ML18059A488
| ML18059A488 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 10/27/1993 |
| From: | Greenman E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| Shared Package | |
| ML18059A489 | List: |
| References | |
| NUDOCS 9311080032 | |
| Download: ML18059A488 (38) | |
See also: IR 05000255/1993020
Text
Docket No. 50-255
Consumers Power Company
ATTN:
Mr. David P. Hoffman
Vice President
Nuclear Operations
1945 West Parnall Road
Jackson, MI
49201
Dear Mr. Hoffman:
October 27, 1993
SUBJECT:
SPECIAL TEAM INSPECTION REPORT 50-255/93020(DRS)
This refers to the special team inspection led by Mr. Robert Lerch of this
office, to follow up events from your Summer 1993 refueling of the Palisades
Nuclear Plant and to observe routine activities. The team was composed of
Michael Parker, Ronald Bailey, Thomas Tongue and David Nelson of the
Region III office and our contractor Carl B~yer of Battelle Northwest*
Laboratories. This refers also to the inspection of your cycle 11 core reluad
plan conducted by Anthony Hsia, Edward Kendrick, James Davis, and
Shih Liang Wu of the Office of Nuclear Reactor Regulation.
At the conclusion
of the onsite inspection, an interim "exit" meeting was held with you and
members of your staff to discuss the inspection findings.
On September 29,
1993, a final exit was held.
Areas examined during the inspection are identified in the enclosed report.
Within these areas, the inspection consisted of selective examinations of
procedures and representative records, interviews with personnel, and
observation of activities in progress.
The teams concluded that your
corrective actions for the damaged 1-24 and stuck SAN-8 fuel assemblies were
adequate for cycle 11 operation.
Your root cause determination efforts were
found sufficiently completed and technically thorough that repetition of those
problems would be prevented this cycle. However, some long term corrective
actions have yet to be determined.
We understand that you plan to collect *
additional information and further evaluate contributing factors for the next
cycle (12).
Our observations of plant activities, though limited, concluded
that your staff was ready and capable to start up and routinely operate the
pl ant.
A review of the regulatory issues.documented in the Augmented Inspection Team
- (All) inspection report, (Report No. 50-255/93018), was also performed.
Based
on these inspection, certain of your activities appeared to be in violation of
NRC requirements as described in the enclosed Notice of Violation (Notice) .
9311080032 931027
---
.PDR
ADOCK 05000255 .
G
~*
Consumers Power Company
2
October 27, 1993
We considered these issues for escalated enforcement pursuant to the NRC
Enforcement Policy, 10 CFR Part 2, Appendix C.
We concluded that escalated
enforcement was not appropriate because some of the issues share common causes
with, and predate, issues previously addressed in our Notice of Violation and
Proposed Imposition of Civil Penalty (EA-93-178) dated September 14, 1993.
The previous escalated enforcement action was taken, in part, to emphasize the
necessity for strict, disciplined control and verification of proper
performance of activities involving reactor components.
We are concerned that
additional violations were observed later in the outage, in which management
controls in the form of procedures were not followed.
Although some of these
violations occurred in the presence of your supervisors and/or auditors, they
did not intervene to stop the activity and correct it.
You are required to respond to this letter and should follow the instructions
specified in the enclosed Notice when preparing your response.
In your
response, you should document the specific actions taken and any additional
actions you plan to prevent recurrence.
After reviewing your response to this
Notice, including your proposed corrective actions and the results of future
inspections, the NRC will determine whether further NRC enforcement action is
necessary to ensure compliance with NRC regulatory requirements .
In your response to these violations, you are specifically requested to
address actions you have taken or plan to take to ensure:
1.
procedures or other controls exist which exactly define the means ~nd
limits for handling reactor components so they will not become damaged;
2.
the controls/procedures are strictly followed;
3.
compliance is carefully monitored and verified; and,
4.
reactor component performance in service is analyzed, acceptance limits '
are established, and proper performance within limits is monitored and
verified.
In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of
this letter and its enclosures will. be placed in the NRC Public Document Room.
The responses directed by this letter and the enclosed Notice are not subject
to the clearance procedures of the Office of Management and Budget as required
by the Paperwork Reduction Act of 1980, Pub.
L~ No. 96.511 .
~*
Consumers Power Company
3
October 27, 1993
We will gladly discuss any questions you have concerning this inspection.
Enclosures: *
2.
Inspection Report
No. 50-255/93020
cc w/enclosures:
David P. Hoffman, Vice President
Nuclear Operations
David W. Rogers, Safety
and Licensing Director
OC/LFDCB
Resident Inspector, Riii
James R. Padgett, Michigan Public
Service Commission
Michigan Department of
Public Health
A. H. Hsia, LPM, NRR
SRI, Big Rock Point
J. Lieberman, OE
R. DeFayette, RA
bee w/enclosure:
PUBLIC 'IE-01
Sincerely,
Original signed by Edward G. Greenman
Edward G. Greenman, Director
Division of Reactor Projects
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-*
Consumers Power Company
3
October 27, 1993
We will gladly discuss any questions you have concerning this inspection.
Enclosures:
1.
2.
Inspection Report
No. 50-255/93020
cc w/enclosures:
Gerald B. Slade, General
Manager
David W. Rogers, Safety
and Licensing Director
OC/LFDCB
Resident Inspector, Riii
James R. Padgett, Michigan Public
Service Commission
Michigan Department of
Public Health
A. H. Hsia, LPM, NRR
SRI, Big Rock Point*
J. Lieberman, OE
R. DeFayette, RA
bee w/enclosure:
PUBLIC IE-01
Sincerely,
Original signed by Edward G. Greenman
Edward G. Greenman, Director
Division of Reactor Projects
Riii
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See Following Page----------~--------------------
Lerch/cg
Bailey
Tongue
Nelson
Jorgensen
10/ /93
10/ /93
10/ /93
10/ /93
10/ /93
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See Following Page-----------------------
W. Dean
OeFayette
Forney
Greenman
10/ /93
10/ /93
10/ /93
10/ /93
~*
Consumers Power Company
2
October 27, 1993
Procedure for NRC Enforcement Actions" (Enforcement Policy), 10 CFR Part 2,
Appendix C.
Accordingly, no Notice of Violation is presently being issued for
these inspection findings.
In addition, please be advised that the number and
characterization of apparent violations described in the enclosed inspection
report may change as a result of further NRC review.
You will be advised by separate correspondence of the results of our
deliberations on this matter.
No response regarding these apparent violations
is required at this time.
In accordance with 10 CFR 2.790 of the Commission's regulations, a copy of
this letter and the enclosed inspection report will be placed. in the NRC
Public Document Room.
We will gladly discuss any questions you have concerning this inspection.
Sincerely,
Edward G. Greenman, Director
Division of Reactor Projects
Enclosure:
Inspection Report
No. 50-255/93020
cc w/enclosure:
David P. Hoffman, Vice President
Nuclear Operations
David W. Rogers, Safety
and Licensing Director
OC/LFDCB
Resident Inspector, Riii
James R. Padgett, Michigan Public
Service Commission
Michigan Department of
Public Health
A. H. Hsia, LPM, NRR
SRI, Big Rock Point
J. Lieberman; OE
R. Defayette, RA
Rill
Riii
Riii
Riii
Riii
See Following Page-------------------------------
Lerch/cg
Bailey
Tongue
Nelson
Jorgensen
10/ /93
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10/ /93
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Consumers Power Company
2
October 27, 1993
enforcement action in accordance with the "General Statement of Policy and
Procedure for NRC Enforcement Actions" {Enforcement Policy}, 10 CFR Part 2,
Appendix C.
Accordingly, no Notice of Violation is presently being issued for
these inspection findings.
In addition, please be advised that the number and
characterization of apparent violations described in the enclosed inspection
report may change as a result of further NRC review.
You will be advised by separate correspondence of the results of our
deliberations on this matter.
No response regarding these apparent violations
is required at this time.
In accordance with 10 CFR 2.790 of.the Commission's regulations, a copy of
this letter and the enclosed inspection report will be placed in the NRC
Public Document Room.
We will gladly discuss any questions you have concerning this inspection.
Enclosure:
Inspection Report
No. 50-255/93020
cc w/enclosure:
David P.
H~ffman, Vice President
Nuclear Operations
David W. Rogers, Safety
and Licensing Director
OC/LFDCB
Resident Inspector, Riii
James R. Padgett, Michigan Public
Service Commission
Michigan Department of
Public Health
A. H. Hsia, LPM, NRR
SRI, Big Rock Point
J. Lieberman, OE
R. DeFayette, RA
R1ir4
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Lerch/cg
10/ty /93
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Bailey
10/ ii.I /93 ~~
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Sincerely,
Edward G. Greenman, Director
Division of Reactor Projects
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Greenman
10/
/93
-*
Consumers Power Company
Palisades Nuclear Plant
Docket No. 50-255
License No. DPR-20
During an NRC inspection conducted on July 8 through July 20, and August 19
through 27, 1993, violations of NRC requirements were identified.
In
accordance with the "General Statement of Policy and Procedure for NRC
Enforcement Actions," 10 CFR Part 2, Appendix C, the violations are listed
below:
A.
10 CFR 50.59 (b)(l) requires in part that the licensee have a
written safety evaluation which provides the bases for
determination that a change in the facility as described in the
Safety Analysis Report does not involve an unreviewed safety
question.
Section 3.3.2.6 of the Updated Safety Analysis Report
describes use of hafnium poisoned assemblies as part of a neutron
fluence reduction program.
Contrary to the above, the licensee's 50.59 evaluation of extended
use of hafnium poisoned I-series fuel assemblies for cycles 10 and
11 did not provide complete bases for the determination that there*
was no unreviewed safety question.
