ML033640026
| ML033640026 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 12/29/2003 |
| From: | Mallett B Division of Nuclear Materials Safety IV |
| To: | Hinnenkamp P Entergy Operations |
| References | |
| EA-03-077, IR-02-007 | |
| Download: ML033640026 (15) | |
See also: IR 05000458/2002007
Text
December 29, 2003
Paul D. Hinnenkamp
Vice President - Operations
River Bend Station
Entergy Operations, Inc.
P.O. Box 220
St. Francisville, LA 70775
SUBJECT: RIVER BEND STATION - FINAL SIGNIFICANCE DETERMINATION FOR A
WHITE FINDING AND NOTICE OF VIOLATION (NRC INSPECTION
REPORT 05000458/2002007)
Dear Mr. Hinnenkamp:
The purpose of this letter is to provide you the final results of our significance determination of
the preliminary White finding identified in the subject inspection report. The inspection finding
was assessed using the Significance Determination Process and was preliminarily
characterized as White, a finding with low to moderate increased importance to safety that may
require additional NRC inspections. This White finding involved a failure to properly lock open
River Bend Station Condensate Prefilter Vessel Bypass Flow Control Valve CNM-FCV200 in
May 2002. This performance deficiency resulted in a loss of feedwater flow to the reactor on
September 18, 2002, when Valve CNM-FCV200 unexpectedly closed following a reactor scram.
At your request, a Regulatory Conference was held on June 23, 2003, to further discuss your
evaluation of this issue. During the meeting, your staff acknowledged the performance
deficiency and described your assessment of the risk significance of the finding. In a
supplemental letter dated July 9, 2003, you provided additional information regarding your risk
evaluation of this event. In your July 9, 2003, letter, you agreed that the failure to control the
position and properly lock Valve CNM-FCV200 was a performance deficiency and a violation of
your Technical Specifications; however, you took exception to certain aspects of NRCs
evaluation of risk associated with this event. After considering all of the information available,
as explained further in the attached enclosures, the NRC has concluded that the finding is
appropriately characterized as White.
In the supplemental information provided on July 9, 2003, you restated your assertion,
presented during the Regulatory Conference, that the risk associated with this event would be
very low, making this a finding characterized at a Green level. This assertion was based on
your belief that it is inappropriate for NRC to use the Individual Plant Examination of External
Events (IPEEE), in concert with "best effort" estimations, for the purpose of determining risk for
inspection findings in todays regulatory environment without more detailed analyses to improve
precision.
Entergy Operations, Inc. -2-
Specifically, you asserted that the overall risk associated with this event, including the change in
core damage frequency from fire, was very low because: (1) the safety systems in the plant
were functional, including the control rod drive system, which would have provided a high
pressure injection source after the first 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; (2) Valve CNM-FCV200 would have failed only
during a plant scram and not during a controlled manual shutdown, as evidenced by the July
2002 plant shutdown; and (3) the fire risk from a fire area is nonexistent for evaluation of this
event, if there is no plant scram caused by a fire in that area. While we took into account the
first two considerations in our independent assessment of the risk of this event, we disagree
with your assertion that fire risk from a fire area is nonexistent for the evaluation of this event.
The basis for our position is discussed in greater detail below and in Enclosure 2 to this letter.
In your supplemental response, you indicated that the NRCs use of your IPEEE results,
together with best effort estimations, was not appropriate for the purpose of evaluating the risk
of inspection findings. However, NRC Manual Chapter 0609, Significance Determination
Process, Appendix A, Attachment 1, step 2.5, Screening for the Potential Risk Contribution
Due to External Initiating Events, states that the impact of external initiators should be
evaluated and could increase the risk significance of a finding by as much as one order of
magnitude. Step 2.5 also states that the evaluation may be qualitative or quantitative in nature.
Qualitative evaluations of external events should, as a minimum, provide the logic and basis for
conclusion and should reference all the documents reviewed. The NRC has qualitatively
assessed the significance of the external events contribution to the risk of this finding.
