ML12208A379
| ML12208A379 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 07/26/2012 |
| From: | Blount T Division of Reactor Safety IV |
| To: | Reddemann M Energy Northwest |
| Hay M (vlm) | |
| References | |
| EA-12-092 IR-12-502 | |
| Download: ML12208A379 (18) | |
See also: IR 05000397/2012502
Text
UNITE D S TATE S
NUC LEAR RE GULATOR Y C OMMI S SI ON
RE G IO N I V
1600 EAST LAMAR BLVD
AR L INGTON , TEXAS 7 60 11 - 4511
July 26, 2012
Mr. Mark E. Reddemann
Chief, Executive Officer
Energy Northwest
P.O. Box 968 (Mail Drop 1023)
Richland, WA 99352-0968
SUBJECT: COLUMBIA GENERATING STATION - NRC BASELINE INSPECTION
REPORT NO. 05000397/2012502, PRELIMINARY WHITE FINDINGS
Dear Mr. Reddemann:
This letter refers to the inspection conducted October 18, 2011, through June 27, 2012 at
Energy Northwests Columbia Generating Station, with onsite inspection February 6-9, 2012.
The inspection reviewed changes made to site dose assessment methods and emergency
action levels between September 2000 and December 2011. The enclosed report presents the
results of this inspection. The preliminary results of this inspection were discussed onsite with
site management on February 9, 2012, and during subsequent conference calls between the
NRC and site representatives on February 24, February 27, March 1, and March 12, 2012. The
results of this inspection were discussed with site management during exit meetings conducted
by conference call on May 16 and June 27, 2012.
This inspection examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations, and with the conditions in your license.
Within these areas, the inspection consisted of a selected examination of procedures and
representative records, observations of activities, and interviews with personnel.
The enclosed inspection report discusses two findings that have preliminarily been determined
to be White findings with low to moderate safety significance that may require additional NRC
inspections. These findings are associated with failure to maintain a standard emergency action
level scheme as required by 10 CFR 50.47(b)(4) and failure to maintain adequate methods for
assessing the potential consequences of a radiological emergency condition in accordance with
the requirements of 10 CFR 50.47(b)(9). These deficiencies were corrected on January 5, 2012
and December 17, 2011, respectively. These findings are also apparent violations of NRC
requirements and are being considered for escalated enforcement action in accordance with the
Mr. Reddemann -2-
Enforcement Policy, which can be found on the NRCs Web site at http://www.nrc.gov/about-
nrc/regulatory/enforcement/ enforce-pol.html. In addition, one apparent traditional violation was
identified and is being considered for escalated enforcement action in accordance with the NRC
Enforcement Policy. The apparent violation is associated with failure to report to the NRC a
major loss of emergency assessment capability in accordance with the requirements of 10 CFR 50.72(b)(3)(xiii).
The preliminary low to moderate safety significance (White) findings were assessed based on
the best available information, using the Emergency Preparedness Significance Determination
Process (SDP) and the NRC Enforcement Policy. The basis for the NRCs preliminary
significance determinations are described in the enclosed report. The final resolution of these
findings will be conveyed in separate correspondence.
In accordance with NRC Inspection Manual Chapter (IMC) 0609, we intend to complete our
evaluation of the White findings, using the best available information, and issue our final
determination of safety significance within 90 days of the date of this letter. The significance
determination process encourages an open dialogue between the NRC staff and the licensee;
however, the dialogue should not impact the timeliness of the staffs final determination. Before
we make a final decision on this matter, we are providing you with an opportunity to: (1) attend a
Regulatory Conference where you can present to the NRC your perspective on the facts and
assumptions the NRC used to arrive at the findings and assess their significance, or (2) submit
your position on the findings to the NRC in writing.
Additionally, as part of the enforcement process for an apparent traditional violation, you will
have the opportunity to request a Predecisional Enforcement Conference, the conference will
afford you the opportunity to provide your perspective on the apparent traditional violation and
any other information that you believe the NRC should take into consideration before making an
enforcement decision. The topics discussed during this conference may include the following:
information to determine whether a violation occurred, information to determine the significance
of a violation, information related to the identification of a violation, and information related to
any corrective actions taken or planned to be taken. In presenting your corrective actions, you
should be aware that the promptness and comprehensiveness of your actions will be
considered in assessing a civil penalty, if any, for the apparent violation.