Specifically, the evaluations
did not include consideration of the effects of the neutron
spectrum at th~ periphery of the core (which was proportionally
higher in "fast" neutrons and lower in "thermal" neutrons compared
to non-periphery locations) on mechanical properties of the
assemblies.
These evaluations were therefore incomplete.
This is a Severity Level IV violation (Supplement I).
B.
10 CFR 50, Appendix B, Criterion XI, requires that a test program
c.
'_be established to assure that all testing required to demonstrate
that structures, systems, and components will perform
satisfactorily in service is identified and performed in
accordance with written test procedures.
Contrary to the above, the licensee's Cycle 10 radiochemistry
testing did not positively identify failed I-series fuel, nor was
additional testing performed between cycles 10 and 11 to assure !-
series fuel would perform satisfactorily for a sixth cycle, which
constituted a unique fuel performance demand.
This i*s a Severity Level IV violation (Supplement 1).
10 CFR 50, Appendix B, Criterion XVI, requires in part that
measures be established to assure that conditions adverse to
quality are promptly identified and corrected.
In the case of
significant conditions adverse to quality the measures shall
assure that the cause of the condition is determined and
corrective action taken to preclude repetition.
--
9311080036 931027
,.
ADOCK 05000255
0
-*
2
Contrary to the above, when the upper guide structure (UGS) was
lifted from the reactor on July 6, 1993, and a fuel bundle from
core location Z-11 stuck onto the UGS, it marked the third
occasion for this same interference between core components.
This is a Severity Level IV violation (Supplement I).
D.
Technical Specification 6.8.1.b requires that written procedures
be established, implemented and maintained for activities covering
refueling operations.
1.
Licensee procedure RVI-M-1, Revision 16, "Removal and
2.
- storage of the Upper Guide Structure," provided instructions
for installation for a Tl-2000 load cell readout device
only.
Contrary to the above, on July 6, 1993, while lifting the
UGS, a J-300 load cell readout device was used.
Licensee procedure RVI-M-1, Revision 16, "Removal and
storage of the Upper Guide Structure," Section 5.3.6.g,
includes a stipulation to follow Work Order No. 24301781 for
steps to use a load cell. Step 3.3.A.7 of the work order
requires that the load cell readout device be zeroed.
Contrary to the above, on July 6, 1993, while p~rforming
procedure RVI-M-1, the load cell readout device was not
zeroed.
3.
Licensee procedure RVI~M-1, Revision 16, "Removal and
storage of the Upper Guide Structure," Section 5.3.14,
specifies an upper load limit of 62,000 pounds.
Contrary to the above, on July 6, 1993, while performing
procedure RVI-M-1, after the UGS was raised approximately
six inches, indicated load reached 62,800 pounds, which is
in excess of the upper load limit.
4.
Licensee procedure FHS0-18, "Recovery of Bundle SAN-8,"
requires in steps 4.2.6 and 5.2.l that chainfall tension be
limited to a combined load of 1500 ~ 1600 pounds.
Contrary to the above, on July 7, 1993, the chainfalls were
tightened to a combined load of 2300 pounds.
This is a Severity Level IV violation (Supplement I) .
-*
3
Pursuant to the provisions of 10 CFR 2.201, Palisades Nuclear Plant is hereby
required to submit a written statement or explanation to the U.S. Nuclear
Regulatory Commission, ATTN:
Document Control Desk, Washington, D.C. 20555
with a copy to the Regional Administrator, Region III, and a copy to the NRC
Resident Inspector at the facility that is the subject of this Notice, within
30 days of the date of the letter transmitting this Notice of Violation
(Notice).
This reply should be clearly marked as a "Reply to a Notice of
Violation" and should include for each violation:
(1) the reason for the
violation, or, if contested, the basis for disputing the violation, (2) the
corrective steps that have been taken and the results achieved, (3) the
corrective steps that will be taken to avoid further violations, and (4) the
date when full compliance will be achieved.
If an adequate reply is not
received within the time specified in this Notice, an order or a Demand for
Information may be issued to show cause why the license should not be
modified, suspended, or revoked, or why such other action as may be proper
should not be taken.
Where good cause is shown, consideration will be given
to extending the response time .
Da~ed a~~len Ellyn, Illinois
th1s35day of October 1993
-*
U. S. NUCLEAR REGULATORY COMMISSION
REGION I I I
Report No. 50-255/93020(DRS)
Docket No. 50-255
Licensee:
Consumers Power Company
Palisades Nuclear Plant
27780 Blue Star Highway
Covert, MI
49043
Facility Name:
Palisades Nuclear Plant
Inspection At:
Palisades site, Covert, MI
License No. DPR-20
Inspection Conducted:
August I9 - 27, and September 29, I993
Inspectors:
Thomas Tongue
Ronald M. Bailey
Michael Parker
Shih Liang Wu
Appr*v*d ay: ~ifl:~rr
Team Leader
Anthony H. Hsia
Edward D. Kendrick
James L. Davis *
Date
Inspection Summary
Inspection on August I9 - 27. and September 29. I993 (Report No. 50-255/93020}
Areas Inspected: Special, announced, team inspection of licensee activities
relating to Summer I993 refueling outage problems, and to plant restart
preparations, including: the safety evaluation for the cycle II core reload,
routine plant operations activities, the impact of the fuel lost from the
damaged fuel assembly, the root cause evaluation for the stuck fuel assembly,
the regulatory issues identified in the All report, and the September 9, I993
public meeting.
Results: Apparent violations resulting from the All *inspection are identified
in paragraph 7 .
9311080041 931027
ADOCK 05000255
G
.i
REPORT DETAILS
1.0
Persons Contacted
G. B. Slade, Plant General Manager
T. J. Palmisano, Plant Operations Manager
K. M. Haas, Radiological Services Manager
R. B. Kasper, Maintenance Manager
D. W. Rogers, Safety and Licensing Director
R. M. Rice, Director, Nuclear Performance Assessment Department (NPAD)
K. E. Osborne, System Engineering Manager
These people and others were present at the exit meetings on July 27, and
September 29, 1993.
R. M. Rice was not present at the September 29, 1993
exit. Other members of the plant staff and NPAD were contacted during the
inspection period.
2.0
Introduction
On July 1, 1993, at Palisades Nuclear Plant, a broken fuel rod was identified
in the tilt pit area of the reactor cavity.
On July 6, 1993, while removing
the upper guide structure (UGS) as part of the investigative activities
associated with the broken fuel rod, a fuel assembly was inadvertently lifted
from the core.
In response to these events, an Augmented Inspection Team
(AIT) was sent to the site to document and validate the relevant facts,
determine the probable causes, and evaluate the licensee's analyses efforts
and review of the events including corrective actions. A public All exit
meeting was held with plant management on July 20, 1993.
The results of the
All inspection were documented in Inspection Report 50-255/93018.
After the All exit, the NRC continued to monitor the licensee's corrective
actions in several ways.
Daily briefings on the licensee's root cause
analyses and recovery efforts were conducted via conference calls with the NRC
until two teams were sent to the site. One team reviewed the safety
evaluation for the core reload, which uses reconstituted L-series fuel
assemblies - see paragraph 3. A technical review of the of the root cause
determinations for the fuel assembly stuck to the UGS was conducted - see-
paragraph 4.
The second team evaluated the plant staff in the performance of
routine activities and a trial installation of the upper guide structure - see
paragraph 5.
A contractor with the second team evaluated the licensee's
search for the fuel lost from the damaged fuel assembly and the consequences
this fuel might have - see paragraph 6.
An assessment of the ftndings from
the All report was conducted to determine which ones represented regulatory
issues - see paragraph 7. A public meeting was held.with the licensee on
September 9, 1993 - see paragraph 8 .
2
-*
3.0
The 50.59 Safety Evaluation of the Cycle 11 Core Reload
The NRR team reviewed the license~'s performance in implementing the-
requirements set forth in 10 CFR 50.59.
The principal measure of this
performance is the quality of the 50.59 safety evaluation prepared by the
licensee for the revised Cycle 11 core design which included the 16
reconstituted L-assemblies in the corner core locations. Also included in
assessment was the documentation of the 50.59 safety evaluation, the
licensee's procedure for safety evaluations, and the PRC review process.
review of the 50.59 safety evaluation included its scope, accuracy, and
thoroughness in both technical content and documentation.
the
The
As a result of the root cause analysis for the damaged fuel assembly, the
licensee revised the original Cycle 11 core design by replacing all the
I-series fuel assemblies with thrice-burned L- series assemblies reconstituted
with 14 stainless steel rods (see Section 3.1 for detailed descriptions).
These modified L-series assemblies were located at the 16 corner core
locations as shown in Figure 2.
According to Palisades Procedure No.3.07,
Rev.7, the licensee performed a safety evaluation of the revised Cycle 11 core
design in compliance with the regulations set forth in 10 CFR 50.59. Their
evaluation was documented in "Palisades Nuclear Plant Safety Review", Plant
Safety and Licensing (PS&L) Log No.93-1025, which concluded that there were no
unreviewed safety questions in accordance with 10 CFR 50.59.
On August 19,
1993, the Plant Review Committee (PRC) reviewed the 50.59 safety evaluation,
agreed with the conclusions and approved the revised Cycle 11 core design.
From the perspective of mechanical design, material, core physics, and thermal
hydraulics, the inspection focused on the licensee's evaluations of safety
significance and whether there was any unreviewed safety question stemming
from the revised Cycle 11 core design.
The inspectors also assessed the
licensee's root cause analysis to date, its plans for continued root cause
analysis, and the changes made to address the root causes.
The team also
agreed with the Pal.isades PRC conclusion that the revised Cycle 11 core design
do~s not constitute an unreviewed safety question and its approval of the
revised cycle 11 core design.