Additionally, quantitative methods used by the staff indicate that external factors would increase
the risk significance of the subject finding by at least a factor of two over risk caused by internal
initiators alone. More detail regarding our evaluation is contained in Enclosure 2 to this letter.
Your supplemental response indicated that it is inappropriate to import the results of the IPEEE
screening method into the Significance Determination Process without fully appreciating the
context in which they were developed. We agree that IPEEE data should be used carefully and
that importing results directly from the IPEEE for those items that were screened in the process
would result in significant overestimation of the risk. However, the results of the IPEEE were
not directly imported for use in our preliminary evaluation. We reviewed your IPEEE to identify
those fire areas in which feedwater was important to risk. In evaluating the change in risk from
fire initiators in those 18 areas, our preliminary evaluation utilized your model of record to obtain
quantitative results as opposed to directly importing the results from the IPEEE evaluation.
Additionally, while industry and NRC tools for evaluating the risk associated with external
initiators are not fully developed, the Significance Determination Process requires that we
evaluate the total risk associated with a finding using the best available information.
For the risk determination under consideration, you contend that the external event contribution
from various potential risk initiators, as well as numerous specific areas within the plant,
screened out as being insignificant based on the IPEEE screening criteria. As a result, you
believe these potential risk initiators and areas should not be used to adjust the risk of a
specific internal event, such as the one in question. We have determined that the contribution
to risk of selected external events, such as high winds, tornados, and hurricanes; transportation
hazards; severe weather storms; and lightning, should not be excluded from consideration
simply because they screened out during the initial development of your IPEEE.
Entergy Operations, Inc. -3-
In your assessment of the risk of the subject event, your staff chose to refine the assumptions
used in the IPEEE for the fire areas that we specifically evaluated in our preliminary
assessment. Your stated purpose was to demonstrate that the original screening criteria were
correct and that these events should be screened as not significant to the risk analysis for this
event. While we agree that increasing the precision of the analyses is appropriate, we have
concluded, as described in our preliminary risk assessment, that the affected areas still
contribute to increasing the overall risk of the event as described in our preliminary risk
assessment. Additionally, your analysis of the impact of fires within the plant did not fully
analyze the potential for fires to cause indirect reactor scrams. Your analysis used
assumptions for fire severity factors, fire sizing, ignition frequencies, and fire modeling that
were not fully supported by the information provided. Also, the increased risk to the plant from
increased probability of human error as a result of the fires was not evaluated. Of the fire areas
at the River Bend Station, only 32 were actually analyzed. The remaining areas were either
quantified using generic industry data, assuming similarities to the 32 analyzed or, in one
instance, was not assessed. Therefore, we conclude that you have provided an insufficient
basis for determining that the increase in risk associated with fires was insignificant. A more
detailed description of our evaluation of your risk assessment is included in Enclosure 2 to this
letter.
After considering the information developed during the inspection, the information presented at
the regulatory conference on June 23, 2003, and the additional information you provided in your
letter dated July 9, 2003, the NRC has concluded that the risk significance of the subject
inspection finding should be based on our preliminary risk assessment further supported by our
assessment described in Enclosure 2. The assessment in Enclosure 2 is intended to address
each of the points presented by Entergy Operations, Inc. during the regulatory conference and
provide our position that those points did not provide a basis for concluding that the issue
should be characterized as Green. Accordingly, NRC has concluded that the finding is
appropriately characterized as White, an issue with low to moderate increased importance to
risk, which may require additional NRC inspections or other NRC actions.
You have 30 calendar days from the date of this letter to appeal the staffs determination of
significance for the identified White finding. Such appeals will be considered to have merit only
if they meet the criteria given in NRC Inspection Manual Chapter 0609, Attachment 2.
The NRC has also determined that the failure to lock open Valve CNM-FCV200 properly is a
violation of Technical Specification 5.4.1.a, as cited in the enclosed Notice of Violation (Notice).