If you request a Regulatory Conference and Predecisional Enforcement Conference, it should
be held within thirty days of the receipt of this letter and we encourage you to submit supporting
documentation at least one week prior to the Conference in an effort to make the Conference
more efficient and effective. If a Conference is held, it will be open for public observation and a
public meeting notice and press release will be issued to announce the conference. If you
decide to submit only a written response, such submittal should be sent to the NRC within
thirty days of your receipt of this letter. If you decline to request a Conference or to submit a
written response, you relinquish your right to appeal the final SDP determination; in that, by not
doing either you fail to meet the appeal requirements stated in the Prerequisite and Limitation
Sections of Attachment 2 of IMC 0609.
If you choose to provide a written response, it should be clearly marked as Response to
Apparent Violations in Inspection Report No. 05000397/2012502; EA-12-092 and for each
apparent violation discussed should include: (1) the reason for the apparent violation, or, if
contested, the basis for disputing the apparent 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 was (will be) achieved. Your response may
Mr. Reddemann -3-
reference or include previously docketed correspondence, if the correspondence adequately
addresses the required response.
Please contact Mr. Michael Hay, Chief, Plant Support Branch 1, at 817-200-1527, within
ten days from the issue date of this letter to notify the NRC of your intentions. If we have not
heard from you within ten days, we will continue with our significance determination and
enforcement decision. Since the NRC has not made a final determination in these matters,
Notices of Violation are not being issued for these inspection findings at this time. In addition,
please be advised that the number and characterization of the apparent violations may change
as a result of further NRC review.
Furthermore, a licensee-identified violation which was determined to be of very low safety
significance is listed in this report. The NRC is treating this violation as a non-cited violation
consistent with Section 2.3.2 of the Enforcement Policy. If you contest this non-cited violation,
you should provide a response within thirty days of the date of this inspection report, with the
basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk,
Washington DC 20555-0001; with copies to the Regional Administrator, Region IV; the Director,
Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-
0001; and the NRC Resident Inspector at Columbia Generating Station.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its
enclosure(s), and your response, if you choose to provide one, will be made available
electronically for public inspection in the NRC Public Document Room or from the NRCs
Agency-wide Documents Access and Management 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 or proprietary information so that it can be made
available to the Public without redaction.
If you have any questions concerning this matter, please contact Mr. Michael Hay, Chief, Plant
Support Branch 1, at 817-200-1527.
Sincerely,
/RA/
Thomas Blount, Acting Director
Division of Reactor Safety
Docket No. 50-397
License No. NPF-21
Enclosure:
Inspection Report 05000397/2012502
w/Attachment
Electronic Distribution for Columbia Generation Station
ADAMS: No # Yes # SUNSI Review Complete Reviewer Initials: MCH
- Publicly Available # Non-Sensitive
Non-publicly Available Sensitive
RIV: DRS\PSB1\EP1 NSIR\DPR\EP DRP\PBA\BC DRP\ Director
DRS\PSB1\SEPI
PElkmann GGuerra ESchrader WWalker KKennedy
/RA/ /RA/ /RA/ /RA/ /RA/
7/16/2012 7/16/2012 7/16/2012 7/18/2012 7/19/2012
NSIR\BC ACES OE DRS\PSB1\BC DRS\Acting Director
R. Kahler RKellar GGulla MHay TBlount
/RA/ /RA/ /RA/ per -Email /RA/ /RA/
7/19/2012 7/24/2012 7/24/2012 7/16/2012 7/25/2012
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket: 50-397
License: NPF-21
Report: 05000397/2012502
Facility: Columbia Generating Station
Licensee: Energy Northwest
Location: Richland, Washington
Dates: October 18, 2011, through June 27, 2012
Paul J. Elkmann, Senior Emergency Preparedness Inspector
Inspectors: Gilbert L. Guerra, CHP, Emergency Preparedness Inspector
Eric Schrader, Emergency Preparedness Specialist, NSIR/DPR/EP
Thomas Blount, Acting Director
Approved By:
Division of Reactor Safety
-1- Enclosure
SUMMARY OF FINDINGS
IR 05000397/2012502; 10/18/2011 - 06/27/2012; Columbia Generating Station, Regional
Report; Emergency Plan Focused Baseline Inspection, 7111404, 7111405
The report covered an announced baseline inspection by region-based inspectors and a
technical specialist from the Office of Nuclear Security and Incident Response. Three apparent
violations were identified. The significance of most findings is indicated by their color (Green,
White, Yellow, or Red) using Inspection Manual Chapter 0609, Significance Determination
Process. The cross-cutting aspect is determined using Inspection Manual Chapter 0310,
Components Within the Cross Cutting Areas. Findings for which the significance
determination process does not apply may be Green or be assigned a severity level after NRC
management review. The NRC's program for overseeing the safe operation of commercial
nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4,
dated December 2006.
