ML092170122
| ML092170122 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 08/05/2009 |
| From: | Khadijah West Division Reactor Projects III |
| To: | Wadley M Northern States Power Co |
| References | |
| EA-09-167 IR-09-010 | |
| Download: ML092170122 (21) | |
See also: IR 05000282/2009010
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION III
2443 WARRENVILLE ROAD, SUITE 210
LISLE, IL 60532-4352
August 5, 2009
Mr. Michael D. Wadley
Prairie Island Nuclear Generating Plant
Northern States Power Company, Minnesota
1717 Wakonade Drive East
Welch, MN 55089
SUBJECT: PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2
NRC INSPECTION REPORT 05000282/2009010; 05000306/2009010
Dear Mr. Wadley:
On July 9, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at
your Prairie Island Nuclear Generating Plant, Units 1 and 2. The enclosed report documents
the inspection findings, which were discussed on July 9, 2009, with you and other members of
your staff.
The inspection examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
This report documents one NRC-identified finding for Unit 1 of very low safety significance
(Green). This finding was determined to involve a violation of NRC requirements. However,
because of the very low safety significance, and because the finding was entered into your
corrective action program, the NRC is treating this finding as a Non-Cited Violation in
accordance with Section VI.A.1 of the NRC Enforcement Policy.
The enclosed inspection report also discusses a finding for Unit 2 that appears to have low to
moderate safety significance (White). As documented in Section 4OA5 of this report, the Unit 2
component cooling water system was inadequately designed to ensure that the system would
be protected from licensing basis events (such as high energy line breaks, seismic and tornado
events) which could occur in the turbine building. The events in the turbine building could cause
a loss of component cooling water inventory from both trains of equipment and a loss of safety
function.
This finding was assessed based on the best available information, including influential
assumptions, using the applicable Significance Determination Process (SDP). The
preliminary safety significance of the finding was determined assuming that the design of the
Unit 2 component cooling water system was inadequate for 1 year.
M. Wadley -2-
This finding was an immediate safety concern, and your staff declared the Unit 2 component
cooling water system inoperable as required by your Technical Specifications. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />,
operations personnel manually closed component cooling water system isolation valves
between the auxiliary building and the turbine building. This allowed the safety-related portion
of the component cooling water system to be returned to an operable status and eliminated the
immediate safety concern.
This finding is also an apparent violation of NRC requirements and is being considered for
escalated enforcement action in accordance with the NRC Enforcement Policy. The current
Enforcement Policy which can be found on the NRCs Web site at http://www.nrc.gov/reading-
rm/doc-collections/enforcement.
In accordance with Inspection Manual Chapter (IMC) 0609, we intend to complete our
evaluation using the best available information and issue our final determination of safety
significance within 90 days of this letter. The SDP encourages an open dialogue between the
staff and the licensee; however, the dialogue should not impact the timeliness of the staffs final
determination.
Before the NRC makes its enforcement decision, we are providing you an opportunity to either:
(1) present to the NRC your perspectives on the facts and assumptions used by the NRC to
arrive at the finding and its significance at a Regulatory Conference, or (2) submit your position
on the finding to the NRC in writing. If you request a Regulatory Conference, it should be held
within 30 days of the receipt of this letter and we encourage you to submit supporting
documentation at least 1 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. The NRC
will also issue a press release to announce the conference. If you decide to submit only a
written response, such submittal should be sent to the NRC within 30 days of the receipt of this
letter. If you decline to request a Regulatory 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.
Please contact John Giessner at (630) 829-9619 within 10 days of the date of this letter to notify
the NRC of your intended response. If we have not heard from you within 10 days, we will
continue with our significance determination and enforcement decision. You will be advised by
a separate correspondence of the results of our deliberations on this matter.
Since the NRC has not made a final determination in this matter, no Notice of Violation is being
issued for this inspection finding at this time. Please be advised that the number and
characterization of the apparent violation described in the enclosed inspection report may
change as a result of further NRC review.
If you contest the subject or severity of a Non-Cited Violation, you should provide a response
within 30 days of the date of this inspection report, with the basis for your denial, to the U.S.
Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-
0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission -
Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of
M. Wadley -3-
Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the
Resident Inspector Office at the Prairie Island Nuclear Generating Plant. In addition, if you
disagree with the characterization of any finding in this report, you should provide a response
within 30 days of the date of this inspection report, with the basis for your disagreement, to the
Regional Administrator, Region III, and the NRC Resident Inspector at the Prairie Island Nuclear
Generating Plant. The information that you provide will be considered in accordance with
Inspection Manual Chapter 0305.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its
enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records (PARS) component of NRC's document system
(ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the
Public Electronic Reading Room).
