ML20198N627
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| ML20198N627 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 12/29/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20198N614 | List: |
| References | |
| 50-302-97-19, NUDOCS 9801210149 | |
| Download: ML20198N627 (95) | |
See also: IR 05000302/1997019
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~'U.S. NUCLEAR REGULATORY COMMISSION !
REGION II .
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Docket No: 50-302 I
. License No: DPR-72- l
: Report'No: -50-302/97-19 ,
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Florida Power Corporation-
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' Licensee:- ;
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Facility: Crystal River 3 Nuclear Station .
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Location: -15760 West Power Line Street
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Crystal River, FL-. 34428-6708
t Dates: October'26-through November 29. 1997.
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Inspectors: S. Cahill, Senior Resident Inspector
T. '~ oper, Resident Inspector
S. C hez, Resident Inspector +
E P. FT . ' . Reactor' Inspector, sections El.2, E8.5,
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E8.-17
T. Johnson; Senior Resident Inspector Turkey Point,
c- Sections 08.1 E8.1-
E. Lea , Project Engineer, sections 08.2, M8.1 E8.21
L. Mellen,~ Reactor- Engineer, -section E8.22
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M. Miller, Reactor Inspector, sections El.1, E8.2 -
E8.4
R. Reyes. Resident Inspector Turkey Point, sections .
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-E8.14 E8.15
- G. Salyers Emergency Preparedness Specialist, section
P8.1
R. Schin, Reactor Inspector, sections 07.1 E8.6 -
E8,10; E8.16
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M. Thomas, Reactor Inspector. sections E8.11 - E8.13
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TApproved by: K. Landis, Ch'ief. Projects Branch 31
Division.of Reactor Projects
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Enclosure 2
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EXECUTIVE SUMMARY
Crystal River 3 Nuclear Station
NRC Inspection Report 50-302/97-19
This integrated inspection included aspects of licensee operations.
engineering, maintenance, and plant support. The report covers a 5-week
period of resident inspection: in addition. it includes the resu s of
announced inspections of open restart items by regional inspectors.
Doerations
The licensee's operational actions to respond to the perturbation in RWP-3A
performance were very good. They displayed an excellent sensitivity to
preserving several options of decay heat removal, took prompt interim
corrective actions and developed a methodical and appropriate action plan to
secure the pump and remove the suction blockage (Section 01.2).
The evolution to establish a pressurizer steam oubble was well controlled and
displayed good coordination among operators (Section 01.3).
The ins)ectors concluded that even though the number of identified clearance
errors las been increasing, the impact of the errors has been reduced and that
the licensee has been identifying the errors mainly through review programs
intended for that function (Section 02.1).
Although Operations' performance problems remained, they were generally minor
and promptly and appropriately corrected. The inspectors have observed that
the licensee has committed to continue to focus on improving )erformance in
this area long beyond resturt of the plant. Additionally, t1e licensee
consistently encouraged self-identification of problems by all groups.
Several of their initiatives required long time frames and were not deterrents
to restart of the plant. The inspectors concluded that overall Operational
performance was adequate (Section 04.1).
The inspector concluded the licensee was performing adequate training on the
new Emergency Operating Procedures and had diligently tracked and responded to
operator problems and questions. (Section 05.1).
The inspectors concluded that the licensee's progress to date on the
Management Corrective Action Plan (MCAP II) continued to be satisfactory and
there was one item related to licensing basis information which remained to be
accomplished prior to plant restart (Section 07.1).
The inspectors observed that Nuclear Quality Assessment (NOA) activities
continued to be appropriately focused. Licensee line management continued to
utilize NQA to follow up on suspected problems within the various departments.
giving management an independent assessment of performance. The licensee
devoted appropriate management attention to screening problems in the
Enclosure 2
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corrective action system. No recent, notable errors were observed (Section
07.2)..
The inspector determined the licensee *s corrective actions appeared
appropriate to address configuration control problems Recent trends have
been positive: however, continuing monitoring for effectiveness and management-
oversight is warranted (Section 08.1).
Maintenance
An instrument technician displayed a poor level of skepticism m J self-
checking when electing not to perform steps-in a procedure. This resulted in
an inadvertent actuation of the pressurizer spray valve that fortuitously had
minimal consequences. However, the iicensee recognized the implications of
the-problem and took appropriate corrective actions (Section M1.1).
The control of work in the protected trains'of engineered safeguards systems
was weak, because it lacked adequate procedural guidance and there was not a
-clear definition as to what constituted a protected train. A weakness was
also noted in changes to modification scope being issued directly to the field, bypassing the operations shift supervisor on duty (Section M2.1). Enoineerina The inspectors concluded that the licensee had implemented and completed superior programs for Generic Letter 96-01 and Decay Heat Closed Cycle System Failure Modes and Effects Analysis. Both programs reviewed were technically adequate and were implemented in accordance with licensee requirements, commitments and NRC regulations (Section E1.1). The inspector concluded that the revised emergency diesel generator loading calculations demonstrated that the generators have the capacity and capability to accept the design basis-loads as required by 10 CFR 50. Appendix A. Criterion 17 (Section E1.2). The inspector concluded that the licensen had performed a very good investigation and root cause determination. although the documentation was difficult to-follow and the corrective actions didn't clearly match the identified causes (Section E1.3). The-inspector concluded that the licensee's final assessment was thorough and adequately resolved any concerns with the potential for loose parts damage.in the RCS (Section E1.4). Several concerns were noted during the functional testing for the B emergency diesel.- Differences existed-between the test limits and precautions and approved annunciator response procedures. -The test allowed generator stator
. temperature levels above the installed meter's maximum reading capability.
The vendor reference manual for the temperature relay for generator stator
temperature was not updated after a change in the installed relay in 1979.
Enclosure 2
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" The.testilogs were weak and did not provide" sufficient detail to reproduce
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Jactions- taken duringLthe testing. -(Section E2.1).
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; A~ Violation (VIO 50-302/97-17-01)_ of 10.CFR 50.59 requirements was identified -
.for-an inadequate safety' evaluation of the modification functional test. !
procedure for the B emergency diesel generator (Section E2.1). j
A Non Cited Violation (NCV 50-302/97-17-02) was identified for performing work- :
=on safety related diesel generator clutch pads without approved procedures or- 4
work instructions (Section E2.1). i
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The. licensee's actions to correct-numerous tank parameters had appropriately
addressed the difficulties associated with the engineering structure, work- ;
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prioritization, and available. resources to )erform the work. The tank
calculations were thorough and the results lad-been appropriately incorporated.
into the required procedures (Section E8.15). l
-- The-inspectors concluded that the licensee's Control Complex-Habitability
Envelope (CCHE) leakage analysis, described in their November 10. 1997 letter 1
to the NRC . failed to recognize the potential for Control Rcom Emergency .
- Ventilation System (CREVS) fans to cause a substantial amount of CCHE-leakage -i
during accident' conditions. Licensee personnel stated that they would address-
the CCHE leakage due to CREVS fans in a revised submittal .to the NRC (Section 1
E8.16). j
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The inspector- agreed that the electrical cable operability evaluation had a
sound basis. An Inspector Follow-up item (IFI 50 302/97-17-03) was ,
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established to ensure NRC review of the final or-long_ term resolution of the
- cable ampacity issue (Section E8.17). ;
i A Violation (VIO 50-302/97-17-04) was identified for inadequate design control
related to thermal relief valves .,a various. heat exchangers (Section E8.20).
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- - An-Inspector Follow-Up Item (IFI 50-302/97-17-05) was established to track the
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resolution of.: improved. Technical Specification setpoint program deficiencies
prior-to entry into Mode 4 (Section E8.22).
- Plant SuDDort
A review of an open item on inconsistent Emergency Action level classification
, determined that the licensee had developed diverse and challenging training-
. scenarios and.had _ adequately. addressed the original concerns (Section P8.1).
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Enclosure 2
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. The inspectors assessed the licensee's performance in the five areas of continuing NRC concern in the following . '
sections: the assessments are limited to the specific issues addressed in the respective sections.
NRC AREA 0F CONCERN ASSESSMENT SECTION
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0 0 0 0 0 M E E E E E E E E E E E E E E E E E E E- E E P
4 7 8 8 8 8 1 1 8 8 8 8 8 8 8 8 8 8 8 8 8 8 8 8 8 8 8 8 -i
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0 1 2 3 4 s 7 8 9 1 2
P.anagement Oversight G G G G A G S G A S G G G G G A A G G G G G G G G A G G .
Engineering Effectiveness G S G G S G G G G A A A G G G A G G A G A 5
Knowledge of Design Basis A G G G G A G G A G G A A A G G G A G A A A S G
Compliance With A G A G A G G G G G G G G G G A A G G G A A G A A A S G
Regulations
Operator Performance A A G A G G A G G
S = Superior G = Good A = Adequate /Acceptabie I = Inadequate i
Blank = Not Evaluated / Insufficient Infcrmation
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Enclosure 2
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04.1 Operator-Performance and Communication Observations.
07,2- Licensee Self-Assessment Activities
08.1 (Closed) VIO 50-302/97-02-01: Failure to follow Equipment Control
Procedure Requirements (FPC Restart Issue 0-13A)
08.2 (Closed) LER 50-302/96-21-00: Delayed Entry Into Technical Specification
Required Action Caused by Inadecuate Documentation of Out-of Service
Equipment Requirements for a Mocification [
08.3 Reportability Program
(Closed) EA 97-094 (4 examples: 01013, 01023, 01033, and 01043): Repeat
Failure to Make Timely Reports to the NRC
(Closed) VIO 97-08-01: Inadequate Corrective Action and Procedure for
External Reporting Requirements
:M8.) (Closed) LER 97-002-01: Out of Calibration Fuel Pool Water Level
Transmitters
(Closed) VIO 97-01-04: Failure to Perform Technical Specification
Surveillance for Spent Fuel Level
E1.1 Design Control Process
El.2 Emergency Diesel Generator Loading Calculations
E8.1 (Closed) VIO 50-302/96-08-01: Failure to Take Timely Corrective Action
to Address Issues and Actions For Makeup System Audit Findings and
Excessive Vibration on a Spent Fuel Pool (SFP) Pump Fan Motor (FPC
Restart Issue OP-24)
E8.2 (Closed) LER 96-011-00 LER 96-025-00, and LER 97-003-00 through 005:
Personnel Errors Caused Testing Deficiencies (GL 96-01)
E8.3 (Closed) VIO 50-302/97-05-03: Incorrect Information in Annunciator
Response Procedure for Inverters.
E8.4~ (Closed) VIO 50-302/97-07-01: Failure to Follow Procedure CP-111 for the
Processing of Precursor Cards (PC)
E8.5 (Closed) URI 50-302/96-201-07: EDG Not Protected Against-Water Spray
-from the Fire Protection System Sprinkler
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E8.6 (Closed)-EA 95-126. VIO I.C.2 (04013): Corrective Actions for an
Inadequate Curve 8 (Two STI's and a Revised Curve 8A and 88) were Also
-Incorrect
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E8.7 (Closed) EA 96-365 C-(03013): -Inadequate Corrective Actions for 10 CFR
50.59 Evaluation Errors for Inadequate-Containment Peneration
Surveillance-
E8.8 (Closed)-EA 97-162 (01013): -Inadequate Safety Evaluations for Added
Operator Actions for Design Basis SBLOCA Mitigation
EB.10 (Closed) LER-96-24-01: Plant Modification Causes Unanalyzed Condition
Regarding Emergency Feedwater
E8.11 (Closed) EA 96-365. EA 96-465. EA 96-527. VIO B (Example 1) (02013): '
Failure to Update Applicable Design Documents to Incorporate Design
Information
E8.12 (Closed) EA 96-365. EA 96 465. EA 96-527. VIO B (Exam)le 2)-(02013):
Failure to Include Applicable Design'Information in t1e Design Input
Requirements for a Modification
E8.13 Followup on Restart Issue Resolution - BWST NPSH Concern (FPC Restart ,
Issue D-18)
E8.14.(Closed) LER 50-302/97-017-00: Personnel Error Caused Inadequate
Electrical Separation Of High Pressure Flow Indicators (FPC Restart
Issue D53A)
E8.15 (Closed) VIO 50-302/EA 95-126 NOV II.B: Failure to take adequate
corrective action for required tank volumes level, and suction points.
(FPC Restart Issue OP-12)
E8.17 (Closed) VIO 50-302/97-01-09: Inadequate Corrective Actions for Cable
Ampacity
(Closed) LER 50-302/97-31-00: Inadequate Cable Sizing Due to
Nonconservative De-rating Factors Could Reduce the Cable Remaining
Qualified Life
E8.18 (Closed) IFI 50-302/97-02-05: Outstanding Issues Associated with the
Emergency Diesel Generator Power Upgrade Modification
E8.19 (Closed) VIO 50-302/97-11-06: Failure to Follow Licensee Procedure NEP-
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E8.21'(Closed) VIO 50-302/96-09-06. Erroneous Calculation Inputs and Inservice
.-Inspection Boundary
(0 pen) LER 50-302/97-038: Engineering Oversight Resulted in Operation
Outside Design Basis of Waste Disposal System
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E8.22 (Closed) EA 95-16: Use of Nonconservative Trip Setpoints for Safety-
Related Equipment
(Closed) LER 50-302/94-006-00 through LER 50-302/94-006-06: Deficiency
in Understanding of Technical Requirements Leads to Nonconservative
Safety Systems Setpoint and Violations of Improved Technical-
Specifications.
P8.1 (Closed) IFI 50-302/97-08-03: Variations in the Classification and
Interpretation of the EALs by the Emergency Coordinators.
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Report Details:
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Summarv of Plant Statusi
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The' unit" remained in Mode 5 through the inspection period; continuing in the- -
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outage that began on September 2. 1996. .The reactor coolant system (RCS).
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started the period filled to a normal- pressurizer level with a nitrogen over *
-pressure of approximately 40 psig. Train "A" of forced decay hoat removal
system flow was operable and in service to support train ~B" maintenance. .
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modifications, and testing.for the B Emergency Diesel Generator (EDG) radiator' :
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uagrade;and other routine emergency equipment train-related work. Both once- !
t1 rough steam generators (OTSG) remained filled to a normal inventory with.a .
nitrogen blanket, and'one was always' preserved as available to support useLas -
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a backup-decay heat sink.Lif needed. 0n November 18, 1997 a vacuum was.
established:in the main condenser using auxiliary steam. On November -20. 1997 ,
'a pressurizer steam bubble was established to control RCS pressure.
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L. 00erations
01: -Conduct of Operations
- 01.1 General Comments (71707)
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. Using. Inspection Procedure 71707 the W pectors performed routine '
reviews of plant-operations which int: Jed shift turnovers, response to'
emergent problems. log reviews, coordi:.ation meetings, and restart
e activities. Significant observations are discussed in the following
paragraphs. ;
W 01.2 Raw Water Pumo Discharae Pressure Perturbation (71707)
On October 26, 1997, an intermittent. low suction alarm was received on
- the operating Nuclear Services and Decay Heat Sea Water (RW) pump RWP-
. 3A. The alarm cleared, but the pump discharge pressure stabilized at a i
lower than expected-value. The operators questioned this, and the
licensee promptly initiated a thorough investigation. They verified
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that vibration levels were normal: temperatures of com)onents cooled by
1 RW were normal, but the discharge pressure was below tie low acce3tance
limit for the RW pump surveillance. Consequently they declared t7e pump
inoperable per Technical Specification 3.4.6 and complied with the-
Limiting Condition for Operation (LCO). The corresponding B train RWP
and Decay Heat Removal System (DH) comaonents were unavailable due to
modification and maintenance work so t1e licensee technically did not
have a. fully operable train.of decay heat. removal. However, they.
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-developed a comprehensive action plan with contingencies-for other
acceptable methodsLof decay heat removal and. developed appropriate
procedural guidance for each of their contingency and action plan items.
They also frequently monitored RWP-3A and verified its parameters were
not. degrading and that it was fulfilling its core cooling function. The
l licensee immediately suspended the B train work, significantly
.]erturbating their outage schedule..and initiated actions-to restore
RWP-38- and :the;B train DH equipment to service. On October 30, 199/.
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after: verifying alternate decay heat removal options were available,
they expeditiously secured RWP-3A. sent divers into the suction pit to
investigate the source of the low discharge pressure, and restored the
pump to service. Discharge pressure increased to tha expected value and
the pump operated normally after removal of a wooden shim with an
attached: lanyard that was partially biccking the suction of the RWP.
The shim had been used earlier in the month for maintenance on the
seawater intake structure that eventually feeds the RWP suction. The
licensee also determined that an identical piece of wood had been noted
ficating in the intake during that work -but no action to account for it
had been taken. This raised significant concerns with the inspector and
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licensee management regarding the adequacy of foreign material exclusion
practices used for that work and why the wood had not been retrieved.
The licensee initiated a root cause investigation which was not
completed at the end of this report period. The inspector will review
the completed investigation to disposition the above ccncerns.
Regardless, the licensee's operational actions to respond to the
perturbation in RWP-3A performance were very good. T1e licensee
displayed an excellent sensitivity to preserving several options of
decay heat removal, took prompt interim corrective actions and developed
a methodical and appropriate action pl a to secure the pump and remove
the suction blockage.
01.3 Establishment of a Steam Bubble Inside the Pressurizer (71707)
On November 20, 1997, the ins)ectors observed activities in the control
room associated with the esta)lishment of a steam bubble inside the
pressurizer using the pressurizer heaters. The evolution observed by
the inspector was slow and well controlled, in part due to several
pressurizer heater groups being out of service for maintenance. These
out of service heaters limited the rate at which the operators could
heat up the water inside the pressurizer. Once the steam bubble was
formed. the operators established saturation conditions inside the
pressurizer by utilizing local vent valve RCV-227 to the reactor coolant
drain tank. Good communication was observed between the control rcom
operators and the operators at the vent valve. Saturation conditions
were maintained at approximately 50 pounds per square inch gauge (psig)
RCS pressure and 297 degrees Fahrenheit pressurizer temperature. The
bulk RCS temperature during the pressurizer bubble evolution remained
constant at approximately 81 degrees Fahrenheit. The inspectors
concluded that the overall pressurizer bubble evolution was performed
effectively.
Enclosure 2
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02 Opetational Status of Facilities and Equipment
02.1 Use of Clearances and Taaaino__0rden
a. Insoection Scoce (71707)
The inspectors performed a follow-up of several recent errors in the
control of clearances in the plant. In addition the inspectors
reviewed the licensee's clearance error trend analysis performed for the
period from May 1997 through October 1997,
b. Observations and Findinos
On November 3. 1997. Precursor Card (PC) 97-7568 was written to document
that the C main condensor water box discharge Amertap screen was found
open contrary to the required position on an active clearance.
Investigations revealed that the clearance required the screens to be
tagged in the "open" position, but the controls for the screens had two
labeled positions: " Operate" and " Backwash". The clearance tag
terminology and the in plant control labels did not match, leading to
confusion on the part of the ]lant operator. The operator tagged the
control switch in the Backwasl position, with the screen open, rather
than in the Operate position, with the screen closed. The licensee
repositioned the screen to the correct position. New tags have been
manufactured to state screen positions in Operate and Backwash modes.
The licensee verified that no manways were o]en on the condensor outlet
waterboxes, preventing a security breach. T1ere was no impact on either
safe operation of the pl6nt, security integrity or personal safety.
On November 7. 1997. PC 97-7641 was written to document an error on
clearance 97-10-017 for work on the B building spray (BS) system. The
licensee discovered that valve BSV-99 was tagged closed on the A BS
train instead of BSV-98 on the B BS train. The licensee verified that
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BSV-98 was locked closed during this period, but was not tagged. This
l error in the development of the clearance order was missed by all
reviewers. At the time of discovery, the licensee was in the process of
releasing the clearance. The licensee verified that the clearance error
did not compromise the safety of any work performed on the system.
The licensee conducted a trend analysis of clearance errors from May
1997 through October 1997. The analysis revealed a slowly increasing
trend in errors during that period. The licensee concluded thac
approximately 60 percent of the clearance errors during that period were
attributable to administrative preparation, approval, and release of
clearances. The inspectors reviewed the trend analysis and determined
that-the licensee *s conclusions and proposed corrective actions were
appropriate,
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Enclosure 2
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c. Conclusions
The number of identified clearance errors is slowly increasing. The
licensee has identified this trend and is taking actions in an attempt
to correct this trend. Th inspectors concluded that even though the
number of identified errors is iricreasing, the impact of the errors has
been reduced and that the licensee is identifying the errvs mainly
through program barriers intended for that function. The errors that
are occurring appear to be mainly personnel attention to detail types of
problems and not a programmatic issue. The inspectors will periodically
monitor the licensee s program to assess the effects of the proposed
improvement initiative.
(" Operator Knowledge and Performance 04.1 Ocerator Performance and Communication Observations
a. In12ection
n Scope (71707)
The-inspectors continued to assess examples of Operations performance
for improvement in operator's questioning attitudes and communications
3ractices. Operations Readiness is a restart restraint item on the NRC
Restart List.
b. Observatioris and Findinas
As discussed in several 3revious reports, minor problems continued to
occur, indicative of weatnesses in Operations * communications with other
departments and inconsistent questioning attitudes. One example was a
late Surveillance Procedure (SP)-157A completion on November 13. 1997
when a manual calculation, necessary to compensate for an out of service
meteorological tower instrument, was informally turned over to an
oncoming shift and not completed within the required surveillance
periodicity. Another example was inappropriate procedural guidance
observed by an inspector in entry 9711.06 to the Operations Study Book
(0SB). An Operability Concerns Report (OCR) had been completed by
engineering to address concerns with spurious trips of molded case
circuit breakers from PC 97-6906. Part of engineering's conclusion was
recommended actions to take for a single spurious trip. Operations
excerpted this guidance and issued it as an OSB entry to direct operator
action for a trip. The OSB is a tool to promulgate information of
interest to operators but not procedural direction. The licensee has
other mechanisms with appropriate reviews to issue direction, such as
Short Term Instructions. Operations imagement recognized the error and
removeo the inappropriate guidance.
An example of poor operator awareness and questioning attitude involved
a malfunction of the control room display computer that caused it to
stop processing data. The only outward indication of this was that the
digital clock on the display monitors did not advance. The malfunction
Enclosure 2
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occurred at approximately 2 a.m. but was not detected until after shift
turnover at 7:30 a.m. by the oncoming crew. The safety sJgnificance of ;
this oversight was minimal because plant conditions had not changed so
the data displayed on the monitors was still relatively accurate and :
main control board instrument indications were available and functioning
for all the computer parameters. While reviewing this problem, the !
inspector observed that several systems in the main control board
contain internal clccks, none of which were synchronized to indicate the
same time. The clocks on the plant display computer, annunciator i
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response computer. and a parameter trending computer all differed by '
several minutes. A prominently displayed digital clock on a monitor !
displaying video camera coverage of main steam relief valve tailpipes
was over 20 minutes incorrect. The Safety Parameter Display System
(SPDS) monitor did not indicate a time, but the inspector considered
it's internal clock likely to be inaccurate also because the operators
'1dicated that computer clocks were rarely synchronized and there was
not a requirement to periodically do so. The inspector considered this
problem to have minimal safety significance in the current plant
conditions, but if operating and a plant trip occurred, the licensee
would experience significant difficulty assembling a valid sequence of ,
events using the data from these systems. This would hinder their
investigation and correction of the cause of the trip. The licensee
recognized the implications of this problem, initiated PC 97-8007 for
corrective action, and was evaluating requirements to synchronize the
clocks automatically or perform a periodic manual synchronization.