A. NRC-Identified Findings and Self-Revealing Findings
Cornerstone: Emergency Preparedness
- TBD. An apparent violation of 10 CFR 50.54(q) was identified involving the
failure to maintain a standard emergency action level scheme in September 2000
and November 2010 in accordance with the requirements of 10 CFR 50.47(b)(4).
The licensee inappropriately calculated and changed Site Area Emergency and
General Emergency radiation monitor threshold values on EAL (Emergency
Action Levels) Table 3, Effluent Monitor Classification Thresholds. These
changes adversely affected the ability of the licensee to properly classify events
involving a radiological release.
The inspectors determined the licensees inaccurate calculation of Site Area
Emergency and General Emergency radiation monitor thresholds in September
2000 and November 2010 were performance deficiencies within the licensees
control. This finding is more than minor because it was associated with the
procedure quality and emergency response organization performance
cornerstone attributes. This finding was evaluated using the Emergency
Preparedness Significance Determination Process and was preliminarily
determined to be of low to moderate safety significance (White) because it was a
degraded risk significant planning standard function. The planning standard
function was degraded because Columbia Generating Station would have been
delayed in recognizing Site Area Emergencies and General Emergencies
because of the inaccurate reactor building stack monitor EAL Table 3 values.
This finding was entered into the licensees corrective action system as Action
Requests AR00244316, AR00244578, and AR00244838 (Section 1EP4).
- TBD. An apparent violation of 10 CFR 50.54(q) was identified involving the
failure to maintain adequate methods for assessing the actual or potential
consequences of a radiological emergency between April 2000 and December
2011 in accordance with the requirements of 10 CFR 50.47(b)(9). The licensee
incorporated inaccurate gas calibration and Xenon equivalency factors into dose
projection software, resulting in inaccurate offsite dose calculations involving
radiological releases measured by the reactor building effluent radiation monitor.
-2- Enclosure
The inspectors determined the failure to maintain a dose assessment process
capable of providing a technically adequate estimate of offsite dose was a
performance deficiency within the licensees control. This finding is more than
minor because it was associated with the emergency response organization
performance and the Facilities and Equipment cornerstone attributes. This
finding was evaluated using the Emergency Preparedness Significance
Determination Process and was preliminarily determined to be of low to
moderate safety significance (White) because it was a degraded risk significant
planning standard function. The planning standard function was degraded
because some methods for assessing the offsite consequences of a radiological
release were inaccurate between April 2000 and December 2011. This issue
has been entered into the licensees corrective action system as Action Requests
AR00244316 and AR00244578 (Section 1EP5).
- TBD. An apparent Severity Level III violation was identified for failure to notify
the NRC of a major loss of emergency assessment capability identified on
October 18, 2011, as required by 10 CFR 50.72(b)(3)(xiii). The licensee failed to
identify that these deficiencies adversely affected the licensees ability to project
offsite dose during a radiological event and therefore constituted a major loss of
emergency assessment capability.
The failure to report was evaluated using the NRC Enforcement Policy and was
determined to be an apparent Severity Level III violation because it was
associated with a Reactor Oversight Program issue of low to moderate safety
significance (White). This issue has been entered into the licensees corrective
action system as Action Requests AR00244578, Revision 2, AR00244838, and
AR00264998 (Section 1EP5).
B. Licensee-Identified Violations
A violation of very low safety significance, which was identified by the licensee, has been
reviewed by the inspectors. Corrective actions taken or planned by the licensee have
been entered into the licensees corrective action program. This violation and corrective
action tracking number is listed in Section 4OA7.
-3- Enclosure
REPORT DETAILS
1. REACTOR SAFETY
Cornerstone: Emergency Preparedness
1EP4 Emergency Action Level and Emergency Plan Changes (71114.04)
a. Inspection Scope
The inspectors performed in-office and on-site reviews of licensee changes to
emergency action level Table 3, Effluent Monitor Classification Thresholds, made
between September 2000 and December 2011. The inspectors reviewed:
- Columbia Generating Station Emergency Plan, Revisions 52, 54, 55, and 56;
- Procedure 13.1.1, Classifying the Emergency, Revisions 28, 29, 35, 39, and 40;
- Procedure 13.1.1A, Classifying the Emergency - Technical Bases, Revisions 7,
and 24; and,
- Action Request AR00244578, Root Cause Evaluation, Inappropriate Emergency
Action Level Modification, Revision. 2, dated October 18, 2011.