The circumstances surrounding the violation are described in detail in the subject inspection
report. In accordance with the NRC Enforcement Policy, NUREG-1600, the Notice of Violation
is considered escalated enforcement action because it is associated with a White finding.
You are required to respond to this letter and should follow the instructions specified in the
enclosed Notice when preparing your response.
Because plant performance for this issue has been determined to be in the regulatory response
band, we will use the NRC Action Matrix to determine the most appropriate NRC response for
this event. We will notify you, by separate correspondence, of that determination.
Entergy Operations, Inc. -4-
In accordance with 10 CFR 2.790 of the NRCs "Rules of Practice," a copy of this letter, its
enclosures, and your response will be made available electronically for public inspection in the
NRC Public Document Room or from the NRCs document system (ADAMS), accessible from
the NRC Web site at http://www.nrc.gov/reading-rm/adams.html. To the extent possible, your
response should not include any personal privacy, proprietary, or safeguards information so
that it can be made available to the Public without redaction. The NRC also includes significant
enforcement actions on its Web site at www.nrc.gov; select What We Do, Enforcement, then
Significant Enforcement Actions.
Sincerely,
/RA/
Bruce S. Mallett
Regional Administrator
Docket: 50-458
License: NPF-47
Enclosure:
1. Notice of Violation
2. NRC Evaluation of Inadequately Secured Condensate Valve
cc w/enclosure:
Senior Vice President and
Chief Operating Officer
Entergy Operations, Inc.
P.O. Box 31995
Jackson, MS 39286-1995
Vice President
Operations Support
Entergy Operations, Inc.
P.O. Box 31995
Jackson, MS 39286-1995
General Manager
Plant Operations
River Bend Station
Entergy Operations, Inc.
P.O. Box 220
St. Francisville, LA 70775
Entergy Operations, Inc. -5-
Director - Nuclear Safety
River Bend Station
Entergy Operations, Inc.
P.O. Box 220
St. Francisville, LA 70775
Wise, Carter, Child & Caraway
P.O. Box 651
Jackson, MS 39205
Mark J. Wetterhahn, Esq.
Winston & Strawn
1401 L Street, N.W.
Washington, DC 20005-3502
Manager - Licensing
River Bend Station
Entergy Operations, Inc.
P.O. Box 220
St. Francisville, LA 70775
The Honorable Richard P. Ieyoub
Attorney General
Department of Justice
State of Louisiana
P.O. Box 94005
Baton Rouge, LA 70804-9005
H. Anne Plettinger
3456 Villa Rose Drive
Baton Rouge, LA 70806
President
West Feliciana Parish Police Jury
P.O. Box 1921
St. Francisville, LA 70775
Michael E. Henry, State Liaison Officer
Department of Environmental Quality
Permits Division
P.O. Box 4313
Baton Rouge, LA 70821-4313
Entergy Operations, Inc. -6-
Brian Almon
Public Utility Commission
William B. Travis Building
P.O. Box 13326
1701 North Congress Avenue
Austin, TX 78701-3326
Entergy Operations, Inc. -7-
Electronic distribution by RIV:
Regional Administrator (BSM1)
DRP Director (ATH)
DRS Director (DDC)
Senior Resident Inspector (PJA)
Branch Chief, DRP/B (DNG)
Senior Project Engineer, DRP/B (RAK1)
Senior Reactor Analyst, DRS (DPL)
Staff Chief, DRP/TSS (PHH)
RITS Coordinator (NBH)
W. Maier, RSLO (WAM)
D. Berean (DMB)
D. Thatcher (DFT)
RBS Site Secretary (LGD)
A. Boland (ATB), OEDO RIV Coordinator
J. Dixon-Herrity, OE (JLD)
G. F. Sanborn, D:ACES (GFS)
K. D. Smith, RC (KDS1)
F. J. Congel, OE (FJC)
OE:EA File (RidsOeMailCenter)
OEMAIL
ADAMS: Yes * No Initials: _rak__
Publicly Available * Non-Publicly Available * Sensitive Non-Sensitive
R:\_RBS\RB2002-07NOV-RAK.wpd
RIV:C:DRP/B SRA:DRS C:PSA/NRR D:DRS D:ACES
DNGraves;dlf DPLoveless MTschiltz DDChamberlain GFSanborn
RAKopriva for /RA/ E - DNGraves /RA/ /RA/
9/24/03 9/25/03 10/14/03 12/1/03 10/30/03
KDSmith FJCongel ATHowell TPGwynn BSMallett
E - DNGraves E - JLDixon-Herrity /RA/ /RA/ /RA/