Sincerely,
/RA by Gary L. Shear for/
K. Steven West, Director
Division of Reactor Projects
Docket Nos. 50-282; 50-306;72-010
License Nos. DPR-42; DPR-60; SNM-2506
Enclosure: Inspection Report 05000282/2009010; 05000306/2009010
w/Attachment: Supplemental Information
cc w/encl: D. Koehl, Chief Nuclear Officer
G. Salamon, Regulatory Affairs Manager
P. Glass, Assistant General Counsel
Nuclear Asset Manager
J. Stine, State Liaison Officer, Minnesota Department of Health
Tribal Council, Prairie Island Indian Community
Administrator, Goodhue County Courthouse
Commissioner, Minnesota Department
of Commerce
Manager, Environmental Protection Division
Office of the Attorney General of Minnesota
Emergency Preparedness Coordinator, Dakota
County Law Enforcement Center
M. Wadley -3-
Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the
Resident Inspector Office at the Prairie Island Nuclear Generating Plant. In addition, if you
disagree with the characterization of any finding in this report, you should provide a response
within 30 days of the date of this inspection report, with the basis for your disagreement, to the
Regional Administrator, Region III, and the NRC Resident Inspector at the Prairie Island Nuclear
Generating Plant. The information that you provide will be considered in accordance with
Inspection Manual Chapter 0305.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its
enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records (PARS) component of NRC's document system
(ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the
Public Electronic Reading Room).
Sincerely,
/RA by Gary L. Shear for/
K. Steven West, Director
Division of Reactor Projects
Docket Nos. 50-282; 50-306;72-010
License Nos. DPR-42; DPR-60; SNM-2506
Enclosure: Inspection Report 05000282/2009010; 05000306/2009010
w/Attachment: Supplemental Information
cc w/encl: D. Koehl, Chief Nuclear Officer
G. Salamon, Regulatory Affairs Manager
P. Glass, Assistant General Counsel
Nuclear Asset Manager
J. Stine, State Liaison Officer, Minnesota Department of Health
Tribal Council, Prairie Island Indian Community
Administrator, Goodhue County Courthouse
Commissioner, Minnesota Department
of Commerce
Manager, Environmental Protection Division
Office of the Attorney General of Minnesota
Emergency Preparedness Coordinator, Dakota
County Law Enforcement Center
See Previous Concurrences
DOCUMENT NAME: G:\Prai\Prairie Island 2009-010 master.doc
Publicly Available Non-Publicly Available Sensitive Non-Sensitive
To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copy
OFFICE RIII RIII EICS RIII RIII
NAME DBetancourt- JGiessner KOBrien LKozak SWest
Roldan:dtp *PL for Per Telecon GLS for
DATE 07/31/09 07/31/09 07/31/09 07/31/09 08/03/09
OFFICIAL RECORD COPY
Letter to M. Wadley from S. West dated August 5, 2009
SUBJECT: PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2
NRC INSPECTION REPORT 05000282/2009010; 05000306/2009010
DISTRIBUTION:
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket Nos: 50-282; 50-306;72-010
License Nos: DPR-42; DPR-60; SNM-2506
Report No: 05000282/2009010; 05000306/2009010
Licensee: Northern States Power Company, Minnesota
Facility: Prairie Island Nuclear Generating Plant, Units 1 and 2
Location: Welch, MN
Dates: June 15 through July 9, 2009
Inspectors: K. Stoedter, Senior Resident Inspector
P. Zurawski, Resident Inspector
L. Kozak, Senior Reactor Analyst
Approved by: J. Giessner, Chief
Branch 4
Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR 05000282/2009010, 05000306/2009010; 06/15/2009 - 7/9/2009; Prairie Island Nuclear
Generating Plant, Units 1 and 2; Inspection of component cooling water system design
deficiency.
This report covers an approximate 1-month period of inspection by the resident inspectors and a
senior reactor analyst. One inspector-identified Green finding and one inspector-identified
preliminary White finding were identified. The Green finding was considered a Non-Cited
Violation of NRC requirements. One apparent violation (AV) was also identified. The
significance of most findings is indicated by their color (Green, White, Yellow, Red) using
Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP).
Cross-cutting aspects were determined using IMC 0305, "Operating Reactor Assessment
Program." Findings for which the SDP does not apply may be Green or be assigned a severity
level after NRC management review. The NRCs 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 and Self-Revealed Findings
Cornerstone: Mitigating Systems
- Green. An inspector identified Non-Cited Violation of 10 CFR Part 50, Appendix B,
Criterion III, Design Control, was identified due to the licensees failure to establish
design control measures to ensure that the design basis for the Unit 1 component
cooling water system (CCW) was correctly translated into specifications, drawings,
procedures, and instructions. Specifically, the licensee failed to ensure that the
safety-related function of the CCW system was maintained following a tornado/high
winds induced failure of the CCW system piping to the 122 spent fuel pool heat
exchanger. Corrective actions for this issue included providing procedural guidance to
isolate the Unit 1 CCW system from the 122 spent fuel pool heat exchanger following
the receipt of a tornado watch and evaluating the need for additional tornado missile
protection for the CCW system piping to the 122 spent fuel pool heat exchanger.
This finding was determined to be more than minor because it impacted the design
control and external events aspects of the Mitigating Systems Cornerstone. The finding
also impacted the Mitigating Systems Cornerstone objective of ensuring the availability,
reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences. This finding was determined to be of very low safety
significance due to the very low probability of the Prairie Island Nuclear Generating Plant
experiencing a high wind condition that could generate a missile large enough to fail the
Unit 1 CCW system piping to the 122 spent fuel pool heat exchanger. The cause of this
finding was related to the cross-cutting element of Human Performance, Decision
Making because the licensee failed to make safety-significant and risk-significant
decisions using a systematic process to ensure that safety was maintained (H.1(a)).