These documents were compared to their previous revisions, to the criteria of NUREG-
0654, Criteria for Preparation and Evaluation of Radiological Emergency Response
Plans and Preparedness in Support of Nuclear Power Plants, Revision 1, to Nuclear
Energy Institute Report 99-01, Emergency Action Level Methodology, Revisions 2 and
4, and to the standards in 10 CFR 50.47(b), to determine if the revisions adequately
implemented the requirements of 10 CFR 50.54(q). The specific documents reviewed
during this inspection are listed in the attachment.
These activities constitute completion of one sample as defined in Inspection
Procedure 71114.04-05.
b. Findings
Introduction. An apparent violation was identified involving inappropriate changes to
EAL Table 3 in September 2000 and November 2010, which adversely affected the
licensees ability to classify a radiological release emergency event.
Description. Two examples were identified in which the licensee inappropriately
calculated Site Area Emergency (SAE) and General Emergency (GE) radiation monitor
threshold values on EAL Table 3, Effluent Monitor Classification Thresholds.
In April of 2000 chemistry personnel failed to properly calibrate the reactor building stack
effluent monitor. The error involved improper positioning of the radiation source
resulting in the technician inappropriately concluding the detector gas calibration factor
had changed from a value of 34.9 to 413.29. This significant change was not questioned
by chemistry personnel and the new value was provided to the emergency preparedness
personnel for use in the offsite dose assessment model. In September 2000 emergency
-4- Enclosure
preparedness personnel changed the Table 3 SAE value for the reactor building stack
effluent monitor from 9.65E2 counts/second (cps) to 9.65E3 cps, and the GE radiation
monitor value from 9.35E3 cps to 9.35E4 cps. These values were calculated using the
dose assessment model after the incorrect gas calibration factor of 413.29 was
incorporated into the model. Both thresholds were a factor of 10 higher than necessary
to indicate a release magnitude associated with a SAE or GE. As a result, the
classification would not have been made when the release magnitude, upon which the
EAL was based, warranted it.
In November 2010 the licensee changed the Table 3 GE value for the reactor building
stack effluent monitor. This threshold was calculated using an incorrect Xenon-133
Equivalent Response value of 0.128, instead of the correct value of 12.8, an error of a
factor of 100. The GE EAL threshold was changed from 9.35E4 cps to 9.35E6 cps.
Given the earlier error by a factor of 10, this EAL threshold was a value 1000 times
higher than necessary to indicate a SAE or GE release. Additionally, the maximum
range for the reactor building stack effluent monitor meter in the control room is 1.0E6
cps, so the as-changed value would be off-scale high. As a result of these two
conditions, the GE classification would not have been made when the radiological
release magnitude warranted it. Emergency Preparedness staff failed to recognize the
error in the Xenon-133 Equivalent Response factor and did not recognize the resulting
GE EAL exceeded the instrument range. The staff did not question an unexpected
change by a factor of 100, and did not validate the change.
The NRC identified during this inspection that the licensee failed to recognize that
changes to emergency action level (EAL) Table 3 decreased the effectiveness of the site
emergency plan. The licensee was unable to provide a 50.54(q) review of the
September 2000 emergency action level change. Licensee staff considered the
November 2010 emergency action level change to be an editorial change and the
licensee found no evidence this change received a 50.54(q) review. These failures are
not being treated under the NRCs Enforcement Policy (traditional enforcement) because
the underlying problems are performance deficiencies in maintaining the emergency
preparedness program being addressed in this report.
The NRC also identified during this inspection that the licensee failed to update
Notification of Unusual Event and Alert effluent radiation monitor thresholds on Table 3 to
ensure accurate classification following changes to the Offsite Dose Calculation Manual
in March 2002. This finding was an additional example of a performance deficiency
related to 10 CFR 50.47(b)(4) because it affected the licensees ability to classify an
event at the Alert emergency classification level. The licensee entered this issue into
their corrective action program as Action Request 00244315.
The licensee corrected the errors on EAL Table 3, Effluent Monitor Classification
Thresholds in Procedure 13.1.1A, Classifying the Emergency, Technical Bases,
Revision 24, dated January 5, 2012.