,
3.1
Description of Changes in the Facility from that Described in the
Final Safety Analysis Report (FSARl
3.1.1
Fuel Assemblies
The M-assemblies, N-assemblies, 0-assemblies, and SAN-assemblies have not been
altered for the revised Cycle 11.
The L-assemblies have been reconstituted_
for the revised Cycle 11 by replacing 14 fuel rods with 14 solid stainless
steel rods.
Eight stainless steel rods were placed in the corner of the fuel
assembly in the shroud corner with the stainless steel rods placed from the
guide bar to the corner plus the location next to the corner location as shown
in Figure 1.
Each remaining corner of the fuel assembly contains 2 stainless
steel rods as shown in Figure 1 .
3
The criteria used by the licensee to select the L-assemblies to replace the
I-Hafnium assemblies were:
1.
8urnup < 37,500 MWD/MTU at EOC 11 for J~assemblies and K-assemblies.
Licensee Justification: This criteria is somewhat arbitrary and less
important if the L- assemblies are used.
The objective is to stay below the
1-24 assembly burnup at the start.of cycle 10 when the 1-24 assembly was
intact. The L-assemblies have the spacers manufactured with a vertical roll
and will grow in that direction with fluence.
The spacer cells will not get
bigger, rather they willget taller; so, the design limit of 46,000 MWT/MTU is
still applicable for the L-assemblies for cycle 11.
2.
Assemblies can not be used that have previously occupied one of the
eight octant symmetric corner shroud positions.
Licensee Justification: This requirement may be conservative for all eight
octant positions.
However, it is advantageous to avoid assemblies that
previously occupied 8-19 and X-19 core locations.
3.
Consider assembly bowing to avoid largely bowed assemblies.
Licensee Justification: Assembly to shroud interference already existed at
Cycle 10 8-19 and X-19 locations. Placing the bowed assembly towards the
shroud in Cycle 11 may*worsen the problem.
4 .
L-assemblies may*have some advantages over the. J-assemblies and the K-
assemblies.
Licensee Justification: The spacer plate material for the L-assemblies was
stamped in a preferential direction to minimize cell opening due to growth.
Also, using L-assemblies will not require ultrasonic examination since they
were examined after their last cycle.
The J-assemb.lies and K-assemblies will
require ultrasonic examination prior to use in Cycle 11.
'
5.
Fluence rates values should not exceed Cycle 9 values.
Licensee Justification: Cycle 9 fluence rates were used for consistency with
the pressurized thermal shock data currently being reviewed by the NRC.
3.2
Mechanical Design
The stainless steel rods are 0.437 inches in diameter while the fuel rods they
replaced were 0.417 inches in diameter.
The licensee proposed that the larger
diameter rods will reduce the probability of fretting damage during Cycle 11
due to an estimated increase in spring force of 0.6 pounds.
The licensee
proposed that the corners with 2 stainless steel rods will be less likely to
fret and that the load on adjacent fuel rods will be symmetrically increased,
decreasing the probability of fretting on the adjacent fuel rods .
4
-*----
The revised L-assemblies and SAN-assemblies have standard design spacers and
all of the remaining assemblies in the core for cycle 11 have the debris
resistant, High Thermal Performance (HTP) spacers.
The I-assemblies that "the
revised L-assemblies replaced also had the standard design spacers. Hence,
the licensee concluded that the use of the revised L-assemblies will not
affect spacer induced flow conditions in the core for the revised Cycle 11.
3.3
Core Physics Design
The revised Cycle 11 core contains 52 twice-burned M-assemblies, 68 once-
burned N-assemblies, 60 fresh 0-assemblies, 8 once-burned N Shield Assemblies
(SAN}, and 16 modified thrice-burned L-assemblies with 14 solid stainless
steel rods inserted in the assembly corners. The difference between the
revised and the original Cycle 11 core designs is the 16 reconstituted
L-assemblies and the relocation of the selected M and N assemblies
(20 assemblies in each batch}.
Figure 2 shows the revised Cycle 11 core
design.
For fluence reduction purposes, the revised Cycle 11 core is a low radial
leakage core incorporating 8 SAN assemblies in the flat peripheral regions and
16 reconstituted L-assemblies in the core corner locations. A higher initial
reactor coolant boron concentration is needed to offset the additional
reactivity in the revised Cycle 11 because of the revised L-assemblies (which
are more reactive as compared to the.I-assemblies), because the 0-~ssemblies
have enrichment higher than previous fuel assemblies, and because of less
gadolinia used in Cycle 11 fuels as compared to previous cycles .
3.4
Impact on the FSAR and the Technical Specifications (TS)
The licensee's 50.59 safety evaluation as documented in PS&L Log No. 93-1025
indicates that many FSAR Sections and TS Sections have been reviewed for
possible changes due to the revised Cycle 11 core design.
These reviews
identified FSAR Sections 3.3.1, and 3.3.4.3, and Tables 1-2, and 3-11 to be
affected by the design changes.
None of the TS Sections was identified to be
affected by the design changes.
The licensee plans to update FSAR Sections 3.3.1 and 3.3.4.3 to delete the
discussions on the boron carbide neutron absorber rods and the Hafnium
clusters that are no longer used, and to add the discussions on the 16
modified L-assemblies and the new assemblies with debris resistant features.
The licensee also plans to update Table 1-2 to include Batch 0 average U-235
enrichment, and to update Table 3-11 for inclusion of L-assemblies and
deletion of* boron carbide neutron absorber rods and Hafnium clusters.
The staff rev~ew also identified FSAR Section 3.3.2.6 which discusses the H-
assembl ies with stainless steel rods in Cycle 8, the I-assemblies with-Hafnium
clusters in Cycle 9, and the SAN assemblies in Cycle 10, as used by the
licensee to reduce neutron fluence on the reactor vessel. This section should
have been identified by the licensee for deletion of the Hafnium clusters and
inclusion of the r~vised L-assemblies.
The licensee plans to make these
additional changes to the FSAR .
5
-*---
The Licensee's 50.59 safety evaluation stated that the radial peaking factor
limits listed in TS Table 3.23-2 were not_ changed by the revised L-assemblies
used to replace the I-assemblies. Although Table 3.23-2 does not specifically
list the limits for the revised L-assemblies and the SAN assemblies, Table 6.1
in the Siemens Power Corporation (SPC) report EMF-92-177, Rev~2, implies that
- the "Revised L" and "SAN" assemblies are bounded by the "M and earlier" and
"N" assemblies, respectively. * The licensee uses the PIDAL code to perform its
weekly monitoring of the radial peaking factors for the 5 different batches of
fuel assemblies (Revised L, M, N, SAN, and 0).
The NRC evaluation of the FSAR and TS Sections, in light of the revised Cycle
11 design, found it necessary for the licensee to amend the TS Tables 3.23-1
and 3.23-2 to include the different assemblies in the core pri?r to operation
above 25% power.
3.5
Root Cause Analysis Evaluation for I-Assembly Failure
The licensee had proposed that there were 8 potential root causes for the fuel
failure in the 1-24 assembly discovered after cycle 10.
These were:
1.
2.
3.
4.
5.
6.
7 .
8.
Damaged during fuel moves during previous cycle(s);
Damaged during EOC 9 Ultrasonic Test Inspection;
Damaged during fuel moves in this refueling outage;
Fuel failure due to a loose spacer grid;
Fuel failure due to increased primary coolant system (PCS). flow;
Fuel failure due to core barrel vibration;
Fuel failure due to a manufacturing defect; and, .
Fuel failure due to a shroud/fuel assembly interface problem.
Number 1 was eliminated because all of the assemblies placed in the core for
cycle 10 were ultrasonically tested (UT) between cycle 9 and-cycle 10.
Number 2 was eliminated since the records indicate that the I-24 bundle was
not rotated during UT examination and could not have been damaged by the UT
test rig.
Number 3 was eliminated by reexamination of the chemistry results
for cycle 10.
Number 4 was eliminated by the licensee as the root cause but
is thought to be a contributing factor.
The licensee considers numbers 5, 6,
and 7 to be less likely as the root cause.
However, the licensee has proposed
that 5 and 6 may be contributing factors. The licensee has proposed that
number 8 is the most likely cause of the I-24 assembly rod failure.
The use of stainless steel rods in the corners of the replacement assemblies
addresses possibility 4, according to the licensee, by increasing the spring
force on these rods.
The use of stainless steel rods in the replacement
assemblies addresses possibilities 5, 6, and 8 by placing sacrificial rods
that do not contain fuel in locations where damage occurred to some rods
during cycle- 10 and which are more suscepUble to potential damages in* future
fuel cycles.
The licensee did not think that the use of stainless steel rods would prevent
the type of spacer,fretting that occurred on 1-24 during cycle 10.
However,
the licensee proposed that the stainless steel rods would be the only rods
affected and no fuel would be lost as a result of spacer fretting .
6
-*--
The team did not complete its assessment of the root cause analysis since the
licensee had not finalized the root cause analysis.
Additional data such as examination of core vibrational data, fuel failure
monitoring, end of cycle {EOC} 11 fuel assemblies examination, and EOC 11
shroud examination will contribute to the understanding of the root causes.
This will be tracked as inspection follow~up item {IFI 50~255/93020-01}
3.6
3.6.1
3.6.1.1
Technical Evaluations
Mechanical Design
Fuel Reconstitution and Generic Letter CGLl 90-02 Supplement 1
The NRC staff issued GL 90-02, Supplement 1, to address the nuclear industry
trend of reconstituting fuel assemblies with dummy {nonfueled} rods of
stainless steel or zircaloy.
The use of dummy rods facilitates the
replacement of failed fuel rods when leakers are detected.
Supplement 1 requires that NRC-approved methodologies be applied for
reconstitution to ensure compliance with General Design Criteria {GDC} 10.