Another exam)le involved a site drain system (SD) valve left e.t of
position on 90vember 1. 1997, due to a skipped procedure ste) following ,
a tank release. However, this example was identified soon t1ereafter by -
an oncoming Turbine Building operator, who recognized the incorrect
position on his rounds. The valve was a cross-tie for release radiation
monitors so its incorrect position did not result in any notable
problem. The o)erator generated PC 97-7539 to investigate and correct
the cause of tie error, ard Operations Management completed a thorough
ir.cident investigation. ' 2 second operator's observation was a good
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example of a questioning attitude, but the original operator's error was
an example of poor work practices.
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A positive example involved questioning of operators regarding an
inspector-identified test deficiency that created conflicting guidance
between the test procedure and the operator's annunciator response (AR)
procedures as fiscussed in Section E2.1. The operators consistently
responded that they wcald follow their AR procedure in the absence of
s)ecific' guidance in the test that allowed exceeding the AR limits.
T11s was also Operations managernt's expectation. The consistent
response indicated to the inspector that Operations' management
expectations were being translated and understood by their personnel.
Enclosure 2
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A last example involved an inadvertent positioning of 6 switch for :
emergency diesel voltage control by an operator during testing on :
November 13, 1997. The significance of the misposition was negligible l
because it transferred control to a local station and was immediately
recognized by the operator (due to an alarm) and corrected. The i
operator informed his supervision of the error and the licensee took l
appropriate action considering the operator self-identified the
deficiency. The inspector observed that the licensee consistently
encouraged self-identification of problems and focused on the solution
to prevent recurrence of the problems, versus punitive discipline. The
inspector considered this appropriete and supportive of their goal to
improve overall performance,
c. Conclusions
The inspectors concluded that these examples indicate that Operations'
performance problems remain but were generally minor and were being
3romptly and appropriately corrected. Examples of poor communications
)etween Operations and other organizations continued to occur, but the
licensee was identifying them and addressing the causes approprhtely.
The inspectors have observed that the licensee consistently encouraged
self-identification of problems by all groups. The inspectors have
observed that the licensee has committed to continue to focus on
improving performance in this area long beyond restart of the plant.
Several of their initiatives required long tima frames and were not
deterrents to restart of the plant. The inspectors concluded that
overall Operational performance was adequate.
The inspector assessed the licensee's performance, with respect to this
restart-related issue, in the five NRC ~ontinuing areas of concern:
. Management Oversight - Good
. Engineering Effectiveness - N/A
. Knowledge of the Design Basis - N/A
. Compliante with Regulations Adequate
. Operator Performance Adequate
05 Operator Training and Qualification 05.1 Emeroency Doeratino Procedure (EOP) and Peaualification Trainino
a. Inspection Scone (71707)
The inspectors observed a session of the licensee's training on E0Ps
done as part of licensed operator requalification training. The E0Ps
have undergone significart overall revision, and all operators were
being trained on the draft procedures, prior to finalizing and
implementing them. The inspector assessed the training for adequacy.
Enclosure 2
_ _ _ _ _ _ _ _ _ _
7
b. Observations and Findinos
The inspector observed strong oversight and ownership of his crew's
performance by the Shift Supervisor. Several malfunctions both minor
and significant, were run on the simulator to assess the crew
performance. The crew responded capably and the inspector observed that
they were very proficient at troubleshooting instrument and control
malfunctions. The training staff emphasized a logical approach to
diagnosing the failures, which the crew readily implemented. The
inspector observed some minor individual performance and communication
deficiencies which were also noted by the training staff and shift
supervisor and appropriately addressed with the individual. The
inspector also o) served some minor uncertainty with the intent and minor
difficulty implementing some of the new F0P steps. However, these were
diligently recorded by the training statt for resolution. Based on
interviews with several of the operators, the inspector determined that
their concerns and cuestions had been consistently recorded, and answers
were always providec to them in a timely manner. The inspector did not
have any notable concerns with the observed items.
c. Conclusions
The inspector concluded the licensee was performing adequate training on
the new E0Ps and had diligently tracked and responded to operator
problems and questions. No concerns were identified.
06 Operations Organization and Administration 06.1 On October 28, 1997, the licensee announced that the Energy Su) ply
Strategic Business Unit (SBU), along with the company's Power iarketing
Group and Purchased Power Resources, will be combined with Nuclear
Operations to form one SBU focused on the company's generation assets.
Senior Vice President of Nuclear Operations Roy Anderson will head the
new organization as Senior Vice President of Energy Supply. The
licensee expectea the change to be fully implemented in January 1998.
07 Quality Assurance in Operations 07.1 Manacement Corrective Action Plan (MCAP 11)
a, la r ction Scope (40500)
The NRC Confirmatory Action Letter to Crystal River of March 4.1997.
ruiuired that FPC achieve satisfactory progress on MCAP 11 before
restart of Unit 3. In September 1997, an NRC inspection of MCAP 11
concluded that the licensee had twelve MCAP 11 items on which additional
progress was needed arior to restart. The results of that inspection
were documented in MC Inspection Report (IR) 50-302/97-13. During this
inspection. the inspectors followed up on the status of those twelve
items.
Enclosure 2
8
b. Observations and Findinas
The inspectors reviewed the licensee *s MCAP 11 files and discussed
certain items with licensee personnel. The inspectors found that the
licensee had made satisfactory progress for restart on eleven of the
twelve items. The one remaining item was to provide adequate licensing
basis information and training to the plant staff to support operability
evaluations. Technical Specifications (TS) interpretations. and 10 CFR
50.59 evaluations. Licensee personnel described plans to accomplish
that item prior to plant restart.
c. Conclusions
The inspectors concluded that the licensee's progress to date on MCAP 11
continued to be satisfactory, and there was one item related to
licensing basis information which remained to be accomplished prior to
plant restart.
The inspectors had previously assessed the licensee's performance as
adequate, relative to MCAP 11. in IR 97-13. That assessment was not
affected by this followup inspection.
07.2 Licensee Self Assessment Activities (71707. 40500)
The inspectors reviewed various licensee self-assessment activities and
corrective action processes which included:
. Routine reviews of Nuclear Quality Assessments (NOA) activities
and surveillance report findings
* Reviews of precursor cards entered in to the corrective action
system
The inspectors observed that N0A activities continued to be
appropriately focused on fulfilling audit requirements and using
discretionary time to inspect suspected problem areas. Licensee line
management continued to utilize N0A to follow up on suspected problems
within departments to giving management an independent assessment of
performance.
The inspectors continued to review PCs entered in the corrective action
system to verify the licensee had addressed problems with screenirg
significance levels. These problems were discussed in previous reports
including IR 50-302/97-16. The inspectors have not identified any
significant errors in screening PCs since those observations. Minor
discrepancies still occurred but were recognized and corrected by
licensee management in their daily reviews of all the screening
committee decisions. The inspectors did not identify any notable
concerns and concluded the licensee is devoting appropriate attention to
the problem.
Enclosure 2
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The inspector assessed the licensee's performance, with respect to this
restert related issue, in the five NRC continuing areas of concern:
. Management Oversight - Good
. Engineering Effectiveness - N/A
e Knowledge of the Design Basis - N/A
e Compliance with Regulations - Good
. Operator Fr formance - N/A
08 Hiscellaneous Operations Issues l 08.1 (Closed) VIO 50-302/97 02-01: Failure to follow Eauioment Control
Procedure Recu1remonts (f PC Restart issue 013A)
a. Insoection Scone (92900.
This item addressed multiple examples of equipment incorrectly
positioned or configured, which revealed inadequate controls to maintain
the appropriate status of operational configuration management.
The inspector reviewed the licensee's root cause and corrective actions,
the violation response, and recent self assessments to monitor the
effectiveness of corrective actions. The inspector also independently
verified a sampling of the corrective actions and assessed the"
effectiveness.
b. Observations and Findinos
The licensee responded to the violation in a letter dated May 23, 1997
(3F0597-29). The licensee agreed with the violation, and stated the
reasons to be: poor work practices, insufficient field supervision, and
3rogram and procedure deficiencies. Additional causal factors assessed
)y the licensee included the following: a lack of self checking: at
times, a false sense of urgency. and an observed preoccupation by and
resultant distractions of personnel; supervision not spending adecuate
time !n the field; weaknesses in compliance procedures (CP)-115, huclear
Plan Tags and Tagging Orders, and in CP-113A. Work Request Initiation
and Werk Package Control; some operating procedure deficiencies: plant
labeling weakr. esses; needed Work Control Center (WCC) facility upgrades
and staffing increases; poor verbal and written communications: and,
inadequate training of operations personnel.
Licensee corrective actions completed and planned included the following
items:
-
Returned the equipment that was found out-of-position to its
required position as stated in the clearance or procedure in
effect
Enclosure 2
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- Developed a configuration control improvement program directed
towards operations performance improvements ,
- Conducted training sessions during a site wide stand down to
address configuration control issues and recent errors
'
- Conducted classroom and on the job training for operations
- Issued a required reading study book item
l
- Issued a night order book item addressing supervisory expectations
in the field
- Increased supervisory time spent in the field monitoring and
coaching
- Evaluated operator administrative duties, and eliminated those
whu could be performed by clerical assistance or WCC personnel
.
- Ash ; W iguration control to operator continuing training
-
Revised Procedures CP-113A and CP-115 .
- Revised those operating procedures which were identified to be in i
error
-
Performed an assessment of the effectiveness of the corrective
actions
-
Plans to upgrade plant labeling
-
Plans to upgrade the WCC facility (work in progress) and to
increase staffing
L
The inspector verified a sampling of the above stated corrective
actions. Observations of the tagout and clearance process were made i
from the control room the WCC, and the field. Selected clearances were
reviewed and walked down in the field. The inspector noted that the
labelling related corrective action has a long lead time, and is not due
for completion until December 1998. The inspector discussed
configuration issues and the above corrective actions with plant and
Operations management, operators and licensing personnel.
-
The inspector also reviewed recent line management and independent
assessments of the program and related enhancements. Recently performed
Nuclear Quality Assurance audits and surveillances concluded that the
configuration control program performance was adequate and recent trends
were improving. This was verified by the inspector's review of the 1997
Nuclear Quality Assurance assessments,
t
Enclosure 2
4
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11
c. Conclusions
The inspector determined the licensee's corrective actions appeared
approariate to address configuration control problems. Recent trends
have 3een positive: however, continuing monitoring for effectiveness and
management oversight is warranted. Based on the above reviews and
inspections, and on assessments of recent performance, the violation was
closed.
The inspector assessed the licensee's corrective action )erformance,
with respect to this restart-related issue, in the five 4RC continuing
areas of concern:
* Management Oversight - Good
. Engineering Effectiveness - N/A
e Knowledge of the Design Basis - Adequate
e Compliance with Regulations Adequate
. Operator Performance - Adequate
08.2 (Closed) LER 50-302/96-21-00: Dalaved Entry into Technical Specification
Reouired Action Caused by inadecuate Documentation of Out-of Service
Eauioment Reauirements f or a Moc i11 cation
a. Insoection Scope (92901)
This issue involved the licensee's delayed entry into a technical
specification required action involving an ino)erable reactor protection
system (RPS) channel. The inspector reviewed _icensee Event Re) ort
(LER) 96-21-00, which was issued in response to the delayed tec1nical
specification entry. The inspector reviewed associated documentation,
and interviewed licensee personnel to determined the adegaacy of the
licensee's response to the issue identified in the LER.
b. Observations and Findinas
The inspector noted that resolution of this LER was being tracked under
licensee Restart Issue 0-15. The inspector reviewed the documentation
associated with the LER and restart package 0-15. Included in the
documentation reviewed were the root cause of why the delayed entry into
the TS occurred and the corrective actions intended to prevent
recurrence. Following a review of associated documentation, the
inspector concluded that the licensee had identified the root cause and
contributing factors, and that the corrective actions were adequate.
The inspector verified that all corrective actions were completed. The
corrective actions included a detailed incident investigation per
Operations Instruction (01)-12. operations department study book entry,
and procedure revisions to 01-7. Control of Equipment and System.
Operating Procedure (0P)-502. Control Rod Drive System and Preventive
Maintenance Procedure (PM)-114. Control Rod Drive Mechanisms -
Electrical Checks.
Enclosure 2
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Although this item is a noncompliance with regulatory requirements, for
the reasons discussed in Inspection Report 97-21. the licensee meets the
criteria for enforcement discretion per Section Vll.B.2 of the NRC
Enforcement Policy as described in NUREG-1600. Consequently this item
is closed and is identified as another example cf Non-cited Violation
NCV 50-302/97-21 01. Examples of Noncompliances in Design Control.10
CFR 50.59 Evaluations. Procedure Adequacy / Adherence. Reportability, and
Corrective Actions That Are Subject to Enforcement Discretion.
C. Conclusions
The inspector concluded that the licensee's corrective actions for this
LER were satisfactory. This item is closed
The inspector assessed the licensee's performance, relative to the
corrective actions for this LER. in the five areas of continuing NRC
concern:
. Management Oversight - Good
. Engineering Effectiveness - N/A
* Knowledge of the Design Basis - Good
. Compliance with Regulations Good
. Operator Performance - Good
08.3 &ngrtabilitv Proaram
(Closed) EA 97-094 (4 examoles: 01013. 01023. 01033. and 01043): Renegl
f allure to Make limely Reports ta the NRC
(Closed) VIO 50-302/97-08-01: Inadeauate Corrective Action and Procedure
for External Reportino Reauirements
a. Insoection Stone (92901. 40500)
To close these items, the inspector reviewed the licensee's open item
closure packages for Restart Items OP-4 and OP-4A. which included their
response to the items and their program for reporting events and
conditions to the NRC as required by 10 CFR 50.72 and 50.73. This
included review of the licensee's procedures and processes, review of
recent reportability determination problems, and discussions with
operations personnel. The first violation was unsuccessfully inspected
for closure in Inspection Reaort 50-302/97-08. A significant deficiency
was f ound in the procedure w11ch resulted in the opening of the second
violation.
b. Observations and Findinos
The inspector reviewed the licensee's current procedures for
implementing the reporting requirements of 10 CFR 50.72 and 50.73: CP-
151. External Reporting Requirements. Rev. 3. dated October 6. 1997 and
Nuclear Operations Directive (NOD)-3. Reporting Requirements Program.
Enclosure 2
_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _
13
Rev. 13. dated August 19. 1995. The inspector noted that the CP-151
procedure definition of discovery time for determining the time limits
'or reporting events had been corrected. The term " Discovery Time" was
eliminated and replaced by " Time Limits for Reporting" which clearly
delineated the regulatory requirements. The inspectors interviewed the
16 clear Shift Managers (NSM) res)onsible for implementing the procedure
and determined they understood tie revised definition. The inspector
also determined CP-151 was generally clearly written and was an
acceptable procedure to sup) ort accurate and timely reportability
determinations. Procedure 100-3 was primarily a matrix summarizing
numerous routinely written reporting regulatory requirements and
responsible licensee personnel for each re) ort. The inspector did not
review the adequacy of the guidance for otler reports, only 10 CFR 50.72
and 50.73. The inspector noted one administrative discrepancy in that
LERs were required to be issued by the Licensing Manager and approved by
the Vice President for Nuclear Production. The licensee's current
3ractice is to have accountable de]artment managers issue the LERs.
3ased on the extended time since t7e last revision to N00-3 and numerous
recent licensee organizational changes the inspector considered it
appropriate to review and revise NOD-3. The licensee agreed and was
evaluating needed revisions.
Numerous reportability dew d iations have been performed since the
identification of the "econd violation. These have generally been
consistently timely aid technically valid. The inspector has observed
that licensee pers;nnel almost always document concerns via the PC
process and forward them to the NSM for reportability review in the same
shift. However, the inspector's review revealed several recent
deficiencies:
* IR 97-08 noted that CP-151 contained a new requirement for
tracking the outstanding re)ortability evaluations by the NSM.
The process prescribed by C)-151 required PCs classified as
potentially reportable items to be tracked. and a final
independent reportability determination to be made by the NSM
after receipt of a written technical evaluation from an assigned
Reportability Review Owner. If determined to be re)ortable. this
removed the PC from the potentially reportable traccing list and
placed it en the pending LER list. PC 97-2055 was originally
screened as a potentially reportable item per 10 CFR 50.73 by the
NSM on May 27. 1997. After receipt of the written technical
evaluation, the NSM upgraded the PC to reportable on June 10.
1997, with May 27, 1997 as the discovery date. However, the
inspector observed that a LER was not issued 30 days from the
discovery date of May 27, 1997, as required per 10 CFR 50.73.
Further investigation revealed that shortly after June 10, 1997
the Licensing and Engineering groups had determined that PC 97-
2055 was not reportable because the issue was covered as part of
the extent of condition for a previously reported LER. Although
this determination was correct, this knowledge caused Licensing to
Enclosure 2
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remove it from the pending LER list without-the PC being-
rescreened by the NSM. The last reportability determination ,
posted against the PC and signed by the NSM was that it was :
reportable. The' licensee generated PC 97-7565 when the inspector
identified this deficiency. and processed a reportability '
evaluation through-the NSM on November 6.1997 to recategorize PC
97-2055 as not re>ortable. The licensee's follow-up to PC 97 7565
determined that tie guidance in CP-151 was adequate but that this ;
deficiency was due to a personnel error. The inspector considered .
this an accurate assessment. l
e- PC 97-2485 was first screened by the NSM on April 6.1997 as not
reportable. An Engineering self-assessment caused the PC to be
re-evaluated and on October 8. 1997, it was determined to be
reportable. However, contrary to CP-151 requirements the NSM i
recorded the " discovery date" as October 8.1997, and not A>ril 6.
-1997. Although the correct discovery date was used in the .ER. PC 1
97-7751 was issued to correct the NSM misconception. l.
e PC 97-4530 was determined to be reportable per 10 CFR 50.72 and ;
50.73 on July 7. 1997. A four hour phone report per 50.72 was j
"
initiated. However, the PC was not tracked for reportability and '
the LER was not issued within 30 days as required. On August 29,
1997 the licensee became aware of the oversight-and initiated ;
*
actions to process the report. However. LER 97-27 was issued
October 3, 1997, which again didn't.make the 30 day limit from :
August 29, 1997. The licensee issued PC 97-6256 to document and 1*
correct the failure. Their apparent cause determination concluded
that their tracking methods were weak and periodic cross-checks
with the NSM logs were non_ existent. Corrective actions included 1
improvements to their tracking process and a weekly review by '
.
Licensing of- all items in various stages of re)ortability
determinations, which were verified against NSi logs to ensure all
items were contained in the database. The inspector verified
these actions and has not noted any similar problems since these -
actions were implemented.
* PC 97-6224 was initiated and screened for reportability on_0ctober '
5, 1997. Although subsequently determined to be not reportable. 3
the licensee's Engineering Manager recognized that the technical a
concern of this PC had been identified approximately a month i
earlier.'but a.PC had not been initiated. Therefore the t
reportability evaluation was late due to the delay in issuance of _ a
~ the PC. The licensee has identified several examples of untimely :
PC generation and has vigorously attempted to correct the problem e
and the resultant delay in reportability evaluations. PC 97-6950- ,
was generated by the licensee to capture these efforts.
. .For a potentially reportable PC that has been evaluated as not l
reportable by the Reportability. Review Owner. CP-151, section
'
.
Enclosure 2- -
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_ . _ . _ .
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15
4.4.5 required the NSM to review the recommendation, make an :
independent reportability determination. and re-perform sections !
4.2 and 4.3. which would entail completing a new Enclosure 3. ;
Reportability Evaluation Worksheet. However, the inspector noted
that several recent PCs that were determined not reportable did
not have new Enclosure 3 forms completed. The NSM indicated his .
concurrence with the re)ortability recommendation by signing the !
memo received from the Reportability Review Owner. The inspector t
also observed this practice on a " Reportable" PC that was
subsequently determined to be "Not Reportable." When questioned
by the inspector, a NSM indicated that signing the memo was
allowed by the procedure and would eliminate confusion that could
result from having two Enclosure 3 forms in the PC file. The
inspector did not agree with this reasoning because the last
Enclosure 3 in the file would be the original "Potentially
Reportable" or "Re
determine the nonreportable
portable" decision
decision,from
andreviewing
an auditorthe would have to
details
of the Reportability Review Owner memo. Also, when the NSM signed
the memo this diluted the apparent independence of the NSM's
reportability review. Per CP-151, the memo was su) posed to
contain a recommendation, but the NSM was responsi)le for making i
an independent determination on reportability. His signature,
indicating concurrence with the memo. gave the appearance that he
was part of the recommendation process and not independent.
Lastly. CP-151 did not allow the signing of the memo as the NSM
indicated was allowed.. It required another Enc. 3 form to be
completed. The inspector discussed these observations with the
Licensing department, who was pursuing corrective action to ensure 6
the NSMs com) lied with the CP-151 requirements. The inspector
considered t11s problem to be administrative with no safety
signi ficance. However, it did indicate a potential disregard for
strict procedural compliance among members of shift management. -
The inspector reviewed the licensee's root cause and corrective actions
for PCs 97-0724 and 0841. which were written for the examples cited in
LA 97-094. The inspector verified that the correctwe actions for these
PCs were appropriate and completed. The licensee's package also
contained a common cause analysis under PC 97-2089 for reportability
errors. The licensee determined that the primary common root cause for
their numerous problems was the lack of ownership of the reportability
process. The inspector's review of their corrective actions revealed
that the Operations department has been clearly delineated as the
process owner, and that Licensing was established as a consistent source
of guidance. The inspector considered the licensee's conclusions and
actions appropriate and the common cause analysis thorough.
. The inspector verified the licensee's closure actions for VIO 97-08-01
that were done under PC 97-4918. This review entailed verifying CP-151
was corrected as discussed earlier. No discrepancies were rated with
the licensee's actions for this item. The inspector also reviewed the
L Enclosure 2
.
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.- - - _v-,. n ,
15 licensee's overall reportability closure package. OP 4. which was initiated in 1996 before the open items were identified. it adequately addressed the overall programmatic issues the licensee had with reportability. The overall licensee reportability process has significantly improved since the current plant shutdown was initiated in September 1996. A team was established by the licensee in Novenber 1996 to develop an integrated plant approach to reportability. Improvements in the corrective action process have ensured that it is the source of virtually all re)ortable items. have resulted in better information provided to the 1SM making the reportability determination. and have improved the timeliness of rompt concern identification and classification. The inspectors Fave also observed that in the spring of 1997, a dedicated point-of-contact person was designated in the Licensing organization. An individual with detailed knowledge of reportability requirements was r.ot previously available to the NSMs for reference and guidance. The result has been an improvement in reportability determination consistency and tracking. Although the process improvements have been significant. deficiencies with reportability still occur due to failures to follow the process. The inspector concluded that these were primarily administrative errors of minimal safety significance, but the licensee was still challenged to ensure their process was correctly implemented.
c. Conclt.sions
The inspector determined the licensee's actions were good improvements
to their reportability process and addressed the 3rogrammatic and
specific causes of the open items. Consequently ]oth of the violations
are closed. However, the inspector and licensee have identified several
examples of personnel errors that can potentially circumvent the
improvements in the process. The inspector concluded the licensee's
reportability program was adequate and acceptable for restart and has
been functioning well during this report period. However licensee
management attention in this area needs to continue to ensure procedure
compliance and that personr 4 errors are minimized.