(Section 4OA5.1)
- Preliminary White. An inspector identified apparent violation of 10 CFR Part 50,
Appendix B, Criterion III, Design Control, was identified due to the licensees failure to
establish design control measures to ensure that the design basis for the Unit 2 CCW
1 Enclosure
system was correctly translated into specifications, drawings, procedures, and
instructions. Specifically, the licensee failed to ensure that the safety-related
function of the CCW system was maintained following initiating events (such as high
energy line break, seismic or tornado events) in the turbine building. This issue has
been preliminarily determined to be of low to moderate safety significance (White).
This issue was entered into the licensees corrective action program as corrective action
document 1145695. Upon identifying this issue, the licensee immediately declared the
Unit 2 CCW system inoperable and entered Technical Specification 3.0.3. The
Technical Specification was exited following the closure of several system isolation
valves approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> later. The closure of the isolation valves prevented the
Unit 2 CCW system from being vulnerable to failure following events in the turbine
building.
This finding was determined to be more than minor because it impacted the design
control and external events aspects of the Mitigating Systems Cornerstone. The finding
also impacted the Mitigating Systems Cornerstone objective of ensuring the availability,
reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences. The initiating events in the turbine building could cause the
CCW piping to fail. Loss of CCW inventory affects both trains of CCW based on the
piping arrangement. The loss of both trains of CCW required a phase 3 significance
determination. The results of the phase 3 assessment showed a delta core damage
frequency of 3.2E-6, White. The cause of this finding was related to the cross-cutting
element of Human Performance, Decision Making because the licensee failed to make
safety-significant and risk-significant decisions using a systematic process to ensure that
safety was maintained (H.1(a)). Since both the Unit 1 and Unit 2 cross-cutting aspects
are from the same performance deficiency and are separated based on the risk
determination, the aspect of H.1(a) counts as one cross-cutting aspect in this report.
(Section 4OA5.1).
B. Licensee-Identified Violations
No violations of significance were identified.
2 Enclosure
REPORT DETAILS
4. OTHER ACTIVITIES
4OA5 Other Activities
.1 (Closed) Unresolved Item 05000306/2008005-02: Component Cooling Water System
Susceptible to High Energy Line Break Interaction
a. Inspection Scope
The inspectors reviewed the circumstances surrounding the licensees failure to
adequately design the component cooling water (CCW) system to ensure that the
system was not vulnerable to failure following licensing basis events (such as high
energy line break, seismic and tornado events) in the Unit 1 or Unit 2 turbine buildings.
The turbine building events could cause a loss of CCW inventory and a loss of safety
function.
b. Findings
Introduction: An inspector identified finding of very low safety significance and a
Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control,
was identified for Unit 1 due to the licensees failure to implement design control
measures to ensure that the design of the Unit 1 CCW system was not vulnerable to
failure during a tornado/high wind event.
In addition, an inspector identified apparent violation (AV) of 10 CFR Part 50,
Appendix B, Criterion III, Design Control, was identified for Unit 2 due to the
licensees failure to implement design control measures to ensure that the design of
the Unit 2 CCW system was adequate to mitigate licensing basis events (such as high
energy line break, seismic and tornado events) which could occur in the turbine building.
The events in the turbine building could cause a loss of CCW inventory and a loss of
safety function.
Description: On July 29, 2008, the licensee initiated corrective action program document
(CAP) 1145695 to document that CCW piping located in the turbine building, and used
to supply water to the chemistry cold lab, passed directly underneath high energy piping
for the 15A and 15B feedwater heaters. As part of the CAP review, operations
personnel requested that engineering personnel complete an operability review to
evaluate the impact that a high energy line break (HELB) could have on the continued
operability of the CCW system.
The current design of the CCW system, a safety related system, includes piping in the
auxiliary building and piping in the turbine building. The piping in the turbine building
supplies miscellaneous, non-safety related loads. Piping in the auxiliary building supplies
safety related loads. Either unit can supply the miscellaneous loads. Based upon this
information, the licensee initially determined that a failure of this piping would have no
impact on the continued operability of the Unit 1 CCW system because the Unit 1 CCW
system was not supplying water to the chemistry cold lab at the time this issue was
identified. The licensee believed that there could be some impact on the Unit 2 CCW
3 Enclosure
system because the 2A train of CCW normally supplied water to the chemistry cold lab.
As a result, the licensee continued to review this issue.
On July 31, 2008, the licensee identified that a failure of a Unit 1 or a Unit 2 turbine
building high energy line could impact the continued operability of the Unit 2 CCW
system. The licensee conducted an additional operability review and determined that
the Unit 2 CCW system was inoperable because initiating events could cause a
complete loss of CCW inventory, if the CCW piping is severed. This would cause the
CCW system to drain since there is no way to separate the safety related loads from the
non-safety related loads. The licensee also determined that the operators ability to
bring Unit 2 to a cold shutdown condition following a HELB and a failure of the CCW
system was impacted. Operations personnel immediately declared both trains of the
Unit 2 CCW system inoperable and entered Technical Specification (TS) 3.0.3. While
making preparations to shut down Unit 2, operations personnel identified and closed
several CCW system manual isolation valves. This eliminated the immediate safety
concern and allowed the safety-related portion of the CCW system to be returned to an
operable status. The non-safety related CCW piping located in the turbine building
remains isolated.
