Part 21 Report - 1998-390
ACCESSION #: 9803190378
LICENSEE EVENT REPORT (LER)
FACILITY NAME: R. E. Ginna Nuclear Power Plant PAGE: 1 OF 7
DOCKET NUMBER: 05000244
EVENT DATE: 02/09/98 LER #: 1998-001-00 REPORT DATE: 03/11/98
OTHER FACILITIES INVOLVED: DOCKET NO: 05000
OPERATING MODE: 1 POWER LEVEL: 100
THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR
LICENSEE CONTACT FOR THIS LER:
NAME: John T. St. Martin - Technical
Assistant TELEPHONE: (716) 771-3641
COMPONENT FAILURE DESCRIPTION:
CAUSE: B SYSTEM: DA COMPONENT: RK MANUFACTURER: B386
REPORTABLE NPRDS: N
SUPPLEMENTAL REPORT EXPECTED: NO
On February 9, 1998, the plant was in Mode 1 at approximately 100% steady
state reactor power. It was discovered that Boraflex degradation in the
Spent Fuel Pool, greater than was assumed in the criticality analysis,
had occurred. This is reportable under 10 CFR 50.73 and 10 CFR 21.
Interim corrective action included removing spent fuel assemblies from
selected degraded storage rack cells and maintaining a high concentration
of soluble boron in the Spent Fuel Pool.
The cause of the Boraflex degradation was high cumulative gamma radiation
exposure and the subsequent dissolution of the boron from the Boraflex
Corrective action to prevent recurrence is outlined in Section V.B.
END OF ABSTRACT
TEXT PAGE 2 OF 7
I. PRE-EVENT PLANT CONDITIONS:
On February 4, 1998, the plant was in Mode 1 at approximately 100%
steady state reactor power. In activities unrelated to plant
conditions, RG&E and contractor personnel began performing "Boron-10
Areal Density Gauge for Evaluating Racks" ("BADGER") testing for
spent fuel storage racks containing Boraflex panels.
Each cell of the Ginna Station Region 2 spent fuel storage racks
contains two Boraflex sheets (panels). Each cell can store one fuel
assembly. These Boraflex panels are 144 inches in length and are
positioned adjacent to the stored fuel assemblies, sandwiched
between stainless steel sheets. The design is such that water
exchange may occur both at the panel edges and up through the
Boraflex region of the rack assembly.
NRC Generic Letter (GL) 96-04 (Boraflex Degradation in Spent Fuel
Pool Storage Racks) was issued June 26, 1996. GL 96-04 requested
that licensees assess the capability of the Boraflex to maintain a
5-percent subcriticality margin and submit to the NRC a plan
describing its proposed actions if this subcriticality margin cannot
be maintained by Boraflex material because of current or projected
future Boraflex degradation. As stated in the Generic Letter,
Boraflex dissolution appears to be a gradual and localized effect
forewarned by relatively high silica levels in the pool water.
Rochester Gas and Electric (RG&E) responded to this GL in a letter
dated October 24, 1996. (Refer to Ginna Docket No. 50-244, letter
dated October 24, 1996, from R.C. Mecredy (RG&E) to USNRC, "Response
to NRC Generic Letter 96-04".) In this response, RG&E committed to
perform blackness testing on selected Boraflex panels to obtain data
on the physical condition of the Boraflex panels in 1997. RG&E
subsequently revised the commitment date to the first quarter of
1998, in a letter dated December 22, 1997. This was done primarily
due to the availability of personnel and equipment from Northeast
Technology Corporation (NETCO), who were contracted to perform the
more rigorous "BADGER" testing, which RG&E felt was more appropriate
than blackness testing. (Refer to Ginna Docket No. 50-244, letter
dated December 22, 1997 from R.C. Mecredy to USNRC, "Revision to
Blackness Testing Schedule Per GL 96-04".)
RG&E contracted with NETCO to perform "BADGER" testing, and
determined which 24 Boraflex panels to perform this "BADGER" testing
on. These panels were selected based on a representative cumulative
gamma radiation exposure, ranging from a low of 9.9 E+8 rads to a
high of 4.19 E+9 rads. Testing started on February 4, 1998.
