ML20056C967
| ML20056C967 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 07/19/1993 |
| From: | Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20056C965 | List: |
| References | |
| 50-458-93-17, NUDOCS 9307300166 | |
| Download: ML20056C967 (12) | |
See also: IR 05000458/1993017
Text
I
..
-
!
l
APPENDIX B
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Inspection Report:
50-458/93-17
Operating License: NPF-47
Licensee: Gulf States Utilities
P.O. Box 220
St. Francisville, Louisiana
Facility Name: River Bend Station (RBS)
Inspection At: RBS, St. Francisville, Louisiana
Inspection Conducted: June 14-28, 1993
_,
inspectors:
W. M. McNeill, Reactor Inspector, Engineering Section
Division of Reactor Safety
Accompanying Personnel:
E. T. Baker, Senior Project Manager, Nuclear Reactor
Regulation
Approved:
)+
.
7-/9-f_?
Thomas F. Westerman, Criief, Enginehing Section
Date
Division of Reactor Safety
Inspection Summary
'
Areas Inspected:
Routine, announced inspection of the 10 CFR 50.59 safety
evaluation program.
.
Results:
I
There was a good definition of scope of the documents to be used as the
.
basis to determine if a change effected the licensing bases and,
-
therefore, required a safety evaluation (Section 2.1).
The consolidation of the safety evaluation process into a single
.
procedure has been slow in development, in that, Procedure RBNP-57 has
been in a draft status since August 1991 (Section 2.1).
Comprehensive training has been recently implemented (September 1992).
l
Based on a limited observation of recent evaluations, the inspectors
noted an improvement in both the level of detail and quality of
'
evaluations (Section 2.2).
,
1
9307300166 930723
{
-
ADOCK0500g8
'
.
.
l
- a
i
-2-
!
It was difficult to locate all the evaluations and screenings because of
the use of different data bases and forms.
Self-assessment activities
by the licensee have identified this as a problem and the licensee had
,
taken steps to assure that all evaluations and screenings are accounted
for and reviewed (Section 2.2).
,
'
The effort to prepare and review safety evaluations and screenings was
good, however, one exception was noted. A violation was identified
i
concerning the failure to comply with Technical Specifications while
establishing the minimum critical power ratio.
In addition, one
deficient evaluation and one deficient screening were identified. The
impact of these deficiencies on plant safety were not significant
,
(Section 2.3).
Summar_y of Inspection Findings:
Violation 458/9317-01 was opened (Section 2.1)
Attachments:
Attachment 1 - Persons Contacted and Exit Meeting
Attachment 2 - Documents Reviewed
,
h
9
i
h
.)
..
,
,
f
-3-
DETAILS
1 PLANT STATUS
During this inspection period the plant was shutdown and in a forced outage.
2 SAFETY EVALUATION PROGRAM, 10 CFR 50.59 (37001)
1
2.1 Program
The inspectors found River Bend Station had established a program for
reviewing changes, tests, and experiments to the facility as described by the
safety analysis report. The program for safety evaluations included
consideration of other documents such as an " Environmental Protection Plan,"
Appendix B to the facility operating license. The program referenced and
followed the guidance of the Nuclear Safety Analysis Center 125, " Guideline
for 10 CFR 50.59 Safety Evaluations," which has not yet been endorsed by the
NRC staff.
The inspectors found the process for evaluations was a three step process.
The first step was to determine if the change, experiment, or test required a
full safety evaluation.
If a hardware type change was made then it was
documented on an initial safety and environmental form.
If a software type
change was made then it was documented on a safety and environmental
evaluation form.
For those changes, experiments, or tests that were not resolved at the first
step, the second step, an evaluation as an unreviewed safety question
.
determination was performed and documented on an unreviewed safety question
determination form or an unreviewed environmental question determination form.
The third step was to determine if NRC approv'. of the change, experiment, or
test was to be obtained.
The third step was called an unreviewed safety
question or unreviewed environmental question.
Currently, design engineering follows Procedure ENG-3-004 for preparation and
review of hardware type changes. The training requirements for both preparers
and reviewers were contained in Procedure EDP-AA-62.
Procedure EDP-AA-62 was
referenced in Eng-3-004. The activities of plant personnel .(system
engineering) in making software changes, such as procedure changes, was
described in Procedure ADM-0003, including the training requirements for the
reviewers only. This procedure could also be used by design engineering
i
personnel.