Analysis. The inspectors determined the licensees inaccurate calculation of Site Area
Emergency and General Emergency radiation monitor thresholds in September 2000
and November 2010 were performance deficiencies within the licensees control. The
finding had a credible impact on the emergency preparedness cornerstone objective
because the licensees capability to implement adequate measures to protect public
health and safety was degraded when emergency action levels were inaccurate. This
-5- Enclosure
finding is more than minor because it was associated with the procedure quality and
emergency response organization performance cornerstone attributes. The finding was
associated with a violation of NRC requirements. This finding was evaluated using the
Emergency Preparedness Significance Determination Process and was preliminarily
determined to be of low to moderate safety significance (White) because it was a failure
to comply with NRC requirements and resulted in a degraded risk significant planning
standard function. The planning standard function was degraded, rather than lost,
because Site Area Emergencies and General Emergencies could still have been
declared, albeit delayed, using EAL thresholds for dose projection results and/or
environmental measurements. Specifically, Columbia Generating Station would have
been delayed in recognizing Site Area Emergencies and General Emergencies because
of the inaccurate reactor building stack monitor EAL Table 3 values. This finding was
entered into the licensees corrective action system as Action Requests AR00244316,
AR00244578, and AR00244838. The finding was not assigned a cross-cutting area
component because the underlying performance deficiencies are not representative of
current performance.
Enforcement. Title 10 of the Code of Federal Regulations, Part 50.54(q), states, in part,
that a holder of a nuclear power reactor operating license shall follow and maintain in
effect emergency plans which meet the standards in 50.47(b). 10 CFR 50.47(b)(4),
requires, in part, that a standard emergency classification and action level scheme is in
use by the licensee, the bases of which include facility system and effluent parameters.
Contrary to the above, between September 2000 and December 2011, Columbia
Generating Station did not follow and maintain in effect an emergency plan using a
standard emergency classification and action level scheme, the bases of which included
facility system and effluent parameters. Specifically, personnel errors in September of
2000 and November of 2010 resulted in the inaccurate calculation of Site Area
Emergency and General Emergency effluent thresholds that were incorporated into
emergency action level Table 3, Effluent Monitor Classification Thresholds. As a result,
these errors adversely affected the licensees ability to classify an emergency event
involving a radiological release: AV 05000397-2012502-01 (Failure to Maintain Accurate
EAL Thresholds).
1EP5 Correction of Emergency Preparedness Weaknesses and Deficiencies (71114.05)
a. Inspection Scope
The inspectors,
- Performed in-office and on-site reviews of site procedures,
- Reviewed AR00244315, Apparent Cause Evaluation, Stack Monitors Non-
Functional for Extended Periods, Revision. 2, dated October 21, 2011;
- Reviewed AR00244578, Root Cause Evaluation, Inappropriate Emergency
Action Level Modification, Revision 2, dated October 18, 2011;
- Evaluated the operability of radiation detector PRM-RE-1C, Reactor Building
Exhaust High Range, by reviewing system work packages and records;
-6- Enclosure
- Reviewed test case calculations performed using the Quick Emergency Dose
Projection System;
- Examined the installed effluent monitoring system (detector PRM-RE-1C);
- Reviewed the licensees Offsite Dose Calculation Manual, Revision 1; and,
- Reviewed the licensees reportability evaluation for Condition Report CR244578,
Inability to project dose using Reactor Building Stack Monitors, dated February
21, 2012.
The inspectors evaluated licensee root cause analyses, apparent cause analyses, and
the response to corrective action requests according to the requirements of procedure
SWP-CAP-1, Corrective Action Program, Revision 24-3, and CDM-01, Cause
Determination Manual, Revision 6-1, to determine the licensee=s ability to identify,
evaluate, and correct problems. Inspectors reviewed corrective actions associated with
the effluent monitoring system, site emergency action levels, and the Quick Emergency
Dose Projection System, initiated between September 2000 and August 2011. Licensee
corrective actions were also compared to the requirements of planning standard
10 CFR 50.47(b)(14) and Appendix E to 10 CFR Part 50. The specific documents
reviewed during this inspection are listed in the attachment.
These activities constitute completion of one sample as defined in Inspection
Procedure 71114.05-05.
b. Findings
.1 Failure to Maintain Adequate Methods to Assess Radiological Consequences
Introduction. An apparent violation was identified involving the failure to maintain
adequate methods for assessing the actual or potential consequences of a radiological
emergency because of erroneous parameters in the dose assessment model, resulting
in inaccurate dose assessments between April 2000 and December 2011.