The ins)ectors assessed the licensee's performance, relative to the
Reporta)ility Program, in the five areas of continuing NRC concern:
. Management Oversight - Adequate
. Engineering Effectiveness - N/A
. Knowledge of the Design Basis - N/A
. Compliance with Regulations - Adequate
. Operator Performance - Adequate
Enclosure 2
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II. Maintenance
M1 Conduct of Haintenance
M1.1 Incorrect Calibration Performance Causes Sorav Valve Doenina ,
a. Inspection Scooe (62707)
On November 5. 1997, a licensee instrument technician did not perform a
procedural step which caused an inadvertent opening of the pressurizer
' spray valve. The inspector reviewed the licensee's response to the
problem and their corrective action plan,
b. Observations and Findinas
Surveillance Procedure 112. Calibration of the Reactor Protection
System. Rev. 57, was being Jerformed on November 5. 1997 to calibrate
RCS pressure transmitters ()T). Steps 4.3.2.6 and 4.3.2.7 required the
instrument technician to have Operations bypass the Smart Analog Signal
Select system (SASS) channel being tested and select the channel not
under test for control. The stops were allowed to be marked not
applicable (N/A) only if certain instruments were being tested. The
technician was nut testing those annotated instruments but he
inappropriately marked these steps N/A. continued on with the procedure,
and did not consult with the control room operator. This resulted in an
inadvertent high pressure signal being sent to the 3ressurizer spray
valve, causing it to open. The impact of this on tie RCS was
inconsequential because the pressurizer spray block valve was closed so
the nitrogen over pressure blanket on the pressurizer was not affected.
The licensee initiated PC 97-7638 for corrective action and assigned it
a grade of "B" which requires a formal root cause be completed. The
licensee assigned the higher grade because they recognized the potential
implications of the procedurt error if plant conditions had been
,
different. The root cause was not completed at the and of the
inspection period, but the ins)ector discussed the preliminary cause and
planned corrective actions wit 1 the instrument technician shop
supervisor. The licensee determined the p.imary causes were personnel
performance and a lack of attention to the details of the procedure.
They were developing appropriate corrective actions. They also noted
.
some ambiguity in tfie wording of the skipped steps and wcre evcluating
I enhanced wording.
c. Conclusions
The inspector concluded the instrument technician displayed a poor level
of skepticism and self-checking when electing not to perform steps in a
l proceJure. This resulted in an inadvertent actuation of plant equipment I that fortuitously had minimal consequences. However, the licensee
recognized the implications of the problem and took appropriate
corrective actions.
Enclosure 2
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H2 Maintenance and Material Condition of Facilities and Equipment
M2.1 Maintenance Activities on_ Protected Enaineered Safeouards (ES) Trains
a. Insnection Stone (62707)
The inspectors investigated the licensee work controls for performing
work on the protected engineered safeguard trains. The circumstances
involved in the issuance of PC 97-7622. involving unauthorized work
performed on a protected ES train observed by a licensee auditor, were
reviewed by the inspectors.
b. Observations arid Findinas
On November 6. 1997, PC 97-7622 was issued to document that work
authorized on the protected A train of 480 V ES motor control center
(MCC) cubicles had been expanded to another cubicle without approval
from the Shift Supervisor on Duty (5500). Craft Jersonnel were
3erforming Work Request (WR) 348172 to implement iodification A) proval
lecord (MAR) 97-06-13-01. to replace a nine point terminal bloct with
two four point blocks. The personnel had notified the Engineering
department that there was insufficient space in the specified cubicle
for the two blocks. Engineering provided a field change notice to the
modification specifying that the terminal blocks be mounted in a
different, spare cubicle. The craft su)ervision opened the spare
cubicle and were sco)ing out the work w1en a technician expressed
concern that the worc was being aerformed on the protected train. At
the time of the original work, t1e A train was considered the protected
and operable ES equipment train. The SS00 had approved work in the
original cubicle but was not aware of the change notice, which expandem
the work outside the bounds of his original approval.
The inspector reviewed the licensee's administrative controls for work
control and modifications. Even though the licensee designated a train
of ES systems as protected, there was neither a procedural definition of
protected nor requirements for controlling activities on these systems.
This weakness created the potential for a threat to the systems that the
licensee was maintaining operable for Technical Specification
requirements.
The inspector reviewed licensee Procedure NEP-251. Preparation. Review,
and Approval of Field Change Notices (FCN). This procedure details the
necessary ap3rovals for development of a FCN. including requiring
Director of luclear Plant Operations approval for im)lementation. The
SS0D was not required to be notified prior ta a FCN 3eing issued to the
field for installation. This weakness allowed the SSOD to be bypassed
and unaware of changes to work scope that had been previously approved.
The licensee took immediate corrective actions t yt suspending the ongoing
work under the FAR. A night order was issued to the SS00 requiring
Enclosure 2
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_ _ _. __ . . _ . . - . - _ - . _ . ..
;
,
19
that, before a 5500 allows work in the protected train, an assessment of
the possible consequences of this work must be performed and approved by
the S500. The information for the evaluation must be sup,1
SSOD in writing, by the individual wanting to do the work.1ed This to the
information was to include. as a minimum, scope of work, reasons why
work must be done during the requested window contingency plans
regarding work, duration of work activity, scaffolding plans, plans for
any required barriers between trains, pre-job briefing requirements,
methods of communication, a single point of accountability at the job
site, and the project manager's name. The forms to perform this work
have been approved since the issuance of the night order and were ,
reviewed by the inspector. Some of the information was observed to be
ambiguous on the form, but the SS00 was aware of the conditions existing
and was able to answer all questions from the inspector.
c. Conclusions
The control of work in the protected trains of engineered safeguards
systems was weak, because it lacked acequate procedural guidance and
there was not a clear definition as to what constituted a protected
train. A weakness was also noted in changes to modification scope being
issued directly to the field, with review and approval of scope changes
not being communicated to the operations shift supervisor on duty.
H8 Hiscellaneous Maintenance Issues M8.1 (Closed) LER 97-002-01: Out of Calibration Fuel Pool Water Level
Transmitters
(Closed) VIO 97-01 04: Failt're to Perform Technical Specification
Surveillance for Soent Fuel Level
a. Inspection Scoce (92903)
This violation and LER involved the licensee's failure to calibrate
Spent Fuel Pool Level Transmitters within their required calibration
intervals. The inspector reviewed documenzation which was generated as a
result of the subject LER and violation, and interviewed licensee
personnel to assess the adequacy of the licensee's corrective actions,
b. Observations and Findinas
The inspector noted that the resolution of the violation and the LER was
being tracked under licensee restart Item H-9. The inspector verified
that the licensee had completed the corrective actions which were
identified as a result of the violation and the LER. The licensee
Enclosure 2
. _ - ._
20
identified the corrective actions in a letter dated June 16, 1997, in
the LER, and in restart issue package M 9. The corrective actions
included:
* Develop a root cause determination
. Develop formal expectation / duties for daily scheduled
surveillances
* Revise appropriate procedures
* Develop new procedures
* Perform an extent of conduction inspection
e Provide training to appropriate personnel
Following the review of documentation associated with the violation and
the LER and interviews of licensee personnel, the inspector concluded
that the licensee had completed the corrective actions identified. The
effectiveness of the corrective actions implemented by the licensee will
be evaluated during future routine inspections,
c. Conclusions
The inspector concluded that the licensee's corrective actions for the
violation and the LER were satisfactory. These items are closed.
The inspector assessed the licensee's performance, relative to
corrective actions for the violation and the LER in the five areas of
continuing NRC concern:
* Management Oversight - Good
. Engineering Effectiveness - Good
. Knowledge of the Design Basis - Good
e Compliance with Regulations - Good
. Operator Performance - Good
III. Enaineerina
El Conduct of Engineering El.1 Desian Control Process
a. Insoection Scope (37550)
The inspectors reviewed two 4RC rest:1rt items that were identified as
engineering programs. GL 96 01. Testing of Safety Related Logic Circuits
and DC Failure Modes and Effects of Loss of DC Power (FMEA) to verify
their completion. These two programs were inspected to verify they were
Enclosure 2
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,
i
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21
technically adequate and were implemented and completed in accordance i
with the licensee's commitments and NRC regulations.
b. Observations and Findinas ,
b.1.(flgicd) GL 96 01. Testino of Safety-Related loaic Circuits (MPA #L601)
'
1R 60-302/97 11 (May 5 9. 1997) referenced IR 50 302/97-07 and
identified that the licensee had completed all the requirements for GL
96 01 except for the following items: 1) reactor protection system ;
validation: 2) closure of eight open PCs: 3) contractor's final
submittal; and 4) final review. approval and closure by the licensee of
all GL 96-01 documents. The final documents for closure were identified
as R01 through R01G.
The inspectors verified that all open items and documents were ,
satisfactorily completed by the licensee. These items included the
revision of 11 surveillance procedures and implementing corrective
action for the eight open PCs. The licensee opened LER 97-003-001
through 005 regarding GL 96-01 testing deficiencies and closure. LER '
97-003-005 encompassed and su)erseded LER 96-025-00 and LER 96 011 00.
which also were concerned wit 1 logic testing deficiencies and GL 96-01.
The licensee sent a Conformation of Completion (TAC 94668) letter to the !
NRC for GL 96 01. Testing of Safety related Logic Circuits, dated l
'
September 5. 1997. The NRC replied by letter dated September 22, 1997
that the licensee had provided the required submittals and responses for ,
their commitments for GL 96 01 and therefore TAC 94668 was closed. The
inspectors concluded the licensee had implemented a superior program to
meet the requirements in GL 96-01.
b.2 (Closed 1 Failure Modes and Effects Analysis (FMEA) for loss of DC Power
FCR3 0.. .7)
LR 50-302/97-11 (May 5-9. 1997) referenced IR 50-302/97-07 and
identified that the licensee had completed all their commitments except
for the following items: 1) address and closecut the 12 open PCs: 2)
complete final review of contractors work: and 3. final review, '
approval, and closure of all FMEA documentation. The inspectors ,
verified that all open items and documents were satisfactorily addressed
and completed by the licensee.
The licensee Conformation of Completion letter dated September 24. 1997,
was in res)onse to the requirement in NRC's Confirmatory Action Letter
dated Marc 1 4. 1997. The licensee's letter stated that the FMEA program
was completed except 'or one recently identified item. A problem with a -
potential 3A battery failure was identified and was being tracked as a
new licensee restart item 0 07A. This potential 3A battery failure ,
condition was described in PC 97-4354 and LER 97-21 as " Loss of Class 1E
Battery A Wito LOOP /LOCA Will Result in failure of EDG-3A to load ES AC '
Buses." The licensee opening of this new restart item for the potential
failure of the 3A battery was being adequately addressed, and closure of
Enclosure 2
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, .-- .v ,, . . , - , , . , , , , ,. . ~ ~ , . . . . . , . , , , , - - . . , __,._..._.-,,.n..e,.. , ---
22
the FMEA program was appropriate. The inspectors concluded the licensee
had implemented a superior FMEA program to meet their requirements,
commitments and NRC regulations.
c. Conclusio.nl
The inspectors concluded that the licensee had implemented and completed
superior programs for GL 96 01 and DC FMEA. Both programs reviewed were
technically adequate and were implemented in accordance with licensee
requirements, commitments and NRC regulations.
The inspectors assessed the licensee's performance, relative to the
design control process, in the five areas of continuing NRC concern:
* Management Oversight - Superior
e Engineering Effectiveness - Superior
e Knowledge of the Design Basis - Good
e Compli6nce with Regulations - Good
. Operator Performance - N/A
El.2 fmeroency Diesel Generator Loadina Calculations
a. Insoection Stone (92903)
The inspector reviewed the licensee's emergency diesel generator loading
calculations. The requirement to have emergency diesel generators and
requirements for the performance of those generators are contained in
10 CFR 50. Appendix A. Criterion 17 - Electric Power Systems. Criterion
17 states that an onsite electric power system shall be provided to
permit functioning of systems important to safety. The safety function
of the onsite power system (emergency diesel generators) shall be to
provide sufficient capacity and capability to assure that the core is
cooled and containment integrity and other vital functions are
maintained in the event of postulated accidents. The onsite electric
power supalies shall have sufficient independence and redundancy to
perform t1eir safety functions assuming a single failure,
b. Observations and Findinas
The licensee's emergency diesel generator loading analysis was contained
in Calculation E-91-0026. EGDG-1A Scenario Based Loading. Voltage Dip.
Frequency Dip and Transient Motor Starting Analysis. Revision 3. dated
October 17. 1997. Calculation E-91-0027 covered the B train emergency
enerator. A major supporting calculation was Calculation M-96-
diesel
0069. E g$ Pump Maximum Flow for EDG Loading. Revision 3. dated
September 15. 1997. As indicated by the title, these calculations
determined loading for nine accidents together with the single failure
of either a diesel generator, the turbine driven emergency feedwater
pump or a battery (control power). The inspector fourd that the
calculations contained the input data to demonstrate that the
requirements stated in the scope section above were met. The inspector
Enclosure 2
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;
23
e
observed that suitable design control measures were employed in
development of the calculation.
In the short term. i.e.. less than one hour following an accident, the - !
'
worst case scenario was a steam line break inside containment with
failure of the B train diesel generator. The calculated load included i
'
all loads powered by the diesel generator in their respective maxinam
flow condition, and the total load was within the diesel generator 30- ;
minute and 200 hour ratings with about 6 percent margin. In the long !
term i.e.. more than one hour following an accident, the worst case was
a small break LOCA with failure of the A train diesel generator and with ,
recirculation in piggy-back mode. This scenario involved operator ,
action to remove and add loads. The inspector observed that the new
operating procedures contained simple, clear instructions to guide the
operator in removing and adding loads in the context of diesel generator
loading. The calculated worst case long term load was within the 200-
hour rating, ,
The ability of the diesel generator to accept the programmed load
sequence was demonstrated with a dynamic type computer program. The !
results of this )rogram showed that peak transient loading was kept
below 3910 kW. tie maximum diesel generator power output capability,
which would minimize frequency drop off associated with suddenly applied ,
relatively large loads. The results of the dyne,11c program showed that
*
the voltage and frecuency excursionE and recovery times remained within
the limits recommenced in Regulatory Guide 1.9. Selection. Design, and
Qualification of Diesel-Generator Units Used as Standby (Onsite)
Electric Power Systems at Nuclear Power Plants. Revision 3. dated 1993.
As an independent check of these computer results. the inspector
reviewed results from the integrated diesel generator sequencing test '
performed under Surveillance Procedure SP-417 on April 25, 1996. The
voltage and frequency traces recorded during this test were consistent
with the results of the E-91-0026 calculation.
The inspector observed that the licensee calculated two sets of load
values within the diesel generator loading calculation: a value based
entirely on manufacturer. originally supplied pump performance curves
(calculated value) and a value determined by recent onsite measurements
of electric power and system parameters made by the licensee for each of
the major loads. The inspector observed that for some loads the
measured value was higher than the calculated value, and for some loads
the measured value was less than the calculated value. The two values
for each load matched within an acceptable deviation for comparison of
test and calculated values. The licensee determined two total loads:
one using all calculated values and one using all measured values. The
,
higher of these two total loads (usually the calculated value) was taken
i as the calculation result and compared to the diesel generator ratings. l Having two values for each load raised the question of which was the l correct load to use in the calculation. The inspector calculated a '
total load using the higher of calculated and measured for each load,
l Enclosure 2 l ;
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r-- - -+,-y e- se- ,s-. --- --w ,- ,m, n
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24
and found that the total load calculated in this manner was about 50 kW
higher than the licensee's calculation result. However the inspector
observed that even with this additional 50 kW the 30 minute and 200 hour
ratings were not exceeded.
The diesel generators had been modified recently to upgrade the 200 hour
and 2000-hour ratings. The results of the calculations were com)ared to
the new ratings. The Updated Final Safety Analysis Report (UFSA1) would
have to be revised to reflect new ratings and hif,her load values. The
licensee provided copies of the revised UFSAR pajes which would be
submitted to the NRC.
c. Conclusions
The inspector concluded that the revised emergency diesel generator
loading calculations demonstrated that the generators have the capacity
and capability to accept the design basis loads as required by 10 CFR
50. Appendix A. Criterion 17.
The inspector assessed the licensee's performance relative to the
revision to the diesel genera +or loading calculations in the five areas
of continuing NRC concern.
. Management Oversight - Good
. Engineering Effectiveness - Good
. Knowledge of the Design Basis - Good
. Compliance with Regulations - Gcod
. Operator Performance - N/A
E1.3 Non-use of T-MARS on Secondary Systems (37551)
a. Inspection Scone
In July of 1997, during repairs to leaking pipes in the site drain (SD)
l
system, the inspector identified a concern with the potential for the
licensee to make modifications to a system and restore it to service
outside of the normal modificat4on approval process. Specifically,
repairs to one of the pipes were delayed and use of the pipe was
urgently needed to support a discharge of the Turbine Building sump
drain tank. SDT-1, which was full and causing the sump to fill to
capacity.
b. Observations and Findinos
Operations and maintenance shift management elected to place a temporary
patch on the pipe by initiating but not fully documenting a WR, and then
placed the SD system in service to discharge SDT-1. They based their
decision on a statement in CP-113A. Work Request Initiation and Work
Package Control, that allowed the use of a maintenance activity in an
emergency without having a documented WR. However, they were not
Enclosure 2
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25 .
i
specifically allowed to restore a system to service per this process. -
were supposed to document the change in the Temporary Alteration Log
which was not done, and were required to document the work afterwards
which should receive the same degree of review as a fully planned WR.
The inspector was concerned that this patch constituted a tem)orary
modification to the SD system which was not processed under t1e
Temporary Modification Approval Record (TMAR) process. Licensee
anagement had identified this problem and also had the same concern. :
They initiated an ap)ropriate investigation and corrective action. The
inspector reviewed t1e results of their root cause determination in PC i
97-5076. The licensee identified another patch on the same line that
had been installed in 1993 under similar circumstances. The PC i
discussed that it had been added by improperly altering the scope of a
WR and had been left in plice without any subsequent documentation. As
an immediate CorreClive action, the licensee performed an engineering
evaluation of both patches and a Commercial Grade Work Request was
develo>ed to document the adecuacy of both patches. They determined
that t1eir work process procecural guidance was weak in involving
Engineering in potential modifications to plant configuration. They
identified appropriate corrective actions to correct the guidance.
c. Conclus12D1
The inspector concluded that the licensee had performed a very good
investigation and root cause determination, although the documentation
was difficult to follow and the corrective actions didn't clearly match
the identified causes. The inspector did not identify any safety
concerns with the use of the temporary patches and determined the
original concerns were adequately resolved by the corrective actions of
the PC. The inspector concluded the licensee's threshold for
identifying temporary modifications was appropriate.
E1.4 Loose Part on A OTSG Unoer Tubesheet (3755D
in IR 50-302/97-11. the inspector discussed the licensee's discovary of
a loose part on the A OTSG tube sheet. The damage from this pari '
significantly expanded the scope of repair efforts for damaged tube ends
and delayed the licensee's schedule. The part was determined to be half
of a 3/4 inch hex nut and caused the licensee to have to repair over
10.500 of the 15,531 tube ends on the A OTSG in order to complete eddy
current inspections. The licensee initiated a root cause investigation
to determine the source of the part address deficiencies with the Loose
Parts Monitoring System (LPMS) that failed to identify the loose part in
the OTSG, and assess the impact of other potential loose parts in the
RCS. The licensee's effort for this investigation was disjointed
because it was tracked by several precursor cards. In September 1997,
the inspector reviewed the completion of PC 97-4269 which addressed the
LPMS problems. The inspector noted that the licensee determined the
LPMS detector for the A OTSG was inoperable following their startup from
the last refueling outage. even though it had been tested acceptable
Enclosure 2
1
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:
26 l
-just prior to startup. This contribut ed to the lack of knowledge of the ;
loose part until the OTSG was opened ior inspection. The licensee !
proposed appropriate corrective actions for the LPMS, but the PC left
*
numerous questions unanswered such as the source of the hex nut, the
potential for the other half of the hex nut to be.in the RCS, and the
potential for other known loose parts in the RCS to cause further damage -
to the repaired tube ends. Licensee Engineering management was unable
to resolve these questions when asked by the ins)ector, although it was !
later determined a separate investigation under )C 97-4440 would resolve
the various issues. The inspector deferred further review of this !
problem until that PC was completed. l
In November 1997 PC 97-4440 was completed. It encompassed PC 97-4269 <
and PC 97-5041 on the OTSG tube sheet damage, to be a final integrated j
assessment of the loose part issue. The licensee's final determination
concluded significant deficiencies in the foreign material exclusion ;
(FME) program over several years had contributed to'the loose part in
the RCS. The inspector verified a signif1 cant upgrade was in process
for the licensee's FME )rogram in response to this and other recently ;
identified problems. 11e licensee determined that the loose part most t
likely originated from a refueling bridge crane and adequately resolved i
'
the source and status of other loose parts. One small loose part was *
known to remain in the bottom of the reactor vessel. This was analyzed
as an acceptable condition in 1994 due to the small size of the part and
remote chance of it relocating. The licensee reviewed that -
' justification to ensure it remained valid and bounded the current
situation. The inspector did not identify any further concerns with the ,
current status of loose parts in the RCS. ,
i
t
c. Conclusions
- Although the final assessment was completed well after the part was
identified in June 1997, the inspector concluded the licensee's final ',
assessment was thorough and adequately resolved any concerns with the
potential for loose parts damage in the RCS. ,
,
E2 Engineering Support of Facilities and Equipment
L E2.1 Emeraency Diesel Generator 1B MAR Functional Test
a. Inspection Scoce (61726. 62707. 92903)
'
The inspectors observed activities associated with the MAR functional
test (ET) of the radiator replacement and power upgrade of emergency t
'
diesel generator (EGDG) IB. The modifications were performed under MAR
97-05 15-0F and 97-05-15-02, radiator re)lacement. MAR 97-05-15 05 for
L the radiator fan drive upgrade, MAR 96-1MS-01 for 7 3 generator 150 kw L >
; upgrade, and MAR 97-04-03-02 for the EDG bt ldin' M ilation system
i 1
L modification. The inspectors observed the perfoi p g of several of the '
.prejob briefings,-test runs of the EGDG-1B. and trouvieshooting
Enclosure 2
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1
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i
activities associated with the MAR FT. A review of the MAR FT procedure
and associated documents was conducted by the inspectors.-
r
b. Observations and Findinas j
The dates, times, and other information for each of the diesel starts
are described in the attachment at the end of the inspection report. l
Prior to the beginning of the MAR functional test, the inspectors
reviewed the test procedure. MAR 97-05-15-01 TP 2. Several areas of
concern were noted. Section 5.0. Limits and Precautions. Step 5.14
stated that generator stator tem)erature and not exceed ISO C. The-step l
'
states that if the generator hig1 temperature alarm actuated at 135 C.
the generator stator temperatures should be closely monitored to ensure !
150 C was not exceeded on the highest reading _ Resistor Temperature . ;
'
Detector-(RTD). The inspector reviewed licensee Procedure AR-902. DGB
Annunciator Response, which stated that for a valid alarm, the operator
shall check for adequate generator airflow, reduce load on the diesel ;
generator, and inform the electrical supervision of protective relay ,
actuation. Discussions with various operations personnel determined i
that if the alarm were received. the personnel would follow the more l '
restrictive annunciator response )rocedure requirements. The developers
of the MAR FT failed to realize tlat the operators would follow the
:
annunciator response procedure and make allowances or provide
justifications in the MAR FT for the operators deviating from the AR !
requirements, j
'
The inspector performed a field walkdown of the EDG. The meter used to
monitor generator stator temperature has a maximum reading of 140 ;
degrees C. The licensee procedure allowance of 150 degrees C could not
be monitored with installed instrumentation, and the procedure
developers failed to validate this change. The licensee changed the
limit to 135 degrees C. after notification by the inspector of the ;
discrepancy. ;
The inspector also identified that referenced temperature relay for !
'
generator stator temperature was a model IRT51A. The vendor information
)f the EDG manual states that this relay was
in thefor
rated controlled
80120 C.cop)he T setpoint on this relay was 135 C.PM-102.
- Calibration of Protective Electrical Relays, stated that the range for ;
this relay was 100-160 C. The licensee located a letter from the
vendor dated June 26. 1979, which stated that the original relay could i
not support raising the setpoint to 135 C. so an extended range model, j
'with revised operator manual. was being supplied. The licensee had not .