Procedures SSP-1-001, QAP 1.1, and QAP 1.0 required procedures
such as training, security, licensing, quality, etc., to be evaluated in terms
of 10 CFR 50.59 when they were changed. The requirements for the evaluation
of condition reports were described in Procedure RBNP-030. There was under
preparation, since August 1991, a Procedure RBNP-057 that consolidated the
safety evaluation process.
Interest and support was observed to be slow in
developing.
!
l
i
_
.
..
,
i
-4-
t
i
The inspectors found by reviewing the procedures and training information
'
that the licensing basis was used for evaluation of changes. The licensing
!
basis was defined as the safety analysis report including the quality
assurance manual, security plan and safeguards plan, environmental report,
safety evaluation report, Technical Specifications, operating license, final
environment statement, emergency plan, radiological environment effluent
monito-ing manual, offsite dose calculation manual, process control program
,
and con-!tments in NRC correspondence. This was considered by the inspectors
to be a good definition of the scope of documents to be used as the basis to
determine a change.
It was also noted that there were provisions in
procedures and training for discretionary margins.
Discretionary margins are
that which can be changed without prior NRC approval.
2.2 Implementation
The inspectors reviewed the " Gulf States Utilities Company Annual Report
Pursuant to 10 CFR 50.59 (b)(2) March 1, 1991 through March 1, 1992." During
that period, the licensee had made 28 changes, 12 of which were hardware-type
changes (design changes).
Fifteen changes were software type changes
(procedure changes) and one was a test.
The inspectors requested a list of modifications completed since March 1992
and found that 90 modifications had an evaluation and 104 had a screening.
In the same period, the inspectors found that 25 prompt modifications
(temporary modifications) were processed,11 of which had evaluations and 14
,
'
had screenings; and 424 condition reports (nonconforming condition reports)
had been issued, 8 of which had evaluations and 416 had screenings.
Comprehensive training had been a recent effort since September 1992.
Observation of some recent evaluations did note an improvement in both the
level of detail and the quality of evaluations.
'
It was difficult to locate all the evaluations and screenings because of the
use of different data bases and forms. The licensee's self-assessment
activities had identified this as a problem. The independent safety
engineering group had taken steps to assure that all screenings and
i
evaluations were reviewed as required by the offsite review committee.
.
!
The inspector noted there was an unreviewed safety question established
,
recently at River Bend Station.
This unreviewed safety question had been the
subject of a violation in NRC Inspection Report 50-458/9311. This was in
'
regard to an inoperable interlock on the containment ~ personnel hatch. There
,
will be no further discussion in this report of this subject.
!
>
2.2.1
Evaluations
The inspectors reviewed the evaluations associated with seven modification
,
requests, four prompt modification requests (temporary modifications), three
document change notices, two condition reports, and the core operating. limits
.'
l
i
.
.
-5-
report. The evaluations reviewed are listed in Attachment 2.
The following
'
are the inspectors comments on specific evaluations:
Prompt Modification Request 90-0005
This temporary modification added a trailer mounted chiller package,
associated piping and valves to the ultimate heat sink in order to maintain
,
the water in the standby cooling tower basin below Technical Specifications'
maximum temperature limits. The evaluation included a seismic analysis of the
-piping, valves, and supports attached to the standby cooling tower. Also
analyzed was a break in the piping outside the tower and the possibility of
syphoning water out of the basin.
Included in Prompt Modification
Request 90-0005 was a drawing, EP-19H-6, showing the suction and discharge
piping terminating at elevation 112 ft. 4 in., which was 6 in. above the
Technical Specifications required minimum water level. Also noted on the
drawing was the minimum water level required for operating the chiller at 114
ft. 2 in. to insure adequate net-positive suction head for the chiller pumps.
,
A maximum water elevation was not specified.
Chapter 9.2.5 of the Updated Safety Analysis Report states that the standby
cooling tower basin has a minimum water elevation of 111 ft.10 in., a normal
water elevation of 113 ft. 4 in., and a maximum water depth of 52 ft. 6 in.
Figure 9.2-11 in the Updated Safety Analysis Report shows the lowest elevation
in the basin as 60 ft. 10 in. Adding the maximum water depth of 52 ft. 6 in.
yields a maximum water elevation of 113 ft. 4 in., the normal water elevation.