Description. Two deficiencies were identified that degraded the licensees ability to
accurately assess the offsite dose consequences of a radiological release. In April of
2000 chemistry personnel failed to properly calibrate the reactor building stack effluent
monitor because the source was not placed in the proper position, and in December
2007 chemistry personnel changed the Xe-133 Equivalent Response Factor used in the
dose assessment model to an incorrect value.
The failure to properly calibrate the effluent monitor in April 2000 occurred because of an
inadequate calibration procedure and human performance errors by chemistry
technicians. The calibration error involved improper positioning of the source, resulting
in the technician inappropriately concluding the detector efficiency had changed. This
error resulted in the licensee changing the gas calibration factor from the correct value of
34.9 to 413.29. Neither chemistry nor emergency preparedness staff questioned an
unexpected change of this magnitude and the new factor was subsequently incorporated
into the dose assessment model. This error resulted in the dose assessment model
overestimating offsite dose by a factor of approximately 12 between April 2000 and
-7- Enclosure
December 2007. As previously discussed this error also resulted in inappropriate
changes to the EAL thresholds for the reactor building stack effluent radiation monitor.
In December of 2007 the dose projection model gas calibration factor of 413.29 was
changed to 12.8. The 12.8 value was listed in the FSAR as the Xe-133 Equivalent
Response factor; however, the change was not appropriate since the gas calibration
factor is derived from the detector calibration. This change was not well documented
and the licensee concluded the individual making the change did not understand the
impact. This error resulted in the offsite dose assessment model underestimating offsite
dose by a factor of approximately 3 between December 2007 and December 2011.
The licensee corrected the dose projection model Gas Calibration and Xe-133
Equivalent Response Factors used in the Quick Emergency Dose Projection System on
December 17, 2011.
Analysis. The inspectors determined the failure to maintain a dose assessment process
capable of providing a technically adequate estimate of offsite dose is a performance
deficiency within the licensees control. This finding is more than minor because it
affected the licensees ability to implement adequate measures to protect the health and
safety of the public, and affected the facilities and equipment and emergency response
organization performance cornerstone attributes. The finding was associated with a
violation of NRC requirements. This finding was evaluated using the Emergency
Preparedness Significance Determination Process and was preliminarily determined to
be of low to moderate safety significance (White) because it was a failure to comply with
NRC requirements and was a degraded risk significant planning standard function. The
planning standard function was degraded because methods to assess the offsite
consequences of a radiological release via the reactor building stack were inaccurate
between April 2000 and December 2011. However, these errors did not affect other
calculations performed by the dose assessment model. This issue has been entered
into the licensees corrective action system as Action Requests AR00244316 and
AR00244578. A cross-cutting aspect was not assigned to this finding because the
performance deficiencies were not reflective of current licensee performance.
Enforcement. Title 10 of the Code of Federal Regulations, Part 50.54(q), requires, in
part, that the holder of a nuclear power reactor operating license shall follow and
maintain in effect emergency plans which meet the standards of 50.47(b). 10 CFR Part 50.47(b)(9) requires, in part, that licensees have adequate methods for assessing and
monitoring actual or potential offsite consequences of a radiological emergency
condition.
Contrary to the above, between April 2000 and December 2011, Columbia Generating
Station failed to follow and maintain in effect adequate methods for assessing and
monitoring potential offsite consequences of a radiological emergency. Specifically,
changes to offsite dose calculation methods using the reactor building effluent monitor
resulted in a process that produced inaccurate offsite doses for the reactor building
stack: AV 05000397-2012502-02 (Failure to Maintain Accurate Methods for Dose
Assessment).
-8- Enclosure
.2 Failure to Report to the NRC a major loss of Emergency Assessment Capability
Introduction. An apparent Severity Level III violation was identified for failure to notify
the NRC of a major loss of emergency assessment capability identified on October 18,
2011.
Description. Licensee root cause analysis AR00244578, Revision 2, dated October 18,
2011, identified that the emergency action level Table 3 radiation monitor threshold
values changed in September 2000 and November 2010 were derived from an
inaccurate Quick Emergency Dose Projection System. The Quick Emergency Dose
Projection System systematically calculated inaccurate offsite doses from the effluent
stack radiation monitor between April 2000 and December 2011. Specifically, between
April 2000 and December 2007 the Quick Emergency Dose Projection System would
overestimate offsite dose by a factor of approximately 12. Between December 2007 and
December 2011 the system would underestimate offsite dose by a factor of
approximately 3.
