ML20055E399
| ML20055E399 | |
| Person / Time | |
|---|---|
| Site: | San Onofre, Rancho Seco  |
| Issue date: | 05/28/1987 |
| From: | Kirsch D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20055C206 | List:
|
| References | |
| NUDOCS 9007110320 | |
| Download: ML20055E399 (14) | |
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MAY 281987 h.cflp i
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MEMORANDUM FOR: Dennis M. Crutchfield Director J 'F E[%
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DivisionofReactorProjectsIII/IV/M@W@p' S
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FROM:i D. F. Kirsch. Director P
a Division of Reactor Safety and Projects i!
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SUBJECT:
. INPUT FOR SAFETY EVALUATION REPORT j
RANCHO SEC0 - DOCKET NO. 50-312-i!
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-II Region V is continuing to' inspect the Sacramento Municipal Utility District's (SMUD) activities relating. to the preparation of Rancho Seco for restart l:
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-following the December 26, 1985 transient.
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During February, March and April of 1986, a special Region V' inspection team.
performed an initial = inspection of the SMUD activities and documented the j
L results :in-Inspection Report Nos. 50-312/86-05, 86-06. - and ~ 86-07.a Several follow-up' inspections to these have been perfonned and were documented. in -
l' Inspection Reports 50-312/86-14, 86-15, 86-16, 86-20, 86-27. 86-37, 86-42.
87-02, 87-05, 87-06, 87-07, 87-08, and 87-11. Copies of applicable po tions, of these reports are enclosed.-
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-Based on these, inspections,' Region V has concluded that the licensee has;
' partially performedideveloped, and implemented the necessary corrective
-action programs to restart Rancho Seco.
Region V staff evaluation of those ll activities;is provided in the attached SER input information.
However, additional / licensee action is still necessary tof address the t
remaining issues associated with the December 26,:1985' transient and.to L
- preclude recurrence-of similar events.
These activities will be further r
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inspected by Region V as stated in the attached SER input. A supplemental SER input will be provided by Region V prio_r to Rancho Seco restart.
Original signed b'9
D F. Kirsch D. F. Kirsch, Director Division of Reactor Safety and Projects o
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9007110320 90062
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Enclosures J As/ stated
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EII/0 ;Z:;t 0;;;T;l Oeek (EiD3) (IE01h ProjectInspector j
-Resident Inspector j
- G. Cook.
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.B. Faulkenberry:
l EJ. Martin d,
, Docket File ' (w/ene) y M.l Smith
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- . G. ~ Knighton. NRR (w/ene)
- =G. Kalman. NRRL (w/ enc) i ic-y.
Region V-
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, REGION V INPUT FOR RANCHO SECO NUCLEAR' GENERATING STATION SAFETY EVALUATION
,e
' Note:
The Region _V SER input for each item is annotated with an asterisk
(*)
p Land a-line in the margin.
1..
Introduction 2~
Backg'round Discussion
'2.1' Brief Discussion of December 26, 1985 Overcooling Event l
2.2 ~ Sumary. of-NRC Actions and Correspondence.
2.31 - Summary, of Sacramento Municipal Utility District's Response -
)
'2.3.1 Plant Performance and Management Improvement Program a.
.ProgramOverview(PBDf6) b.
Program Evaluation (Region V)
The licensee's Plant Performance and Management.
s Improvement Program is still being evaluated by Region V.
The. Region V evaluation will be'.provided as a supplement.
3.
Resolution:of Identified Areas of Concern Related to the December 26, 1985 Overcooling. Event 3.1 Plant Electrical, Instrumentation and Control Systems Issues 3.1.1
-Integrated Control System-(ICS) and Non-Nuclear Instrumentation (NNI) - General Description and Operational Experience a.
ICS and NNI System Descriptions and' History of Loss of ICS or NNI Power Events
.b.
ICS and NNI Power Distribution Systems.
c.
Power Monitor Designt and Operation u
-d.
Identification of Systems / Components Controlled by ICS and NNI 3.1.2 Loss of ICS or NNI DC Power a.
Root Cause for December 26, 1985 Loss of ICS DC Power-andCorrectiveActions(RegionVtoverify)-
The licensee's determination of root cause for the i
December 26, 1985 Loss of ICS and.00 power was 1
inspected by Region V.