The inspectors noted that the safety evaluation did not analyze for raising
the water level to an elevation of 114 ft. 2 in. to provide net-positive
suction head for the chillers. The concern was if the standby cooling tower
basin was seismically qualified when filled to that elevation.
Based on
further review and discussion with the licensee, the inspectors determined
that the basin was seismically qualified to an elevation of 117 ft.10 in.
In
addition, the control room received an alarm when water reached an elevation
of 116 ft.
While the safety elevation was deficient, additional review indicated there
was no unresolved safety question associated with the plant change. The
licensee will revise the Updated Safety Analysis Report to reflect current
limitations on maximum water depth in the cooling tower basin.
!
Core Operating Limits Report
On June 19, 1992, General Electric submitted a Part 21 report notifying that
'
an error existed in the General Electric Standard Application for Reactor Fuel
Report (GESTAR-II/NEDE-240ll-P-A). The original approved version of GESTAR-II
,
concluded there was no need to perform cycle-specific evaluations of a rotated
'
fuel assembly for a C-lattice plant. The change in critical power ratio was
analyzed at that time to be insignificant because of the fuel design
configuration. This conclusion was based on a generic analysis performed in
,
1982. However, the rotated bundle analysis remained. a requirement of the NRC
j
approved methodology documented in GESTAR-II.
l
I
-6-
The Part 21 report stated that recent studies had shown that the conclusion of
,
the generic analysis was sometimes inappropriate for modern-typ fuel designs.
For BWR-6 S-lattice plants, the delta critical power ratio for ie rotated
i
bundle was larger than the delta critical power ratio calculated for the
identified limiting transient. Additionally, if a BWR-6 S-lattice plant was
operating at the operating limit minimum critical power ratio, the postulated
rotated bundle condition could exceed the Technical Specifications' safety
limit minimum critical power ratio. The Part 21 also stated that General
Electric would perform cycle-specific rotated bundle analyses for future C-
and S-lattice fuel designs.
General Electric performed the reload analysis for Reload 4, Cycle 5, for
River Bend Station using the methods documented in Revision 10 of GESTAR-II.
General Electric submitted the results of the reload analysis to the licensee
in the Supplemental Reload Licensing Report, 23A7181, Revision 0, dated August
1992. General Electric performed a cycle-specific rotated bundle analysis as
part of the reload analysis and concluded that the operating limit minimum
critical power ratio for this cycle was 1.22.
In preparing the core operating limits report, the licensee performed an
evaluation which analyzed the impact of adding 200 bundles of Fuel
Type GE8-P8SQB334-10GZ-120M-4WR-150-T in the core during Reload 4.
In that
evaluation, the licensee analyzed the probability of having rotated fuel and
considered the method of verifying placement of fuel bundles. The licensee
concluded that a rotated bundle did not exist and that the .04 minimum
critical power ratio penalty for rotated bundles was not applicable to River
Bend Station. The licensee documented in the core operating limits report
that the operating limit minimum critical power ratio applicable to River Bend
Station for Reload 4, Cycle 5, was 1.18.
However, neither the screening nor evaluation addressed Technical
'
Specification 6.9.3.2, which required that the analytical methods used to
'
calculate the core operating limits be those described in the latest approved
version of GESTAR.
In establishing the operating limit minimum critical power
ratio as 1.18, rather than the 1.22 calculated by General Electric, the
licensee violated Technical Specification 6.9.3.2.
This was identified as a
violation (458/9317-01).
The licensee reviewed the operating history for River Bend Station from
September 8,1992, when the plant started up from the refueling outage, to
April 18, 1993, the last day the plant operated with an operating limit
minimum critical power ratio of 1.18. The licensee identified one instance on
January 8,1993, when the plant operated below an operating limit minimum
.
'
critical power ratio of 1.22.
However, the plant was below 1.22 for less than
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, which met the action statement for Technical Specification 3.2.3.
i
i
-
.
L
_7_
,
2.2.2
Screenings
The inspectors reviewed the screenings of five modification requests and two
prompt modification requests. The screenings reviewed are listed in
'
Attachment 2.
The following are the inspectors' comments on specific
screenings reviewed.