11/21/03 12/17/03 12/1/03 12/24/03 12/29/03
OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
ENCLOSURE 1
NOTICE OF VIOLATION
Entergy Operations, Inc. Docket: 50-458
River Bend Station License: NPF-47
During an NRC inspection concluded on November 14, 2002, a violation of NRC requirements
was identified. In accordance with the "General Statement of Policy and Procedure for NRC
Enforcement Actions," NUREG-1600, the violation is listed below:
Technical Specification 5.4.1.a requires that written procedures be established,
implemented, and maintained covering the applicable procedures recommended in
Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.
Regulatory Guide 1.33, Revision 2, Appendix A4, "Procedures for Startup, Operation,
and Shutdown of Safety-Related BWR Systems, Item n., lists "Condensate System
(hotwell to feedwater pumps, including demineralizers and resin regeneration)."
System Operating Procedure SOP-0007, Condensate System, Revision 21, required
Condensate Prefilter Vessel Bypass Flow Control Valve CNM-FCV200 to be locked
open.
Contrary to the above, on September 18, 2002, Valve CNM-FCV200 failed closed as a
result of not having been properly locked open, as required by System Operating
Procedure SOP-0007, "Condensate System." As a result, the feedwater flow transient
resulting from a reactor scram on September 18, 2002, caused Valve CNM-FCV200 to
close unexpectedly, causing a complete loss of feedwater flow to the reactor pressure
vessel.
This violation is associated with a White Significance Determination Process finding.
Pursuant to the provisions of 10 CFR 2.201, Entergy Operations Inc. is hereby required to
submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555 with a copy to the Regional Administrator,
Region IV, 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; EA-03-077" and should
include for each violation: (1) the reason for the violation, or, if contested, the basis for
disputing the violation or severity level, (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. Your response may reference or include
previous docketed correspondence, if the correspondence adequately addresses the required
response. If an adequate reply is not received within the time specified in this Notice, an order
or a Demand for Information may be issued as to 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.
-2-
If you contest this enforcement action, you should also provide a copy of your response, with
the basis for your denial, to the Director, Office of Enforcement, United States Nuclear
Regulatory Commission, Washington, DC 20555-0001.
Because your response will be made available electronically for public inspection in the NRC
Public Document Room or from the NRCs document system (ADAMS), accessible from the
NRC Web site at http://www.nrc.gov/reading-rm/adams.html, to the extent possible, it should
not include any personal privacy, proprietary, or safeguards information so that it can be made
available to the public without redaction. If personal privacy or proprietary information is
necessary to provide an acceptable response, then please provide a bracketed copy of your
response that identifies the information that should be protected and a redacted copy of your
response that deletes such information. If you request withholding of such material, you must
specifically identify the portions of your response that you seek to have withheld and provide in
detail the bases for your claim of withholding (e.g., explain why the disclosure of information will
create an unwarranted invasion of personal privacy or provide the information required by
10 CFR 2.790(b) to support a request for withholding confidential commercial or financial
information). If safeguards information is necessary to provide an acceptable response, please
provide the level of protection described in 10 CFR 73.21.