In mid-September 2008, the licensee completed an apparent cause evaluation (ACE)
and determined that the interaction between the turbine building high energy piping and
the CCW system had been identified in July 2006. However, no actions were initiated to
assess the continued operability of the CCW system. The September 2008 ACE further
stated that as part of a resolution to the July 2006 CAP, the licensee planned to perform
a study to eliminate CCW as the cooling medium for the chemistry cold lab. While
reviewing other CAPs associated with this issue, the inspectors found a draft June 2007
and the finalized January 2008 study report (attached to a CAP initiated in 2004). Both
reports concluded that none of the CCW piping to or from the chemistry cold lab had
been analyzed for susceptibility of failure due to a HELB or a failure from another
system. In addition, the study report indicated that other piping near the CCW piping
were high energy lines that could damage the CCW piping. The study concluded that if
the HELB occurred near the CCW piping the CCW system would fail in approximately
6 minutes.
The inspectors reviewed several historical documents as part of this inspection. The
inspectors determined that the licensee had identified potential design deficiencies with
the CCW piping located in the turbine building multiple times. However, the licensee
failed to properly prioritize the resolution of these deficiencies. During this inspection,
the NRC also questioned the licensee regarding the need to include HELBs that could
be induced from seismic and tornado/high wind events as part of their evaluation. After
several discussions, the licensee agreed to include these external initiators as part of
their evaluation.
On March 23, 2009, the licensee initiated CAP 1174370 to document that a portion of
CCW piping that supplies water to the 122 spent fuel pool heat exchanger was not
adequately protected from missiles generated during tornado/high wind events. The
inspectors considered this an additional example of a failure to adequately design the
CCW system to ensure that it was adequately protected from failure. This issue was
specific to the Unit 1 CCW system as this was the system that was normally aligned to
supply cooling water to the 122 spent fuel pool heat exchanger.
4 Enclosure
On April 15, 2009, CAP 1178236 was initiated to document that a turbine building
HELB could result in internal flooding of safety-significant areas due to the release
of feedwater/condensate from the pipe break, consequential failure of a cooling
water (service water) line, and potential actuation of the fire protection system
sprinklers/deluge system. This issue remained under licensee review at the
conclusion of the inspection period.
Analysis: The Prairie Island licensing basis requires that the CCW safety related system
is designed to ensure no loss of safety function occurs from natural phenomena such as
earthquakes or high winds/tornados. In addition, the CCW safety function is required to
be maintained for all licensing basis events including HELBs.
(1) Unit 1 Risk Analysis
The inspectors determined that the licensees failure to design the Unit 1 CCW system
such that it would be protected during licensing basis events (such as HELB, seismic
and tornado events) was a performance deficiency that required evaluation using the
This finding was determined to be more than minor since it impacted the design control
(initial design) and protection against external factors (seismic and weather) aspects of
the Mitigating Systems Cornerstone. In addition, the finding impacted the cornerstone
objective of ensuring the availability, reliability, and capability of mitigating systems
equipment used to respond to events and prevent core damage. Since the failure of the
CCW piping could potentially represent a loss of CCW system safety function following
an external event, Table 4a of IMC 0609 required the completion of a phase 3 SDP
evaluation. The senior reactor analyst (SRA) conducted a phase 3 evaluation to
determine the risk contribution from a tornado/high wind event that could cause a failure
of the Unit 1 CCW system. The SRA determined that the frequency of tornado/high wind
events that could create missiles large enough to damage the Unit 1 CCW piping to the
122 spent fuel pool heat exchanger was less than 1E-6/yr. As a result, this issue was
determined to be of very low safety significance (Green).
The cause of this finding was related to the cross-cutting element of Human
Performance, Decision Making because the licensee failed to make safety-significant
and risk-significant decisions regarding the design of the Unit 1 CCW system using a
systematic process to ensure that safety was maintained (H.1(a)). Since both the Unit 1
and Unit 2 cross-cutting aspects are from the same performance deficiency and are
separated based on the risk determination, the aspect of H.1(a) counts as one cross-
cutting aspect in this report.
(2) Unit 2 Risk Analysis
The inspectors determined that the licensees failure to design the Unit 2 CCW system
such that it would be protected during licensing basis events (such as HELB, seismic
and tornado events) was a performance deficiency that required evaluation using the
SDP.
The finding was determined to be more than minor since it impacted the design control
(initial design) and protection against external factors (seismic and weather) aspects of
the Mitigating Systems Cornerstone. In addition, the finding impacted the cornerstone
5 Enclosure
objective of ensuring the availability, reliability, and capability of mitigating systems
equipment used to respond to events and prevent core damage. Since the failure of the
CCW piping could be caused by an internal or an external event, and potentially
represent a loss of CCW system safety function, a phase 2 SDP evaluation was
required.