Modification Request 91-0013
,
This modification added fencing and a detection system to the radwaste and
auxiliary control buildings that had been removed previously. The security
plan which was reportedly used as part of the bases for answers to the
screening was not identified in the documentation.
Procedures require the
bases for answers to the screenings be documented. The inspectors reviewed
the security plan and found it did not detail the fencing and detection system
and, therefore, an unreviewed safety question did not exist. This screening
was considered deficient.
2.3 Conclusions
In general, the inspectors found the effort to prepare and review safety
evaluations and screenings was good, however, one exception resulted in the
identification of a violation.
In addition, one deficient evaluation and one
d2ficient screening were identified. The impact of these deficiencies on
-
plant safety were not considered significant.
-
9
k
!
1
-
.
ATTACHMENT 1
1 PERSONNEL CONTACTED
i
1.1 Licensee Personnel
- K. Barrow, Board of Directors
- J. Booker, Manager Safety Assessment & Quality Verification
J. Brescher, Senior Electrical Engineer
- G. Bysfield, Electrical & Special Projects Supervisor
- T. Crouse, Director Engineering Support Services
- W. Curran, Cajun Site Representative
- D. Dorbonna, Assistant Plant Manager Operations
- L. Dietrich, Supervisor Nuclear Licensing
J. Egan, Senior Nuclear Safety Engineer
- D. Galle, Nuclear Safety Assessment Comittee Member
- K. Gardner, Licensing Engineer
- K. Gladrusich, Director Quality Assurance
- P. Graham, Vice President - River Bend
- J. Hamilton, Manager Engineering
W. Herman, Senior Technical Specialist
- T. Hoffman, Supervisor Civil / Structural Plant Engineering
- P. Hughes, Senior Licensing Engineer
A. Kahan, Senior Electrical Engineer
- G. Kimell, General Maintenance Supervisor
K. Klamert, Senior Mechanical Engineer
- L. Leatherwood, Supervisor Core Engineering
J. Leavines, Supervisor Nuclear Safety Assessment Group
J. Maher, Licensing Engineer
D. Melear, Senior Design Engineer
N. Pressler, Equipment Qualification Specification Engineer
- L. Rougeux, Senior Independent Safety Group Engineer
G. Scronce, Senior Nuclear Fuels Engineer
- W. Simmons, Senior Licensing Engineer
- B. Smith, Mechanical Maintenance Supervisor
- J. Spivey, Senior Quality Assurance Engineer
- M. Stein, Director Plant Engineering
- K. Suhrke, Manager Site Support
- J. Thompson, Supervisor Balance of Plant
- H. Woodson, Nuclear Safety Assessment Comittee
,
S. Woody, Director Nuclear Station Security
- J. Vachon, Senior Compliance Annalist
P. Vo, Senior Nuclear Fuels Engineer
.
r
1.2 NRC Personnel
- W. Smith, Senior Reactor Inspector
In addition to the personnel listed above, the inspectors contacted other
personnel during the inspection.
- Denotes personnel that attended the exit meeting.
.
'
.
+
-2-
2 EXIT MEETING
An exit meeting was conducted on June 18, 1992.
During this meeting, the
inspectors reviewed the scope and findings of the report. The licensee did
not identify as proprietary any information provided to, or reviewed by, the
inspectors.
,-
i
e
f
f
I
.
.
'
R
.
,
ATTACHMENT 2
i
DOCUMENTS REVIEWED
PROCEDURES
River Bend Nuclear Procedures Manual, RBNP-057, " Safety and Environmental
Evaluations," Revision draft, August 12, 1991
River Bend Nuclear Procedure Manual, RBNP-030, " Initiation and Processing of
Condition Reports," Revision 1, April 10, 1992 with Change Notices 1-4
River Bend Nuclear Procedure Manual, RBNP-027, " initiation of a Change to a
Licensing Document," Revision 3, March 15, 1991
.