Dated this 29th day of December 2003
ENCLOSURE 2
River Bend Station
Evaluation of Inadequately Locked Condensate Valve
Conclusions:
On the basis of the reconsideration of the NRCs preliminary risk evaluation and the additional
information regarding the risk evaluation that was submitted by the licensee following the
regulatory conference, the NRC staff concluded that the licensees risk assessment was
incomplete and contained nonbounding assumptions. Specifically, the licensees risk
assessment: (1) excluded external events other than fires, (2) used nonbounding assumptions
to evaluate 32 fire areas, and (3) assumed that the quantitative risk of the remaining fire areas
was similar to the risk determined for the first 32 areas without a clear basis. In the NRCs
view, appropriate consideration of the risks associated with external events and internal fires
would result in an increase in the estimated risk.
Accordingly, the NRC concluded that the final risk significance of the subject finding should be
made on the basis of the NRCs preliminary risk evaluation, which included the qualitative
assessment of external events, including internal fires. The following assessment, supported
by quantitative methods, indicated that the risk associated with external initiators is at least the
same as that of the internal events analysis previously documented in the preliminary risk
evaluation. Therefore, the analysts best estimate of external events was adjusted to 7.7 x 10-7,
which is equal to the internal event risk documented in the preliminary evaluation. This resulted
in a final lower-bound best estimate for total risk associated with the finding of 1.5 x 10-6,
indicating that the finding is of low to moderate risk significance (WHITE). The basis for this
conclusion, including quantitative methods utilized in supporting this qualitative assessment is
documented in the following discussion.
Basis for Conclusion:
NRC Inspection Manual Chapter 0609, "Significance Determination Process," Appendix A,
Attachment 1, step 2.5, "Screening for the Potential Risk Contribution Due to External Initiating
Events," states that the impact of external initiators should be evaluated and that accounting for
these initiators could result in increasing the risk significance attributed to an inspection finding
by as much as one order of magnitude. Furthermore, step 2.5 states "This evaluation may be
qualitative or quantitative in nature. Qualitative evaluations of external events should, as a
minimum, provide the logic and basis for the conclusion and should reference all of the
documents reviewed." The NRC has qualitatively assessed the significance of the external
events contribution to the risk of this finding. Additionally, quantitative methods used by the
staff indicate that the risk associated with external initiators would be at least equal to that of
the risk associated with internal initiators.
The information that the licensee provided did not change the NRCs view regarding risk
significance of the finding, nor did it change the basis for the risk significance determination.
The external events evaluation discussed at the regulatory conference and documented by the
licensees assessment of internal events changed from 7.7 x 10-7 to 5.3 x 10-7. However, this
did not affect the final outcome of this evaluation. Some of the information provided by the
licensee would affect the limited quantitative external events evaluation developed by the NRC
staff in support of the preliminary significance determination. However, the qualitative
-2-
evaluation by the NRC staff indicates that the external initiators portion of the total risk
associated with the finding is at least equal to the estimate of internal risk provided in the
preliminary assessment. Using the licensees revised internal event estimation (5.3 x 10-7) and
increasing the internal risk by a factor of two, the total risk estimate of 1.1 x 10-6 still indicates
that the finding is of low to moderate risk significance.
Weaknesses in the licensees evaluation can be separated into the following three categories:
(1) Exclusion of Certain External Initiators
The licensee used design basis or IPEEE thresholds as their bases for determining that
the risk associated with external events, other than internal fire, were negligible.
External events are typically evaluated separately from internal events because of their
ability to affect widespread areas of the plant and/or randomly affect isolated areas.
The licensees response did not take this into account for the following specific external
events that were qualitatively considered by the analyst:
a. Under the heading "High Winds, Tornados, Hurricanes," the licensee stated that
Seismic Category 1 structures are designed to withstand high winds.
Additionally, the licensee stated that the feedwater and condensate systems are
not designed to operate during a design basis tornado. Therefore, they
assumed that, during any severe wind event, the systems would fail.
The analyst agrees that, given the design of the plant, the probability of failure of
safety-related mitigating systems without a loss of feedwater is small. However,
the precision of this assessment is on the order of parts per 10 million.