For the phase 2 SDP evaluation, the SRA used the transient with the loss of the power
conversion system worksheet to represent the event which could result in a loss of CCW
for Unit 2. The worksheets were solved assuming the condition existed for an exposure
period of 1 year and that all the functions supported by the CCW system were
unavailable. The CCW system provides cooling for all emergency core cooling system
(ECCS) pumps, the CCW heat exchangers, and one method of reactor coolant pump
(RCP) seal cooling. The unavailability of the CCW system affected the high pressure
injection, high pressure recirculation, feed and bleed, and RCP seal cooling functions.
The result of the phase 2 evaluation was a Red finding using the counting rule.
However, the SRA determined that the results were overly conservative because the
worksheet did not account for the fact that the initiating events in the turbine building that
would result in the loss of the CCW function were a subset of all transient with the loss of
the power conversion system events. The SRA determined that a phase 3 SDP
evaluation was necessary.
For the phase 3 SDP evaluation, the NRC staff determined that a number of initiating
events could cause a break in the non-safety-related CCW piping in the turbine building
and the subsequent failure of the Unit 2 CCW system. These initiating events included
Unit 1 HELB events, Unit 2 HELB events, earthquakes, and tornadoes. The SRA
primarily used the Risk Assessment of Operational Events (RASP) handbook and its
references and the results from a licensee risk evaluation in the phase 3 SDP
evaluation.
To estimate the frequency of HELB events that could result in the loss of the Unit 2
CCW system, the SRA used the generic pipe rupture frequency of 1.2E-10/ft-hr from
Table 3A-2-1 of the RASP handbook for External Events. This value is taken from
EGG-SSRE-9639, Component External Leakage and Rupture Frequency Estimates.
The licensee provided the length of HELB piping that, upon rupture, could impact the
CCW system. The total linear feet of Unit 1 high energy piping from the feedwater or
condensate system that could rupture and impact the Unit 2 CCW system was 167 feet.
The total length of Unit 2 HELB piping that could cause the same effect was 78 feet.
Using this reference and the pipe lengths provided by the licensee, the initiating event
frequencies for Unit 1 and Unit 2 were estimated to be 1.8E-4/yr and 8.6E-5/yr,
respectively.
To estimate the conditional core damage probability (CCDP) given the HELB-induced
loss of CCW, the SRA used the licensees risk evaluation results. The calculation
assumed that either a Unit 1 or Unit 2 HELB impacted the CCW system piping to the
chemistry cold lab resulting in a CCW system pipe break and draining of the Unit 2 CCW
surge tank in approximately 6 minutes. This resulted in a failure of the Unit 2 CCW
system that was not recoverable. The main feedwater/condensate, reactor water
makeup, instrument air systems and auxiliary feedwater system crosstie were also
assumed to be impacted by the initiating event and were considered to be unavailable in
the CCDP calculation. The licensee calculated a CCDP for Unit 2 under these
6 Enclosure
conditions of 1.2E-2. The dominant sequences involved a loss of CCW event, followed
by the failure of the charging system, which resulted in the loss of all RCP seal cooling
and a RCP seal loss of coolant accident (LOCA). If the seal LOCA was large enough
(greater than 76 gallons per minute) it would not be able to be mitigated because all of
the ECCS pumps were unavailable. This led to an unmitigated LOCA. When the
initiating event frequency was combined with the CCDP, a total estimated change in
core damage frequency (CDF) of 3.2E-6/yr was estimated. The SRA assumed that the
baseline CDF was negligible compared to the CDF related to the finding. Therefore, the
change in CDF was approximately equal to the CDF of the finding. The SRA assumed
that if no design deficiency existed then the impact of the postulated non-safety-related
failures would be limited to a transient with a loss of feedwater event rather than a
transient with the loss of CCW function and as a result, baseline risk would be much
lower than risk calculated given the plant condition created by the performance
deficiency.
The SRA performed the same CCDP calculation using the Prairie Island Standardized
Plant Analysis Risk Model (revision 3.47) and the same assumptions as the licensee.
The Standardized Plant Analysis Risk Model produced similar results to the licensees
risk evaluation.
The postulated loss of Unit 2 CCW from a Unit 1 HELB during times when Unit 2 was
shutdown and on shutdown cooling (SDC) was qualitatively evaluated. A loss of CCW
under these conditions would ultimately result in a loss of SDC. Given that the exposure
period of this condition was limited, decay heat was lower, and additional options for
using the charging system for injection or recovering CCW were available, the risk
contribution from a shutdown scenario was considered to be much less than the internal
event at-power scenario.
The SRA also considered the risk contribution from earthquakes and tornadoes causing
a failure of the non-safety-related CCW piping. The SRA determined that the frequency
of seismic events or high wind events damaging the CCW line was less than the
frequency of HELB events and as a result the risk contribution from external events was
less than 1E-6/yr.
The impact of turbine building flooding as a result of the postulated HELB event was not
explicitly evaluated in this phase 3 SDP. Recent information from the licensee indicated
that a subset of the postulated HELB events that impacted the CCW system could also
result in flooding the turbine building if a cooling water pipe was also ruptured and the
fire deluge systems were initiated. If sufficient flooding occurred, the auxiliary feedwater,
direct current (DC) power, and emergency diesel generator systems could also be
impacted, in addition to the CCW system. Quantitative consideration of this impact
would increase the estimated delta CDF of this finding. Note that if the turbine building
roll-up doors were open (usually during summer months), then flooding would not impact
these other systems.
