preparing for a periodic station housekeeping
j activity. !
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- V. Manaaement Meetinas a
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j X1 Exit Meeting Summary ;
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The inspector presented the inspection results to members of plant management at the
conclusion of the inspection on December 16,1996. The plant manager acknowledged
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the inspectors' findings. The inspectors asked whether any materials examined during the
j inspection should be considered proprietary. No proprietary information was identified.
l X2 Review of UFSAR Commitments
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- A recent discovery of a licensee operating their facility in a manner contrary to the UFSAR
- description highlighted the need for a special focueed review that compares plant practices,
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procedures and/or parameters to the UFSAR description. While performing the inspections
discussed in this report, the inspector reviewed the applicable portions of the UFSAR that
related to the areas inspected. The inspector verified that the UFSAR wording was
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consistent with the observed plant practices, procedure and/or parameters, in addition, the !
inspectors reviewed portions of the Emergency Plan, since the UFSAR does not specifically l
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include emergency preparedness matters. The inspectors specifically evaluated Section 5 1
- concerning laboratory facilities and Section 6 concerning local support service personnel
- training for proper implementation. No discrepancies between Plan wording and
i implementation were noted.
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