Therefore, determining that the potential hazard is small is insufficient to rule out
high winds as a contributor to the external risk of this finding.
On page 20 of their response, the licensee states that SECY-00-0162 "allows
one to exclude from a PRA analysis things that screen out." However, the
SECY, in page A1-13, states that an external event may be screened out if "it
can be shown using an analysis that the mean value of the design-basis hazard
used in the plant design is less than 10-5/year, and that the conditional
core-damage probability is less than 10-1, given the occurrence of the
design-basis hazard." The licensee did not show that the high wind hazard at
River Bend Station is less than 10-5/year, even though River Bend Station is
located in an area in which high wind conditions are prevalent.
b. Under the heading of "Transportation Hazards," the licensee stated that the
IPEEE demonstrated that the transportation of hazardous material has
decreased or remained mostly constant. They, therefore, concluded that the risk
of these events is expected to be negligible.
While this type of screening analysis was sufficient to look for design
vulnerabilities of the plant, the scope is insufficient to determine the total impact
of these events on the subject finding.
-3-
c. Under the heading "Severe Weather Storms," the licensee stated "As stated in
the IPEEE, severe weather storms generally result in either a partial or complete
loss of offsite power event and are fully analyzed in the internal events PSA."
These events are analyzed in the internal events PSA only if the storm causes a
loss of offsite power and no other damage. The qualitative judgment that these
storms "generally" cause a loss of offsite power does not indicate that the
likelihood of storms causing a plant scram, without a loss of offsite power and a
loss of safety-related equipment, is less than 10-5/year.
d. Under the heading of "Lightning," the licensee stated "since RBS [River Bend
Station] does not have a history of frequent lightning strikes causing plant
scrams, lightning is eliminated from further analysis."
River Bend Station has been operating for approximately 18 years. Therefore,
even if River Bend Station had never had a scram caused by a lightning strike,
one would not be able to draw the conclusion that the likelihood of a scram and
additional equipment damage would be less than 10-5/year.
e. Under the heading of "Spurious of Inadvertent Fire Suppression Activation," the
licensee concluded "the inadvertent actuation of suppression systems alone, will
not render safe shutdown of the plant inoperable."
This design basis argument does not indicate that the failures would not result in
increased core damage frequency. Should internal flooding cause a reactor
scram in addition to failure of risk significant equipment, it would increase the risk
associated with the finding.
Any of these external events would need to be fully analyzed to obtain a plant-specific
result. The precision of the IPEEE screening was insufficient to determine if the
feedwater system is important to these external events. Qualitatively, the analyst
concluded that these external events represented risk contributors that would increase
the overall risk significance of the finding if fully analyzed.
(2) Nonbounding Assumptions Used to Evaluate 32 Fire Areas
The licensee grouped the 22 fire areas that were screened by crediting feedwater in the
IPEEE and the seven unscreened areas, as well as three other areas that had specific
fire area ignition frequencies for further analysis. Of these areas, the licensee
determined that 8 areas had cables that could result in a direct scram. These 8 areas
were then reanalyzed using the licensees revised internal events model. The resulting
assessment indicated that the finding represented a change in core damage frequency
of 8 x 10-9 over the 126-day period for those 8 areas. The licensee assumed that, if a
direct scram did not result, the change in risk caused by a fire in the area was negligible.
The NRC staff identified the following concerns and weaknesses during the review of
the licensees response.
-4-
a. Fire Induced Scrams
On page 7 of Attachment 12 in their response, the licensee stated "no cables
that would directly cause a scram were found." The licensee stated that they did
not conduct a review to determine if an indirect scram could occur.
Indirect scrams do occur and are expected to be caused by certain fires. Fires
affecting control room instrumentation and/or those that are large enough to
concern licensed operators may result in a manual reactor scram.