The staff is reviewing SPC topical report ANF-90-082, entitled "Application of
ANF Design Methodology for Fuel Assembly Reconstitution." This report is
expected to be approved soon.
The licensee's reconstitution for the revised
Cycle 11 core is not covered explicitly by ANF-90-082, as discussed in
licensee's 50.59 evaluation. However, SPC reanalyzed the Cycle 11 core with
the 16 new reconstituted assemblies as described in the safety analysis report
EMF-92-177, Revision 2.
The results showed that the Chapter 15 analyses were
still bounded by the previous analyses.
The team considers the reanalyses of
Cycle 11 core adequate.
3.6.1.2
Oversized Stainless Steel Rods
The*stainless steel rods in the reconstituted L assemblies ~re slightly larger
in diameter, than the fuel rods.* The larger stainless steel rods in the corner
will exert more force on the lantern spring, thereby tightening up the
adjacent fuel rods. This results in higher spring force on fuel rods to
compensate the spring relaxation during irradiation. The fuel vendor's
analysis and testing results confirmed that there was higher spring force on
- the adjacent fuel rods. Since SPC has confirmed that the bimetallic spring
relaxed as expected, the use of larger stainless steel rods tends to reduce
the potential for fretting between the fuel rQds and spacer springs. Thus,
the team considered the use of larger stainless steel rods acceptable.
3.6.1.3
Spring Retention
The rod withdrawal force data frQm fuel assemblies I, J, K, H, and L have
shown that the spring force relaxed significantly during Cycle 10.
In some
cases, the rod cell force was completely diminished.
However, this result was
not unexpected by SPC .
7
-*--
SPC incorporated the Palisades data into the spring force relaxation versus
burnup_data from other SPC assemblies.
The results showed that the bimetallic
spring relaxed with higher burnup as expected for Palisades' spacer springs.
SPC's analysis has demonstrated that the spring force will remain higher than
-the vibration force until the end of Cycle 11.
The team understands that the
licensee's current fuel design uses the high thermal performance (HTP) grid
spacer.
The HTP grid spacers do not use bimetallic springs and should be less-
prone to the fretting damage.
Because the Palisades spring data were within
the bounds of the ~nalysis and the larger stainless steel rods were used to
increase spring retention, t~e team ~oncludes that the spring retention has
been adequately addressed.
3.6.1.4
Fretting Wear Against Core Shroud
Although the licensee has not yet determined the root cause, some possible
causes were surveyed and examined.
The licensee speculated that the most
probable cause was due to interference between the core shroud and fuel
assembly, indicated by the missing or torn grid spacers and the corresponding
wear indications on the core shroud.
While the Palisades fuel failure root
cause analysis is continuing, the licensee has taken actions to prevent such a
phenomenon from recurring by placing stainless steel corner rods in the
reconstituted L-assemblies which will be placed in the core corner locations.
This is a solution that is similar to that used in resolving the Westinghouse
reactor baffle jetting problem.
The reconstituted L-assemblies will reside in
the Cycle 11 core for only one cycle, and will be replaced by new shielding
assemblies in future reloads.
Based on the reconstituted assemblies, the team
concludes that the licensee has taken appropriate action in mitigating the
consequence of interference between fuel assembly and core shroud.
3.6.1.5
Wear of the Control Rods
The revised L-assemblies are placed in the core with fluence induced bow
toward the center of the core.
The licensee has proposed that orienting the
revised L-assemblies in this manner minimizes the possibility of interactions
of the fuel assemblies with the shroud corner.
The revised L-assemblies were
selected based on those L-assemblies with the most uniform fluence.
The
maximum projected burnup gradient across the revised L-assembly at the end of
Cycle 11 was calculated by the licensee to be approximately 6,400 MWD/MTU.
Eight of the revised L-assemblies are next to a control rod.
The maximum
projected gradient for an L-assembly next to a control rod is calculated to be
about 4,600 MWD/MTU.
The licensee reports that during previous cycles,
assemblies with gradients as high as 7,000 MWD/MTU were oriented towards a
control rod resulting in no interaction between the assembly and the control
rod.
The licensee proposes that the burnup gradients for the revised
L-assemblies is expected to be bounded by past operating history of SPC fuel
at Palisades.
The space between fuel assemblies is 0.365 inches and the width of the blades
on the control rod is 0.180 inches leaving a gap of 0.093 inches between the
fuel assembly and the control rod blade on each side of the blade.
The
licensee has proposed that the bow in the assembly will be less that 0.093
inches based on past operating history and calculations by the licensee's fuel
8
supplier.
The NRC team concurs with the licensee's analysis that the fuel
assemblies next to the control rods will not affect the functionality of the
-
3.6.2
Core Physics Design
As stated in the SPC Letter, to the licensee dated August 17, 1993, HGS:312:93
the evaluation of the core physics design changes is documented in Revision 2
(August 17, 1993) of EMF-92-177, "Palisades Cycle 11 Safety Analysis Report,"
and in Revision 1 (August 23, 1993) of EMF-92-178, "Palisades Cycle 11:
Disposition and Analysis of Standard Review Plan Chapter 15 Events,"
The physics characteristics evaluated in Revision 2 of EMF-92-177 include
power distribution, control rod reactivity, and moderator temperature
coefficient {MTC) considerations.
The Cycle 11 core loading configuration was
redesigned by SPC to minimize radial peaking factors; specifically, to remain
within the approved Technical Specification {TS) limits of:
Assembly Radial Peaking Factor, FrA limit of 1.76 {Batch 0), 1.66
{Batch N), and 1;57 (Batches Mand L)
Total Radial Peaking Factor, F; limit of 2.04 {Batch 0), and 1.92
{a 11 others)
-Linear Heat Generation Rate {LHGR) li'mit of °15.28 kw/ft to 60%
core height; linearly decreasing to 14.21 kw/ft at 100% core
height
The maximum calculated_values for Fril and F; were 1.575 and 1.851,
respectively; which, when combined with the TS uncertainties, are within the
TS power peaking limits. The largest calculated LHGR is 11.63 kw/ft which is
also within the TS limiting value with the TS uncertainty included.
Shutdown margin calculations were performed for the modified Cycle 11
configuration, yielding a cycle minimum shutdown margin of 2.56% delta-rho at
End of Cycle {EOC}, Hot Full Power {HFP) conditions, which is above the TS low
limit of 2.00%.
The moderator temperature coefficient {MTC) was evaluated for
the revised Cycle 11 core configuration at both Hot Zero Power {HZP) and HFP
for Beginning of Cycle {BOC) and EOC conditions.
The calculated MTC values
are within the safety analysis limits of +0.5xl0-
4 delta-rho/degree F and -
3.5x10-4 delta-rho/degree F.
-
The minimum departure from nucleate boiling ratio (MDNBR) evaluation for the
revised Cycle 11 core configuration was performed with the ANFP CHF
correlation for the TS limiting radial peaking factor values.
The -re-analysis -
of eight Standard Review Plan (SRP) Chapter 15 events, as reported in
Revision 1 of EMF-92-178, shows that calculated DNB margins were improved with
the exception of Events 15.4.2 - Uncontrolled control rod bank withdrawal at
9
-*----
power, 15.4.3(5} - Control rod misoperation: single rod withdrawal, and
15.6.1 - Inadvertent opening of pressure relief valve, which showed a slight
DNB margin degradation. All FSAR event acceptance criteria, however, were met
for the revised Cycle 11 core.
3.6.3.
Assembiy Flow Vibration And Core Barrel Vibration
The licensee considered core barrel vibration and assembly flow vibration as
two possible contributi~g factors to the fretting wear problem.
The core
barrel vibration problem was discovered early in the plant life and was
addressed by adding a hold-down ring to the UGS.
There was no indication in.
subsequent operations that the core barrel vibration occurred again.
The
licensee included the core barrel vibration in its root cause analysis and
concluded that it was a possible contributing factor.
The team considered the
licensee's effort in addressing the concern of core barrel vibration adequate.
For the assembly flow vibration, the licensee has reconstituted the
L-assemblies with larger stainless steel rods in the corner locations. The
reconstituted assemblies are designed to mitigate the potential fuel rod
damage due to assembly flow vibrations.
The team considered_ the licensee's
effort in addressing the concern of assembly flow vibration adequate.
- Recently, a few other PWR plants also experienced similar phenomena of fuel
failure and fretting wear near the ~ore shroud.
The ~ssemblies involved were
of a newer spacer design located near the core shroud.* Subsequent
examinations and flow testing uncovered that there was a natural vibrational
frequency that existed for.a combined condition of a particular spacer design
and restricted flow.
Those licensees involved have modified the fuel
assemblies to dampen the flow ~ibration for the short term.
The fuel design
and flow testing procedures have been modified to take into account this
effect. Since this phenomenon is rather new and unique, the staff is
preparing an Information Notice to alert all licens~es of this type of flow
induced vibrational fretting.
The effects of the modified L-assembly oversized stainless steel rods.on local
assembly flow distributions and the intra-assembly and inter-assembly cross
flows was evaluated by SPC and was found to be insignificant .. The effects of
the revised core loading configuration on the core-wide flow distribution were
analyzed by SPC using the XTG computer code, including the simulation of the
two spacer designs {bi-metallic for the L-assemblies and HTP for the remaining
assemblies}.
The results showed that the revised core design had
insignificant effects on the core wid~ flow distribution.
3.6.4
Materials
The licensee analyzed the irradiation-induced spring force relaxation. The
spacer springs are designed such that they will not damage the cladding during
installation nor will they damage the cladding during operation due to
differential thermal expansion.
The spring force should be sufficiently high
to prevent fretting of the cladding, and to suppress as-fabricated and thermal
bow of the fuel rods and to resist flow induced vibration of the fuel rods .