River Bend Nuclear Procedure Manual, RBNP-026, " Evaluating Potentially
Reportable Conditions Pursuant to 10 CFR 21," Revision 2, August 11, 1992
Station Support Manual Procedures, ENG-3-004, " Safety and Environmental
Evaluations," Revision 1, November 21, 1989, with Change Notices 1-3
Station Support Manual Procedures, ENG-3-026, " Document Change Notices (DCN),"
'
Revision 0, Nuvember 11, 1991
Station Support Manual, TTP-7-031, "10 CFR 50.59 Reviewer Training Program,"
Revision 0, April 14, 1993
Engineering Department Procedures Manual, EDP-AA-62, " Training Requirements
for Preparers, Reviewers and Approvers of Unreviewed Safety / Environmental
Question Determinations," Revision 3, March 17, 1990 with Interim Procedure
Change 1
Station Operating Manual, ADM-0003, " Development, Control and Use of
Procedures," Revision 18C, September 9, 1992 with Change Notices 93-0370 and
92-1347
Training Program Lesson Plan Number SET-721-01, "10 CFR 50.59 Reviewer"
EVALUATIONS AND SCREENINGS ASSOCIATED WITH MODIFICATION REQUESTS (MR) AND DATE
HR-88-0088, Modification of the Leak Off of Valve Packing, September 7,1988
i
(Evaluation)
MR 91-0013, Add Fence and Detection System to West Walls of Rad Waste and Aux
!
Centrol Buildings, October 2,1991 (Screenings)
MR 91-0044, Converts Temporary Piping Installed under Prompt Modification
!
Requests 90-0005 for Cooling the Standby Cooling Tower to a Permanent
Installation, January 29,1992 (Evaluation)
HR 91-0059, Service Water Piping Replacement Install Flanges on Alternate
Cooler, July 21, 1991 (Evaluation)
4
'
-
.
b
'
-2-
MR 91-0079, Provides a Source of Demineralized Make Up Water to the Closed
Loop Service Water System, January 28,1992 (Evaluation)
MR 91-0087, Install Permanent Taps in the Reactor Water Clean Up System for
Decontamination, February 27, 1992 (Evaluation)
HR 92-0010, Revise Documents to Spare Cables and Junction Box that Had Been a
Temporary Modification, February 28, 1992 (Screenings)
MR 92-0011, Provide Alternate Model and Associated Mounting and Termination
Details for Replacement Differential Pressure Switches, January 23, 1992
!
(Screenings)
MR 92-0036, Revises Setting for Pressure Regulators, March 29, 1992
(Screenings)
HR 92-0047, Place a New Isolation Valve on the Condensate Demineralizer,-
July 16,1992 (Evaluation)
HR 92-0048, Replace Motor on Feed Water Containment Isolation Valves,
April 30,1992 (Evaluatic, -)
.
MR 92-0053, Upgrade Pressure Switch, July 18, 1992 (Screenings)
EVALUATIONS AND SCREENINGS ASSOCIATED WITH PROMPT MODIFICATION REQUESTS (PMR)
AND DATE
PMR 90-0005, Installation of Temporary Chillers for Standby Cooling Tower,
April 30, 1990 (Evaluation)
PMR 92-0014, Provide Temporary Bulk Temperature of the Reactor Vessel Water in
Mode 5, July 18, 1992 (Screenings)
PMR 92-0025, Install Different Model Chemical Feed Pump in the Normal Cooling
i
Tower Make Up Water Treatment System, December 16,1992 (Evaluation)
PMR 93-0008, Terminate All Flow to the Condensate Storage Tank from the
Control Rod Drive System, March 16, 1993 (Evaluation)
PMR 93-0012, Installation of Furmanite Clamp Around a Flange, March 20, 1993
(Screenings)
PMR 93-0015, Install Temporary Pressure Transmitter, April 23, 1993
(Evaluation)
EVALUATIONS ASSOCIATED WITH DOCUMENT CHANGE NOTICES (DCNs) AND DATE
DCN 91-00189, Revise P&ID to Show Extant of ASME Boundary of Hydraulic Control
Units, undated
,
.'
,
.
'
-3_
DCN 92-00413, Revise P&ID 6-1B and FSK-6-IC to Refit .t As-Built Configuration,
November 24, 1992
DCN 93-0082, Revisc System Design Requirements Document P-21 "Feedwater and
Auxiliaries," February 14, 1993
EVALUATIONS ASSOCIATED WITH CONDITION REPORTS (CR) AND DATE
CR 92-0572, Feed Water Check Valve IB21*A0VF032A, Surface Indications on Valve
Seat, August 8, 1992
CR 92-0675, Reactor Pressure Vessel Closure Stud Ultrasonic Indication, August
18, 1992