-updated the diesel vendor manual to reflect this change. l
'Further review of the MAR FT procedure development documentation. CP- i
134. Preparation and Aparoval of MAR Functional Test Procedures. l
Enclosure 4 disclosed tlat the test was to be performed in conjunction :
with Attachment A. SP-354B. Monthly Functional Test of the Emergency +
Enclosure 2
.
9-
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__. _. _ _ . _. _ _ _ _ _ . . _ _ _ .
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28
'
Diesel Generator EGDG 18. The safety assessment performed for the MAR
FT procedure stated that the operation of the EGDG for this test would ,
be the same as that performed for normal periodic surveillances and i
functional verification with the additional load performance
verification for the modifications. The inspector reviewed Attachment A
to the procedure and determined that it was not the approved revision of
SP 354B but was a proposed revision that had not been reviewed and
approved by the licensee. The safety analysis did not address the
differences in the acceptance criteria contained in the attachment from
the approved revision of SP-354B. The licensee's screening evaluation
concluded that an unreviewed safety cuestion determination and Plant
Review Committee review were not neeced. The inspectors reviewed the ,
safety evaluation and determined that the safety analysis for the EGDG- *
IB procedure was, with minor changes, identical to the safety analysis
for the previously performed EGDG-1A MAR FT procedure. However, the A
diesel procedure used the approved revision of SP-354A as Attachment A.
The B diesel procedure used an unapproved revision to SP-3548, but this
was not addressed in the safety analysis.
The Regulation. 10 CFR 50.59, state that a licensee may make changes in
procedures as described in the safety analysis report or may conduct
tests not described in the safety analysis report, without prior NRC
approval, unless the proposed change or test involves a change in the
technical specifications incorporated in the license or an unreviewed
safety question. The licensee must maintain records of changes to
procedures or of tests conducted, including a written safety evaluation
which provides the bases for the determination that the change, test. or '
experiment does not involve an unreviewed safety question. The
licensee's failure to recognize that the incorporation of new acceptance
criteria into the attachment to the MAR FT procedure in the safety
analysis resulted in a failure to assess that the changes did not
constitute an unreviewed safety question. This failure to conduct an
adequate unreviewed safety question determination was identified as a
violation. VIO 50-302/97-17-01. Failure to Conduct an Adequate .
Unreviewed Safety Question Evaluation for a Modification Functional
Test.
When the inspector identified that the licensee had incorporated an
unreviewed revision to SP-354B into the MAR FT procedure, several
licensee representatives were notified. However, the licensee failed to
take prompt corrective actions and the test was begun the following day,
without changes to the safety evaluation or procedure. At that time. '
the inspector notified Operations management of the situation, who
stop)ed the performance of the procedure until the appropriate changes ,
had Jeen made.
l The inspectors witnessed portions of a number of the diesel starts and ,
runs.- The initial start of EGDG-1B occurred on November 9, 1997. using
l Maintenance Procedure (MP)-499. Emergency Diesel Generator Engine !
Inspection / Maintenance. This was a slow start with an unloaded run to
l l Enclosure 2 l - . . . - . . .- .- --
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!
!
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29 ,
allow for air flow measurements and visual inspection for maintenance ,
check out. including an examination for coolant and oil leaks. A review :
of the Test Log for the MAR FT procedure revealed that at 1:15 am on ,
November 9,1997, Operations was nearly ready to start the 1B diesel for ,
the MP-499 run. The next log entry was at 3:55 am, which stated that l
while attempting to complete the MP-499 unloaded run, the diesel (EGDG- ;
~
18) developed a large jacket coolant leak and testing was stopped. The
inspector reviewed the reactor operator log and determined that the ,
diesel had been started and stopped three times during that time period: ,
the first three starts and stops due to fan drive clutch slippage and on
the fourth run the diesel was stopped for the jacket coolant leak. The
test log did not reflect start or stop times, or that the diesel was -
'
started.and stopped four times for the initial MP-499 runs. Overspeed
trip testing was performed during a test run on November 10. 1997 The
test log states at 4:30 cm that the engine was warm enough to start the
'
MP-499 run. At 6:30 am, the test log states that the unloaded run was completed satisfactorily. The Reactor Operator (RO) log states that the diesel was started at 5:35 am and was shutdown at 6:06 am on ovet speed trip testing. The test log did not state the start or stop times of the November 10, 1997 run. nor did it state that overspeed trip testing was comaleted as part of the run. Licensee procedure CP-134A, Performance of %R Functional Test Procedures, Section 4.1.5.1. Test Log, provided detailed instructions as to the type of entries to be entered into the test log. However, the detail provided did not include specific directions as to entering such data as diesel start and stop times during this MAR FT procedure. Even though not specifically called for, the procedure stated that the test log should contain, but was not limited to. the information contained in section 4.1.5.1. The test log for the EGOG-1B MAR FT exhibited weaknesses, resulting in a failure to record accurately actions taken during performance of the test. At 5:04 pm on November 10, 1997, the B diesel was started for a test run. At 5:43 pm. the diesel was shut down for high vibrations. A troubleshooting run was conducted November 12. 1997 in an attempt to determine the cause of the vibration. At the time. It was identified that the new clutch pads on the fan drive were not balanced. Engineering directed maintenance personnel to balance the pads without providing the technicians with written procedure or work instructions. The technician assumed that in order to accomplish the work, he should grind all of the pads to the size of the smallest. When this work had been accomplished, a maintenance supervisor discovered what was taking place. The work was stopped and installation of the pads was not allowed. The original clutch pads, removed prior to the modifications.
, were reinstalled. The licensee issued PC 97-7600 to document the event,
: and an investigation was initiated to determine the cause and take
additional corrective actions.
'
The licensee's investigation determined that the engineering personnel
inappropriately provided verbal instructions to the craft personnel.
The craft personnel did not follow the verbal instructions and did not
Enclosure 2
.
e sem. ---er--,r- ,e <-- g
__ __ _ __ _ - - _ . ._ _ _ ._ . __
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30
question the lack of written, approved work instructions. Corrective i
actions have been developed but not yet implemented to addresses these ,
areas. ,
10 CFR 50. Appendix B. Criterion V states that activities affecting j
cuality shall be prescribed by documented instructions, procedures, or ^
crawings, of a type appropriate to the circumstances, and shall be
accomplished in accordince with these instructions. procedures or
drawings. Instructions. procedures, or drawings shall include
appropriate quantitative or qualitative acceptance criteria for ;
determining that important activities have been satisfactorily :
accomplished. Contrary to these requirements, the Engineering personnel
involved in the diesel engine troubleshooting verbally directed the
mechanical maintenance personnel to perform maintenance on the clutch
pads and the work was accomplished without any written, reviewed, and '
approved instructions or procedures. However consistent with Saction
Vll.B.1 of the NRC Enforcement Policy, this licensee identified and
corrected violation treated as a Non Cited Violation. This issue was
identified as NCV 50 302/97-17-02. Maintenance Performed on Safety
Related Components Without Approved Procedures or Work Instructions.
The inspectors observed com)letion of a selection of the testing '
performed as part of the mal FT. The final loaded test run was
concluded at 6:40 am on November 24. 1997. Inspections were performed
on the diesel during the next several days and an unloaded run was
performed on November 27, 1997 as part of PM-123. Periodic Electrical
'
Checks of Emergency Diesel Generators. On November 2.1997 SP-354B
was performed and EGDG-1B was declared operable for mode 5 conditions.
c. Conclusions
The inspectors concluded that the development of the MAR functional test
pnadure was weak including inadequate safety analysis and the failure
to recognize conflicts between the existing annunciator response
3rocedure and the functional test procedure. The Operations staff.
10 wever, were consistent in stating that the annunciator response
procedure, being more conservative, would be followed, if the alarm was
received. The implementation of the test log was weak, omitting detail
necessary to reconstruct the sequence of events during the test.
l
The inspectors were concerned with the actions taken by the Engineering
and Maintenance personnel with the fan clutch pads. However.
Maintenance supervision appropriately identified the problem and
- ' prevented the installation of the clutch pads.
l The inspectors reviewed the completed t3sts and determined that the , acceptance criteria and commitments made to the NRC for EGDG-1B testing l
were successfully completed.
Enclosure 2
- . .-
_ _ - _ _ . __. _._ _ - . - . . _ _ _ _ _ _ - - - _ _ _ . _ _ - _ _ _ _ _ _ _ _ - _
,
f I 'h
.
131' > I
The inspector assessed the licensee's performance, relative to'this violation.- l
oin the five areas of continuing NRC concern:
l
. Management Oversight - Inadequate.
1
1
.
. Engineering Effectiveness - Inadequate'
. Knowledge of Design' Basis - Inadequate !
, Compliance with Regulation - Inadequate ;
. Operator Performance - N/A
.;
E8 Miscellaneous Engineering Issues !
E8.1 (Closed) VIO 50-302/96-08 01f Failure to Take Timely Corrective Action i
"o Address issues-and Actions For Makeuo System Audit Findinas and
.
i
ixcessive Vibration on a Soent Fuel Pool (SFP) Pumo Fan Motor (:PC -
Restart issue OP-24):
'
a. .Insoection'Scoce (92903) q
,
This item addressed two examples where the licensee'had failed to take
timely corrective action for self-initiated make-up system audit- <
'
findings and excessive vibration on a SFP pump fan motor. These
degraded equipment conditions were self-identified: however, the- '
licensee had not pursued prompt action to correct the deficient
conditions.
b. Observations and Findinas
The SFP pump fan (AHF 8A) was o)erated with higher than normal and
increasing vibration levels. Tie ' apparent cause of the vibration was a
lack of stiffness in the fan housing and a mismatch between the fan
drive belt and the drive sheave. A licensee make-up system audit had
identified discrepancies during piping walk downs including drawing
errors.
The licensee responded to and accepted the violation in a letter dated
October 14. 1996 (Letter 3F1096-07). The licensee concluded that the
following causal factors resulted in the violation: personnel error'on
the part of a system enginetr. a lack of management oversight and
. accountability, weaknesses in the deficiency reporting and tracking
systems, and a lack of engineering sensitivity.
Corrective actions included the following:
= *- Audited the deficiency tracking system and found no other issues
= Reviewed the: extent 'of the condition for vibration issues, and found
'two'other. examples.' This included the deca heat closed cycle cooling
g ns1(AHF 15A/B);and the non-safety re ated sump pumps (WDP-
Enclosure 2
.
t
[[-<. I e ,
32
* Revised the problem and deficiency reporting and tracking systems
* Conducted training for engineering personnel relative to self-
checking, timeliness, and sensitivity to degraded equipment
conditions
e Enhanced the root cause analysis program
. Corrected the vibration issues on the fans and sump pumps
* Corrected the piping audit issues
The inspector reviewed documentation including applicable precursor
cards, completed work orders, maintenance related information, revised
arocedure CP-111 Processing of Precursor Cards for Corrective Action
3rogram, and the violation response. The inspector discussed these
issues with licensee management and assessed recent performance in this
area. The inspector also walked down selected equipment in the field
and independently verified that the vibration issues were appropriately
addressed.
c. Conclusions
The inspector determined the licensee's actions were appropriate to
address the above degraded equipment conditions, including the extent of
the related conditions. Recent performance in these areas has been
noted to be acceptable, and the licensee has demonstrated improvements
in these areas. Based on the above reviews, inspections and
assessments, the violation was closed.
The inspector assessed the licensee's corrective action aerformance,
with respect to this restart-related issue, in the five iRC continuing
areas of concern:
* Management Oversight - Adecuate
. Engineering Effectiveness - Gooc
e Knowledge of the Design Basis - Adecuate
. Compliance with Regulations - Gooc
. Operator Performance - N/A
e Operator Performance - N/A
Enclosure 2
33
E8.2 (Closed) LER 96-011-00: Personnel' Error Causes Testina Deficiency
Resultina in Condition Prohibited by Imoroved TS
,
(Closed) LER 96-025-00: Personnel Error Causes Testina Deficiency
Resultina in Condition Prohibited by TS
(Closed) LER 97-003-00 throuah Rev. 05: Personnel Errors Caused Testina
Deficiencies (GL 96-01)
a. Insoection Scope (92903)
The three LERs identified testing deficiencies that did not meet the
requirements in the technical specification. The licensee committed to
1"1plement the requirements in GL 96-01. Testing of Safety-Related Logic
Circuits, as corrective action for the LERs. The inspectors reviewed
the GL- 96-01 program to verify corrective action was implemented.
b. Observations and Eindinas
The GL 96-01 program discussed in Section El.1 addressed the
deficiencies in the three LERs listed above. The inspectors verified
that the licensee had implemented and completed a superior 3rogram to
meet their commitments and the requirements in GL 96-01. T1e completion
of GL 96-01 by the licensee was satisfactory corrective action to close
these LERs.
c. Conclusions
The inspectors concluded the licensee's corrective action (GL 96-01) had
been implemented. LERs 96-011-00. 96-025-00, and 97-003-00 through 05
are closed.
The inspectors assessed the licensee's )erformance, relative to
corrective actions for these LERs. in t1e five areas of continuing
concern:
* Management Oversight - Superior
o Engineering Effectiveness - Superior
. Knowledge of Design Basis - Good
e Compliance with Regulations - Good
* Operator Performance - N/A
E8.3 (Closed) VIO 50-302/97-05-03: incorrect Information in Annunciator
Response Procedure for Inverters.
a. Insoection Scope (92903)
Annunciatrr response procedure AR-701. SSF P Annunciator Response, was
not changed to identify the location of the sensor input for " Battery
lapplyirg Load Alarm" after modification MAR 93-05-07-03 was
Enclosure 2
34
implemented. The licensee moved the location of the sensor from
monitoring the battery input current shunt to monitoring the inverter's
rectifier DC input voltage. The inspectors reviewed the licensee's
response to verify corrective action was implemented.
b. Observations and Findinas
The licensee corrected Procedure AR-70: n rovide the correct location
[
of the sensor. However, the alarm did n Anttion as xpected; it
alarmed during momentary voltage transie a. Th( licensee determined it
was necessary to relocate the sensor back to its original location. The
inspectors verified that )lant modification MAR 93-05-07-04 was being s
implemented to relocate t1e sensor back to monitoring the battery input
current shunt. and AR-701 was revised accordingly. This violation was
closed,
c. Conclusions
The inspectors concluded that satisfactory currective action was
implemented to close Vi0 50-302/97-05-03
The inspectors assessed the licensee's performance, relative to
1 corrective actions for this violation, in the five areas of continuing j NRC concern:
* Management Oversight - Good
* Engineering Effectiveness - Good
. Knowledge of Design Basis - Good
. Compliance with Regulations - Good
+ Operator Performance - N/A
E8.4 (Closed) VIO 50-302/97-07-01: Failure to Follow Procedure CP-111 for the
Processina of Precursor Cards (PC)
a. Insoection Scone (92903)
This violation involved three areas where tne licensee did not follow
Procedure CP-111: 1) the 20 day time limit for completion was exceeded:
2) non-qualified personnel were acting as the " Root Cause Team" leaders
and the " Apparent Cause Evaluators"; and 3) precursor cards were
4 improperly graded. The inspectors reviewed the licensee's corrective j action to verify :mplementation was completed,
b. Observations and Findings
gt n
The inspectors reviewed he licensee's response to the violation stated
in FPC's letter 3F0897-1. dated August 5. 1997. The corrective actions
taken by the licensee ant verified by the inspectors were: 1) CP-111 was
revised to clarify mar;dge.mnt expectations regarding PC timeliness: 2)
30 additional personnel we e trained during the week of July 28. 1997:
Enclosure 2
_
- - _ _ _ . .
..
35
and 3) a followup assessment. Quality Programs Surveillance (OPS)-97-
0129. dated September 22, 1997 was completed where all PCs were reviewed
by quality assurance for improper grading. This assessment identified
43% of the PCs were improperly graded.
The licensee opened restart item OP-2B. dated September 22, 1997, to
address the improperly graded PCs identified in OPS-97-0129.
c. Conclusions
The inspectors concluded satisfactory corrective actions were
implemented to close VIO 50-302/97-07-01.
The inspectors assessed the licensee's performance, relative to
corrective actions for this violation, in the five areas of continuing
NRC concern:
. Management Oversight - Good
. Engineering Effectiveness - Good
. Knowledge of Design Basis - Adequate
. Compliance with Regulations - Good
. Operator Performance - N/A
E8.5 (Closed) URI 50-302/96-201-07: EDG Not Protected Against Water Soray
from the Fire Protection System Sprinkler
a. Insoection Scone (92903)
This unresolved item states: Sheild the deluge valves be disabled by a
seismic event, water impingement could occur on both redundant diesel
generators which could cause serious damage to both diesel generators.
During this inspection, the inspector reviewed the design of the
sp'inkler system in the diesel generator rooms and the licensee's
a tions to resolve this issue. The inspector also determined whether
any violations of NRC requirements had occurred.
b. Observations and Findinas
The sprinkler system in the diesel generator area, which includes the
engine room and the control panel air compressor room, is a dry, air
pressurized system. A' control valve. FSV-101, located in the diesel
generator radiator room, receives electrical signals from heat detectors
te actuate the system. The sprinkler piping downstream of Valve FSV-101
runs into both A train and 8 train diesel generator areas. The
sprinkler piping was not originally designed with seismically qualified
supports. In order to get spray down of both diesel generators as a
result of a seismic event one wculd have to postulatc failure of valve
FSV-101 to maintain its pressure boundary and failure of certain branch
piping in both diesel generator rooms.
Enclosure 2
36
The inspector determined that there was no specific requirement for the
sprinkler piping to be seismically designed at Crystal River. The
lic msee presented evidence that the emergency diesel generators were
designed to operate during actuation of the sprinkler system. The
inspector observed that the diesel generators were constructed ir such a
way that water spray from above would be diverted away from openings to
internal parts. The inspector also observed that the control panels had
a drip hood installed similar to outdoor construction.
The licensee obtained documentation from the manufacturer that valve
FSV-101, a Multimatic Model A-4 by Grinnell Co.. was a seismically
qualifled style of valve. The documentation showed it was seismically
qualified by test to remain closed during a seismic event, and to open
upon receipt of a valid signal following the seismic event. The
licensee implemented modification MAR 97-06-11-01. Fire Service Pipe
Near FSV-101 Seisrr.ic Qualification, to harden the supports for the
sarinkler piping in the immediate vicinity of the valve. The purpose of
t11s modification was to prevent flooding of the diesel generator
radiator room as a result of a postulated seismic event leading to
failure of the sprinkler piping in that area. The inspector verified by
inspection of the piping that the new supports were installed,
c. Conclusions
The inspectors determined that the original design of the sprinkler
piping at the diesel generators met the requirements. The licensee took
actions to improve the design in terms of postulated seismic events.
Consequently this item is closed.
The inspector assessed the licensee's performance relative to this
unresolved item in the five areas of continuing NRC concern:
. Management Oversight - Good
. Engineering Effectiveness - Good
. Knowledge of the Design Basis - Good
. Compliance with Regulations - Good
. Operator Performance - N/A
E8.6 (Closed) EA 95-126. VIO I.C.2 (04013): Cotrective Actions for an
nadeouate Curve 8 (Two STI's and a Revised Curve 8A and 8B) were Also
: ncorrect
a. Inspection Scooe (92903)
This violation involved inadequate corrective action for an inadequate
curve for maximum makeup tank operating pressure versus level. The
inspectors followed up on the licensee's corrective actions for this
violation.
Enclosure 2
_ _.
37
b .- Observations and Findinos
The inspectors reviewed the licensee's response to this violation, -
interviewea licensee engineering and operations personnel, and reviewed
the following procedures.
,
-- OP-103B, Plant Operating Curves. Rev. 17 dated October 18, 1996:
including:
-
Curve 8A, Maximum MUT Operating Pressure vs. Level Wide Range
-
Curve 88, Maximum MUT Operating Pressure vs. Level. Operating
Range
- and Curve 8C, Maximum MUT Operating Pressure vs. Level. Preferred-
Range
'
- 01-6, Shift Orders. Rev. 3, dated November 14, 1996
- AI-400 C. New Prccedures and Procedure Change Process. Rev. 22.
dated July 22, 1997
- AI-400 F. New Procedures and Procedure Change Process for E0Ps,
APs, and Supporting documents. Rev. 4, dated June 19. 1997
-
CP-213, Preparation of a Safety Assessment and Unreviewed Safety
Question Determination (10 CFR 50.59 Safety Evaluation). Rev, 4
dated September 10. 1997
- N00-45. Management Self Assessments and Performance Monitoring.
Rev. 7. dated August 5, 1997
- AI-500. Conduct of 0)erations. Operations Department Organization
and Administration, Rev. 94, dated April 30. 1997
-
AI-1700. Conduct of Nuclear Engineering and Projects. Rev. 2.
dated August 21, 1997
The inspectors verified that the revised Curve 8 in OP-1038, for .naximum
makeup tank overpressure, provided additional margin and clearly
delineated acceptable and unacceptable operating regions. The
inspectors also verified that the 3rocess dealing with procedure
revisions was revised to require tlat engineering and other
interdisciplinary reviews be 3erformeo. Also, the inspectors noted that
the arocess for issuance of S1 ort Term Instructions (STIs) was revised
so t1at STIs would not be used in place of the normal procedure revision
process.
The ',nnectors verified that management oversight of the operations and
enginaring interface had been strengthened through enhanced processes
Enclosure 2
38
and additional personnel. Inspectors also verified that the licensee
had e.stablished a single point of contact, called an issue manager, for
import 6nt technical issues. Inspectors verified that a new operations
management position had been established to provide increased oversight
of plant o)erations and interface with operators. Also, ins)ectors
verified tlat a special Rapid Engineering Response team had )een
established. In addition, inspectors verified that expectations for
engineering and operations personnel had been strengthened and
communicated
c. Conclusions
The inspectors concluded that the licensee's corrective actions had been
implemented. included actions to prevent recurrence of the violation,
and were effective improvements. EA 95-126. VIO I.C.2 is closed.
The inspectors assessed the licensee's performance, relative to
corrective actions for this violation, in the five areas of continuing
NRC concern:
. Management Oversight - Good
. Engineering Effectiveness - Cood
. Knowledge of the Design Basis - Good
. Compliance witt Regulations - Good
. Operator Performance - Good
E8.7 (Closed, 2 96-365. VIO C (03013): Inadeauate Corrective Actions for 10
CFR 50.59 Evaluation Errors and for inadeouate Containment Penetration
Surveillancn
a. Inspection Scone (92903)
This violation included two examples where engineers were involved in
inadequate corrective actions related to emergency diesel generator
loading and one example where engineers were involved in inadequate
corrective actions related to containment surveillance procedures. The
inspectors reviewed the licensee's response to this violation and
followed up on the licensee's corrective actions for this violation.
b. Observations and Findinas
The inspectors verified that the licensee had implemented a revised
corrective action program and a revised 10 CFR 50.59 program. The NRC
had inspected those programs and documented the results in irs 50-
302/97-07. 97-08, 97-11 and 97-13. The NRC had also inspected the
Quality Programs monitoring of the new corrective action process and
documented the results of that inspection in IR 50-302/97-11. In
addition, the NRC had inspected the licensee's MCAP 11 programmatic
actions addressing engineering aerformance and documented the results of
that inspection in IR 97-13. T1e previous NRC inspections determined
Enclosure 2
39
that the revised 50.59 process was adequate. Also, the revised
corrective action process for Grade A and B Precursor Cards was
adequate. However, a determination on the effectiveness of the revised
corrective action process, for Grade C and D Precursor Cards, was not
made pending the results of further NRC followup of IFI 50-302/97-11-04.
Corrective Actions for Approximately 4000 Precursor Cards Not Tracked to
Completion,
c. Conclusions
The inspectors concluded that the licensee's corrective actions had been
implemented and included actions to prevent recurrence of the violation.