Initial results were reported to RG&E on February 9, 1998.
II. DESCRIPTION OF EVENT:
A. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES:
o February 9, 1998: Event date.
o February 9, 1998, 1445 EST: Discovery date and time.
o February 9, 1998, 1528 EST: NRC is notified of this
condition per 10CFR50.72 (b) (1) (ii) (B).
TEXT PAGE 3 OF 7
On February 9, 1998, the plant was in Mode 1 at approximately
100% steady state reactor power. In activities unrelated to
plant conditions, the preliminary results of "BADGER" testing
were reported to RG&E personnel.
During Spent Fuel Pool "BADGER" testing, degradation beyond the
four (4) inch gap assumption of the criticality analysis was
noted on selected boraflex panels. This data indicates that
some panels have undergone dissolution beyond expected levels.
One panel had experienced up to 100 inches of dissolution.
This is considerably more than was previously identified by
The Ginna Station Updated Final Safety Analysis Report (UFSAR),
Section 18.104.22.168.12, states: "Fuel storage racks using nuclear
poisons additional to those inherent in the structural
materials shall be designed and fabricated in a manner to
prevent inadvertent removal of the additional poison by
mechanical or chemical action."
Specific results of the "BADGER" testing were as follows:
o 16 panels with cumulative gamma exposures greater than or
equal to 2.47 E + 9 rads had different degrees of
degradation. 11 of these panels showed small degrees of
dissolution around gaps and panel edges. Two other panels
showed edge dissolution in the lower 60 inches of the
panel. The remaining three panels had gaps and
dissolution ranging from 20 inches to 100 inches, which
exceeded the assumptions of the criticality analysis of
o Eight panels had cumulative gamma exposures lower than
2.47 E + 9 rads. These panels showed either uniform boron
content or had only slight dissolution along the edges.
o Preliminary assessment indicates that up to 184 storage
rack cells may have one or more adjacent panels with
cumulative gamma exposures above 2.4 E+9 rads.
The Plant Operations Review Committee (PORC) met on February
10, 1998, to review these results. PORC directed that
administrative controls be established to maintain a high
concentration of soluble boron in the Spent Fuel Pool (SFP).
C. INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT
TO THE EVENT:
D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:
TEXT PAGE 4 OF 7
E. METHOD OF DISCOVERY:
This event was self-identified by RG&E personnel after review
of the initial data from the "BADGER" testing, as provided by
F. OPERATOR ACTION:
Primary Systems Engineering personnel notified Operations
supervision, who notified the Control Room operators. The NRC
Senior Resident was notified at this time. At approximately
1445 EST on February 9, 1998, plant staff determined that a
non-emergency one hour notification, per 10CFR50.72
(b)(1)(ii)(B), should be made to the NRC Operations Center.
The Shift Supervisor made this notification at approximately
1528 EST on February 9, 1998.
G. SAFETY SYSTEM RESPONSES:
III. CAUSE OF EVENT:
A. IMMEDIATE CAUSE:
The immediate cause of the plant being in an unanalyzed
condition was the Boraflex degradation identified by the
B. INTERMEDIATE CAUSE:
The intermediate cause of the Boraflex degradation was
dissolution of the boron from the Boraflex matrix.
C. ROOT CAUSE:
The underlying cause of the dissolution of boron from the
Boraflex matrix is attributed to a high cumulative gamma
radiation exposure, aggravated by washout-accelerated
dissolution of the Boraflex, caused by pool water flow through
the panel enclosures.
TEXT PAGE 5 OF 7
When Boraflex is subjected to gamma radiation in the pool
aqueous environment, the silicon polymer matrix becomes
degraded and silica filler and boron carbide are released.
Since irradiated Boraflex typically contains 46 percent of
crystalline silica and 50 percent of boron carbide, in a 4
percent silicone rubber matrix (polydimethyl siloxane polymer),
the presence of silica in the pool indicates the likely
depletion of boron carbide from Boraflex. The loss of boron
carbide is characterized by slow dissolution of the Boraflex
matrix from the surface of the Boraflex and a gradual thinning
of the material.