The SRA quantitatively bounded the potential risk impact associated with the turbine
building flooding by assuming that for 9 months of a year when the turbine building roll-
up doors were closed (cold weather months) a HELB-induced loss of CCW for Unit 2
would also result in severe turbine building flooding 10 percent of the time. The result of
this bounding evaluation which included the potential flooding effects was a delta CDF of
less than 1.0E-5/yr. The SRA determined further quantitative evaluation of the internal
7 Enclosure
flooding effects related to this finding was unnecessary because the result was not likely
to change the overall conclusion of this SDP evaluation.
The large early release frequency (LERF) impact was negligible. Using the insights from
IMC 0609, Appendix H, Containment Integrity SDP, LERF was not impacted because
the dominant core damage sequences did not involve an inter-system LOCA or a steam
generator tube rupture event.
The conclusion of the phase 3 SDP was an estimated delta CDF of 3.2E-6/yr which
represented a finding of low-to-moderate (White) safety significance. The dominant
sequences involved a HELB-induced loss of CCW event, followed by the failure of the
charging system, which results in the loss of all RCP seal cooling. If the seal LOCA was
large enough it could not be mitigated because all ECCS were unavailable due to the
loss of CCW.
The cause of this finding was related to the cross-cutting element of Human
Performance, Decision Making because the licensee failed to make safety-significant
and risk-significant decisions regarding the continued operability of the Unit 2 CCW
system using a systematic process to ensure that safety was maintained (H.1(a)).
(3) Old Design Issue Review
NRC IMC 0305, Operating Reactor Assessment Program, Section 4.11 defined an
old design issue as an inspection finding involving a past design-related problem in the
engineering calculations or analysis, associated operating procedure, or installation of
plant equipment that does not reflect a performance deficiency associated with existing
licensee programs, policy, or procedures. IMC 0305 stated that the NRC can refrain
from considering safety significant inspection findings in the assessment program for a
design-related finding as long as the following statements were true:
- The issue was licensee identified as a result of a voluntary initiative such as a
design basis reconstitution;
- The performance issue was or will be corrected within a reasonable period of
time following identification;
- The issue was not likely to have been previously identified by routine efforts
such as normal surveillance or quality assurance activities; and
- The issue does not reflect a current performance deficiency associated with
existing licensee programs, policy, or procedures.
With regards to Unit 1, the inspectors determined that this issue did not qualify as an old
design issue due to being inspector identified. Specifically, the NRCs request that the
licensee include tornado/high wind induced piping failures as part of their analysis
resulted in identifying the design vulnerability associated with the Unit 1 CCW piping to
the 122 spent fuel pool heat exchanger. In addition, this issue was not identified as part
of a voluntary initiative. Instead, it was identified as part of the extent of condition review
completed for CAP 1145695.
8 Enclosure
Although the Unit 2 CCW design deficiency was not likely to have been identified
through routine efforts, licensee documentation indicated that CCW design issues
associated with the CCW system piping to the chemistry cold lab piping was identified in
the 1990s. Due to the long time that had elapsed since the initial identification of this
issue, the inspectors concluded that the performance deficiency discussed above had
not been corrected within a reasonable period of time due to the failure to properly
prioritize the implementation of corrective actions. In addition, the failure to properly
prioritize the implementation of corrective actions was reflective of a current performance
deficiency associated with the licensees corrective action program. Finally, there was
operating experience including generic communication from the NRC which provided
information which the licensee should have evaluated in the corrective action process.
The licensee had also listed corrective action effectiveness as one of the six drivers of
their recent performance decline. As a result, the inspectors concluded that the Unit 2
CCW design deficiency did not meet the criteria for being considered an old design
issue.
Enforcement
(1) Unit 1 Enforcement
10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that
measures be established to assure that the design basis for safety-related functions of
structures, systems, and components are correctly translated into specifications,
drawings, procedures, and instructions.
The Prairie Island licensing basis requires that the CCW safety related system is
designed to ensure no loss of safety function occurs from natural phenomena such as
earthquakes or high winds/tornados. In addition, the CCW safety function is required to
be maintained for all licensing basis events including HELBs.
Contrary to the above, as of March 23, 2009, the licensee had failed to establish
measures to assure that the design basis for the Unit 1 CCW system had been correctly
translated into specifications, drawings, procedures, and instructions. Specifically, the
licensee failed to ensure that the Unit 1 CCW system was protected from failure
following a tornado/high wind event. In addition, the events in the turbine building could
have resulted in a loss of CCW inventory and a loss CCW safety function if the Unit 1
CCW system had been aligned to supply cooling water to the chemistry cold lab.
However, because this violation was of very low safety significance (Green) and was
entered into your CAP as CAP 1174370, it was treated as an NCV consistent with
Section VI.A.1 of the Enforcement Policy (NCV 05000282/2009010-01). Corrective
actions for this issue included providing procedural guidance to isolate CCW from the
122 spent fuel pool heat exchanger following the receipt of a tornado watch and
evaluating the need for additional tornado missile protection for the CCW system piping
to the 122 spent fuel pool heat exchanger.