10
-*
/
The springs start with a spring force ranging from 2.6 to 4.5 pounds.
The
minimum spring force is the force required to overcome flow induced vibration
force, which is 0.08 pounds.
The minimum spring force at the end of life will
be a 90 percent relaxation of the 2.35 pound beginning of life spring force or
0.2 pounds, which exceeds the flow induced vibration force of 0.08 pounds.
The team reviewed the licensee's iubmittal dated August 16, 1993, responding
to the NRC request for additional information on the fuel failure event, and
had the following comments.
Examination of the I-assembly fuel rod withdrawal
data showed that many of the individual spacers had apparent spring forces
less than 0.20 pounds with many of the spring forces being 0.0 pounds.
There
was no evidence of fretting damage at many of the spacers with low recorded
spring forces.
On the other hand, there were spacers with spring forces of
1.43, 1~32, 0.69, and 0.59 pounds that gave minor eddy current indications,
and spacers with spring forces of 0.69, 0.68, 0.59, 0.48, 0.34, and 0.29
pounds that gave severe eddy current indications.
Eddy current indications
are indicative of fretting damage.
The average spring forces for all of the
spacers appear to confirm the licensee's analysis; however, individual spacer
data are contradictory.
The average Palisades rod withdrawal force relaxation versus burnup data
agrees with data from other SPC assemblies.
The team agreed with the
licensee's determination of a 90 percent relaxation of spring forces at end of
life as being conservative based on the data presented.
The data presented by the licensee indicates that fast flux exposure for a
fourth cycle on the reconstituted L-assemblies will not affect the performance
of the assemblies.
The licensee does not intend to use the reconstituted L-
assemblies past cycle 11.
The team concurred that fast flux exposure for a
fourth cycle will not adversely affect the L-assemblies.
Furthermore, the
preferential stamping of the L-assemblies grid spacers so that the cells will
not increase in size with exposure is a marked improvement for the
reconstituted L-assemblies over earlier assembly designs.
3.6.5
, Burnup Considerations
Initially,, the I-shielding assemblies were planned for three cycles {Cycles 9,
10, and 11) in the core periphery.
The I-assemblies would have been in the
core for a total of six cycles, taking into account the previous three cycles
of normal operation. During the Cycle 9 outage, the licensee examined the
whole core with ultrasonic testing {UT).
There were no leakers among the !-
assemblies.
During Cycle 10 operation, some I-assemblies were damaged by the
flow vibrational fretting. Thus, it is prudent practice that, after three
cycles of normal power operation, an assembly should not go beyond one more
cycle {the fourth cycle) when used for shielding. The licensee plans to use
the reconstituted L-assemblies for only one cycle, Cycle 11, and plans to
replace the reconstituted L-assemblies with a new design of shielding
assemblies for future reloads.
The selection criteria used for replacement candidates for the I-assemblies
included a target EOC 11 burnup for J and K assemblies of less than the BOC 10
burnup for assembly 1-024 {37,500 MWD/MTU).
The assemblies finally chosen
11
-*
were from the L-assemblies with an allowable maximum burnup limit of 46,000
MWD/MTU due to the improved grid spacer design.
The projected batch average
EOC 11 exposure for the L-assemblies is approximately 36,500 MWD/MTU (with a
range of 33,321 to 40,028). Thus, the team concluded that, based on burnup
considerations, the reconstituted L-assemblies were acceptable.
3.6.6
Fluence Reduction
The design criteria .for the revised Cycle 11 core loading configuration
includes a requirement that fluence rates not exceed the documented Cycle 9
values. According to the Palisades reactor engineering staff, the criteria
used by SPC to meet this requirement was that the bundle powers in the
peripheral core locations be less than or equal to the Cycle 9 core peripheral
bundle powers. A preliminary scoping analysis was then performed by Palisades
staff using their in-house DOT-IV model which confirmed that the corresponding
fast fluence (>1.0 Mev) values were less than the Cycle 9 values at the
critical weld orientations. The final fluence evaluation calculations *will be
performed by Westinghouse, to be consistent with the Cycle 9 results
previously furnished to the NRC.
3.7
Plant Review Committee (PRC) Review and Approval
The PRC reviewed and approved the 10 CFR 50.59 submittal on the modified core
reload plan on August 19, 1993.
The PRC review was observed by members of the
NRC team.
The Palisades team that prepared the 10 CFR 50.59 review were
questioned extensively by the PRC personnel prior to receiving approval.
.
Questions were raised about the chromium plating on the SAN-8 upper tie plate
alignment pin hole inside diameter, about the cycle 11 fuel load integrity,
about the plans to detect a low power fuel rod failure during cycle 11, about
the plans to determine if the shroud is interacting with fuel assemblies
during cycle 11, and numerous additional questions. All of the questions were
addressed adequately by the licensee staff.
3.8
Documentation of the Licensee's 50.59 Safety Evaluation
The 50.59 safety evaluation was documented according to Procedure No.3.07,
Rev. 7, "Safety Evaluations".
However, the audit team found the 50.59
evaluation package designation confusing in that the current revised Cycle 11
core was identified as FC-934, Rev. 0, with an associated safety evaluation
SE Rev. 1 whereas the original Cycle 11 core was also identified as FC-934,
Rev. 0, with an associated safety evaluation SE Rev. 0.
The licensee
indicated that the difference between *these Cycle 11 designs was documented in
an Engineering Design Change (EDC).
The existence of the associated EDC was
not mentioned in the current FC-934, Rev. 0 documentation for the reader to
realize that the current FC-934, Rev.a is different from the original Cycle 11
core design bearing.the same Item Identification Number.
The team reviewed the documentation of the 50.59 safety evaluation for its
technical depth and thoroughness and found it acceptable but improvements are
needed.
The licensee's 50.59 evaluation adequately documented the revised
Cycle 11 core physics design, the various assembly mechanical designs,
I-assembly replacement criteria, detailed L-assembly modifications, assembly
12
-*
reconstitution considerations, qualitative thermal hydraulic DNB
considerations, and the evaluation of the projected bowing from the revised
L-assemblies.
The documentation should have included or referenced the thermal hydraulic
analysis addressing the core average and subchannel flows as a result of the
new design, its impact on core barrel and assembly vibrations; the loose grid
spacer, spacer spring retention forces and its impact on fretting wear; and a
discussion of how the fluence reduction criteria are satisfied by the new
design. These issues were addressed in other documentation as discussed
above.
3.9
Conclusions
_The team completed its audit review of the licensee's 50.59 safety evaluation
for the revised Cycle 11 core design and concluded that the licensee followed
their procedure in performing the safety evaluation and the evaluation and its
associated documentation are acceptable.
The team also agreed with the
Palisades PRC conclusion that the revised Cycle 11 core design does not
constitute an unreviewed safety question and its approval of the revised
cycle 11 core design.
No violations, deviations, unresolved or inspector followup items were
identified.
4.0
Stuck Fuel Assembly Root Cause Evaluation
On July 6, 1993,, fuel assembly SAN-8 was inadvertently partially lifted with
the upper guide structure (UGS) from core position Z-11.
Two previous
incidents involving inadvertent lifting of a fuel assembly from core position*
Z-11 with the UGS were experienced in 1988 and 1992.
Initial followup of the
July 6, 1993, stuck fuel assembly was performed by the resident inspectors and
documented in inspection report No 50-255/93017(DRP). Subsequent followup of
the event was performed by an augmented inspection team (All) and documented
in inspection report No. 50-255/93018.
The All performed an extensive review
of the event including evaluation of the root cause for the stuck fuel
assembly.
At the conclusion of the All inspection the licensee had not
identified any single root cause for the stuck assembly; however, several
potential contributors had been identified.
The potential contributors to the lifting of the stuck fuel assembly with *the
UGS were:
Undersized upper tie plate pin holes in SAN-8 .
Deformation of core shroud creating an interference between the
UGS fuel alignment pins and the fuel assembly.
Fuel assembly bow .
13
UG5 fuel alignment pins loose or returned to a bent state during
operation.
UG5 fuel alignment pins out of position or the core support plate
alignment holes out of position.
Debris causing hang-up of the fuel assembly in the UG5 .
UG5 not level during lift, to the extent that there was
interference between the UG5 and the fuel assembly.
Core support barrel mis-located .
Damage to alignment pins or to the lower alignment plate lifting
or setting of UG5.
Degraded surface condition of the UG5 alignment pins at core
location Z-11 which could have promoted sticking within fuel
assembly upper tie plate holes.
Loss of preload on cap screws and alignment pins, which hold the
UG5 together, resulting in a significant loss of structural
rigidity.
The licensee developed an action plan to determine the role of each of the
potential contributor~, during the lifting-of the fuel assembly with the UG5;
As a result, the licensee identified 75 action items relating to the 11
potential contributors,. for further followup.
Followup inspection on the action items has resulted in the licensee's root
cause investigation team identifying the following:
UG5 fuel alignment plate pins bent in location Z-llN (1.56
degrees), Z-115 (0.41 degrees), and Z-165 (0.99 degrees).
A single gage of 0.995" diameter could not be placed on pins Z-llN
and Z-165, indicating the pins to be curved.
Analyses indicated that angular misalignment between UG5 fuel
alignment pins and fuel assembly upper tie plate of 1-1.5 degrees
was sufficient to lift fuel assemblies.
Qualification test results and analyses indicate that bent and
straightened pins have a gap between alignment pin* shoulder and
the UG5 lower alignment plate. Bent and *straightened pins are
less rigid and less resistant to bending than originally installed
pins.
14
--*
SAN-08 upper tie plate had two distinctive peen marks offset from
the center of upper tie plate holes by approximately 0.5 inch.