The licensee determined on February 21, 2012, that systemic inaccuracies in the Quick
Emergency Dose Projection System were not reportable to the NRC. The NRC
determined that long-term systemic inaccuracies in dose assessment methods did
constitute a major loss of emergency assessment capability that should have been
reported to the NRC after being identified on October 18, 2011. After discussions with
the NRC the licensee reevaluated the reportability decision of this deficiency and
reported to the NRC a major loss of emergency assessment capability on June 7, 2012.
Analysis. The inspectors determined the inadequate assessment of deficiencies in
methods for offsite radiological assessment was a performance deficiency within the
licensees control. The performance deficiency was evaluated using the NRC
Enforcement Policy and was determined to be a Severity Level III violation because it
was associated with a Reactor Oversight Program issue of low to moderate safety
significance (White). This issue has been entered into the licensees corrective action
system as Action Requests AR00244578, Revision 2, AR00244838, and AR00264998.
Enforcement.
Title 10 of the Code of Federal Regulations, Part 50.72(b)(3)(xiii) states that a licensee
shall notify the NRC as soon as practical and in all cases within eight hours of any event
that results in a major loss of emergency assessment capability. Contrary to the above,
on October 18, 2011, the licensee did failed to notify the NRC a within eight hours of any
event that results in major loss of emergency assessment capability. Specifically, the
licensee failed to recognize longstanding inaccuracies in the Quick Emergency Dose
Projection System as a major loss of emergency assessment capability, and failed to
report these deficiencies in radiological assessment methods to the NRC: AV 05000397-
2012502-03 (Failure to Report a Loss of Emergency Capability).
-9- Enclosure
4. OTHER ACTIVITIES
4OA6 Meetings
Exit Meeting Summary
On February 9, 2012, the inspectors discussed the onsite inspection of the licensees
radiological emergency action levels and dose assessment capabilities with Mr. B. Sawatzke,
Chief Nuclear Officer, and other members of the licensees staff. The licensee acknowledged
the issues presented. The inspectors asked the licensee whether any materials examined
during the inspection should be considered proprietary. No proprietary information was
identified.
On May 16, 2012, the inspectors conducted an exit meeting with Mr. B. Sawatzke, Chief
Nuclear Officer, and other members of the licensees staff by conference call, to communicate
the inspection results regarding the licensees radiological emergency action levels and dose
assessment capabilities. The licensee acknowledged the issues presented.
On June 27, 2012, the inspectors conducted an exit meeting with Mr. W. Hettell, Vice President
Operations, and other members of the licensees staff by conference call, to recharacterize the
inspection results regarding changes to the licensees radiological emergency action levels and
the failure to report to the NRC a major loss of emergency assessment capability. The licensee
acknowledged the issues presented.
4OA7 Licensee-Identified Violations
The following violation of very low safety significance (Green) was identified by the licensee and
is a violation of NRC requirements which meets the criteria of Section 2.3.2 of the NRC
Enforcement Policy, NUREG-1600, for being dispositioned as an NCV.
.1 Failure to Maintain Radiation Detector PRM-RE-1C
Title 10 of the Code of Federal Regulations, Part 50.47(b)(8) requires, that adequate
emergency facilities and equipment to support the emergency response are provided
and maintained. Contrary to the above, between April 2000 and February 2012, the
licensee failed to provide and maintain adequate emergency equipment to support an
emergency response. Specifically, radiation detector PRM-RE-1C, Reactor Building
High Range Exhaust, was not adequately maintained. The Reactor Building High Range
Exhaust monitor was unavailable 38 of 149 months and corrective actions were
ineffective in restoring the availability and reliability of the monitor. The finding had a
credible impact on the emergency preparedness cornerstone objective because it
affected the cornerstone attributes of emergency response organization performance
(program elements meet the 50.47(b) standards) and facilities and equipment
(maintenance surveillance and testing). The finding is more than minor because the
licensees ability to implement adequate measures to protect the publics health and
safety is degraded when equipment used to assess the consequences of a radiological
event is not adequately maintained. This finding was evaluated using the Emergency
Preparedness Significance Determination Process and was determined to be of very low
safety significance because the planning standard function was degraded, causing
potential delays in the associated key emergency response organization functions of
- 10 - Enclosure
classification and radiological assessment. The issue was entered into the licensees
corrective action system as Action Request AR00244315, Stack Monitors Non-
Functional for Extended Periods, Revision. 2.