The results of the inspection were documented as item Region V-E-5 in Inspection-Report 50-312/86-07 (Attachment A-1). The licensee's trouble shooting, subsequent to the December 26, 1985 L
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j event.f resulted in the discovery of a loosely crimped
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terminal lug-for the' irs power monitor which caused -
intermittent-high resistance, which in turn caused' the ICS-power supply to trip. The licensee
,f corrective actions for the root cause of the event 1
were inspected by Region V.
The results of the o
inspection were documented as items Region V-MA-3 and Region-V-MA-5 in: Inspection Report 50-312/86-07 (Attachment.A-1) and' Item Region-V_-MA-2 in' Inspection 1
c Reports 50-312/86-07c(AttachmentA-1)and 50-312/87-08-(Attachment-A-2). The licensee replaced the terminal lug _ identified as the root cause for the 1
ICS power-supply trip.
In addition, the'11censee:..
l performed an' inspection for similar conditions ~on all-critical: terminations.in the control room. The 4
licensee inspection did nott identify a programmatic
'f
. problem with Amp connectors or crimping.
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Accordingly, the inspection was not expanded beyond the control room.
Region V considers that the c
licensee's actions have satisfactorily resolved this-issue.
l b.
System / Component Response to Loss of.ICS or NNI DC Power (PEICS) c.
ICS and NNI Backup' Instrumentation and Controls' (PEICS)
-d.
Indication / Annunciation,of Loss of ICS or NNI Power (PEICS)
L e.
Interactions Between ICS or NNI and Safety Related 1-Systems (PEICS)'
0 f.
ICS and:NNI failure Modes and Effects: Analyses-(PEICS)-
g.
iProposed Modifications Prior to Restart (PEICS) h.-
ICS and NNI. Maintenance, Surveillance and. Testing a
'(PEICS)
L 1.
OperatorResponse/ Procedures (PEICS) j.
Root Cause of Discrepancy Between 0TSG Level Strip.
4 Charts and-SPDS (Region V)
The licensee'sL detennination of the root cause for the discrepancy between OTSG level strip charts-and-t SPDS has been inspected by Region V!and.theLresults f
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of the inspections documented as item Region _V-E-13 J
in inspection reports 50-312/86-07(AttachmentA-1) and 50-312/87-08 (Attachment A-2). The licensee-review of this issue identified an incorrect 1
algorithm for temperature compensation in the SPDS i
L manual. The licensee'is still in'the process of establishing the reliability of information contained I
fi in the SPDS manual. The root cause of the J
discrepancy has therefore yet to be clearly established by the licensee. Licensee action will be further inspected and the results of the inspection h
will be provided as a supplement prior to restart.
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- 3.1.3 Restoration of ICS or NNI DC Power a._
System / Component Response to Restoration of ICS ore NNI' Power
/n 3.1.4 Additional ICS and.NNI Issues
-t a.
System / Component Response to Loss and Restoration of ICS or NN! AC. Power b..
ICS and NNI Response to Loss of: Instrument Air and Effect on Plant Operation
,1 Ec.
SMUD Response to ICS or NNI Concerns Identified as
- Part of the B&WOG Reassessment
=
d.
Loss of Offsite Power Coincident with Loss of ICS or NNI Power or Loss of Instrument Air 3.1.5 Availability of Remote and Local Indicatione on Loss of ICS and NNI Power y
-a.
R.Gi 1.97 Instrumentetion b.-
Safety Parameter Display System (SPDS) c.
Interi:n Data Acquisition and Display System (IDADS)
- y d.
' Computer / Annunciator? Systems e.
Other Indications 3.1.6 Emergency Feedwater Initiation and Control (EFIC):
a.
EFIC System Design and Operation-b.
Independence of EFIC from the ICS and NNI c.
AFWS/EFIC Failure Analysis d.
0.TSG Overfill Protection
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3.1.7 Main Feedwater' System (MFWS) a.
MFWS' Response to ICS or'NNIiFailures and'Effect on
+.
Plant Operation b.
OTSG Overfill Protection Circuits "3.1.8 Main Steam System oy a.
Atmospheric Dump Valve Operation and Response to-ICS or NNI Failures-L b.
Turbine Bypass Valve Operation and Response to ICS'or NNI Failures c.