(2) Unit 2 Enforcement
10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that
measures be established to assure that the design basis for safety related functions of
structures, systems, and components are correctly translated into specifications,
9 Enclosure
drawings, procedures, and instructions. Further, Criterion III requires that the design
control measures shall provide for verifying or checking the adequacy of designs.
The Prairie Island licensing basis requires that the CCW safety related system is
designed to ensure no loss of safety function occurs from natural phenomena such as
earthquakes or high winds/tornados. In addition, the CCW safety function is required to
be maintained for all licensing basis events including HELBs.
Contrary to the above, as of July 31, 2008, the licensee had failed to implement design
control measures to ensure that the design basis for the Unit 2 CCW system was
correctly translated into specifications, drawings, procedures, and instructions.
Specifically, the licensee failed to ensure that the design of the Unit 2 CCW system was
adequate to mitigate licensing basis events (such as high energy line break, seismic and
tornado events) which could occur in the turbine building. The events in the turbine
building could cause a loss of CCW inventory for both trains of CCW. This would result
in a loss of the CCW safety function. This is an apparent violation of 10 CFR Part 50,
Appendix B, Criterion III pending the completion of the final significance determination
.2 (Closed) Licensee Event Report 05000282/2009-003-00: Component Cooling Water
System Vulnerability to Tornado Missile Hazard
This licensee event report discusses the Unit 1 CCW vulnerability discussed above. No
new information was provided in the report. The inspectors determined that this issue
constituted a finding of very low safety significance and an NCV of 10 CFR Part 50,
Appendix B. This licensee event report is closed.
4OA6 Management Meetings
.1 Exit Meeting Summary
On July 9, 2009, the inspectors presented the inspection results to M. Wadley and other
members of the licensee staff. The licensee acknowledged the issues presented. The
inspectors confirmed that none of the potential report input discussed was considered
proprietary.
ATTACHMENT: SUPPLEMENTAL INFORMATION
10 Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
M. Wadley, Site Vice President
B. Sawatzke, Director Site Operations
K. Ryan, Plant Manager
J. Anderson, Regulatory Affairs Manager
L. Clewett, Business Support Manager
B. Flynn, Safety and Human Performance Manager
R. Hite, Radiation Protection and Chemistry Manager
D. Kettering, Site Engineering Director
J. Lash, Operations Manager
R. Madjerich, Production Planning Manager
J. Muth, Nuclear Oversight Manager
S. Northard, Performance Improvement Manager
M. Schmidt, Maintenance Manager
J. Sternisha, Training Manager
NRC
J. Giessner, Reactor Projects Branch 4 Chief
T. Wengert, Office of Nuclear Reactor Regulation Project Manager
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
Opened
05000282/2009010-01 NCV Failure to Ensure Design Measures Were Appropriately
Established for The Unit 1 Component Cooling Water
System (Section 4OA5.1)05000306/2009010-02 AV Failure to Ensure Design Measures Were Appropriately
Established for The Unit 2 Component Cooling Water
System (Section 4OA5.1)
Closed
05000282/2009010-01 NCV Failure to Ensure Design Measures Were Appropriately
Established for The Unit 1 Component Cooling Water
System
05000306/2008005-02 URI Component Cooling Water System Susceptible to High
Energy Line Break Interaction
05000282/2009-003-00 LER Component Cooling Water System Vulnerability to Tornado
Missile Hazard
1 Attachment
LIST OF DOCUMENTS REVIEWED
The following is a list of documents reviewed during the inspection. Inclusion on this list does
not imply that the NRC inspectors reviewed the documents in their entirety, but rather, that
selected sections of portions of the documents were evaluated as part of the overall inspection
effort. Inclusion of a document on this list does not imply NRC acceptance of the document or
any part of it, unless this is stated in the body of the inspection report.