Peen marks are in the orientation and separated by about the
spacing of UGS fuel alignment plate pins.
SAN-08 video camera inspection in the spent fuel pool identified a
piece of debris stuck to the bottom of one foot. Debris could
have made it harder to fully seat SAN-08.
An object 0.05 inch
thick under an assembly foot would dislocate the upper tie plate
by approximately 0.5".
With only SAN-08 removed from location Z-11, the core support
plate was inspected with no debris identified. Several days
later, with five assemblies removed from around Z-11 location,
debris was observed.
Reactor head alignment pin at 0 degree location was incorrectly
installed.
\\
Measurement of UGS levelness on successive lifts indicates
significant variation of direction and magnitude of out-of-
levelness.
Analysis indicated assembly bow may contribute to angular
misalignment between.UGS alignment pins and upper tie plate.
Close up video camera inspection of UGS alignment pins at location
Z-11 indicated a greater degree of engagement in the upper tie
plate alignment holes.
Tip of UGS fuel alignment plate pin at location Z-16S was observed
to be distorted during video camera inspection.
The above findings resulted in the following potential contributors to a stuck
fuel assembly being eliminated:
UGS fuel alignment pins out of position or the core support plate
alignment holes out of position.
UGS fuel alignment pins loose or returned to a bent state during
operation.
Deformation of core shroud creating an interference between the
UGS fuel alignment pins and the fuel assembly.
Undersized upper tie plate holes in SAN-8 *
Loss of structural rigidity of the UGS .
Damage to the UGS fuel alignment plate .
15
-*
Debris between the fuel assembly upper tie plate holes and the UGS
fuel alignment plate pins.
Mis-location of core support barrel .
The remaining potential contributors have been categorized as to their
significance in contributing to the stuck bundle.
Potential contributors found highly likely to solely result in bent fuel
alignment pins and/or stuck fuel assemblies were:
Tilted/unlevel UGS .
UGS with bent fuel alignment pins .
Fuel assembly not properly positioned or seated .
Potential contributors having moderate magnitudes and probabilities and which
could combine with other contributors to bend fuel alignment pi~s and/or stick
fuel assemblies were:
Fuel assembly bow.
Fuel alignment pins. with degraded surface conditions.
- The licensee has taken the following actions to specifically address the five
remaining potential contributors identified above:
Replaced UGS fuel alignment plate pins Z-llN, Z-llS, and Z-16S.
Replaced the fuel assembly (SAN-08} upper tie plate for core
location Z-11.
The upper tie plate modified design reduces the
potential for interference between tie plate alignment holes and
UGS altgnment pins.
Implemented methods to evaluate fuel assembly elevations and
relative positions after reactor core reloads.
Modified equipment and revised procedures to assure UGS lift
rig/UGS levelness is established.
Modified procedures to assure UGS is level within acceptable
limits prior to lift of UGS.
In reviewing the licensee's root cause analysis, corrective actions to prevent
. reoccurrence, and. trial insertion/removal, the inspectors noted that the
licensee had expended significant resources to address the root cause for the
fuel assembly lifting. The licensee's root cause investigation team was
composed of highly dedicated individuals who utilized extensive problem
solving techniques.
Management was actively involved throughout the
investigation. Although the licensee had not completed all the long term
16
-*
actions, the actions taken were demonstrated adequate for cycle 11 operation.
The following long term actions were being evaluated by the licensee.
-
.
UGS levelness.
Replacement of RV head/UGS alignment pins.
Minimized fuel assembly bow.
Centering of crane.
Fuel assembly height/levelness/core verification.
Modification of upper tie plate for all fuel assemblies.
Camera/ljghtning/water clarity.
Dedicating specific load cell for UGS removal/insertion.
Future gauging of fuel alignment pins.
UGS key/keyway measurements.
Procedure upgrades, as. necessary.
No violations, deviations, unresolved, or inspection followup items were
identified in this area .
5.0
Observations of Activities
The team observed selected activities and interviewed licensee personnel to
evaluate the effectiveness of communications in the plant staff and workers
understanding of management expectations. Overall, the results were positive
with regards to employee readiness for operations.
At the time of the
inspection ~owever, there was limited activity in the plant; the outage work
was mostly completed and the.licensee was concentrating on finalizing the
efforts of the root cause assessment teams and final engineering restart
issues.
5.1
Operations
The team observed the conduct of operations personnel both inside and outside
of the main control room during major evolutions. This observation included
equipment testing, surveillantes and maintenance activities being conducted in
support of the outage.
Based upon these observations, the team concluded that
the-operations personnel demonstrated a ve~*good*awareness of plant
conditions and were able to communicate their knowledge through proper
coordination and control of plant activities in a safe manner. Additionally,
the operation's shift turnover activities were performed in a professional and
competent manner which ensured the appropriate transfer of information to
oncoming shift personnel .
17
Due to the low frequency of assigned work orders and surveillance tests in
progress, the team was unable to properly evaluate the adverse effects of
maintenance and testing activities on the operators' ability to control
support activities and maintain safe plant operations. However, a limited
number of activities that were observed appeared to be conducted in a
controlled manner which did not affect the operators' ability to perform
normal duties.
In general, procedures and administrative controls were in place to adequately
control and direct the safe startup and continued operation of the plant .. The
team reviewed selected operations procedures and controls (i.e. tagouts) to
verify the current plant conditions with existing procedural requirements with
minimal discrepancies noted. This verification included the system walkdown
of one safety system tagout and daily plant tours observing equipment status.
The team determined through interviews and observation of on-the-job
performance that operations personnel were knowledgeable and capable of
performing their licensed duties. Shift manning was maintained in accordance
with Technical Specifications at all times.
The team concluded that management actions to improve performance were in
place and.efforts undertaken to date were effective. The team interviewed
selected personnel in the management and non-management staff positions for
licensed operators.
In general, the non-management position operators felt
that management goals, directives and policies were adequately represented to
them but the importance of incorporating operator feedback into the
improvements was not clear. Management position operators felt that
management goals, directives and policies emphasized the professionalism with
which they conducted their jobs and were making improvements in plant
performance.
5.2
Maintenance
The team observed work in all the maintenance disciplines and saw that. work
orders and procedures were available~ adequate and followed, spare parts and
tools were proper and available, and that the knowledge and training of the
personnel involved was adequate for the job. Observations were made of shift
briefs, and pre-job briefs, which were thorough, with ~ood discussion.
Management expectations were made known.
Post-maintenance testing was also
observed and found to be appropriate. The test results met the acceptance
criteria and were properly documented and trended.
The team noted the
presen~e and involvement of first line supervisors and system engineers.
Interdepartment cooperation was good.
NPAD assessors were also present
occasionally.
In summary, in light of the limited activities in progress, the
maintenance observed was performed in an acceptable manner with adequate
resources and oversight, and with appropriate consideration for safety.
5.3
Engineering
The team discussed engineering efforts in progress and readiness for startup
with the engineering staff. For the issues discussed, the engineering
deliberations were thorough and conservative with regard to achieving complete
18
and final resolutions. Nuclear engineers were prepared for startup having
identified personnel assignments, training, and the procedural approach to
startup testing. It was noted however, at the time of the inspection, that
engineering was involved in resolving several significant operability issues
the licensee had identified. The issues were being appropriately identified,
discussed, and evaluated for corrective action.
5.4
Radiological Services
During the inspection, Radiological Services Department {RSD} performance
during the refueling outage was assessed and RSD plans for supporting start-up
activities evaluated.
In addition, the RSD plan for responding to the
presence of fuel in the primary coolant and corresponding systems, once start-
up activities begin, was reviewed. Conclusions drawn were based on interviews
with RSD management and involved personnel, observations of work activities in
radiologically controlled areas and reviews of pertinent documents.
The RSD participation in the outage had four specific elements, each of which
was assessed during the inspection.
RSD planning and scheduling activities both before and during the
outage were excellent. Jobs were performed on schedule and the
RSD had ample staff to provide coverage when needed.
Even after
the master schedule was changed, following the stuck control rod
incident and the discovery of a failed fuel pin, the RSD planners
were able to plan new work requests in a timely manner and work
with the schedulers to insure that critical jobs were not delayed .
RSD pre-job briefs needed improvement.
Interviews with
individuals who had attended pre-job briefs indicated that there
were numerous distractions {doors opening and closing, people
talking and phones ringing} in the areas were the briefs were held
and the quality of the presentation was dependent on the
technician giving the brief. Prior to the inspection, the*RSD had
, been made aware of the deficiencies and had taken steps to correct
them.
The RSD tried to find quieter areas to hold the briefs and
had developed a check-off sheet to be used by the presenter to
insure that all the relevant information was passed on to the
worker during the brief.
Radiation safety technician performance during the outage was
generally very good.
The technicians appeared to be technically
competent and well trained. There were, however, problems with
contractor technicians setting poor examples for other workers .. A
Nuclear Performance Asses.sment Department {NPAD} survei 11 ance
reported that a number of contractor technicians had used poor *
radiological practices. Those practices included improper
placement of dosimetry and wearing scrubs in areas where scrubs
were not allowed.
The poor practices were brought to the
attention of RSD management and immediate corrective action was
taken.
The surveill~nce concluded that, in general, technicians
19
did a good job of-keeping workers informed of radiological
conditions, providing advice about radiological conditions, and
controlling access to various radiological areas.
The inspectors
who had worked with the RSD technicians during the inspection
concurred with this conclusion.
Post-job briefs were held in accordance with station procedure and
appeared to be effective.
In conclusion, the RSD performance during the outage was effective. The one
deficiency identified during the outage, pre-job briefs, had been noted by the
RSD and corrective action had been taken.