OTSG Isolation Capability (Main Steam Line Failure-L _.,;
Logic) 3.1.9 Sumary of Plant Modificat%ns Designed to Prevent /Hitigate Transients aesulting from Loss of ICS or NNI 3.2 Plant Mechanical System Issues 3.2.1 Water Supply to Makeup /HPI Pumps
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. Root Cause of Makeup /HPI Pu'ap'FailureL(RV)l
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- The licensee's determination of.the root cause of the makeup /HPI pump failure was. inspected by Region V.
7
.The results of-the inspection were documented as item Region V-E-15 in Inspection Report 50-312/86 j (AttachmentA-1). The pump failure was attributed to w
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the absence of adequate procedures for recovery from k Q, Safety Features Actuation System (SFAS) initiation.
'4' In addition, lack of operator understanding led to i
personnel error regarding pump: operation. The licensee has performed a. satisfactory root cause j>
analysis for the make-upipump failure, and has-n
~'
completed training and procedure changes to prevent.
recurrence of this so%1fic item. Region V considers this item satisfactorily resolved.
b.
Assurance of Water Supply Sources (RSB) c.
Makeup Pump Repair (RV) y j,
q This item was originally established to obtain a -
commitment for repair or replacement =of this pump in i
an expeditious manner following restart of. Rancho Seco. The subsequent decision to have the: pump in' service prior-to restart of the plant made the issue moot. The makeup pump has been repaired. This was identified.as item Regien V-E-17 in Inspection Report -
50-312/86-07-(AttachmentA-1).-RegionVconsiders a
this' item satisfactorily resolved.
9 3.2.2 Effects of Overcooling Event on Reactor Vessel and Steam:
Generatorse (y
a.
Analysis of-Transient-on. Vessel and Steam Generators-b.
Technical Basis for PTS Guidelines c..
Potential for Core' Lift 3.2.3 Operation 'of Radiation M'onitoNng Systems Following-Containment Isolation 4,
a.
Root Cause of Radiation Monitor System Damage T
V b.
Effects of Containment Isolation on Systems Required-4 to Operate following ESFAS-Actuation x
M 3.2.4 Steam Generator Overfill and Flooding of the Main Steam d.,' '
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a.
Evaluation of Steam Header Supports (EB)
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Steam Line Support Inspection (RV)-
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The licensee's stress evaluation of the Main Steam
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line was reviewed by Region V.
In addition, the results of the licensee's walkdown of the piping and 7
supports for the main steam line and the main steam j!
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J bypass line to the condenser for possible water i
hammer damage were reviewed by Region V; The Region V review was documented as item Region V-E-8 in ~
t Inspection Report 50-312/86-07(AttachmentA-1). The licensee did not identify any piping damage and the main steam lines were determined to be acceptable:for r
operation.- Region-V concluded that the licensee has e
performed an adequate inspection.and analysis of the l
Main Steam Header and pipe supports, and this issue 4
is resolved satisfactorily.
3.3 Plant Maintenance 3.3.1 Maintenance Program Evacuation y
3.3.2-Valve Preventive Maintenance Program The licensee's Valve Preventive Maintenance Program has
'[
been: inspected by Region V.'
The results of the
. inspections have been identified as item RV-MA-1.in.
Inspection-Reports 50-312/86-07(AttachmentA-1).and 1
50-312/87-08'(AttachmentA-2). The licensee has~not yet developed and. implemented an acceptable. valve preventive maintenance program. Additional licensee' action will be.
further inspected by Region V and an evaluation will be provided by Region V as an SER supplement prior to
- restart, 3.3.3 Operabi. ity Program for Manual, and Remote Operated Valves j
The licensee's operability program for canual and remote
. operated valves has been' inspected by Region V.
The results of the inspections have been identified as item '
RV-MA-4 in inspection reports 50-312/86-07(Attachment-A-1) and 50-312/87-11 (Attachment-A-3). The licensee has not yet implemented an: acceptable operabil.ity program for-manual and remote operated valves. - Region V's concerns-regarding'the adequacy of the licensee's program and baseline inspections are discussed in. Inspection Report 50-312/87-11.
Furthter licensee action will be inspected-by Region V and the results will be provided as'a supplement prior to restart.
3..'. 4 Previous Maintenance Deficiencies -
Identification / Resolution 3.3.E Troubleshooting / Root Cause Determination Program s
3.4 Train ng and Operator Performance 3.4.1 Adequacy of Operator Training a.
ProgramOverview(MTB) e
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ICS'0ff Normal Operation R) c.