- 11 - Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
D. Brown, Manager, Operations
G. Davis, Quality Auditor
M. Davis, Manager, Radiological Services
Z. Dunham, Supervisor, Regulatory Affairs
C. England, Manager, Organizational Effectiveness
R. Fahnestock, Manager, Emergency Preparedness
R. Garcia, Engineer, Regulatory Affairs
D. Gregoire, Manager, Regulatory Affairs
W. Hettell, Vice President, Operations
A. Javorik, Vice President, Engineering
C. King, Assistant Plant Manager
S. McCain, Emergency Management Consultant
B. MacKissock, Plant General Manager
C. Moon, Manager, Training
M. Reddemann, Chief Executive Officer
B. Sawatzke, Chief Nuclear Officer
D. Swank, Assistant Vice President, Engineering
R. Torres, Manager, Quality
L. Willliams, Acting Supervisor, Regulatory Affairs
NRC Personnel
J. Groom, Senior Resident Inspector
M. Hayes, Resident Inspector
LIST OF ITEMS OPENED AND CLOSED
Opened
05000397-2012502-01 AV Failure to Maintain Accurate EAL Thresholds (1EP4)
Failure to Maintain Accurate Methods for Dose Assessment
(1EP5)05000397-2012502-03 AV Failure to Report a Loss of Emergency Capability (1EP5)
-1- Attachment
LIST OF DOCUMENTS REVIEWED
Section 1EP4: Emergency Action Level and Emergency Plan Changes
NUMBER TITLE REVISIONS /
DATE
AR00244315 Apparent Cause Evaluation, Stack Monitors Non-Functional Revision. 2,
for Extended Periods October 21, 2011
AR00244578 Root Cause Evaluation, Inappropriate Emergency Action Revision. 2,
Level Modification October 18, 2011
PPM 13.1.1A Classifying the Emergency, Technical Bases 22, 23
PPM 16.14.2 Offsite Dose Assessment Manual Revision 1,
March 2002
Surveillance CSP-PRMRE-X302 3-8
Calculation NE-020-09-02, CGS Emergency Action Levels November 8,
Technical Bases 2011
Work Order 01012603-03 April 27, 2000
Chemistry Calculation 96-01 March 19, 1996
Problem Evaluation Request 296-0176 March 6, 1996
Condition Report 2-07-04069 May 8, 2007
Condition Report 183656, Action 2 (EC9323) July 20, 2009
Condition Report 202259, Action 6 April 19, 2010
SWP-LIC-01 Regulatory Commitment Management
SWP-LIC-02 Licensing Basis Impact Determinations
SWP-LIC-03 Licensing Document Change Process
LBDM-01 License Basis Documents Review and Maintenance Manual
EPI-16 Emergency Plan Change Processing
Section 1EP5: Correction of Emergency Preparedness Weaknesses and Deficiencies
NUMBER TITLE REVISIONS /
DATE
AR00244315 Apparent Cause Evaluation, Stack Monitors Non-Functional Revision. 2,
for Extended Periods October 21, 2011
AR00244578 Root Cause Evaluation, Inappropriate Emergency Action Revision. 2,
Level Modification October 18, 2011
PPM 13.8.1 Computerized Emergency Dose Projection System 20 - 30
Operations
PPM 13.9.1 Environmental Field Monitoring Operations 25 - 40
PPM 13.14.9 Emergency Program Maintenance 18 - 28
OD23.17 Emergency Dose Projection System Manual Revision 0,
October 31, 1989
OD23.17 Emergency Dose Projection System Manual Revision 1,
August 1994
QEDPS 2.0, A Near-Field Dose Assessment Model for March 1998
Emergency Response at the WNP2 Nuclear Facility
PRM-RE-1C Instrument Operating History 2000 - 2011
System Health Reports - Process Radiation Monitors
CGS System Description, Vol. 8, Ch. 6, Process Radiation
-2- Attachment
Section 1EP5: Correction of Emergency Preparedness Weaknesses and Deficiencies
NUMBER TITLE REVISIONS /
DATE
Monitors
Letter, W.A Macon (NRC) to Mr. J.V. Parrish (Energy August 31, 2004
Northwest) approving emergency plan changes
Letter, C.F Lyon (NRC) to Mr. M.E. Reddeman (Energy November 3, 2010
Northwest) approving emergency plan changes
ACTION REQUESTS (CORRECTIVE ACTIONS)
00020836 00021004 00024065 00037594 00120718
00121367 00190759 00191685 00216221 00221079
00244315 00244838 00246173 00246178 00246607
00247162 00264998
MISCELLANEOUS
MISCELLANEOUS DOCUMENTS
NUMBER TITLE REVISION
Columbia Generating Station Emergency Plan 56
-3- Attachment