EA 95-365. VIO C is closed.
The inspector assessed the licensee's performance, relative to
corrective actions for this violation, in the five areas of continuing
NRC concern:
* Management Oversight - Good
. Engineering Effectiveness - adequate
. Knowledge of Design Basis - adequate
. Compliance with Regulation - Goood
. Operator Performance - N/A
E8.8 (Closed) EA 97-162 (01013): Inadeauate Safety Evaluations for V _j
Ooerator Actions for Desian Basis SBLOCA M1tiaation
a. Insoectiori ?one (92903)
This violation involved inadequate 10 CFR 50.59 safety evaluations for
procedure and FSAR changes that added four required operator actions,
and changed one, for mitigation of a design basis small break loss of
coolant accident (SBLOCA). The inspectors reviewed the licensee's
corrective actions for this violation.
b. Observations and Findings
The inspectors reviewed the licensee's response to this violation,
interviewed licensee engineering and operations personnel, and reviewed
the following procedures and training records:
-
CP-213. Preparation of a Safety Assessment and Unreviewed Safety
Question Determination (10 CFR 50.59 Safety Evaluation). Rev. 4
dated September 10. 1997
-
Nuclear 0]erations Engineering Standard AI-1700/0ES-03, 10 CFR
50.59 SA/ JSOD Expectations. Rev.1. dated May 9.1997
Enclosure 2
_ .- _
a
;
'
~40
-
Nuclear Operations Training Department Lesson Plan. Special
Technical Training 10 CFR 50.59 Safety Evaluation. SA/US0D
Lessons Learned Training. NUCST-0067LL. Two Hours. Rev. O. dated
July 23, 1997
The inspectors noted that the NRC had inspected the 10 CFR 50.59
program, including
'
the Safety Analysis Group and training and
documented results of that inspection in IR 97-08. During this
:s )ectio- ..le inspectors verified the licensee had again revised the
.) CFR SC. J procedure and had conducted training on the " lessons
. learned" from this violation. The inspectors also verified that the
licensee had submitted TS Change Request 2W to the NRC and that it
adequately addressed the unreviewed safety questions of this violation.
A scheduled NRC inspection of emergency operating procedures, prior to -
plant restart. will verify that the E0Ps include operator actions as
described by the licensee in TS Change Request 210 and that these
actions are effective in mitigating an SBLOCA.
During this review, the inspectors had comments on information recently
added to Procedure CP-213. Section 4.6.3.2 stated that only a Safety
' Assessment, and not an Unreviewed Safety Question Determination (US00).
is required for relocation of information from the FSAR into a .
referenced program (e.g., fire protection), because it already exists in
an NRC reviewed / approved licensing basis document (e.g., relocation of
existing information by reference to docketed material that has been
reviewed / approved by the NRC). The inspector noted that this could
improperly allow removal of significant information from the FSAR
without having an NRC Safety Evaluation Report (SER) s)ecifically
approving the removal and without performing a US00. Once the
information was removed from the FSAR. subsequent modifications or
3rocedure changes could potentially be rade without performing a US00.
)ecause they no longer represented changes to the )lant or procedures n
described in the SAR, CP-213 define 6 the SAR as t le latest issued FSAR.
Sending FSAR changes not yet incorporated into the controlled FSAR, and
4RC SERs.
Section 4.6.3.7 stated that only ' Safety Assessment, and not a US00, is
required for organizational char >s made as a specific NRC commitment
(e.g. , GL, BL NOV. or LER resp c. , . The inspector noted that NRC
review of an LER or NOV response does not constitute review and approval
of a change to the licensing basis of the plant. Therefore. these
documents were not an appropriate basis for bypassing the 10 CFR 50.59
requirement for performing a US00. The licensee planned to review the
inspectors * comments and revise CP-213 as needed.
c. Conclusions
The inspectors concluded that the licensee's corrective actions had been
implemented-and included actions to prevent recurrence of the violation.
EA 97-162 is closed.
Enclosure 2
41
The inspectors assessed the licensee's performance, relative to
corrective actions for this violation. in the five areas of continuing
NRC concern:
* Management Oversight - Adequate
* Engineering Effectiveness - Adequate
* Knowledge of the Design Basis - Adequate
* Compliance with Regulations - Adequate
* Operator Performance - Adequate
E8.9 (Ocen) EA 97-330 (01013): Unreviewed Safety Question Involvino Added EDG
Protective Trios
a. I soection Scooe (92903)
This violation involved an inadequate 10 CFR 50.59 safety evaluation for
a modification that had been made to the EDGs in 1987. The modification
had added five protective trips to the control circuit for each EDG that
were not bypassed during emergency operation and did not have two out of
three coincidence logic. The inspectors reviewed the licensee's
corrective actioas for this violation.
b. Observations and Findinos
The inspectors reviewed the licensee's response to this violation,
interviewed licensee engineering and operations personnel, and reviewed
the procedures and training records fcr improving the 10 CFR 50.59
process as listed above for violation EA 97-162.
In addition, the inspectors reviewed the modification Jackage for MAR
97-08-61-01. EDG Protective Trips, and compared it witi License
Amendment Request 219: Emergency Diesel Generator 'rotective Relays
Unreviewed Safety Question. dated September 12, 1997. The modification
was to reroute the added EDG protective trips so that they would trip
the EDG output breaser instead of tripping the EDG. The ins]ectors
noted that the modification was appropriately described in t1e license
amendment request.
The inspectors noted that the 10 CFR 50.59 safety evaluation for MAR
97-08-01-01 was for installing the modification while the affected EDG
was out of service, and did not provide for returning the EDG to
service. It stated that another 10 CFR 50.59 evaluation would have to
be performed before returning the EDG to service. At the end of this
inspection, the licensee had not received a] proval of the license
amendment request from the NRC. However, tie MAR was installed on the B
EDG, which was scheduled to be returned to service in four days. In
response to inspector questions, the licensee stated that they if they
did not receive an NRC approval in time, they would complete the
appropriate Justification for Continued Operation (JCO) evaluations to
return the EDG to service when it was ready.
Enclosure 2
. 42
c. Conclusions
This violation remains open for further review of the licensee's-
resolution of the unreviewed safety question and installation of MAR 97-- ,
08-01-01.
E8.10 (Closed) LER 96-24-01: Plant Modification Causes Unanalyzed Condition
Reaardina Emeraency Feedwater
a. Insoection Scooe (92903)
This LER involved a modification that removed the automatic start signal
from the A side of the Emergency Feedwater Initiation and Control (EFIC)
system to the turbine-driven emergency feedwater pump. The modification
failed to recognize that a previous modification, which had installed an
automatic trip of the turbine-driven emergency feedwater pump at 500
psig RCS pressure (decreasing), had relied upon the automatic start of
the emergency feedwater pump from the A side of the EFIC. The inspectors
followed up on the licensee's corrective actions for this violation.
b. Observations and Findinos
The inspectors reviewej the licensee's stated corrective actions in the
LER and reviewed the following procedures and records: ,
-
Nuclear Operations Engineering Standard OES-4. System Assignment
Expectations. Rev. O. dated January 22. 1997
-
Nuclear Operations Engineering Standard OES-1. Design Review Board
Expectations. Policies, and Practices. Rev. O. dated January 22,
1997
-
Operations Instruction 01-41. System Operator Program. Rev. 2.
dated May 31. 1996
-
Nuclear Plant Technical Support Manual. Rev. 9. dated December
1995
-
NEP-104. Interface Design Control. Rev. 7. dated March 31. 1997
-
NEP-210. Modification Approval Records. Rev.16. dated March 31,
1997
-- NEP-211. Commercial Grade Design Control. Rev. 17. dated March 31.
.
.1997
Nuclear Operations Department Manual N00-38: Planning. Budgeting,
and Scheduling Project Controls. Rev. 4
-
Training Lesson Plan NUCST-2011. Solution Sets Bases for TS
Enclosure 2
-- ,
_ _ - _ - - ___
i l
43
Changes (Including SBLOCA Analysis) 6 hours
- Extent of Condition Review for Omission of Information from Design
Basis Document Temporary Changes, dated March 14, 1997
The inspectors verified that nuclear engineering management had provided
interim direction stating that all design modifications will receive a
10 CFR 50.59 evaluation, and that the 10 CFR 50.59 evaluations will be
" stand-alone" documents. Also. the inspectors verified that a separate
group had been established for reviewing 10 CFR 50.59 evalu6tions.
Inspectors reviewed the extent of condition review for the omitted
temporary change to the EFW/EFIC design basis document. ond noted that
the licensee found that one additional similar error had been made
during the period of 1988 through 1993. Inspectors verified the
issuance of expectations for and assignment of system ownership teams,
expectations for project teams, and expectations for review of
modification pachges. In addition, inspectors verified the
establishment of a Design Review Board. Inspectors verified that
training had been provided to operators regarding SBLOCA design basis
accident analysis and that engineers were scheduled to receive the
training.
c. Conclusions
The inspectors concluded that the licensee's corrective actions had been
almost all implemented and included actions to prevent recurrence of the
event. LER 96-24-01 is closed.
The inspectors assessed the licensee's perf, mance, relative to
corrective actions for this violation, in t ie five areas of continuing
NRC concern:
* Management Oversight - Adequate
. Engineering Effectiveness - Adequate
* Knowledge of the Design Basis - Ade tate
. Compliance with Regulations - Adequate
. Operator Performance - Good
E8.11 (Closed) EA 96-365. EA 96-465. EA 96-527. VIO B (Examole 1) (02013):
Failure to Uodate Anolicable Desian Documents to Incorporate Desian
Information
a. Insoection Scoce (37550. 92903)
This violation involved the licensee's failure to update the Final
Safety Analysis Report (FSAR), the Enhanced Design Basis Document
(EDBD). and the Improved Technical Specifications (ITS) Bases with
regard to operation of the steam turoine 'iriven auxiliary feedwater pump
(EFP-2) for certain accident scenarios. The inspector followed up on
the licensee's corrective actions by reviewing procedure changes.
Enclosure 2
_- _____
44 training records internal licensee correspondence, and interviewing engineering personnel,
b. Observations and Findinos
The inspector noted that resolution of this violation was being tracked under licensee restart item D-39. Related corrective actions im)lemented to address this issue were being tracked under MCAP II and otler licensee restart items, which included D-40 and OP-6. The inspector reviewed some of the actions addressed under restart item OP-6 and documented the results in IR 50-302/97-07. The corrective actions addressed by OP-6 included revisions to various Nuclear Engineering Procedures (NEPs) to enhance the design control process. Changes to the NEPs included, but were not limited to, the incorporation of Procedure CP-213 requirements, additional guidance regarding design inputs, and guidance regarding prompt revision to design basis documents following implementation of a plant modification. During this current inspection, the ins)ector verified that the licensee had initiated documentation to update t1e FSAR. EDBD and ITS Bases to reflect operation of UP-2 for certain accident scenarios. Licensee interoffice corresponder.ce (IOC) 97-0165, dated February-25. 1997, transmitted the FSAR chansa package from Nuclear Operations and Engineering (NOE) to Nuclear Licensing. Temporary changes 552, 553, and 554 had been incorporated into the EDBD. The ITS and ITS Bases changes were addressed in TS Change No. 210 dated June 14. 1997. The inspector also reviewed the actions taken by the licensee to strengthen the expectations regarding the use of the FSAR. EDBD and the ITS Bases. These actions included implementation of procedure N00-55. Control of Design Basis Information, and development of the training course ST- 1222. Plant Design Basis and Configuration Management. This training course was being 3rovided to NOE. operations. Nucleai degulatory Assurance, and otler selected personnel involved in preparing modifications or performing safety assessments or US00s.
c. Conclusions
The inspector concluded that the licensee's corrective actions for this
violation were satisfactory. This item is closed.
The inspector assessed the licensee's parformance, relative to
corrective actions for this violation. the five areas of continuing
NRC concern:
. Management Oversight - Good
. Engineering Ef fectiveness - Good
. Knowledge of the Design Basis - Good
. Compliance with Regulations - Good
. Operator Performance - N/A
Enclosure 2
~
45
E8.12 (Closed)'EA 96 365. EA 96-465. EA 96-527. VIO B (Examole 2) (02013):
f ailure to Include AoDlicable Deslan Information in the Desian Inout
Reau1rements for a Modification.
a. Insoection Scone (37550. 92903)
This violation involved a failure to include desi;n basis information in
the design input requirements for the MAR which disabled the automatic
opening of valve ASV-204. The inspector followed up on the licensee's
corrective actions by reviewing procedure changes, training records,
internal licensee correspondence, and interviewing er.gineering
personnel.
b. Observations and Findinas
The inspector noted that resolution of this VIO example was being
tracked under licensee restart item D-40. Related corrective actions
implemented to address this issue were being tracked under the MCAP II
and other licensee restart items which included D-39 and OP-6. The
inspector reviewed some of the actions addressed under restart item OP-6
dnd documented the results in NRC IR 50-302/97-07. The corrective
actions addressed by OP-6 included revisions to various NEPs to enhance
the design control process.
During this current inspection, the inspector verified that the licensee
had incorporated design basis information into MAR 96-11-01-01. This
MAR was prepared to reinstall the automatic open signal to valve ASV-
- 204. The field work for this MAR was completed on July 10, 1997. All
of the testing associated with this MAR had not been completed at the
conclusion of this current inspection. The inspector also riated that
the licensee had initiated documentation to update the FSAR. EDBD. and
ITS Bases to reflect this MAR. IOC 97-0165. dated February 25, 1997,
transmitted the FSAR change package from NOE to Nuclear Licensing.
Temporary changes 552, 553. and 554 had been incorporated into the EDBD.
The ITS and ITS Bases changes were addressed in TS Change No. 210. dated
June 14, 1997. The inspector also reviewed the actions taken by the
licensee to strengthen the expectations regarding the use of the FSAR.
EDBD. and the ITS Bases. These actions included implementation of
procedure N00-55. Control of Design Basis Information: training course
ST-1222. Plant Design Basis and Configuration Management. This training
course was being 3rovided to NOE. operations. Nuclear Regulatory
Assurance, and otler selected personnel involved in preparing MARS or
performing safety assessments and/or USQDs.
Enclosure 2
__ _. .
46
c. Conclusions
The inspector concluded that the licensee's corrective actions for this
violation were satisfactory. This item is closed.
The inspector assessed the licensee's performance, relative to the
corrective actions for this violation, in the five areas of continuing
PRC concern:
. Management Oversight - Good
. Engineering Effectiveness - Good
. Knowledge of the Design Basis - Good
. Compliance with Regulations - Good
. Operator Performance - N/A
E8.13 Followuo on Restart issue Resolution - BWST NPSH Concern (FPC Restart
Issue D-18)
a. Insoection Scone (92903)
The inspector followed up on the licensee's actions to resolve a concern
regarding net positive suction head (NPSH) for the emergency core
cooling system (ECCS) pumps when the spent fuel pumps (SFP) were running
in recirculation to the borated water storage tank (BWST).
b. Observations and Findinas
The licensee had documented this concern in problem report (PR) 96-0360
and PC 97-0085. The licensee determined that this problem would be
resolved prior to restart from the current shutdown. The resolution of
this problem was being tracked as licensee Restart Issue D-18. The
inspector noted that the licensee had completed the corrective actions
to address this concern. These corrective actions included. but were
not limited to using SFP-2 instead of SFP-1B as the preferred method
for BWST recirculation: revisions to numerous calculations for the ECCS
to demonstrate that the flow rate for the SFP-2 would have a negligible
impact on the operability of the associated ECCS pumps: determination of
the flow rate to be used to revise the calculations: and revisions to
various procedures and design basis documents. etc. The inspector noted
that the following calculations and procedures had been revised:
M94-0013 Building Spray System Hydraulic Spray, Rev. 4
M93-0047 CR3 Makeup System Hydraulic Analysis. Rev. 3
M94-0047 CR3 Decay Heat Removal System Hydraulic Studies. Rev. 2
M95-0004 Makeup Pump NPSH Evaluation During Post LOCA Cooling and S/0
to the RB Sump. Rev. 2
Enclosure 2
47
M95-0016 BWST Swapover and Minimum Allowable level Evaluation. Rev. 2
M96 0010 Head Loss in BWST to Makeup Pump. Rev. 1
M95-0005 Minimum BWST Level Necessary to Prevent Vortexing During
Drawdown Rev. 4
M94-0053 Allowable MUT-1 Indicated Overpressu:e vs. Indicated Level.
Rev. 5
M97-0043 Head at Tie-In of Makeup TsiiK Surge Line to Makeup Pump
Suction Line. Rev. 0 (Note: This calculation stated that it
was performed to determine past operability and should be
used for historical information only)
The inspector discussed this issue with licensed operators in the main
control room and determined that they were aware of this issue and the
changes to the applicable procedures. In addition to the above
calculations, the inspector verified that the follcwing procedures had
been revised to address operation of SFP-2 for recirculation of the BWST
OP-406 Spent Fuel Cooling System. Rev. 57
SP-320 Availability of Boron Injection Sources and Pumps. Rev. 67
c. Conclusions
The inspector concluded that the licensee had taken satisfactory
corrective actions to resolve this issue 3rior to restart.
Consequently, this item is closed on the VRC Restart List.
The inspectors assessed the licensee's performance, relative to the
corrective actions to resolve this issue in the five areas of
continuing NRC concern:
. Management Oversight - Good
. Engineering Effectiveness - Good
. Knowledge of the Design Basis - Good
. Compliance with Regulations - Good
. Operator Performance - Good
E8.14 (Closed) LER 97-017-00: Personnel Error Caused Inadeouate Electrical
Seoaration Of Hiah Pressure Flow Indicators (FPC Restart Issue D53A).
a. Insoection Scooe (92903)
This item addressed multiple examples of the incorrect electrical
insulation material used during installation of modifications in the
main control boards.
Enclosure 2
. .
48
~ The inspector reviewed the licensee's root cau5e and corrective actions.
The inspection included reviews of licensee procedures relating to the *
use of electrical insulation material, and interviews with licensee
personnel-. The insSector independcolly verifled a sampling of the
-corrective actions )y performing a main control board walk down, and
inspection of the rework associated with this LER.
b. Observations and Findinas
During installation of a modification in the main control boards, the
licensee noticed that the electrical insulation material on the cables
of a previously installed modification differed from the material which
they were presently using. The licensee subsequently identified that
the wrong sleeving insulation material had been used on-three
modi fications. The incorrect insulation material which had been used
was Nextel. The qualified material for the application was Siltemp.
The inspector noted that the appearance of the Nextel and the Siltema
looked similar in color and were both braided materials. However, t1e
Nextel, which was 1/8-inch in diameter, was smaller in diameter than the
Siltemp. The Nextel was to be used as a tie cord material and not for
sleeving electrical' cables / conductors. Licensee corrective actions
completed included the following: ,
-
performed an extent of condition review to identify other areas
where Nextel may have been used:
- reworked and re) laced the Nextel with Siltemp insulation on areas
identified whic1 had incorrect use of Nextel:
-
clarified the procedure describing the use of Nextel and Siltemp,
and
-
developed a procedure for Siltemp sleeving to include an in-
process verification of the installed Siltemp.
The inspector found that an extent of condition review and a walk down
had been completed by the licensee for the purpose of identifying any
other misuse of the Nextel. The inspector noted that the walk down did
not include all pieces of equipment where the electrical separation was
required. However, the walk down did include the areas where most of
the design and modification-activity had taken place and where it was
most probable that the Siltemp material would have been used. The
licensee identified that Nextel had incorrectly been used on three
modifications: 1) MAR 96-02-09-01. High pressure flow indicators: 2) MAR
96-03-12-01. Emergency diesel generators Kilowatt indicators: and 3) MAR
-
91-08-26-04. Relay wirir6 separating non safety related signals from
- safety related signals.
Enclosure 2
. - -
49
The inspector reuewed the rework packages for the three modifications
and noted that there was in-process verification to identify that
Siltem) material had been used. In addition. Procedure MP-405A, used to
descri)e application of Nextel and Siltemp. had been clarified and gave
explicit instructions describing the correct use of Nextel. The
inspector reviewed the reworked cabling in the main control boards and
noted that the Nextel had been removed and replaced, as required in the
rework packages. In addition, the inspector walked down the rest of the
main control boards and did not identify any other misuse of Nextel.
The licensee had communicated to the technicians the correct use of
Nextel through de)artment meetings and via a memo. The inspector
interviewed a tecinician, who had done modification work at the main
control boards, and found that the technician was well versed with the
Nextel issue and knew the correct insulation material to use. Although
this item is a noncompliance with regulatory requirements, for the
reasons discussed in Inspection Report 97-21. the licensee meets the
criteria for enforcement discretion per Section VII.B.2 of the NRC
Enforcement Policy as described in NUREG-1600. Consequently this item
is closed and is identified as another example of Non-cited Violation
NCV 50-302/97-21-01. Examples of Noncompliances in Design Control 10
CFR 50.59 Evaluations. Procedure Adequacy /Adnerence. Reportability, and
Corrective Actions That Are Subject to Enforcement Discretion.
c. Conclusions
The inspector determined that the licensee had a3propriately addressed
the issues with the incorrect use of Nextel and lad completed the
required corrective actions. Based on the inspection finding and
observations this LER is closed.
The inspector assessed the licensee's corrective action ]erformance,
with respect to this restart-related issue. in the five 4RC continuing
areas of concern:
. Management Oversight Good
. Engineering Effectiveness - Adequate
e Knowledge of the Design Basis - N/A
e Compliance with Regulations - Adequate
. Operator Performance - N/A
E8.15 (Closed) VIO 50-302/EA 95-126 NOV ll.B: Failure to Take Adecuate
Corrective Action for Recuired Tank Volumes. Level and Suction Points
(FPC Restart Issue OP-12)
a. Insoection Scone (92903)
This item addressed multiple examples identified by the licensee and the
NRC in which safety related and technical specification related tank
levels, suction position, volume requirements, and their respective
relationships, were incorrectly being used by the licensee.
Enclosure 2
. - . . -_ . . - . _. .- .- - . . . - . . _
-.
,
"
4
50 j ,
TheIlicensee'_s response to the violation was-reviewed. The root'cause-
E
}
- andicorrective actions were reviewed andLverified by the~ following: W ,
'
_ procedures. FSAR, and Technical Specifications reviews: 2) walk-in and-
= scheduled interviews with engineering management, supervisors, staff.-
and other licensee-personnel:-3) review of selected tank level, suction
point and volume calculations; and 4) field ins)ections of tank'
- construction. geometries and specifications, tanc gauge locations and +
. measurement methods, and tank suction points. _;
b. Observations and Findinas :
By letter-(3F0996-01) dated September 9, 1996, the licensee replied and l
agreed with the violation. The licensee stated that the reason for the -
violation was inadequate prioritization and management of. existing and i
changing work loads associated with a rapidly changing environment. !
Examples included: a change in interna'l philosophy
in-house, reduction of permanent staff, and the nges cha.to perforrt
associated more work
with-
relocating the engineering staff to the plant site.
Licensee correctTve' actions completed included the following:
,
e an organizational structure with the required resources con istent
with the work load- ,
.
e work loads and prioritiu consistent with safety significance and
available resources
e increased design margins thereby decreasing the amount of direct
_ engineering effort. and
e comprehensive review of tank calculations associated with level.
' instrument errors, suction points, and usable volume
The inspector reviewed and verified specific examales of each corrective-
action. Relocation of the engineering staff to tie site had-been
completed, and the design engineering staff had been increased. In
'
addition. the use of_ specialty contractors haa increased commensurate
^ with emergent work loads-and required specializations. To address short
~ term issues. there had been a rapid response team (RRT) created. The '
team's charter included: 1) rapid engineering _ support to other plant
- organizations. 2) support to reduce the backlog of request for
engineering assistance (REA), and 3) design support on minor
modifications, commercial grade work, and plant equivalency replacement
evaluations. The inspector reviewed the experience of the team's
members. There was a total of eight engineers which made up the- team.