Makeup System Operation V
d,
_ AFWThrottling(RegionV j
e.'
AFW Pump Trip Criteria ( V) i f.
Consnunications h.
ADV,TBVOperation(RV).
1.
Differences Between B&W Simulator and Rancho Seco Plant (RV) j.
Operation of Manual Valves (Incliiding Hand Jacks):
(RV) k.
Plant Modifications, Procedural Changes and Additions 1
e (RV) 1.
OperatorRetrainingDuetoLongTermShutdown-(RV)-
The 11censee's operator retraining for this item has-
+
s' been inspected by Region V.
The results of the inspections have been identified-as. items RV-0-9.
RV-0-10, RV-0-12 and RV-0-14 in Inspection Reports:
1 50-312/86-07 (Attachment A-1) and 50-312/87-06 (AttachmentA-4). Region V concluded that retraining' in the above noted areas, adequately addressed each 1
t~aining weakness. identified above as a result ofcthe
- 10f2/85 and 12/26/85 events.
J s
In addition, NRC Inspection Report 50-312/86-06'-
-i (Attachment A-5) identified numerous instances where the actions required by the,11censee's Emergency Plan implementingprocedures(EPIPs)werenotperformed.
Training was considered to be~the. primary cause of.
. [
these problems.. The licensee's corrective action for 7
these ' problems consisted of specific training-for the operating crews.in the principles of command and i
control, specific actions. required for an unusual-event,;and required emergency notification. This D'
supplemental' training was completed on. September 4, 1986. -The licensee's corrective actions were
~_
evaluated during the 1986 annual exercise and considered adequate. Prior to start-up, Region V J
will reevaluate'the licensee's Emergency Response.
Training Program, including emergency response
. training for the control room staff. This
.infonnation till be added as a. supplement to this z
SER.
Finally, NRC Inspection Report 50-312/86-06 identified deficiencies in the licensee's L
implementation of their emergency procedure for L
notification. The failure to update State and local agencies and to notify plant personnel of degraded plant conditions was identified as a violation.
Training was also considered the' primary cause of this problem.
The licensee's corrective action was to improve the training conducted for emergency p '
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notificationprocedure(AP-506)andtoprovide H
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supplemental training to further emphasize the O
importance of keeping State and local agencies-y>'
apprised of_ plant conditions. The supplemental' training was completed on April 1,11986.- Based upon 1
E the licensee's performance during the 1986 annual exercise, the licensee's corrective action in this area appears satisfactory.
3.4.2 Minimum Staffing Requirements 3.4.3 Incapacitated Operator The licensee's review of the fitness for' duty of a Senior
_0perator who collapsed during the December 26, 1985r event-was inspected by Region Y.
The results of the inspection J
was identified as item RV-0-11 and closed in Inspection
(
- Report 50-312/86-07(AttachmentA-1). The @gion V inspection concluded that the collapse wL mt related to
- any' underlying condition that would prohibit the y
individual from control room work. Region V. considers-this' issue to be satisfactorily resolved.
3.4.4 Potential. Security / Safety Interface Issues-
"' 6 NUREG-1195, the Incident Investigation Team. Report of the
_i December 26,:1985 Event at Rancho Seco, addressed two ls security / safety interface issues which have-since been
' corrected by the licensee, and verified in Inspection i
Report 50-312/87-071(AttachmentA-6):
a.
Controlled Area Fence The (non continuous)~
a chain-link fence-segment located. inside the Tank' Farm (avital! area) marked," Caution:-
a controlled Area Boundary, No Admittance",
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inhibited the movement of operations personnel-i During this unusual event, one nonlicensed operator found it expeditious to climb over this non-securit A AFW (ICS)y fence as he moved from theB to the This fence, originally installed l
as a radiation control barrier, has since been removed as it was no longer necessary. Region V m
considers this issue to be satisfactorily resolved.
P b.
Lost Security Badge After assisting in a
isolating the makeup pump, a nonlicensed operator noted he had lost his security badge and as such, was no longer able to operate (vital area) doors that required a badge for 3
J entrance. He was escorted to the control room and waited approximately 20 minutes to be issued a visitor badge. The licensee has since issued
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8 special "Z" (security) badges to the shift supervisor for his immediately issuance, as needed, to authorized operations personnel.
These (non-pnoto) Z-badges allow immediate access to all plant vital areas.