Following the discovery of a broken fuel pin in the Reactor Cavity Tilt Pit,
the RSD developed a RSD Fuel Failure Response Plan to identify, track the
status of, and document RSD actions taken as result of the fuel failure.
The
plan was a living document and once completed would document the basis for the
program .wit~ regard to failed fuel.
The plan contained assumptions made about
the risks associated with failed fuel and detailed the organizational
structure for implementing the plan. Actions within the plan were assigned to
specific individuals and target dates were set for completing the assigned
tasks. Of the more than 56 action items identified in the plan a number were
directly related to RSD start-up activities and RSD plans for tracking data *
points to indicate failed fuel in the future.
Those actions directly related
to start-up activities included:
Review and revision of the whole body counter's (Fast Scan)
library to accommodate fuel material
Reevaluation of the frequency of surveys during start-up and
normal operations
Reviewed derived air concentration (DAC) calculation and skin dose
methodology to account for new fuel nuclide mix
Evaluation of TLD (Panasonic) algorithms for response to the
presence of fuel
Evaluati~n of PCM-lb (Personnel Contamination Monitor), PM-7
(Portal Monitor) and frisker energy distribution of calibration
sources
Tr~ining the-RSD staff in preparation for start-up activities
The whole body counter, PCM-lb and PM-7 and DAC evaluations had been performed
and completed.
The RSD decided to increase the frequency of surveys during
start-up activities, electronic dosimeters would be posted in critical areas
throughout the plant and the data collected every four hours during start-up.
Following start-up the electronic dosimeters would stay in place and the data
they supply tracked .
20
RSD plans for tracking data points for an indication of failed fuel included:
Part 61 samples were collected and sent to the licensee's vendor
laboratory for analysis. Future analyses would be tracked for the
presence of fuel.
The hot spot program was reevaluated to include the possibility of
finding fuel fragments.
Criteria for determining when to perform
gamma spectral analyses on newly discovered hot spots would be
developed.
That evaluation was due to be completed by October 31,
1993.
The licensee would review the data collected from personnel
contamination incidents, whole body counts, air samples, and
massilin survey smears to determine if it could be used in the
future to indicate fuel failure. That evaluated is due to be
completed by October 31, 1993.
In conclusion, the response plan was comprehensive in scope and provided a
good basis for integrating the RSD response to operational events during
start-up activities and implementing plans for tracking various parameters
during normal operations for the presence of failed fuel.
The RSD appeared
fully prepared to support the facility during start-up activities.
The Nuclear Performance Assessment Department performance during the outage
with regard to the RSD activities was also assessed.
One surveillance and a
number of Field Manito~ Reports were reviewed.
Following field assessment
activities, NPAD assessors issue Field Monitor Reports to report their
findings.
If deficiencies are noted during the assessment they are recorded
in the department's database and the NPAD director meets with the plant
manager once a week to discuss the previous week's findings.
If a deficiency
warrants management attention the assessor will normally issue a Deficiency
Report. A review of the weekly reports indicated that while many of the*
deficiencies identified during the assessments had been brought to
management'~ attention, none of them had been documented in the RSD deficiency
reporting system (radiological deficiency reports).
NPAD uses its database to record deficiencies and track corrective actions.
In principle the RSD is responsible for correcting its own deficiencies,
however, if those peficiencies are not reported in the RSD system NPAD assumes
that responsibility. For example, during th~ dutage NPAD conducted a
surveillance to assess health physics technician performance during backshift.
In general, the assessors found that the technicians had demonstrated good
performance, however, several technicians were*observed using poor
radiological practices.
In the surveillance, NPAD reported that the RSD had
been informed of the observed practices and had taken corrective action.
The
deficiencies, however, had not been documented in the RSD deficiency system
and plant management had not been given a copy of the surveillance.
Under
this system NPAD, not the RSD, had been responsible for insuring that the
deficiencies had been corrected and the actions taken documented.
This was a
weakness in the program .
21
In summary, the RSD performance during the outage was good to excellent. The
pre-job briefs needed improvement and steps were taken to require use of the
briefing check list. The RSD had developed a comprehensive plan in
preparation for start-up activities and the plan provided a good basis for
supporting those activities. The NPAD system for reporting deficiencies and
documenting corrective actions needed improvement.
5.5
Trial installation of the UGS
The licensee decided to perform a trial insertion and removal of the UGS to
demonstrate successful removal of the UGS without a fuel assembly.
This.
action was also used to demonstrate/identify the interaction of the UGS with
the fuel assemblies and the core support barrel by closely monitoring this
activity with underwater cameras.
The licensee performed the trial insertion
and removal on August 21 and August 22, 1993, respectively.
The inspectors
observed both the insertion and removal activities. The observations are
addressed below.
The inspectors observed the upper guide structure {UGS) set onto the core on
August 21, 1993, and also the lift from the core on August 22, 1993.
Both
activities were performed in accordance with the applicable sections of
procedure RVI-M-12, "Final 1993 Installation of Upper Guide Structure."
The inspector attended the pre-job brief for both activities. The briefs were
comprehensive, and covered the procedure steps that were to be performed in
detail. The inspector verified that all personnel with assigned
responsibilities were present .
Proper radiological protective meaiures were established for personnel once
inside containment.
Proper dosimetry and protective clothing were worn by
each indivjdual, and coverage by the radiation protection technician assigned
to monitor the job was good.
The inspector observed that proper communications were established ,between the
control room and the senior reactor operator/shift supervisor directi.ng the
evolutions from containment.
Cameras, lights, video equipment, and other
measuring devices were properly aligned and _staged.
The actual lift evolution was satisfactory. The UGS was lifted to the six
inch elevation above the t6p of the fuel assemblies and hel~ for data
gathering and visual inspection. During the lift, the fuel alignment plate.~n
the bottom of the UGS was monitored for attached fuel assemblies and none were
observed.
Particular attention was paid to core location Z-11.
The lift of
the UGS was smooth and no swaying or tilting of the UGS was observed.
Levelness of the UGS was checked acceptable.
The load cell indicated that the
UGS was within its expected weight.
The UGS*was then lifted to the three foot elevation and a thorough camera
inspection was performed to verify that there were no attached fuel
assemblies.
Following this inspection the UGS was transferred to its storage
location and set on its pads with no apparent problems .
22
No violations, deviations, unresolved or inspector followup-items were
identified .
6.0
Evaluation of the Fuel Lost from the 1-24 Assembly
The Palisades Nuclear Power Plant experienced a fuel failure in assembly 1-24
at core location 819 during cycle 10 of operation. This failure resulted in
significant fuel loss from one fuel rod to the primary system during operation
and additional fuel loss from this rod during handling of the 1-24 assembly
during the outage.
The purposes of this inspection were to 1) evaluate the
licensee's efforts to find the missing -900 g of U02 from the failed rod, 2)
evaluate the loose parts in the primary coolant, and 3) evaluate the
licensee's ability to detect fuel failures during cycle 11 operation ..
6.1
Search For Missing Fuel
Higher than normal activities were found on primary system and core components
but these activities account for only 6 to 7% of the total fuel lost frQm
assembly 1-24. The core was examined to the maximum extent possible without a
full core off-load.
The reactor cavity tilt pit was where the failed fuel rod was stripped from
the assembly during the fuel handling operations. However, a significant
quantity of the fuel may have been lost from the fuel rod prior to its
transport to the tilt pit. The bottom of the tilt pit was examined, however,
a thorough visual examination was not possible because of hoses and machinery
at the bottom of the pit. The licensee intends to drain the pit to a lower
water level and see if they can find fuel fragments with activity detectors.
The tilt pit was vacuumed and drained after the rod was stripped from the
assembly in the tilt pit. A survey of the filters following this first
vacuuming of the pit resulted in higher than normal activity levels but only
accounted for less than 1% of the lost fuel.
The majority of the piping
{>90%) that leads from this drain was surveyed and while some small increases
in activity were noted they were not at the high levels expected if large
fragments ~f fuel were present. However, the less than 10% of piping not
surveyed was located in concrete nearest the drain location.
From the piping
surveys, and surveys of the filters from three separate vacuuming efforts on
the bottom of the tilt pit, less than 1% of the total fuel loss had been
accounted for in the tilt pit.
The team concluded that it was not likely that the licensee would find
significant additional quantities of the missing fuel.
The fuel appeared to
be efther hiding in inaccessible locations in the primary system, the drains
of the reactor cavity tilt pit and the adjacent spent fuel pool, or, more
likely, a combination of the above.
6.2
Loose Parts in the Primary Coolant
There are parts missing from the failed fuel assembly that are either in the
primary system or in the reactor cavity tilt pit.
Among the missing parts are
1) a half diameter piece of cladding approximately 20-inches in length,
2) three to four pieces of the spacer grid including a lantern spring {each is
23
estimated to be less than an inch in length and less than one-half inch in
width), and 3) an insulator disk (approximately the size of a fuel pellet) .
These loose parts have the potential of causing further fuel failures due to
debris fretting if they reside in the primary system.
They may also contrib-
ute to damage of other primary system components.
However, historically,
debris in the primary system has primarily been a fuel failure problem rather
than significantly impacting other components.
It should be noted that of
201 fuel assemblies in the cycle II core, 136 assemblies are debris resistant
assemblies by design.
6.3
Ability to Detect Fuel Failures in Cycle II
If a significant quantity of the fuel missing from the failed rod (>10%)
resides in the reactor coolant system, the ability to detect further fuel
failures will be hampered.
The licensee had significantly improved their
failed fuel detection capabilities for cycle II operation.
They expanded
their activity measurements to include additional radioactive isotopes that
help in identifying fuel failures with high tramp (background) activity
levels. The licensee has also improved their analytical capabilities.
Even
with these increased measurements and capabilities the licensee will most
likely have difficulty in detecting any small failures from a small number of
failed rods; however, they will be able to detect fuel failures with large
defect sizes such as those experienced in cycles 9 and 10.