The potential for indirect scrams caused by fire was not evaluated by the
licensee. The analyst assumes that indirect scrams that affect the significance
of this finding would occur in certain fire scenarios and that these scrams would
cause an increase in the conditional core damage probability for the postulated
fire. On the basis of discussions with fire protection engineers, the analysts
assumed that potentially 10 percent of the fire areas could cause an indirect
b. Severity Factor
In Attachment 8 of their response, the licensee calculated severity factors of
between .01 and .24. Additionally, the severity factors for the fire areas that
were ultimately quantified ranged from 0.014 to 0.02. These values are lower
than typically used. The staff determined that factors between 0.1 and 0.3 were
typical for the type of fire areas described.
If a low-range expected industry severity factor of 0.15 was used for these,
instead of the licensees values, then it would change the significance of
feedwater to the first 32 fire areas from 8 x 10-9 to 7.6 x 10-8 for the 126-day
period.
c. Fire Size
In Attachment 1, page 14, of their response, the licensee indicates that the River
Bend Station electrical cabinet fire heat release rates were assumed to be
90 btu/second for 30 minutes. This value appears to be appropriate for low
voltage (~120V) electrical cabinets but, for 480V and greater, the current SDP
uses heat release rates of approximately 190 btu/s. Also, a shorter more intense
fire may be more destructive than the 30-minute fire considered by the licensee.
Additionally, no energetic faults were considered. However, the NRC staff
determined that they should have been included.
To fully evaluate the effect on risk of these assumptions would require significant
resources to reanalyze the fire areas and then quantify the risk. However,
qualitatively, the risk would increase as a result of the higher heat release rates.
-5-
d. Fire Ignition Frequency
In Attachment 13 of their response, the licensee lists the fire ignition frequencies
between ~1.6E-2 to ~3E-6. The smallest Mean Fire Frequency identified by
RES/OERAB/S02-01 (the NRCs Office of Nuclear Regulatory Research fire
frequency calculation), is for the cable spreading room and has a value of 8.4E-4
for power operation. The low end frequencies assumed by the licensee appear
to be very low.
An increase in fire ignition frequency in any area would result in a direct increase
in the change in core damage frequency analyzed by the licensee for this
finding.
e. Fire Modeling
In Attachment 1, page 15, of their response, the licensee used the
COMPBRN IIIe model with openings closed. This configuration creates an
oxygen limited fire scenario which may not be worst-case as the licensee
assumes. They concluded that there is no propagation within a compartment.
This conclusion is affected by not assuming the compartment is closed. The
licensee assumes that doors and dampers would be closed, but there would be a
period before the doors and dampers close where the fire would not be oxygen
limited, resulting in greater fire growth.
To evaluate fully the effect on risk of these assumptions would require significant
resources to reanalyze the fire areas and then quantify the risk. However, the
analyst determined that the risk would increase as a result of the increase in
oxygen available to the fire.
f. Human Error Probabilities
Attachment 1, page 16, of their response, indicates that "no adjustment of error
rates to account for fire environments" were made. No description of actions
required in or near fire environments were provided.
Large fires usually take significant crew resources for fighting the fire.
Therefore, human reliability and system recovery modeling in the internal events
model would normally need adjustment to account for the additional loads on the
remaining operators. Additionally, recovery of equipment in or near the fire
areas is hampered.
Using the SPAR HRA method, a significant decrease in control room and plant
operator resources would result in an increase by an order of magnitude in the
failure rates for recovery. Given that fire scenarios usually require a significant
operator response in the plant, adjusting the human factors basic events would
result in a significant increase in the analyzed risk.
-6-
The analyst determined through expert judgment that the differences between the
licensees assumptions and the industry norm are enough to justify the difference
between the NRC's and licensee's risk characterizations.
(3) Inferred Quantitative Assessment of Remaining Fire Areas:
Out of the 164 fire areas at River Bend Station, only 32 were actually analyzed. Of the
remaining fire areas: 1 area was not assessed, but grouped with others; 53 areas were
quantified by assuming similarities to the first 32; and 78 areas were quantified using
generic industry data related to fire induced scrams. The following groups of fire areas
were assessed:
a. Group 1, Fire Areas 1-32
The licensee grouped the 22 fire areas that were screened by crediting
feedwater in the IPEEE and the 7 unscreened areas, as well as 3 other areas for
further analysis. The evaluation of the licensees analysis is discussed in
Section 2 above.