Region Y considers that this issue is satisfactorily resolved.
3.5 Plant Normal and Emergency Procedures 3.5.1 Need for Event-Related Procedures 3.5.2 Adequacy of AT0G Procedures 3.5.3 Adequacy of Health Physics Procedures k
The adequacy of health physics procedures has been
(
reviewed during seven NRC Region V inspections since the Ri=
December 26, 1985 event, inspection Report Nos.
50-312/86-05, 06, 16, 20, 27, 37 and 87-05. These inspections have found that procedures are adequate to support restart.
Region V considers that this issue is satisfactorily resolved.
3.5.4 Adequacy of Annunciator Procedures Manual The adequacy of the annunciator response procedures was
=
inspected by Region V.
The results of the inspection are documented as Item RV-0-15 in Inspection Report 50-312/86-07(AttachmentA-1)andclosedinInspection Report 50-312/87-08 (Attachment A-2). At the time of the inspection documented in Inspection Report 50-312/87-08 the licensee had completed approximately 45% of the 272 procedures that were being rewritten. Region V considers this issue is satisfactorily resolved on the basis of the adequacy of the procedures reviewed during that inspection, and licensee's commitment to complete all procedure upgrading prior to restart.
3.5.5 Methodology for Procedure Updating The licensee's program for assuring that procedure changes are made and training is completed when plant modifications are made was inspected by Region V.
The inspection findings are documented as part of item RV-0-12 in Inspection Report 50-312/86-07 (Attachment A-1).
The inspection verified that an acceptable program existed but it apparently had not been adequately implemented in that training was not effective. The inspection further concluded that the failure to adequately train may have contributed to either the 10/2/85 or 12/26/85 events.
Region V considers this issue to be satisfactorily resolved.
In addition, the licensee's operator retraining for specific plant modifications is addressed in paragraph i-iein---mmi eiimumm-isemm-mmm-mmi
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3.4.1.k. and is also considered to be satisfactorily w
resolved.
S 3.5.6 Adequacy of Emergency Procedures The current edition of the licensee's Emergency Operating Procedures was inspected.by Region V.
The inspection-findings were documented as item RV-0-4 in Inspection Report 50-312/86-07(AttachmentA-1)andinInspection Report 50-312/87-08 (Attachment'A-2).- In addition, a Headquarters Augmented System Review and Test Program Inspection Team reviewed the licensee's Emergency Operating Procedures.
These inspections have found that-many selected procedures had.yet to be rewritten by the
-licensee.
In view of their complexity and importance, further NRC review of the Emergency Operating Procedures is necessary and an evaluation of the finalized emergency procedures will.be provided as a supplement prior.to restart.
3.6 Human Engineering Considerations 3.6.1.
Simplified Schematics for Switches S1 and S2 l-p 3.6.2 Valve Position Indication 3.6.3 Control Room HVAC Noise s
3.6.4.
. Alarm forLICS
'3.6.5-Control Room Modifications
-3.7-. System Review and Test Program t
3.7.1 Program Overview 3.7.2 Program Evaluation 3.7.3 Review of System Testing The licensee's system: testing is currently being inspecte'd L
by Region V.
However, the licensee has not completed a l
.significant fraction of the system tests and a mr.aningful evaluation of'the licensee's systems testing cannot be performed at this time..Further Region V inspection and evaluation of the licensee's systems testing will be l
performed and provided as a supplement prior to restart.
3.8 Management and Organizational Considerations (F0B has lead, Region Y i:
will provide input to F08)
The. licensee's management organization is.being evaluated by Region V, in addition to the lead review by NRR. The licensee has selected permanent SMUD employees for most key positions. The Chief i
10'
.N Executive Officer, Nuclear was selected and assumed the top on-site J
management position on May 4', 1987. Other key management positions such as the Assistant General Manager for Nuclear Operations, the Quality. Assurance Manager, and the Licensing Manager have either just been recently selected'or are still to be selected.
Further Region V evaluation will be performed when these positions are filled'and organizational relationships have stabilized and will be
.i provided as a supplement prior to rectart.
i 3.9 Retrospective considerations 3.9.1 Evaluation of FSAR Accident Analyses that Presumed Availability of Non-Safety Systems i
3.9.2 Reevaluation of Responses to Previous Reports on B&W-j Transients and Operating Experience NUREG-0560,
</
NUREG-0667, BAW-1564,Bulletin 79-27 3.9.3 Probability of Pressurized Thermal Shock (PTS) Events (BAW-1791)
- j i
3.9.4 EFIC System History
.NUREG-0737,'ItemII.E.1.2[ustification r
a.