These were well
below Technical Specification limits.
The licensee had an action plan for addressing increased coolant activity
levels. However, this plan left out some of the actions recommended in EPRI
report EPRl-NP5521.
In addition, the highest activity level that can be
achieved before the licensee will consider shut down, or derating, was close
to the Technical Specification limit for coolant activity. These action
levels were being reevaluated by the licensee.
No violations, deviations, unresolved or inspector followup items were
identified in this area.
7.0
AIT Report Findings Compliance Review
A review was conducted to evaluate findings of the NRC Augmented Inspection
Team (All), as documented in Inspection Report 50-255/93018(DRS)~ against
applicable regulatory requirements.
The review identified a number of
examples of noncompliance with requirements. These are discussed further
below, organized by their applicability to the three broad conclusions reached
by the All.
7.1
The Licensee's Organization Had a Less than Questioning Attitude
Operating Cycles 10 and 11 involved the use of previously burned fuel
assemblies (from the I-series fuel) to provide reactor vessel neutron flux
reduction shielding. At the beginning of Cycle IO, the I-series assemblies
had operated three full fuel cycles. The licensee had little or no experience
operating fuel assemblies beyond three cycles. A review of the proposed
extended use of I-series bundles was performed pursuant to 10 CFR 50.59. This
24
review was required to fully consider whether the proposed action involved any
Unreviewed Safety Questions (USQ).
The conditions of the proposed use
involved placing the assemblies at the periphery of the reactor core, where
the neutron exposure spectrum (and other parameters) were different from other
core locations.
In particular, the neutron spectrum at the periphery of the
core was proportionally higher in "fast" neutrons and lower in "thermal"
neutrons compared to non-periphery locations. A full consideration of the
potential that a USQ might be involved required that the effects of the unique
neutron spectrum on the I-series assemblies be specifically evaluated.
The licensee's documented 50.59 analyses did not include consideration of the
effects of the unique neutron spectrum on the I-series assemblies.
Those
analyses were therefore incomplete.
The written safety evaluation for the
subject use of the I-series assemblies did not provide complete bases for the
determination that there was no unreviewed safety question.
This is an
- apparent violation of 10 CFR 50.59.(b)(l}, which requires such written bases
(Violation 50-255/93020-02).
During Cycle 10, it became apparent that fission products and transuranics
were present in the primary coolant.
The licensee did not have or develop
procedures to evaluate this condition with such scrutiny as. to correctly
identify the root cause.
The cause was ascribed to "tramp" uranium/fuel left
in the coolant from fuel cladding failures during the previous cycle, Cycl~ 9.
This was not correct.
As a consequence of failure to identify that a low power peripheral fuel rod
had failed in service, the fuel was not inspected for damage after Cycle 10 .
The damage to fuel assembly 1-24 was not detected, and the assembly was
returned to the reactor for Cycle 11.
Utili~ation of primary coolant radiochemistry testing procedures which did no~
assure that reactor fuel was performing properly in service, is an apparent
violation of 10 CFR 50, Appendix B, Criterion XI, "Test Control"
(Violatitin 50-i55/93020-03a).
7.2
Conservative Decisions were not Made When Warranted
As noted above, the I-series fuel assemblies were not inspected between
Cycles 10 and 11.
This was despite radio-chemistry evidence of fuel in the
primary coolant system and despite the fact they were the oldest assemblies in
the reactor.
Furthermore, the I-series assemblies were known to have been
fabricated with grid straps which were susceptible to relaxation due to strap
growth from prolonged exposure to radiation. This potential relaxation would
impose a risk of fuel pin vibration and fretting against the grid strap, which
could damage the cladding of the pin. This is potentially what occurred.
Failure to verify by test or inspection that I-series fuel assemblies would
perform acceptably in service after five previous cycles in the reactor is
another apparent violation of 10 CFR 50, Appendix B, Criterion XI, "Test
Control" (Violation 50-255/93020-03b) .
25
When the Upper Guide Structure (UGS) was lifted from the reactor on July 6,
1993, and a fuel bundle from core location Z-11 stuck onto the UGS, it marked
the third occurrence of this problem.
Root cause analyses and corrective
actions for the previous events failed to prevent recurrence.
Following the
event of September 3, 1988, no corrective actions were taken with respect to
the alignment pins. After the event recurred on February 29, 1992, the
alignment pins at location Z-11 were determined to be bent and they were
straightened.
Procedure changes were made to provide protection for the
alignment pins during handling of the UGS, so they would not become bent
again.
These changes failed. After the event of July 6, 1993, the alignment
pins were found to be bent once more.
The pins have now been replaced and a
series of actions taken to eliminate more potential causes of pin damage, as
well as other potential causes of interference between the UGS and the fuel.
Failure to preclude repetition of interference between core components, which
caused fuel mishandling events, is an apparent violation of 10 CFR 50,
Appendix 8, Criterion XVI, "Corrective Action" (Violation 50-255/93020-04).
7.3
Management Expectations were not Effectively Applied
Controls which were developed for the lift of th~ Upper Guide Structure (UGS)
were not positively and effectively applied.
The load cell equipment used for
the lift on July 6, 1993, was not the equipment specified by the applicable
procedure, No. RVI-M-1, Revision 16, "Removal and storage of the Upper Guide
Structure."
The procedure specified a Tl-2000 load cell but a J-300 device was actually
used.
A procedure stipulation (Section 5.3.6.g) to follow Work Order No.
24301781 for steps to use the load cell was violated in that the readout
device was not zeroed as required by step 3.3.A.7.
In addition, the specified
upper load limit of 62,000 pounds (Section 5.3.14) was exceeded by the
indicated load of 62,800 pounds after the UGS was raised about six inches. A
licensee supervisor was present during this evolution and did not intercede
effectively to enforce compliance.
Failures to implement requirements of the UGS lift procedure as described
. above are apparent examples of violations of Technical Specification 6.8.1.b
(Violation 50-255/93020-05a).
The activities associated with recovery of the fuel assembly which stuck to
the UGS on July 6, 1993, though successful in safely returning the assembly to
the reactor without damage, were not positively controlled in all respects.
Specifically, the RWP and procedure FHS0-18, "Recovery of Bundle SAN-8" were
not properly implemented.
The limitations of procedure FHS0-18 steps 4.2.6
and 5.2.1, in instructing that chainfall tension be limited to a combined load
of 1500-1600 pounds, were violated.
The chainfalls were tightened to a
combined load of 2300 pounds.
Failure to implement requirements of the stuck fuel assembly recovery
procedures as described above are an additional apparent example of violation
of Technical Specification 6.8.1.b (Violation 50-255/93020-05b) .
26
Seven apparent violations, some involving more than one example, were
identified.
No deviations, unresolved items, or inspector followup items were
identified.
8.0
Public Meeting on September 9. 1993
A public meeting was held between managers of Consumers Power Company and the
NRC on September 9, 1993.
The meeting was held in the Holiday Inn, Route 94,
in Benton Harbor, Michigan.
The NRC was represented by Hubert J. Miller,
Deputy Regional Administrator of Region III and members of his staff, and
James G. Partlow, Associate Director for Projects for the Office of Nuclear
Reactor Regulation, and members of his staff. Consumers Power Company was
represented by David Hoffman, Vice President for Nuclear Operations, Gerald
Slade, General Manager of the Palisades Plant, and members of his staff. The
purpose of the meeting was for the licensee to present the results of the root
cause assessments done for the damaged fuel bundle 1-24 and the fuel bundle
lifted with the UGS.
Mr. Miller opened the meeting, the licensee made the
presentation, then Mr. Miller closed the meeting and responded to questions
from the public.
Licensee representatives also remained to respond to
questions.
9.0
Exit Meetings
Interim exit meetings were conducted with licensee representatives* on
August 21, 1993, and August 27, 1993. A final exit was conducted
September 29, 1993.
The first exit was by the team of inspectors reviewing
the cycle 11 core reload safety evaluation.
The second exit interim was by
the team which observed routine activities. The final exit summarized the
results of the NRC review of the AIT findings for regulatory issues. The
inspectors summarized the scope and findings of the inspection.
Tbe licensee
acknowledged the statements made by the inspectors,
The inspectors also
discussed the likely informational content of the inspection report with
regards to documents and processes reviewed by the inspectors during the
inspection and the licensee did not identify any such documents or processes
as proprietary .
27
SIOEO
1
2
3
4
5
8
7
8
9
10
11
12
13
14
15
A
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G
FUEL ASSEMBLY ROD LOAD HAP
CYCLE 11 L-ASSEHBLY DESIGN
PALISADES
SIOEA
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EXAMPLE PATTERN SHOWN FOR THE LOADING OF 14 STAINLESS STEEL ROOS INTO AN L
ASSEMBLY BASED ON ASSEMBLY TOP VIEW .
SYMMETRIC PATTERN USED FOR All 16 l ASSEMBLIES WITH THE 8 STAINLESS STEEL ROD
CORNER ALWAYS POSITIONED TO BE AT THE CORE SHROUD CORNER.
Figure 1 - Revised L~assembl y for Cycle _11 *
Figure i
Revised
CYCLE II CORE PLAN
PALISADES NUCLEAR PLANT
N
A BCD EFG
AST VWX
Z
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4
5
6
7
8
9
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19
20
21
22
23
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21
22
23
Batch 0: New Fuel 14 Batch N: Once Burnt rt SAN: Once Burnt
Batch M: Twice Burnt 8 Batch L: Three Times"Burnt
SAN assemblies contain Stainless Steel pins
Cycle 11 Core Plan uses ~ core rotational symtnetry
Figure 2 - Revised Cycle 11 Core Design
E