With the exception of two design engineers, the rest were degreed
,
- engineers.-and included two registered professional engineers. There
was good variation amongst the disciplines, i.e., mechanical,
electrical, and nuclear. The years of-experience ranged from 10 to 20
'
. Enclosure 2
.
e
-
,-
rr r y- ~ r4=-' y- f ' - ** e- tw*2-y
.
51
years and one individual had 36 years of experience. . The experience
was very diversified including: operations. design, testing, systems, ,
maintenance, licensing, and plant construction -However, the inspector
found that the RRT was not presently fully activated. Team members were
on loan to special start-up projects. Engineering management indicated
that the team would return to its normal charter after startup.
Efforts to increase design margins were evident. Examples included:
modification of the reactor building sump screen, and the emergency
diesel generator upgrade. However, the inspector found that these items
were being tracked separately by the NRC and the licensee.
The communications amongst the engineering groups appeared to be good.
The information provided by the various engineering groups was
consistent. Engineering management had a daily morning meeting and a
weekly meeting. The agenda of the daily morning meetings included:
prioritization of activities, review of modifications matrix, status
review of set point calculations for E0Ps, and review of emergent work.
The ins 3ector reviewed the REA backlogs. The inspector found that the
REA baccloas have been continuously decreasing. The M cklog was
reviewed on a weekly basis at the weekly engineering managers meeting.
Engineering management appeared to have a good handle on REA activity.
Backlog indicators were printed weekly, and tracked on a daily basis.
There was a strong focus on planning, accountability, and the continuous
review of the required resources. including specialist contractors to
achieve the REA workload - Engineering was strongly committed to
achieving a goal of less than a 200 REA backlog prior to restarting.
Additionally, engineering indicated the organization goals would be to
maintain a maximum REA backlog of 200. The inspector questioned the use
of tha engineering contractors after the restart workload had been
completed. Engineering management indicated that after the startup they
planned on decreasing the contractor resources but would increase the
permanent staff.
The inspector reviewed the licensee's activities related to the tank
calculations. The licensee had embarked on a comprehensive program to
review the calculations relating to tank requirements and had compiled a
list of tanks in order of safety significance. The calculations
included analysis for usable tank volume and the corresponding required
tank levels, taking into account inaccuracies associated with instrument
errors. Analysis on net positive suction pressure and vortexing was
included where appropriate. Data from vendor tank prints and technical
information. information notices, and industry reports was used, as well
'
as licensee experience with the tanks. The inspector found that
portions of the analysis were performed by contractors. However, the
project was owned and driven by the licensee's engineering group. This
provided good oversight on technical specifics (design specifications)
relating to the plant's applications, i.e.. tank tilt, materials, and
modification history. Appropriate independent verification was noted on
Enclosure 2
. . .
52
'
the tank calculations. In addition, management reviews and approvals
were required prior to accepting results.
The inspector reviewed the design basis. FSAR. and Technical
Specifications for selected tanks, and verified the licensee had
incli.ded the various requirements in the analyses. In addition, the
inspector reviewed assumptions made on calculations and enalysis' methods
and found them to be consistent and appropriate. The inspector had
various discussions with engineering relating to specifics on tank
calculations =. assumptions, equations, and analysis methods and found
toat the licensee was very well versed with the tank requirements and
overall analysis. Field inspections and control room board walk down
were made on the borated water storage tank, condensate storage tank,
emergency feed water tank, and the emergency diesel generator day-tanks
to verify data used in the calculations and that surveillance procedures
appropriately addressed Technical Specification requirements. Gauge
pressure tap locations, suction heights, piping size, and flow path
losses used in the calculations were verified. Accessability of gauges-
used in taking measurements for surveillance was reviewed and found to
be appropriate. Surveillance procedures had been approariately updated
and reflected the required heights-on the tanks and tooc into
consideration the volume analysis results and instrument errors.
However, the inspector found that the height requirements on the
surveillance procedures for the emergency diesel generator day tank were
too conservative. For example. Technical S
level to be >/- to 245 gallons of fuel oil.pecification 3.8.1.4
Based on the required
licensee's
tank calculation result (taking into consideration the instrument error)
this would equate to a height >/- to 22.2 inches. The surveillance
procedure. SP-345A, however, had an administrative limit of >/= 24
inches. and that height equated to 348.7 gallons. The inspector later
found that these differences were associated with the modification to
the emergency diesel generator and that this was being tracked under a
separate issue number.
c. Conclusions
The licensee's actions to correct numerous tank parameters had
appropriately addressed the difficulties associated with the engineering
structure, work prioritization, and available resources to perform the
work. However, continued monitoring after startup is warranted. The
tank calculations were thorough and the results had been appropriately
incorporated into the required procedures. Based on these observations
and findings, this violation is closed.
i L l
Enclosure 2
l l
_ _ _
__
53
The inspector assessed the licensee's corrective action performance,
with respect to this restart-related issue, in the five NRC continuing
areas of concern:
* Management Oversight - Good
. Engineering Effectiveness - Good
. Knowledge of the Design Basis - Adequate
. Compliance with Regulations - Adequate
. Operator Per formance - N/A
E8.16 (Onen) URI 50-302/95-02-02: Control Room Habitability Enveloce Leakaae
a. Insoection Scooe (92903)
On November 10, 1997, the licensee submitted a letter to the NRC titled:
" Control Room Habitability. NUREG-0737. Item III.D.3.4." The purpose of
the letter was to provide a description of the recent Control Complex
Habitability Envelope (CCHE) in-leakage testing and the revisions to the
calculational methodology used by the licensee for determining control
room operator dose. The licensee stated that they planned to use the
results of the testing and calculations to demonstrate operability of
the CCHE and Control Room Emergency Ventilation System (CREVS) prior to
plant restart from the current outage. The inspectors reviewed the
technical content of the letter, reviewed the test conditions and data,
inspected CCHE conditions in the )lant performed independe analyses,
and discussed CCHE in-leakage witi licensee engineering and licensing
personnel,
b. Observations and Findinas
In the November 10 letter, the licensee stated that since the CCHE is
maintained at a neutral pressure by the CREVS. the only mechanism for
developing an in-leakage under maximum hy)othetical accident conditions
is the effect of the outside wind on the Juilding. The licencee also
stated that they had examined every wall of the CCHE, assessed the
condition of each penetration. and sealed penetrations to eliminate
potential in-leakage sites that may have existed previously. In
addition, the licensee stated that testing has demonstrated that in-
leakage is extremely low for a structure the size of the CR-3 control
complex.
The licensee *s CCHE test concluded that there was in-leakage of 462
cubic feet per minute (CFM) with the following test conditions: CREVS
in emergency recirculation, a combination of auxiliary building exhaust
and supply fans running to generate a negative pressure of 0.17 inches
w.g in the auxiliary building, and turbine building and intermediate
building pressures at zero inches w.g. (equal to atmospheric pressure).
During the test, pressure in the control room was measured to be at
negative 0.045 inches water gauge (w.g.) By using the test results and
a calculation method different than that previously approved by the NRC,
Enclosure 2
l
__
. . . . . . . .. . .
. . .. _.
t ) l
54
the licensee concluded that the CCHE leakage during a maximum
hypothetical accident (with loss of offsite power and no auxiliary
building fans running) would not cause operator dose limits to be
exceeded. In that operator dose analysis, the licensee assumed that the
only cause of CCHE in-leakage would be outside wind.
The. inspectors used the licensee's test data to calculate what the CCHE
leakage would be if the NRC-approved Standard Review Plan (SRP) method
were used. The SRP requires determining CCHE leakage with a 0.125
inches w.g. pressure in the CCHE and dividing that number by two, adding
10 CFM for opening and closing of doors and also adding leakage through
CREVS closed boundary damaers due to pressures caused by the CREVS fans
running. The test data slowed that, with 0.125 inches w.g. in the CCHE
and zero 3ressure in the auxiliary building, the out-leakage from the
CCHE to t1e auxiliary building would be 462 CFM. (The differential
pressure from the CCHE to the auxiliary building during the test was
0.17 minus 0.045 inches w.g. , which equals 0.125 inches w.g. Also,
since the auxiliary building was the only area with a lower test
pressure than the CCHE. essentially all of the 462 CFM of CCHE
outleakage must have gone into the auxiliary building.) During the
test, pressure in the CCHE was stable. Therefore, in-leakage from the
turbine building, intermediate building, and outside atmosphere into the
CCHE had to equal the out-leakage to the auxiliary building of 462 CFM.
Since the air leakage rate varies as the square root of the differential
3ressure. SRP in leakage across all CCHE walls other than the auxiliary
auilding wall, with 0.125 inches w.g. pressure in the CCHE would be
about: 462 times the square root of (0.125 divided by 0.045), which
equals 770 CFM. Adding the 462 CFM for the auxiliary building wall to
the 770 CFM for all other walls iesults in a total CCHE leakage of 1232
CFM. Then, dividing the 1232 CFM by 2 equals 616 CFM of SRP leakage.
Adding 10 CFM for doors and zero for CREVS boundary dampers (because
CREVS fans were running during the test and because the new boundary
damaers are to have zero leakage) results in an a) proximate SRP CCHE
leatage value of 626 CFM. The inspectors noted t1at 626 CFM was
substantially higher than the 355 CFM maximum SRP CCHE leakage that had
been stated by the licensee in their letters in the 1980s to the NRC in
response to NUREG-0737. Item 111.D.3.4. Since the licensee's analysis
based on 355 CFM of SRP CCHE leakage resulted in a calculated operator
dose that was near the regulatory limit of 30 REM to the thyroid, the
inspectors assessed that a similar analysis based on approximately 626
CFM of SRP CCHE leakage would result in a calculated operator dose that
would exceed the regulatory li'it.
The inspectors used the test data to estimate the cumulative size of
leakage ')aths in the CCHE walls. Since each square inch of leakage path
will leac about six CFM at 0.125 inches w.g. differential pressure. the
inspectors calculated that the cumulative size of leakage paths in the
CCHE to auxiliary building wall was approximately 462 CFM divided by 6
CFM per square inch, or approximately 77 scuare inches. Similarly, the
leakage paths of the other walls of the CCFE were calculated to be
Enclosure 2
i
l
______ __ _.__ _ _ _ _ _ __ __
_.
.
,
55-
approximately 770 CFM divided by 6 CFM per square . inch, or approximately_
-
128 square inches. The total cumulative CCHE leakage )aths were
therefore approximately 77 plus _128, or 205 square incies.
The inspectors reviewed this evaluation with licensee engineers and
incuired about the locations of the approximately 20^ square inches of
CCFE leakage paths and the difficulty in sealing them. The engineers
found no fault with the inspector's analysis and stated that the leakage
'
paths were primarily in cable trays that penetrated the CCHE walls. The-
spaces between the cables stacked in the trays were not fully sealed,
resulting in detectable air leakage during the test. Most of these
cable tray penetrations were from electrical rooms in the lower levels
of the CCHE to the auxiliary building and to the turbine building. The
-inspectors looked at some of these cable tray penetrations in the plant
and observed that abysical access to them was difficult. Licensee
engineers stated t1e licensee was testing a'new fire retardant sealant
and applicator, and planned to apply _ the sealant to the cable tray
penetrations in early 1998.
The inspectors noted that the licensee's recent CCHE leak test data
included measured differential pressures, both before and during the
tests, between various rooms in the CCHE and the auxiliary building or
-the turbine building. Using the October 11, 199? test data, the
inspectors plotted the pressures in the vcrious rooms with respect tc
atmospheric pressure. The plot revealed that there were substantial
differences in pressures among the various rooms. The pressures in some
of the CCHE rooms with auxiliary building pressure at minus .04 and
minus .17 inches w.g., are listed below.
Pressure (in. w.o.)
Auxiliary Building -0.04- -0.17
CCHE Room
. Control Room +0.06 -0.04
'
~B" 4160 Volt SWGR -0.01 -0.10
Cable Spreading Room -0.07 -0.16
EFIC/480 Volt SWGR -0.10 -0.20
In each case. the pressure in the control room was higher than in any
other CCHE room. The electrical rooms below the control room were each
at a substantially lower pressure than the control room. At least one.-
the EFIC/480 Volt SWGR room, was at a lower pressure than the auxiliary
building. The inspectors confirmed the above CCHE room relative
)ressures with the CREVS in recirculation and the normal auxiliary
3uilding ventilation operating by entering each CCHE room and noting
the velocity of air flow past each room's doors when cracked open. The
inspectors concluded that the CREVS fans were causing substantial
differences in pressures among the CCHE rooms. The inspectors also
noted that when the auxiliary building pressure was reduced by 0.13
inches w.g. (from .04 to .17), the pressure in each room in the CCHE
Enclosure 2
56
went down by about 0.10 inches w.g., and the differences in pressures
among rooms remained approximately constant. This provided further
indication that the differences in pressures among CCHE rooms were-
caused by CCHE fans and not by forces originating outside the CCHE. As
a result, there were also significant pressure differences between
certain CCHE rooms and the auxiliary building or the turbine building.
These pressure differences existed in rooms that had known cable tray
leakage )aths to the auxiliary building and turbine building. It was
likely tlat a substantial portion of the CCHE leakage that was measured
during the recent test was caused by the CREVS fans. Also, contrary to
the licensee's analysis in their November 10, 1997 letter, it was likely
that a substantial amount of CCHE leakage would be caused by the CREVS
fans during a maximum hypothetical accident. Licensee personnel stated
that they would address the CCHE leakage due to CREVS fans in a planned
revised submittal to the NRC.
c. Conclusions
The inspectors concluded that the licensee's CCHE leakage analysis, in
their November 10, 1997 letter to the NRC failed to recognize the
potential for CREVS fans to cause a substantial amount of CCHE leakage
during accident conditions. Licensee personnel stated that they would
address the CCHE leakage due to CREVS ' fans in a revised submittal to the
NRC. URI 50-302/95-02-02 remains open for further licensee analysis and
NRC review.
E8.17 (Closed) VIO 50-302/97-01-09: Inadeauate Correct've Actions for Cable
Amoacity
(Closed) LER 97-31-00: Inadeauate Cable Sizino Due to Nonconservative
De-ratirq Factors Could Reduce the Cable Remainino Oualified Life
a. Insnection Scoce (92903)
The violation involved a case where a potential non-conformance with
regard to the ampacity of electric cables had been identified by the
licensee, but the potential non-conformances were not corrected nor
satisfactorily resolved through analysis. The licensee's res)onse dated
April 23. 1997, stated that the technical issues which were t1e subject
of the violation would be resolved through implementation of Restart
Issue D-22. The ins)ector reviewed the licensee's analysis performed
under Issue D-22. T1e violation indicated weaknesses in the licensee's
Corrective Action Program, however. that program has been substantially
upgradedandhasbeenthesubjectoftworecentNRCinspections.
Toerefore. this. inspection focused on the technical issue only.
Enclosure 2
.
57
b. Observations and'Findinos
/ts part of the resolution.of Issue D-22, the licensee )erformed Case
Study CSEH-97-0012A. Electrical Calculation E-91-0020 M Assessment,
which' reviewed all the known potential problems with the-ampacity of
power cables-and reviewed all relevant engineering documents issued
since initial plant startup, to determine whether those documents
contained information affecting the ampacity calculation for power
cables. The ampacity calculations consisted of a separate analysis
sheet for each cable. Within the case study, after the reviews stated
above were completed, ampacity calculation sheets were marked up with
correct information as necessary. The case study was subjected to
independent review and was approved by the responsible supervisor.
Therefore. the case study hel)ed ensure that potential problems were
satisfactorily analyzed and tlat the ampacity calculation accurately
reflected the as-built configuration. There were separate case studies
for control cabler and vital ac cables as well.
The ampacity calculations, (i.e. E-91-0020) were performed using ICEA
Publication P-46-426. Power Cable Ampacities, as a guide. This standard
determines ampacity of cable in cable tray by applying a derate factor
based on the total number of conductors in the tray to the ampacity of
cable in free air. In cases where the cable tray is wrapped with
Thermo-Lag. Mecatiss or a combination of the two, an additional fairly
severe derate factor must be applied. The fire wrap derate factor was
recently revised, and this sub-issue will be discussed later in this
section. A number of cables in a number of tray sections had
insufficient ampacity when the ICEA and fire wrap derate factors were
a] plied. This did not mean that the cable size was inadequate, because
t7e ICEA derate factor is correct when all the conductors in the tray
continuously carry their maximum allowable current. Since many of the
cables in the tray sections in question carry currents much less than
maximum allowable, the derate factor was conservative. The licensee's
approach to resolving this problem was to use a thermodynamic computer
program to estimate actual temperatures of conductors.
Thethermodynamicanalysisconsistedofmodelingeachtraywrabpedin
Thermo-Lag and certain non-wrapped tray (e.g. trays 105, 505, O7 and
508 to analyze cable DCL-1). Actual load currents were entered as input
data'and the most heavily loaded cables were grouped at the center of
the tray. This conservative modeling was necessary to deal with cases
where the actual location of each cable in the tray under study was not
known.
Temperature detectors were installed in certain trays and wired to
recorders which also recorded the status (on - off) of the associated
loads, In this manner, the actual temperature of the outer surface of a
particular cable was determined. Then the actual conductor temperature
could be calculated.
Enclosure 2
.__ _ _
58
The results of_ this analysis with the conservative modeling was that- .
certain cables in seven trays had calculated conductor temperatures in ,
excess of the rated 90 C. All of these trays contained circuits
operating at 480 V or 120 V. There were no ampacity issues with -
circuits operating at voltages higher than 480 V. For four of the
trays, the calculated temperature was only slightly above 90*C.
Remaining life calculations showed that the cables in these four trays
had a remaining life of from 14 to 27 years. No further analysis was
performed on these. For three of the trays (one containing all.480 V
power circuits and two containing primarily 120 V control circuits with
a few low-level power 480 V ), the calculated conductor temperature of
certain cables was considerably above 90*C. Remaining life-calculations
yielded inadequate remaining life. These cables were physically
examined and subjected to a non-destructive Indenter Polymer Aging
Monitor test. The indentor test was developed by the Electric Power
Research Institute (EPRI)(refer to publication EPRI TR-104075). The
indentor is a computer controlled test device that presses an anvil
against the side of a cable at a constant velocity. The anvil is
instrumented to provide position and force measurement. The slope of
the force vs position curve is the " Indentor Modulus" which indicates
the hardness of the polymer and directly relates to aging due to heat.
The examination and the indentor test indicated that the cables had not
been overheated in past operation. This indicated that the most heavily
loaded cables were evenly distributed rather than tightly grouped as
modeled in the computer program. I'. was also determined that past
operation was not less severe than accident scenarios in terms of the
currents carried by the cables. LER 97-31-00. Inadequate Cable Sizing
Due to Nonconservative De-rating Factors Could Reduce the Cable
Remaining Qualified Life, dated September 26, 1997. reported the
situation with the later three cable trays.
LER 97-31 also discussed potential problems with the ampacity of the
feeder cables for the rector building cooling air handling fans AHF-1A.
18 and IC. The problem was that the calculation indicated that the full
load current of 170 A was carried by two conductors per 3hase, but
actually this current was carried by one conductor per plase.
Contributing to this error was the fact that the fans were two speed
type and two cables were in fact run to the motor. Upon further
analysis the cable was still acceptable. The correct configuration was
pr;oably understood at the time of original plant design. It was the
r ew calculation performed in 1992 that made the error. This error in
the calculations did not represent a violation of NRC requirements
because it was discovered as aart of the corrective action for Violation
97-01-09. The inspector checced the marked-u) ampacity calculation
sheets for these cables, and verified the as-Juilt configuration of the
associated trays and conduits in a walkdown inspection. LER 97-31 was
closed.
Enclosure 2
. _ _ _
_ m _ .. . . _ .. _ _ _ _ _ . .
:
y-
:
;
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59 1
'
JThe th'ermodynamic analy' sis and the' remainingilife ca'culations were-
~
contained in Electrical. Cable Operability Evaluation (Cable Ampacity . ;
Concerns) Revision 1, dated November 15, 1997.
F
Ongoing plant modifications _were putting MecatissJon certain trays and ,
'
. conduits. The ampacity considerations associated with adding Mecatiss-
were addressed in Calculation E-96-0003, Revision 0, dated May 9, 1996.'
The derate factors applied-in reviewing the ampacity of cables in-the
'Mecatiss wrap)ed raceways were determined in a test conducted by
Underwriters .aboratory. (UL). A report on this test was submitted.to
. NRC- Headquarters for review pursuant'to-Generic Letter.92-08. This.
report was reviewed by NRC and its consultant, and a request for- -
additional information was'sent to the licensee 11n a letter dated May -
22, 1997. The licensee responded to the request in a letter dated July'
~3, 1997. The inspector reviewed these two~ letters and the consultant's
report,.and concluded that the NRC had no further questions regarding '
the acceptability of the derate factors. The test also included
determination.of derate factors for use with Thermo-Lag and Thermo-Lag-
Mecatiss combination wrap. As explained above, avhen the new derate . .
factors.for Thermo-Lag were applied to existing installations. certain
cables became potential problems with regard to ampacity, and this was
addressed in an operability evaluation. The analysis to support recent
. modifications to install Mecatiss applied correct derate factors, and
. cables were rerouted-as necessary to achieve an installation consistent
'with the ICEA standards. An example of a cable being re outed due to
the addition of Mecatiss was cable AHC-957, which was r noved from tray
643 by MAR 96-01-05-01. The inspector verified that this cable had in
fact been re-routed by on-site inspection of the new installation.
The Electrical Cable Operability Evaluation concluded that cables were
g either acceptable for the life of-the plant (i.e. another 20 years) or
.certain cables (order of magnitude 100 cables) were acceptable for at
t least one operating cycle. For these cables further analysis would be
performed in an attempt to demonstrate longer qualified life. The
present o)erability of these cables was based primarily on examination
'
of the ca)le, an indentor test and remaining life calculations, which
showed 14 years remaining life,
After review of the Case Study (which was really an extent of condition
type review) and the operability evaluation performed under Restart
. Issue:D-22, the inspector found that the corrective actions stated in
4
the response to Violation 97-01-09 were satisfactorily performed.
~
Therefore the violatior, and related LER 97-31 were closed. The licensee
:was in the process-of pre)aring a justification for continued operation
.for oneifuelicycle in lig1t of the cable ampacity sit"ation described
above. 'Further analysis would-be performed to determine a final
resolutionLofLthe ampacity issue. -In order to ensure NRC review of the
-
e
. final'or long-term resolution, Inspector Follow up Item IFI 50-302/97-
17-03. Review of Cable Ampacity Issue, was established.
.
Enclosure 2
4
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vi
-
( yr. T9 9 ,my y .is. y v. . , . , - -
60
c. Conclusions
The licensee performed sufficient review and analysis to define the
extent of any potential problems with cable ampacity. A number (f
problems were permanently resolved. Problems with one set of cables
were not permanently resolved, but an operability evaluation supported
operability for at least one fuel cycle (about two years). The
inspector agreed that the electrical cable 03erability evaluation had a
sound basis. An Inspector Follow up item (1:1 50-302/97-17-03) was
established to ensure NRC review of the final or long term resolution of
the cable ampacity issue.
The inspector assessed the licensee's performance relative to cable
ampacity issue in the five areas of continuing NRC concern.
. Management Oversight - Good
. Engineering Effectiveness Good
. Knowledge of the Design Basis - Good
. Compliance with Regulations - Good
. Operator Performance - N/A
E8.18 (Closed) IFI 50-302/97-02-05: Outstandina Issues Associated with the
Emeraency Diesel Generator Power Unorade Modification
a. Inspection Scone (92903)
The inspectors reviewed the actions taken to address the issues
identified in IFI 50-302/97-02-05.
b. Observations and Findinas
Three issues were identified in the refercqced IFI regarding the
emergency diesel generator power upgrade modification.