The licensee then used the change in core damage frequency for these 32 areas
and adjusted the number to estimate the change in core damage frequency for
those areas. The nonbounding assumptions used by the licensee in evaluating
Group 1 areas are discussed in detail in Section 2 of this enclosure.
b. Group 2, Fire Areas 33-62
In Areas 33-62, the licensee stated that they adjusted the core damage
frequency from the first 32 areas by taking credit for manual suppression of the
fire. Therefore, they reduced the core damage frequency for Group 1 areas by a
factor of 10 to estimate the frequency for Group 2 areas. First, the licensee did
not indicate why the fire risk in other fire area groups would be similar in risk to
Areas 1-32. However, this was the licensees assumption. In addition, fire
severity factors are typically developed from fire data bases by identifying fire
ignition rates and determining if the fire becomes large. This method of
quantifying the severity factor inherently includes both the probability that a fire is
self-extinguishing and the probability that manual suppression was effective. If a
fire was extinguished early either by manual suppression or by the
characteristics of the component or fire, then it would not have grown to a large
fire and would have been used to reduce the severity factor. Therefore, the
analyst determined that it was inappropriate to give additional credit for manual
suppression as the licensee did for these fire areas.
c. Groups 3 and 4, Fire Areas 63-86
In Areas 63-86, the licensee stated that they took credit for manual suppression
as well as for the failure of the fire to spread from one cabinet to another. Again,
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the licensee assumed that a fire in these areas would be similar in risk to
Areas 1-32. The analyst, as before, determined that the additional credit for
manual suppression was inappropriate.
d. Group 5, Fire Areas87-164
In Fire Areas87-164, the licensee quantified the risk using generic fire induced
scram data. The licensee assumed that a fire in these areas would cause either
a scram or damage to safe shutdown equipment, but not both. Therefore, they
qualitatively assumed that a fire in these areas would cause a scram and used
the conditional core damage probability for scrams with loss of feedwater to
quantify these areas. This was an appropriate approach to quantifying these
areas.
Only 32 of 164 fire areas were actually analyzed. The assumption that Groups 2
through 4 fire areas can be represented as similar in scram frequency, ignition
frequency, and severity, appears unfounded. Additionally, the application of a manual
suppression factor is inconsistent with the method used for developing the fire severity
factors and with the assumption that all groups are similar to Group 1. Therefore, the
analyst concluded that the licensees assessment was incomplete and that the risk
quantified by the licensee should be a factor of 10 higher for Groups 2 through 4
because of the inappropriate use of a manual suppression factor.
Summary
The analyst identified three areas of concern: the licensee excluded all external initiators with
the exception of internal fires; the licensee only analyzed Group 1, representing 32 of the
164 fire areas at River Bend Station; and the licensee made the unsupported conclusion that
Groups 2-4 fire areas were similar in risk to Group 1. Of these 32 areas, only seven were
determined to have scram potential by direct means. There were several weaknesses in the
licensees assumptions related to the quantification of those seven. These weaknesses
indicated a potential error of several orders of magnitude in the licensees quantification. The
remaining fire areas were evaluated by making the basic assumption that Groups 2-5 were
similar in risk to Group 1. The licensee then inferred a risk quantification starting with the
assessment of the first seven fire areas. That assumption aside, the licensees quantification of
Groups 2-4 applied quantitative factors that reduced the risk associated with those groups.
One of these assumptions, that manual suppression in those fire areas was independent of the
fire severity factors, was determined to be erroneous. Therefore, the analyst concluded, based
on qualitative assessment of the licensees factors, that the subject finding remained of low to
moderate risk significance as determined in the preliminary risk assessment of the finding.