The rate of silica release from Boraflex is influenced by the
water exchange to and around the Boraflex panels. The rate of
dissolution also increases with higher pool temperature and
gamma exposure. Experimental data indicates that once silica
reaches an equilibrium value, the rate of dissolution is
dramatically reduced. An increase in pool water flow past the
Boraflex panels can disturb any localized silica equilibria.
This event is NUREG-1022 Cause Code (B), "Design,
Manufacturing, Construction / Installation".
IV. ANALYSIS OF EVENT:
This event is reportable in accordance with 10 CFR 21 and in
accordance with 10 CFR 50.73, Licensee Event Report System, item (a)
(2) (ii) (A), which requires a report of, "Any event or condition
... that resulted in the nuclear power plant being ... In an
unanalyzed condition that significantly compromised plant safety".
The amount of Boraflex degradation is greater than that assumed in
the criticality analysis, which placed the SFP in an unanalyzed
An assessment was performed considering both the safety consequences
and implications of this event with the following results and
There were no operational or safety consequences or
implications attributed to the Boraflex degradation because:
o RG&E had been maintaining a high concentration, greater
than 2300 parts per million (PPM), of soluble boron in the
Spent Fuel Pool (SFP). This value is being monitored
o Calculations show that 1450 PPM of soluble boron is
required to compensate for a complete absence of Boraflex
in all the panels in Region 2, while maintaining the
reactivity condition K sub eff < 0.95 under all postulated
off-normal conditions (i.e., fuel misload accident).
o There are no credible sources of boron dilution that would
be expected to decrease SFP boron concentration below the
required 1450 PPM.
TEXT PAGE 6 OF 7
o RG&E is developing strategies to minimize or eliminate the
need to credit Boraflex in Region 2 of the SFP. There are
several possible options being considered or implemented:
a. Region 1 of the SFP will be re-racked with new
storage racks that incorporate borated stainless
steel. This work is scheduled for completion prior
to the 1 999 refueling outage. The Region 1 storage
racks do not contain Boraflex, and the re-rack will
provide additional storage space in non-Boraflex rack
b. Licensing actions to obtain credit for soluble boron
in the SFP may be pursued in accordance with
NRC-accepted methodology. This credit could justify
analyses to assure that the fuel would remain
subcritical, even if this boron were replaced with
pure water. If pursued, this will require a separate
c. Licensing actions to obtain credit for neutron
absorber material (control rods, absorber rodlets,
and/or absorber panels) may be pursued. Such
absorber material could be strategically placed in
Region 2 locations to support the criteria outlined
above. If pursued, this would also require a license
d. More restrictive storage patterns can be utilized in
Region 2. If pursued, this would also require a
license amendment request.
Based on the above, it can be concluded that the public's health and
safety was assured at all times.
V. CORRECTIVE ACTION:
A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL
o Procedures have been changed to ensure administrative
controls are in place to verify at least 2300 PPM of
soluble boron is maintained in the SFP.
o Spent fuel assemblies were removed from selected degraded
storage rack cells, so that for Boraflex panels with
cumulative gamma radiation exposure greater than 2.47 E +
9 rads, the configuration is bounded by the current
criticality analysis of record.
TEXT PAGE 7 OF 7
B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:
o Administrative controls will be established to prevent
storage of spent fuel assemblies in designated cells, as
determined by the Reactor Engineer or designee, to ensure
the configuration is bounded by the criticality analysis.
o Long term strategies will be pursued that do not credit
Boraflex in Region 2 of the SFP.
VI. ADDITIONAL INFORMATION:
A. FAILED COMPONENTS:
The Boraflex panels in the Ginna Station SFP were manufactured
by Brand Industrial Services Corporation (BISCO), and have been
in the SFP since 1984.
B. PREVIOUS LERs ON SIMILAR EVENTS:
A similar LER historical search was conducted with the
following results: No documentation of similar LER events with
the same root cause at Ginna Station could be identified.
C. SPECIAL COMMENTS:
Due to the amount of Boraflex degradation, the industry will be
notified of this event via Nuclear NETWORK.
*** END OF DOCUMENT ***
Page Last Reviewed/Updated Friday, January 31, 2020