Precursor Card 97-0996 was issued on February 28, 1997, to document that
a discrepancy was identified with the emergency diesel (EDG) fuel oil
volume requirements. The TS requirements for the EDG day tank (each)
and underground storage tanks (combined) is 245 gallons and 37.177
gallons, respectively. The licensee had identified that during the
revision to Calculation M89-0012. as part of the EDG upgrade
modification, a non-conservative assumption was identified in the
assumed American Petroleum Institute (API) specific gravity for the fuel
oil. During the calculation review cycle, it was not identified that
the calculation had impacted the existing TS limits. During the
subsequent System Readiness Review, the impact was identified. The
licensee's immediate corrective action was to revise SP-354A and SP-
354B. the EDG monthly surveillance procedures, to require that a more
conservative minimum value was maintained in the fuel oil tanks. The
licensee performed a review of tank volumes between the period of
January 1986 to March 1997 and identified four instances where an
Enclosure 2
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l
l
61 !
underground tank did not contain the minimum volume. but considering the
volume in the day tank and the other tank, the total minimum j
requirements were met.
To control future volumes, the licensee revised the purchase
specification to include a maximum limit for API gravity. SP-7461. !
Diesel Fuel Oil Testing Surveillance Program: New Diesel Fuel Receipt. '
has oeen revised to ensure that new diesel fuel receipts meet the new
purchase specification. Technical Specification change request 210
includes proposed changes on the emergency diesel generator. including i
the new. more restrictive values for EDG fuel oil requirements, l>
Precursor Card 97-1501 was issued on harch 4,1997, to identify the ,
potential for the EDG exciter to exceed potentially the design *
-
limitation of 54 amperes at maximum loading. based on testing performed i
'
on February 1. 1997. On March 14 and 15, 1997. further testing was
conducted to evaluate the capability of the EDG exciter. The licensee
concluded that the original data were inconclusive as the field amperage
data were not obtained at the same time as the watt and power factor
data. The second set of data was gathered with coordinated .
communications to allow simultaneous recording. The licensee concluded '
that the exciter was designed for 54 amperes operation at 3250 KW with a
O.8 power factor and had a nameplate of 3000 KW. with a 0.8 power factor.
The test data demonstrated that the exciter operated acceptably within
the nameplate rating. The calculations based on data gathered resulted .
'
in acceptable values. even for the power upgrade. During the
modification functional test for EDGD-18. the design values for exciter -
amperage was not exceeded. The licensee has closed the PC based on the
new calculations. The conclusions reached were supported by the testing
following the modifications on the diesels,
Precursor Card 97-0999 was issued on March 4. 1997, to discuss the
assessment of data obtained during the February 1.1997, test which
demonstrated that the EDG engine room temperature could potentially
er.ceed design maximum temperature of 120 F. at the maximum loading.
Precursor Card 97-3300 was issued on June 23. 1997, to identify concerns >
with the recirculation of hot radiator air exh6ast which directly
affects supply air temperature to the EDG room. The basic impact
identified was an increase in the effective inlet temperature of up to
15 F, On June 26. 1997. LER 50-302/97-013 was issued to address these
same issues.
c. Conclusions
The licensee has addressed the concerns with the fuel oil API Gravity
through administrative controis, pending final issuance of a Te@nical
-Specification amendment on minimum fuel oil requirements. Testing has .
'
Enclosure 2
.
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vv yw -,
m .g we+ . ---w - -
r- e
62
dispositioned concerns with the EDG exciter amperage. These items are
closed. The third concern, with the EDG elevated temperatures will be
addressed in the fLilow up to LEP 50-302/97-013. IFl 50-302/97-02-05 is
closed.
The inspectors assessed the licensee's )erformance, relative to
corrective actions for this issue, in t7e five areas of continuing NRC
concern:
* Management Oversight - Good
* Engineering Effectiveness Adequate
* Knowledge of Design Basis - Adecuate
* Compliance with Regulations - A: equate
* Operator Performance - N/A
E8.19 (Closed) VIO 50-302/97-11-06: Fajjure to Follow licensee Procedure NEP-
2fL4
a. Insoection Scone (929031
This violation involved a determination by the inspectors that the
licensee had inappropriately perfortrui modifications to the building
spray pump im]ellers under the Plant Equipment Equivalency Replacement
Evaluation (PE.ERE) 3rocess rather than the MAR process as required by
licensee procedure 1EP-254. Plant Equipment Equivalency Replacement
Evaluation. The licensee responded to the violation in a letter dated
October 8. 1997. The inspectors reviewed the licensee's corrective
actions as stated in the response.
b. Observations and Findjptqi
The licensee performed a review of completed PEEREs and determined that
both PEERE 1497 and PEERE 685 should have been processed as MARS. Both
PEEREs were voided, with the work previously accomplished under PEERE
1497 being reassessed and documented under MAR 97-09-08-01. PEERE 685
had not been implemented, so no further work had been performed on it.
The inspector reviewed the results of the licensee review and discussed
the findings with the licensee. Based on the licensee's review and the
results on a review of PEEREs conducted by the licensee, no additional
concern was identified with existing PEEREs.
The licensee issued an interoffice corrcspondence to all engineering and
modification personnel or, September 11. 1997, describing the issue and
providing interim guidance on the correct use of the PEERE process,
while clarifications were made to NEP-254. On November 7. 1997, a
revision to NEP-254 was issued which included a clarification for the
use of the PEERE process and a checklist to guide the personnel through
the decision making.
Enclosure 2
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% -
63
Training classes were conduced fo,' engineering and modifications
personnel 07 September 16. ;397. The violation was discussed with the
personnel and corrective actions were outlined with the personnel. The
inspectors reviewed the attendance sheets and noted that the inajority of
the required personnel were in attendance for the training.
The licensee has scheduled a self assessment conducted by the
engineering department to determine the effectiveness of the corrective
actions for this issue prior to July 15. 1998.
c. Conclusions
The inspectors have verified that all required corrective actions have
been completed. In addition. the licensee has scheduled a self-
assessment to assess the effectiveness of the corrective actions after a
period of implementing the nc 9 process. This violation is closed.
The inspectors assessed the 1.censee's p. iormance, relative to
corrective actions for this violation, in the five areas of continuing
NRC concern:
. Management Oversight - Good
. Engineering Effectiveness - Good
. Knowledge of Design Basis - Adecuate
. Compliance with Regulations - Acequate
. Operator Performance - N/A
E8.20 (Closed) IFl 50-302/95-15-05: Relief Valves Removed From Heat Exchanaers
(92903)
a. InspectionStone
This follow up item was identified during the service water system self-
assessment inspection. A design change rackage (MAR 80 04-13-01) had
been issued to allow the removal of rel;ef valves from various sets of
heat exchangers in the Nuclear Services Closed Cycle Cooling (SW) and
Decay Heat Closed Cycle Cooling (DC) systems. The modification was made
to address a chronic problem with leakage and because these valves were
tho"]ht to be redundant. The heat exchangers affected included the
reactor building coolers and fan motor coolers; the reactor coolant pump
motors, bearings, and seal coolers; and the makeup and purification pump
motor coolers. Code USAS B31.1-1967. Section 122.6.1 requires that no
intervening stop valves shall separate a protected component from its
overpressure protection device. Contrary to this requirement. MAR 80-
04-13-01 allowed inter" =ing stop valves between the relief devices and
the components being pro m ed.
.
Enclosure 2
64
b. Observation and Findinas
To address the issue raised during the service water inspection. MAR 96-10-04-02 was developed to re-install the relief valves that were removed by MAR 80-04-13 01. Also per MAR 96-10-04-02, all NuPro thermal re:ief valves were to be replaced with Anderson-b eenwood valves because the NuPro valves' relief capacity was derated by the vendor. As part of the process to close IFl 95-15-05, the inspector requested a walkdown with the system engineer to verify that the valves had been replaced /re- installed. During the walkdown on November 19. 1997, a DC system thermal relief valve (DCV-109) was identified as not having been replaced. DCV-109 was a NuPro valve, PC 97-7920 was written to document this discrepancy and assigned a grade level "C." Following discussions with the project engineer associated with MAR 96- 10 4-02 it was determined that a previous failure to update the Configuration Management Information System (CMIS) and a personnel error both contributed to DCV 109 not being replaced. Specifically, CMis was not updated when the original Texsteam valve was rerlaced by the NuPro valve in the early 1980s. The project engineer overlooked valve DCV-109 when preparing the work request to rep' ce the NuPro valve with an Anderson-Greenwood valve, and therefon. DC',109 was not included in MAR 96-10-04 02. This oversight was further perpetuated when the design change package was closed-out and submitted in a restart issue folder for NRC inspection. A Field Change Notice was issued to install an Anderson / Greenwood relief valve and will be completed prior to plant restart. The inspector determined that the original failure to recognize the requirements and establish the design basis for multiple relief valves to protect components from thermal overpressurization and the omission of a valve identified as required to be replaced and the failure to recognize this omission during closecut of the design change package. constitutes a violation of 10 CFR 50. Appendix B. Criterion 111. L,esign Control. This will be tracked as VIO 50-302/97-17-04 Inadequate Design Control.
c. Conclusion
The inspector assessed the licensee's performance, relative to this
violation. in the five areas of continuing NRC concern:
* Management Oversight - Inadequate
. Engineering Effect veness - Inadequate
e Knowledge of Design Basis - Inadequate
* Compliance with Regulation - Inadequate
e Operator Performance - N/A
Enclosure 2
65
E8.21 fClosed) VIO 50-302/96-09 06. Erroneous Calculation Inouts and Inservice
: nspection Boundarv
(Ooen) LER 50-302/97-038: Enoineerina Oversicht Resulted in Doeration
Outs 1de Deslan Basis of Waste Disposal System
a. Inspection Scone (92903)
This violation involved a failure of the licensee to assure that
applicable regulatory requirements and design basis were correctly
transl6ted into specifications, drawings, procedures and instruction.
The violation involved two examples of calculation errors and an example
of an erroneously located Inservice inspection boundary. The inspector
reviewed documentation, and interviewed licensee oersonnel to assess the
adequacy of the licensee's corrective actions,
b. Observations and Findinos
The inspector noted that resolution of this violation was being tracked
under licensee Restart Issue OP-31A. The two parts of this violation
were as follows:
* The design input currently used in calculations for safety
related battery charger (MAR 93-05-07-01) and 4160/480 volt
transformer (MAR 95-08-22-01) replacements was incorrect.
The input currently used in the calculations was 56 amperes,
whereas the correct value was 62 amperec
. The Inservice inspection class 2/3 makeup system boundary
shown on FS\R drawing FD-302-661. Sheet 4, was not moved
from Valve MUV-64 to Valve MUV-65 in 1984. when the
Engineered Safeguard signal was removed from MUV-64. With
that change. MUV-64 could no longer be considered a
boundary, as it was open and would not automatically close
to provide a boundary.
The inspector reviewed the corrective actions identified in the
licensee's letter dated May 20. 1997. In this letter, the licensee
committed (at the request of the NRC) to address the seismic
classification break problems that had been identified subsequent to the
issuance of VIO 96 09-06. The inspector reviewed the affected
documentation and interviewed licensee personnel to determine if the
corrective actions were implemented and to assess the adequacy of the
corrective actions for item 1 and item 2 of the violation. The
ins]ector reviewed sections E8.3 of NRC IR 50-302/97-07 and section El.1
of 4RC 1R 50-302/96-03. Followino the review of the inspection reports
the inspector concluded that the Ticensee had implemented adequate
corrective actions to address the concerns identified in item 1.
Therefore, item #1 of VIO 50-302/96-09-06 is closed.
Enclosure 2
66
The inspector also reviewed documentation which addressed concerns
associated with Item #2, The inspector reviewed selected drawings,
procedures and documentation submitted to the NRC. NRC inspection
reports and interviewed licensee personnel to determine if adequate
corrective actions had been implemented for those concerns identified in
item #2. After reviewing NRC 1R 50-302/97-16. the inspector found that
additional concerns associated with Item #2 were identified. Specifically, it was concluded in IR 97-16 that. ". . .there was a weakness in the corrective actions for VIO 96 06-06 in that the extent of condition review did not include all of the WDS tanks and associated piping."
in a supplemc. cal response to VIO 96 06-06, dated October 30, 1997, the licensee provided the results of its extent of condition review of the
design basis requirements for seismic and inservice inspection (ISI) classification of piping systems, with focus on the interface requ?rements which govern the transition between class boundaries. In the licensee's October 1997 response. the discrepancies identified by the NRC in IR 97-16 were addressed. The letter also addressed discrepancies 1dentified from the licensee's own extent of condition effort. In the letter dated October 30. J97. the licensee stated that the result of their review effort and any corrective actions will be re)orted to the NRC under LER 97-038-000. The inspector reviewed the su] ject LER to determine what additional corrective actions had been identified, and if the corrective actions had been implemented. Following the review of the LER the inspector determined that the licensee was continuing its effort to address all discrepancies and implement the necessa y corrective actions. In LER 97-038-000, the licensee made the fol,owing commitments to resolve the issues.
* The liquid waste outlet piping for the haste Gas Decay
Tanks. Miscellaneous Waste Storage Tank. Spent Resin Storage
Tank, and Neutralizer Tank will be upgraded to Seism : Class
1 - Refueling Outage (llR).
. Reactor Coolant Drain Tank Liquid waste outlet alping is
being evaluated by FPC, in accordance with 10 C R 50.59. as
a change from seismic to non-seismic. The FSAR will be
amended as appropriate - Refu2 ling Outage (11R).
* A Justification for Continued Operation for the WD system is
being Developed by FPC, consistent with Generic Letter 91-
18. Revision 1. Prior to MODE 4.
Based on the corrective actions Presently c *pleted by the licensee and the comitment made in the LEP. the inspector concluded that the licensee's actions addressing #2 of VIO 50-302/96-09-06 are acceptable for restart. The NRC wili track the implementation of the remaining corrective actions under LER 97-038-00. Although this item is a noncompliance with regulatory requirements, for the reasons discussed in
Enclosure 2
67
Inspection Report 97-21. the licensee met the criteria for enforcement
discretion per Section Vll.B.2 of the NRC Enforcement Policy as
described in NUREG-1600. Consequently this 1 tem is closed and is
identified as another example of Non-cited Violation NCV 50 302/97-21-
01 Examples of Noncompliances in Design Control. 10 CFR 50.59
Evaluations. Procedure Adequacy / Adherence. Reportability, and Corrective
Actions That Are Subject to Enforcement Discretion.
c. Conclusions
The inspector concluded that the licensee's corrective actions for this
violation were adequate. However, it sk ald be noted that the licensee
implemented several of the corrective actions only after additional
discre:>ancies were identified by the NRC. The ir.:pector concluded that,
even t1ough several of the corrective actions' implementation are not
presently completed, the time table established by the_ licensee for
completion of the actions is acceptable. Therefore, this violation is
closed.
The inspector assessed the licensee's performance, relative to the
corrective actions for this violation, in the five areas of continuing
NRC concern:
. Management Oversight - Adequate
* Engineering Ef fectiveness - Adequate
e Knowledge of the Design Basis - Adequate
e Compliance with Regulations - Adequate
. Operator Performance - N/A
E8.22 (Closed) EA 95-16: Use of Nonconservative Trin Setnoints for Safety-
Related Eauipment
(Closed) LER 50-302/94-006-00 throuch LER 50-302/94-006 06: Deficiency
in Understanding of Technical Reou1rements leads to Nonconservative
Safety Sv11 ems Setnoint and Violations of Imoroved Techn1 cal
Specifications
a. Inspection Scone (92903. 37550)
As part of the continuing review of corrective actions for EA 95-16 the
inspectors reviewed several new instrument loop uncertainty setpoint
calculations and LER 50-302/94-006. Deficiency in Understanding of
Technical Requirements Leads to Nonconservative Safety Systems Setaoint
and Violations of Improved Technical Specifications. Revisions 0 t1 rough
6. In IR 50-302/95-06 the inspectors found that some safety-related
trip setpoint calculations did not follow the methodology specified in
Instrument Society of America (ISA) 67.04. Part 11. as referenced by
instrumentation and controls Design Criteria Instrument String Error and
Setpoint Determination Methodology. To assess the progress the licensee
had made in this area, the inspector reviewed a sample of the most
Enclosure 2
l
68
recent instrument string error /setpoints. Various discrepancies with
the instrument loop u.1 certainty calculations were reported in LER 50-
302/94-006. Revision 0 throuah 6. The ins)ectors reviewed the
licensee's actions associate 0 with this LER and it's subsequent
revisions.
b. Observations and Findinc-
The inspector reviewed a representative sample of instrument loop
uncertainty setJoint calculations. These calculations were well
documerited. wit 1 well founded assumptions, and followed the methodology
specified in ISA 67.04. Part 11. a referenced by instrument and
controls Design Criteria Instrument String Error /Setpoint Determination
Methodology. Through a field walkdown, the inspector verified the
instrumentation installed in the field was appropriately included in the
setpoint calculations.
However the licensee had not completed the field installation for some
of the setpoint loo) uncertainty calculations. In addition the licensee
had not completed t7e Alarm Response Procedures. Surveillance
Procedures. or Operating Procedures for several setpoint loop
uncertainty calculations. These included:
* Setpoint Calculation 188 0022. RC (T Hot) Temperature Loop
Accuracy. RC-4A-TEl RC-4B-TE4. Rev. 6
-
Procedures requiring revision for this calculation:
SP-161A. Reactor Coolant T m and T,nia Calibration. Rev.19
e Setpoint Calculation 190-0019. Reactor Bldg. Pressure Loop
Accuracy (BS-16/17). Rev. 1
-
Procedures requiring revision for this calculation:
SP-162. Post Accident Monitoring Instrumentation Channel
Calibration. Rev. 33 (See 195-0017 for Procedures Changed
to use Narrow Range RB Pressure)
* Setpoint Calculation 191-0012. BWST Level Accuracy. Rev. 3
-
Procedures requiring revision for this calculation:
AR-303. Esc Annunciator Response. Rev. 26
SP-162, Post Accident Monitoring Instrumentation Channel
Calibration. Rev. 33
SP-300. Operating Daily Surveillance Log. Rev. 139
Enclosure 2
_ _ _________ ___ ____ _- -_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _
69
SP-301. Shutdown Daily Surveillance Log. Rev. 104
OP-103 . Tank Volumes Rev. 9
e Setpoint Calculation 191-0021. RC Flow Loop (NNI) Accuracy. Rev.
1
-
Procedures requiring revision for this calculation:
AR-502. ICS J Annunciator Response. Rev. 10
SP-112. Calibration of the Reactor Protection System. Rev.
57
SP-162. Post Accident Monitoring Instrumentaticn Channel
Calibration. Rev. 33
SP-300. Operating Daily Surveillance Log Rev. 139
. Setpoir,t Calculation 195-0017. Reactor Building Narrow Range
Pressure. Rev. 0
-
Procedures requiring revision for this calculation:
SP-135A, Engineered Safeguards Actuation Channel 1 System
Response Time Test Rev. 12
SP-135B. Engineered Safeguards Actuation Channel 2 System
Response Time Test. Rev. 15
SP-135C. Engineered Safeguards Actuation Channel 3 System
Response Time Test Rev. 13
SP-162. Post Accident Monitoring Instrumentation Channel
Calibration, Rev. 33
SP-300. Operating Daily Surveillance Log Rev. 139
SP-347. ECCS and Boration Flow Paths. Rev. 45
SP-356. ES Manual Actuation Channel Functional Test for RB
Isolation and Cooling Rev. 20
SP-357. ES Manual Actuation Channel Functional Test for High
Pressure Injection and Low Pressure injection. Rev. 19
SP-456. Re weling Interval Equipment Response to an ESAS
Test Signal Rev. 18
Enclosure 2
70
e Setpoint Calculation 185-0004. Dedicated EFW Tank Alarm Settings.
Rev. 5
-
Procedures requiring revision for this calculation:
AR-403. PSA H Annunciator Response. Rev. 29
OP-103F. Tank Volumes, Rev. 9
OP-450. Emergency feedwater System. Rev. 20
SP-162. Post Accident Monitoring Instrumentation Channel
Calibration, Rev. 33
SP-300. Operating Daily Surveillance Log. Rev. 139
SP-338. Remote Shutdown and Post Accident Monitoring Channel
Check, Rev. 27
* Setpoint Calculation 187-0003. EFW Flow Control & Interlock, Rev.
5
- Procedures requiring revision for this calculation:
SP-193B. EFW Flow Transmitter Channel Calibration. Rev. 2
c. fancJusions
The inspectors concluded that the licensee continued to make progress in
resolving the Imaroved Technical Specifications (ITS) setpoint program
deficiencies. T1e calculations reviewed were well documented, with
well founded assumptions and followed the methodology specified in ISA
67.04 part 11. as referenced by instrumentation and controls Design
Criteria Instrument String Error /Setpoint Determination Methodology.
The licensee's actions were adequate to close the open items. The items
that are not complete are scheduled to be completed prior to entry into
Mode 4. The completion of these items will be tracked as IFI 50-302/97-
17 05. Resolution of Improved Technical Specification Setpoint Program
Deficiencies Prior to Entry into Mode 4.
The inspectors assessed the licensee's performance relative to lack of
design control for assumptions in instrument setpoint calculations, in
the five areas of continuing NRC concern:
* Manarement Oversight - Good
. Engineering Effectiveness - Superior
e Knowledge of the Design Basis - Superior
* Compliance with Regulations - Superior
. Operator Performance - N/A
Enclosure 2
71
IV. Plant Sucoort
P8 Hiscellaneous EP Issues P8.1 (Closed) IFI 50-302/97-08-03: Variations in the classification and
interpretation of the Emeraency Action levels (EAls) by the Site
Emeroency Coordinatqrs (SECS) (92904)
a. Inspection Secae
The inspector evaluated applicable areas of Florida Power Corporation's
September 4. 1997 response to IFI 50-302/97-08-03. In the response, the
licensee committed to:
e develop an EAL Interpretation Guide.
. conduct classroom training. focusing on a review of the EAls, with
all Emergency Coordinators.
e conduct semi-annual EAL/ Protective Action Recommendation exercises
for each SEC to ensure consistent application, and
e target April 1998 as the date for submitting it's NUMARC EALs to
the NRC
b. Observations and Findinas
As part of the EAL and initial EAL Interpretation Guide training, the
licensee developed a questionnaire consisting of 19 scenarios. The
questionnaire was given to the SECS during their classroom training to
evaluate the SEC's ability to use the EALs and the guide to classify
scenarios. The inspectors reviewed the questionnaire and noted that the
scenarios were similar to the scenarios used by the inspectors during IR
50-302/97-08. but the licensee's scenarios were more diverse and
challenging. The licensee stated that during the EAL and initial EAL
Interpretation Guide training, the questionnaire facilitated active
discussions between the SECS and the class instructor. These
discussions produced a greatly improved EAL Interpretation Guide over
the licensee's initial Interpretation Guide. Also in the revision, the
licensee had annotated the applicable EAL with a number that
corresponded to the numbered interpretation in the EAL Interpretation
Guide.
The ins)ectors verified that all 25 SECS had received classroom training
on the EAL Interpretation Guide and a focused review of the EALs, by
verifying a SEC roster against the training attendance sheets.