4.
Resolution of concerns Unrelated to the December 26, 1985 Overcooling Event i
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4.1 Post Accident Sampling System (PASS) a.
System Modifications b..
Procedures and Training.
c' Testing The licensee has not yet confirmed that they have a PASS-system
.i that will fulfill the Technical Specification requirements.
NRC Inspection Report No.~50-312/87-05 provides a detailed j
review of system status (Attachment.A-7). Since that j
inspection. the licensee has started testing the containment i-4 air sample: portion of the system and plans to continue with other test to demonstrate system performance capabilities.-
1 Region V will review these test,,and provide a. supplemental L
evaluation prior to restart.
1/,
4 j
4.2 Control Room / Technical Support Center HVAC System l
a.
Adequacy of Design and Installation-
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b.
Modifications The licensee has not completed testing of its modified HVAC system for the Control Room and Technical Support Center.
Region V will inspect the licensee's efforts and intends to perform. independent measurements of the modified Control h
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b.
g3-Room / Technical L pport Center HVAC system. Region Y will provide the results of the inspections and test in a l-supplemental evaluation, p-4.3 Radioactive Liquid Effluent Releases a.
Offsite Contamination b.
Technical Specification Deficiencies l_
c.
Long Term Resolution Inspection Report No. 50312/86-15(AttachmentA-8) documented L
significant deficiencies in the licensee's management of radioactive liquid effluents. The NRC Office of Investigations has an open investigation of this matter. The licensee met with NRC representatives an March 3, and May 5,1987, to l
H discuss their planned actions. NRC expects the licensee will submit a request to revise their radiological effluent portion of their Technical Specifications prior to restart. Following l
submission of the Technical Specification amendment Region V L
will inspect to confirm that necessary actions are being taken and Region Y will provide a supplemental evaluation prior to restart.
L l
4.4 Emergency Plan a.
Meteorology I
A safety evaluation of the Onsite Meteorology Program was i
performed by NRR staff.
The safety evaluation report'(TAC i
No.55153)wascompletedandtransmittedtotheProjectManager i
on March 20, 1987. The safety evaluation report concluded that, based on exisH ng and proposed modifications to the mettorological are N m and t1e completion of the planned imptovements, t.ie Rancho Seco Facility will satisfy the minimum l
meteorological requirements of 10 CFR 50 and the guidelines of i
Regulatory Guide 1.23 and 1.97, Revision 2, at the time of L
re-s ta rt. - Region V considers these proposed changes to.the program to be a satisfactory resolution of this issue; however, a confirmatory follow-up inspection will be required. Region V will provide a supplemental evaluation prior to restart.
b.
Training Inspection Report 50-312/86-14(AttachmentA9) identified numerous violations in the licensee's emergency response training program.
The licensee's corrective action for those violations consisted of numerous initial and long term actions.
s NRC Inspection Report 50-312/87402(AttachmentA-10), conducted in March of 1987, evaluated the licensee's initial corrective actions and determined them to be adequate. This report also identified numerous areas that still require improvement.
Prior to start-up, a reinspection of the emergen:y response training program m 1 be performed.
The statut of the long term corrective actions and the improvement items will also be j
1j
. ~
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12 examined. This information will be added as a supplement to this SER, c.
Emergency Plan implementing procedures and Dose Assessment Presently the licensee is in the process of the revising the Emergency Plan and is also making major revisions to the EPIPs.
The licensee's procedures for dose assessment, training, classification, and drills and exercises are in revision at this time.. Prior to start-up, the licensee's Emergency Plan and implementing procedures will be evaluated and this information will be provided as a supplement to this SER.
4.5 Regulatory Guide 1.97 a.
Implementation 4.6 Safety Parameter Display System (SPDS) a.
Upgrade to Safety Grade b.
Isolation Devices c.
Modifications (Format) 4.7 TDI Diesel Generator Qualification a.
TDI Diesel Generator Qualification b.
Technical Specifications c.
Evaluation of Associated Class 1E Electrical System Diesel Generator Building Design, and Fire Protection System 5.-
Summary and Conclusions Appendices EDO Action L.etter Regarding IIT Report References i
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