To assess independently the effectiveness of the EAL Interpretation
Guide and SEC training, the inspector interviewed SECS from the training
roster. The inspector asked the SECS to classify sixteen inspector
Enclosure 2
72
prepared scenarios. The interviewees used the EALs and EAL Interpretation Guide to accurately classify the scenario. Because of the SECS training in October 1997, the next semi-annual EAL/ Protective Action Recommendation exercises for each SEC would not be required until April 1998. The ins:>ectors noted that additional clarification could improve some EALs. Tlese EALs were discussed with the licensee during the inspection and during the exit meeting. The licensee's managemerit acknowledged and took note of the inspector's comments. The issues discussed were as follows: . The lack of supporting documentation to provide a basis for
" Containment Gross Gamma monitor reading exceeding limits" of
greater than 1000 R/hr was to be classified as a Site Area
Emergency (SAE) and 10,000 R/hr for a General Emergency (GE).
. in the Site Area Emergency EAL for "All Alarms Lost". the word
" Transient" was defined. The definition of " Transient" in the
latest revision of the "EAL Interpretation Guide" did not include
" greater than 10 percent thermal power oscillation" as a
" transient"
. The titles " Safety Related Equipment" and " Safe Shutdown
Equipment" resulted in an apparent inconsistency in two EALs. The
title " Safety Related Equipment" was used in the EAL for " Fire
within the Protected Area", and the title, " Safe Shutdown
Equipment" was used in the EAL for " Missile Impact". If a fire
damaged a Containment Spray Pump, it would be classified as a SAE.
If a missile (nozzle broken off of a nitrogen bottle) damaged a
Containment Spray Pump, it would not be classified as an
emergency.
. The EAL for " Loss of Main and Emergency Feedwater" did not include
" Auxiliary Feedwater" The EAL had not been modified since the
licensee had determined that Auxiliary Feedwater (turbine driven)
was not a " Safety Related System". and would be considered a
separate non-safety related independent feedwater system.
. The word " Imminent," as defined in the "EAL Interpretation Guide."
did not clarify the intent of the word.
. The SAE and GE classification criteria for a " Steam Generator tube
leak" and a " Steam Generator tube rupture with a loss of offsite
power" were the same.
Enclosure 2
~ . . _ . - ~ . _ _ _. _ _ _ _ _ _ _ _ . _ . _ _ . . . _ . - _ - . ._
,
,
1
1
q
"
73 l
:
c. Conclusion j
- l
The inspectors concluded that the revised guide was considerably more
4
comprehensive and more precise than the initial guide and adequately
- addressed the inspectors' concerns in IFl 50-302/97-08-03. ;
-
:
. Management Oversight - Good :
e Engineering Effectiveness - N/A !
.. ' Knowledge of Design Basis - Good !
'
.- Compliance with Regulations - Good
c e Operator Performance - N/A j
'
V. Manaaement Meetinas
X1 - Exit Meeting Summary .
' ; 1
The inspection scope and findings were summarized on December 1.- 1997.
Proprie b / information is not contained in this report.- Dissenting
-
.
'
,
comments were not received from the licensee.
r
X3- Management Meeting Summary
- - X3.1 A meeting was held on October 30. 1997 at the FPC Training Center to *
. discuss Engineering issues. A separate meeting summary was issued on
4 November 21. 1997. : '
X3.2 A Public Meeting was held on October 31, 1997 at the FPC Training Center
to discuss the licensee's progress on readiness for restart. A separate
meeting summary was issued on November 21. 1997.
PARTIAL LIST OF PERSONS CONTACTED l
:
Licensees
R. Anderson.' Senior Vice President. Nuclear Operations -
,-- J.'Baumstark. Director. Quality Programs
"
J. Cowan. Vice President. Nuclear Production
R. Davis. Assistant Plant Director. Operations and Chemistry
'
-
R. Grazio. Director. Nuclear Regulatory Affairs .
- G, Halnon. Assistant Plant Director. Nuclear Safety -
- B. Hickle. Director, Restart-
J.- Holden. Site Director ,
D.:Kunsemiller. Manager, Naclear Licensing
-
M. Marano.-Director. Nuclear Site & Business Support t
, - - C. Pardee. Director. Nuclear Plant Operations
W. Pike.: Manager. Nuclear Regulatory Compliance ->
- M.- Rencheck. Director. Nuclear Engineering
M. Schiavoni. Assistant Plant Director. Maintenance .
:
T. Taylor.- Director. Nuclear Operations Training _'
o
,_ l
.
Enclosure 2 {
'; t
;
. . _ - . m u._ a m . ..u,- ._.u.-
-
. _ - . . _ ,u w - -,2_.--s-. , us . . ., n - ,,, a -
_ _ _ _ _ _ _ _ _ _ _ __ - __ _____ _ - __ - _ ____ __ _ _ _ _ - _ _ _ _ _ _ - - _ _ _ _ ____ _
l
l
!
.
!
_74_. _
l
. g ,
i Christensen. Engineering. Branch Chief.' Region-II (0ctober 30 through 31. _
!
1997) .
.S. Collins, i)irector. NRR (October 31, 1997) :
P.!Fillion. Reactor: Ins
-
!
November 17 through1997) 21.pector. Region II (October 27 through November 5
'
F. Hebdon. Directorate 11 3 NRR-(October 30 through 31. 1997); -
.. ,
J.- Jaudon. Director. Division of Reactor Safety. Region II.(October.30.through-
31. 1997) ]
,
J. Johnson Director.' Division of Reactor Projects. Region II (October 30; 1
through 31, 1997) !
1
T ' Johnson. Senior Resident Inspector, Turkey Point (November 3 through 7, "
1997)
e - C. Julian. Technical Assistant. Region II. (November 18 through 21' 1997)
.K. Landis. Branch Chief. Region II-(October 30-through 31, 1997)
E. Lea. Project Engineer.-Region II (November 17 through 21. 1997) -
[
LL..Mellen.' Reactor Engineer Region II-(November 23 through 25, 1997) l
-M.~ Miller. Reactor Inspector. Region II (October 27 through 31. November 17
. ;
through'21, 1997)
'
> '
W. Miller. Reactor Inspector. Region II (November 17 through 21. 1997) !
S. Ninh,' Project' Engineer Region II (October 30 through 31. 1997) i
L. Raghavan. Project Manager. NRR (October 30 through 31, 1997)
L. Reyes. Regional Administrator. Region II (October 30 through 31, 1997) !
R. Reyes. Resident Ins)ector. Region II (November 10 through 14. 1997) '
G. Salyers. Emergency )reparedness Specialist. Region II (November 10. 1997) !
'
R. Schin. Reactor Inspector. Region II (November 3 through 7. November 17
.through.21, 1997) -
-M. Thomas Reactor Inspector. Region II (November 3 through 7, 1997) .
G. Wiseman. Reactor Inspector. Region II (November 17 through 21. 1997)
.
INSPECTION PROCEDURES USED
IP 37550: Engineering
IP 37551: Onsite Engineering .
IP 40500: Effectiveness of Licensee Controls in Identifying. Resolving and 1
Preventing Problems :
>
IP 61726: Surveillance Observations
IP 62707- Conduct of Mainten3nce ,
i IP 71707_: Plant Operations 1
'
e :IP 92901: Followup - Operations
. IP 92903: Followup Engineering
IP -9290.4:_ - Followup - Plant Support-
-
c +
-
P
\
Enclosure 2 ]
-
!
.
. ,
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-, - - - . . - . .- .. . . - . . - ~ . . . . - - . . . - . _ - .
,
-
.
I
'
!
75 -l
.!
ITEMS' OPENED, CLOSED, AND DISCUSSED. j
.:
Opened !
.,
lyng Item Number Status Descriotion and Reference -
: VIO_ ~50-302/97-17-01 Open Failure to Conduct ~an Adequate
Unreviewed Safety Question i
Evaluation-for a Modification :
Functional Test. (Section E2.1) .
IFI 50-302/97-17-03 Open -Review of Cable Ampacity Issue.
(Section EB.17) .
;
- VIO 50-302/97-17-04 Open . Inadequate' Design Control. (Section
-
i
E8.20) :
- IFI 50-302/97-17-05 Open Resolution of Improved Technical :
Specification Setpoint Program i
Deficiencies Prior to Entry Into
Mode 4. (Section E8.22) .
l
-Closed
hag item Number Statm Descriotion and Reference i
VIO 50-302/97-02-01 Closed Failure to Follow Eouipment Control
Procedure Requirements. (Section
08.1)
.
LER 50-302/96-21-00 Closed Delayed Entry Into Technical
Specification Required Action Caused
by Inadequate Documentation of Out-
of Service Equipment Requirements ;
for a Modification. (Section 08.2)
'
VIO EA 97-094 01013. Closed Repeat Failure to Make Timely
01023. 01033. Reports to the NRC. (Section -
'
01043- 08.3)
VIO 150-302/97-08-01 Closed Inadecuate Corrective Action and i
'
Procecure .for External Reporting :
,
Requirements.-(Section 08.3)
LER 50-302/97-002-01 Closed Out of Calibration Fuel Pool Water
Level Transmitters. (Section M8.1)
,, ,
Enclosure 2 :
,
. " ! -
.
, ; ,
..
_ ',
? 'ii l!I. * ,
nn c.s . - __ .. _ ,~, _. ,__ _ _ , _._ __
76
V10 50-302/97-01-04 Closed Failure to Perform Tech" cal
Specification. Surveillance for Spent
-Fuel Level. (Section M8.1)
NCV 50 302/97-17-02 Closed Maintenance Performed on Safety
Related Components Withcut Approved
Procedures or Work Instructions.
(Section E2.1)
V10 50-302/96-08-01 Closed Failure to Take Timely Corrective
Action to Address Issues and Actions
For Makeup System Audit Findings and-
Excessive Vibration on a Spent Fuel
Pool Pump Fan Motor. (Section E8.1)
LER 50-302/96-011-00 Closed Personnel Error Causes Testing
Deficiency Resulting in Condition
Prohibited by Improved TS. (Section
E8.2)
LER 50-302/96-025-00 Closed Personnel Error Causes Testing
Deficiency Resulting in Condition
Prohibited by TS. (Section E8.2)
LER 50-302/97-003-00- Closed Personnel Errors Caused Testing
50-302/97-003-05 Deficiencies (GL 96 01). (Section
E8.2)
VIO 50-302/97-05-03 Closed Incorrect Information in Annunciator
Response Procedure for Inverters.
(Section E8.3)
VIO 50-302/97-07-01 Closed Failure to Follow Procedure CP-111
for the Processing of Precursor
Cards. (Section E8.4)
URI.50-302/96-201-07 Closed EDG not Protected Against Water
Spray from the Fire Protection
System Sprinkler. (Section E8.5)
V10 EA 95-126. 1.C.2 Closed Corrective Actions for an Inadecuate
(04013) Curve 8 (Two STI's and a Revisec
Curve 8A and 8B) were Also
incorrect. (Section E8.6)
,
Enclosure 2
.
.. . - _ - - .- . . . _ _
77 !
V10 EA 96-365. C Closed Inadequate Corrective Actions for :'
(03013) 10 CFR 50.59 Evaluation Errors and
for Inadequate Containment '
Penetration Surveillances. (Section
E8.7)
VIO EA 97-162 (01013) Closed inadequate Safety Evaluations for
Added 0)erator Actions for Design
Basis $3LOCA Mitigation. (Section
E8.8) :
- LER 50-302/96 24-01 Closed Plant Modification Causes Unanalyzed
Condition Regarding Emergency
Feedwater. (Section E8.10)
EA 96-365. 96-465 Closed Failure to Update Applicable Design
96 527. V10 B Documents to Incorporate Design '
Ex. 1 (02013) Information (Section E8.11)
EA 96 365. 96-465 Closed failure to Include Applicable Design
96 527. V10 B Information in the Design Input
Ex. 2 (02013) Requirements for a Modification
(Section E8.12)
.
LER 50-302/97-017-00 Closed Personnel Error Caused Inadequate
Electrical Separation Of High -
Pressure Flow Indicators. (Section 1
E8.14)
VIO EA 95-126 Closed Failure to Take Adequate ;
'
NOV 11.B Corrective Action for Required Tank
Volumes. Level, and Suction Points.
(Section E8.15) ;
VIO 50-302/97-01 09 Closed Inadequate Corrective Action for
Cable Ampacity. (Section E8.17)
LER 50-302/97-31-00 Closed Inadequate Cable Sizing Due to
Nonconservative De-rating Factors
Could Reduce the Cable Remaining
Qualified Life. (Section E8.17)
'
IFI 50 302/97-02-05 Closed Outstanding Issues Associated with
the Emergency Diesel Generator Povier
Upgrade Modification. (Section
E8.18)
i
Enclosure 2 i
:
s ..
w _ - . - - - y ,
. - -. .. - -. . . .. --
1
I
78
1
VIO 50-302/97-11-06 Closed Failure to Follow Licensee Procedure i
NEP-254. (Section E8.19)
IFI 50-302/95-15-05 Closed Relief Valves Removed From Heat !
'
Exchangers. (Section E8.20)
VIO 50-302/96-09-06 Closed Erroneous Calculation inputs and
Inservice Inspection Boundary.
(Section E8.21)
VIO EA 95-16 Closed Use of Nonconservative Trip
Setpoints for Safety-Related
Eouipment. (Section E8.22)
LER 60-302/94-006-00- Closed Deficiency in Understanding of ,
'
50-302/94-006 06 Technical Requirements Leads to
Nonconservative Safety Systems
Set'oint
) and Violations of Improved
Tec1nical Specifications. (Section
E8.22)
IFl 50-302/97-08-03 Closed Variations in the classification and
interpretation of the EAls by the
Emergency Coordinators. (Section
P8.1)
Discussed
Iype Item Number Status pescriotion and Reference
V10 EA 9/-330 (01013) Open unreviewed Safety Question involving
Added EDG Protective Trips.
(Section E8.9)
IFI 50-302/97-11-04 Open Corrective Actions for Approximately
4000 Precursor Cards Not Tracked to
Completion. (Section E8.7)
URI 50-302/95-02 02 Open Control Room Habitability Envelope .
Leakage, (Section E8.16)
Enclosure 2
4
.. ._ . - , _
,, _ .,, .
- - ._ - . -
,
t
i
!
79 i
'
-LER 50-302/9/ 038 00 Open Engineering Oversight Resulted in !
Operation Outside Design Basis of
Waste Disposal System. (Section ,
E8.21) :
LIST OF ACRONYMS USED
1
Al - Administrative Instruction '
AP - /bnormal Procedures *
API - American Petroleum Institute
AR - Annunciator Response '
BAST - Boric Acid Storage Tank
BL - (NRC) Bulletin ,
BS - Building Spray
BWST - Borated Water Storage Tank
CCHE - Control Complex Habitability Envelope
CFM - Cubic Feet per Minute
CFR - Code of Federal Regulations
Cr'T Core Flood Tank
CMIS - Configuration Management Information System
CP - Compliance Procedure
CREVS - Control Room Emergency Ventilation System
CR3 - Crystal River Unit 3 !
CT - Current Transformers
DBD - Design Basis Document
DH - Decay Heat
DHP - Decay Heat Pump
DHV - Decay Heat Valve
DNPO - Director Nucle - Plant Operations
ECCS - Emergency Core coling System
.
-EDBD -Enhanced Design Basis Document
EDG - Emergency Diesel Generator
EFIC - Emergency Feedwater Initiation and Control
EFW - Emergency feedwater !
E0P - Emergency Operating Procedure
EPRI .- Electric Power Research Institute
ES - Engineered Safeguards
ESOPM - Environmental and Seismic Qualification Program Manual
FCN - Field Change Notice
FLA - Full Load Amperes
FME - Foreign Material Exclusion
FMEA - Failure Modes and Effects Analysis
FPC- - Florida Power Corporation *
FSAR- - Final Safety Analysis Report
FSP - Fire Service Pump ,
FT- - Functional Test
GL' - Generic Letter '
HPI - High Pressure Injection -
HVAC -- Heating Ventilation and Air Conditioning
I&C - Instrumentation and Control
Enclosure 2
P
5
, . + - e -
---w , , . +. - , , w
-- .
.
--- - _ - . . -
80
~
IFl - Inspection Followup Item- '
IOC - - Interoffice Correspondence
IP - Inspection Procedure
IPAP - Integrated Performance Assessment Process
IR - Inspection Report -
ISA - Instrument Society of America
ISI -Inservice Inspection
ITS - Improved Technical Specifications t
JC0 - Justification for Continued Operation
kw . - Kilowatt
LC0 - Limiting Condition for Operation
LER - Licensee Event Report-
LOCA - Loss of Coolant Accident
LOOP - Loss of Offsite Power
LPI - Low Pressure Injection
LPMS - Loose. Parts Monitoring System
MAR - Modification Approval Record
MCAP - Management Corrective Action Plan
MCC - Motor Control Center
MP - Maintenance Procedure
HSLB - Main Steamline Break
MUT - Makeup Tank
MUV - Makeup Valve
NCV - Non-cited Violation
NEP - Nuclear Engineering Procedure
N00 - Nuclear Operations Directive
N0E - Nuclear Operations and Engineering
NOV - Notice of Violation
NPSH - Net Positive Suction Head
NP&SM - Nuclear Procurennt and Storage Manual
NOA Nuclear Quality Assessments
NRC - Nuclear Regulatory Commission
NRR - Office of Nuclear Reactor Regulation
NSM - Nuclear Shift Manager
OCR - Operability Concerns Resolution
01 Operating Instruction ;
OJT - On The Job Training
Or - Operating Procedure
05B - Operations-Study Book
OSTI - Operational Safety Team inspection
OTSG Once Through Steam Generator
PC - Precursor Card ,
PEERE - Plant Equipment Equivalency Replacement Evaluation
PM - Preventive Maintenance
PMRG - Plant Modification Review Group
PMT - Post-Maintenance Test
PORV .- Power Operated Relief Valve
! PR - Problem Report
PRC - Plant Review Committee
PSIG - Pounds Per Square Inch Gauge
Enclosure 2
.. -. . . _ _ . - . - ._. , .-
- _ - - - . - - - - - - ..- - - . . _ _ .
i
!
81
PT - Pressure Transmitters i
~
QPS - Quality Programs Surveillance
RCA - Radiologically Controlled Area
-
RCBT - Reactor Coolant Bleed Tanks
RCP - Reactor Coolant Pump
RCS - Reactor Coolant System
REA - Request for Engineering Assistance
RG - Regulatory Guide
R0 - Reactor Operator
RPS - Reactor Protection System
RRT - Rapid Response Team
RTD - Resistor Temperature Detector
RW - Raw Water
SA -Safety Assessment
SAR - Safety Analysis Report
SASS - Smart Analog Select System
SBLOCA - Small Break Loss of Coolant Accident
SBU - Strategic Business Unit '
SD - Site Drain System
SDBI - Suspected Design Basis Issue
SEC - Site Emergency Corrdinators
SEL - Security Event Log
SER Safety Evaluation Report
SFP - Spent Fuel Pump
SIR - Security Information Reports
SLUR - Setond Level Undervoltage Relays
SM - Shift Manager ,
SP - Surveillance Procedure
SPDS - Safety Parameter Display System
SR - Surveillance Requirement
SRO - Senior Reactor 0)erator
SRP - Standard Review )lan
SSOD - Shift Supervisor on Duty
STI - Short Term Instruction
TC - Temporary Change
TDBD - Topical Design Basis Document
IMAR - Tem)orary Modification Approval Record
15 - Tec1nical Specifications
UFSAR - Updated Final Safety Analysis Report
UL - Underwriters Laboratory
URI - Unresolved Item
US00 - Unreviewed Safety Question Determination
VIO - Violation
WCC - Work Control Center
WG - Water Gauge
WI - Work Instructions
WR - Work Request
.
'1
., . . _ - , _.
. . - . . . . . - . - - _ _ - . -.. .- - .- - - -- . - - _ . _ - . - -. - _ - . _ . -
,
i
!
!
Summary of EGDG-1B Test Runs
Type of Run Start Stop Parameter (s) of Interest
Date/ Time Date/ Time
Unloaded maintenance 11/9/97 11/9/97 Stopped due to fan drive ;
run per MP-499 2:18 am 2:19 am clutch slippage :
Unloaded maintenance 11/9/97 11/9/97 Stopped due to fan drive
run per MP-499 3:00 am 3:07 am clutch slippage
Unloaded maintenance 11/9/97 11/9/97 Stopped due to fan drive
run per MP-499 3:22 am 3:36 am clutch slippage
Unloaded maintenance 11/9/97 11/9/97 Stopped due to jacket
run per MP-499 3:45 am 3:55 am coolant leak
Unloaded maintenance 11/10/97 11/10/97 Overspeed trip testing
run per MP-499 5:35 am 6:06 am
Unloaded maintenance 11/10/97 11/10/97 Secured due to high
run per MP-499 5:04 pm 5:43 pm vibrations
Unloaded run per MP- 11/12/97 11/12/97 Run with fan clutch
531 to troubleshoot 4:15 am 4:33'am disengaged. Secured when
vibration problems high temperature alarms
received.
Slow start. 11/12/97 11/12/97 Breaker 3210 closed from
Loaded 2625 - 2825 kw 6:04 pm 10:24 pm 6:31 pm until 10:20 pm.
for approximately 2 -
3 hours
Slow start. 11/13/97 11/13/97 Breaker 3210 closed from
loaded 2625 - 2825 kw 6:39 am 11:09 am 7:08 am until 11:04 am.
for approximately 2 - Second slow start with 2 - 3
3 hours hour loaded run, following
adjustment of fan blade
pitch.
Slow start 11/15/97 11/15/97 Breaker 3210 closed from
Loaded 2625 - 2825 kw 1:42 pm 2:28 pm 2:10 pm until 2:28 pm.
for 24 hours Breaker opened and diesel
secured due to fan drive
clutch slippage. Loaded only
to 1680 kw.
Slow start 11/15/97 11/16/97 Breaker 3210 closed from
Loaded 2625 - 2825 kw 5:33 pm 10:16 pm 5:56 pm on 11/15/97 until
for 24 hours 10:12 pm on 11/16/97.
l Fast start 11/17/97 11/17/97 Breaker 3210 closed from
Loaded 2625 - 2825 kw 4:35 am 11:00 am 5:05 am until 10:55 am,
for 4 hours
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Type of Run Start Stop Parameter (s) of Interest
Date/ Time Date/ Time
Slow start 11/18/97 11/19/97 Breaker 3210 closed from !
Loaded 2625 - 2825 kw 3:17 am 8:22 pm 3:37 am on 11/18/97 until ,
for 24 hours 8:19 m on 11/19/97. :
Multi le attempts to raise !
-
load bove 2825 kw resulted
in high vibration conditions ,
'
. in the pedestal bearing.
Slow start 11/20/97 11/21/97 Breaker 3210 c1osed from l
Loaded 3100 - 3138 kw 1:41 am 5:23 am 2:01 am on 11/20/97 until .
for 14 hours 5:21 am on 11/21/97. !
Slow start 11/22/97 11/23/97 Breaker 3210 closed from !
Loaded 3300 - 3375 kw 10:38 pm 10:37 pm 11:01 pm on 11/22/97 until :
20 hours at 3300 - 10:34 pm on 11/23/97. '
-
3375 kw
2 hours at 3325 - -:
3375 kw
Hot condition fast 11/23/97 11/24/97 Breaker 3210 closed from
start 10:42 pm 12:55 am 10:51 pm on 11/23/97 until
Loaded 2625 - 2825 kw 12:48 am on 11/24/97. :
for 1 hour
Fast start 11/24/97 11/24/97 Breaker 3210 closed from
Loaded 2625 - 2825 kw 4:51 am 6:40 am 5:01 am until 6:36 am. ;
for 1 hour-
Maintenance run 11/27/97 11/27/97 Run to support performance ,
unloaded 11:46 am 12:08 pm of PM-123. Periodic r
Elett: 'al Checks of
Emergency Diesel Generators
SP-354B Operability 11/29/97 11/29/97- Breaker 3210 closed from
Run 10:25 am 3:49 pm 10:52 am until 3:43 pm
>2625 kw EGDG 1B declared Operable
5:21 am on 11/30/97.
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