4OA5 Other Activities
- CAP 1162511; Missed Opportunities to Identify HELB and CC System Interaction;
December 15, 2008
- CAP 1145695; CC Piping Adjacent to HELB Location in Turbine Building; July 29, 2008
- ACE 1145695-04; September 4, 2009
- 478-AI-01; USAR Appendix I - Flooding; Revision 22
- 478-AI-03; USAR Appendix I - Pipe Stress and Pipe Whip; Revision 22
- CAP 737382; Non-Seismic Equipment in CC System Pressure Boundary; August 2, 2004
- NF-39297-1; Sampling System Units 1 and 2; Revision R
- FC-61-350; Component Cooling Water to H2 Generator as Built Print Record; Revision 2
- DC-496; Component Cooling Design Change; June 14, 1974
- FC-61-348; H2 Generator Cooling Return; Revision 2
- PI-233-39P23A; Pipe Stress Analysis - CC System, Part 2A; Revision 0
- PI-233-39P23A; Pipe Stress Analysis - CC System, Part 2B; Revision 0
- PI-233-39P23A; Pipe Stress Analysis - CC System, Part 2C; Revision 0
- PI-233-39P23A; Pipe Stress Analysis - CC System, Part 2D; Revision 0
- ENG-CS-278; Seismic Qualification of Components in CC System Pressure Boundary;
Revision 1
- CAP 1163206; OBM for CAP 737382; December 19, 2008
- ACE 1163206;
- C1.1.14-1; Unit 1 Component Cooling System; Revision 24
- LER 2-08-01; Unanalyzed Condition Due to Both Trains of Component Cooling Being
Susceptible to a Postulated High Energy Line Break; September 29, 2008
- NF 39297-3; Sampling Systems - Unit 1 & 2; Revision XX
- NF 39246-1; Unit 2 Component Cooling System Flow Diagram; Revision S
- NF-39246-2; Unit 2 Component Cooling System Flow Diagram; Revision G
- USAR Appendix I; Postulated Pipe Failure Analysis Outside of Containment; Revision 29
- SE 487-AI-01; USAR Appendix I Review - I.5.5 Flooding; Revision 1
- 2C14 AOP1; Loss of Component Cooling; Revision 16
- 1C14; Component Cooling System Unit 1; Revision 26
- C1.1.14-1; Unit 1 Component Cooling System; Revision 24
- 2C14; Component Cooling System Unit 2; Revision 27
- C1.1.14-2; Unit 2 Component Cooling System; Revision 29
- 2C12.1 AOP1; Loss of Reactor Coolant Pump Seal Injection; Revision 3
- 2C3 AOP2: Loss of Reactor Coolant Pump Seal Cooling; Revision 6
- 2E-0; Reactor Trip or Safety Injection; Revision 25
- 2E-1; Loss of Reactor or Secondary Coolant; Revision 22
- 2E-4; Core Cooling Following Loss of Residual Heat Removal Flow; Revision 11
- CAP 31550; Turbine Building HELB Analysis; February 18, 2002
- Sulzer Pump Letter dated February 5, 2009; SI Pump Loss of Cooling Water for Lube System
Review
- CAP 34876; Turbine Building HELB Analysis; December 21, 1969
2 Attachment
- CAP 53226; Delay HELB Analysis Corrective Action Due to Budget Does Not Comply with
5AWI3.15.5
- CAP 5530065; Evaluate Turbine Building HELB Analysis; November 23, 2003
- Operability Evaluation 1145695; Revision 0
- Project Review Group Meeting Minutes; September 12, 2008
- OPR000509; Non-Seismic Equipment in CC System Pressure Boundary; August 3, 2004
- Completion Notes for CAP 737382-15; No Date
- Review of Seismic Qualification of Components Identified in CAP 037749; No Date
- CE005702; Condition Evaluation for Non-Seismic Equipment in CC System Pressure
Boundary; No Date
- Results of July 15, 2005, Meeting Held on Qualification of CC Piping 1-CC-79, 1-CC-80, and
1-CC-138 Routed to Sample Coolers in the Cold Lab and to 123 Nitrogen Compressor
- CAP 826114; Perform Seismic Analysis of 1-CC138 up to CC-71-1 and CC-71-2;
March 29, 2005
- CAP 870304; Update USAR to Incorporate Turbine Building HELB Analysis; July 26, 2005
- CAP 1002268; HELB Project Cost Overruns; October 28, 2005
- CAP 1162511; Missed Opportunities to Identify HELB and CC System Interaction;
December 15, 2008
- CAP 1143812; Turbine Building HELB Funding Delays Could Affect Project Success;
July 10, 2008
- 03Q0418-C-002; Assessment of Turbine Building for N-S Tornado; Revision 1
- 03Q0418-C-003; Assessment of Turbine Building for E-W Tornado; Revision 1
- Component Cooling Piping Project Engineering Work Scope; no date
- NF-38500; Architectural Ground Floor Plan Elevation 695-0; Revision P-76
- NF-38501; Architectural Ground Floor Plan Elevation 715-0; Revision AE
- NF-38502; Architectural Ground Floor Plan Elevation 735-0; Revision P-76
- Prairie Island Response to Generic Letter 87-02; November 20, 1995
- SL-11973-014; Chemistry Lab Component Cooling Study; January 2008
- P9160S-001; Simulator Exercise Guide Scenario #1 (CC HELB Scenario); April 7, 2009
- RM-4; Job Performance Measure - Switch Reactor Makeup Tanks on Degas; March 20, 2009
- WO 379471-01; 22 Reactor Makeup Pump; March 18, 2009
- WO 379471 50.59 Screening - Measure Unit 2 RMU Flows to CC Surge Tank and VCT for
PRA; Revision 0
- WO 379471 Reactivity Management Screening Checklist; No Date
- SI-3; Job Performance Measure - Switchover to Recirculation Per 2ES-1-2, Attachment K;
June 16, 2009
- USAR Section 10.2.5; Reactor Makeup Water Deoxygenation System; Revision 25
- NF 39242; Flow Diagram Unit 1 and 2 Reactor Make-up and Demineralized Water Systems;
Revision BI
- CAP 1174370; No Tornado Protection for Component Cooling Water Piping to the 122 Spent
Fuel Pool Heat Exchanger; March 23, 2009
- OPR 1174370-01; Operability Review for CAP 1174370; March 24, 2009
- ACE 1174370-08; Apparent Cause Evaluation for CAP 1174370; June 11, 2009
- Abnormal Procedure AB-2; Tornadoes, Thunderstorms, and High Winds; Revision 33
- Temporary Procedure Change Request 033B; Add Steps to AB-2; May 22, 2009
- Procedure 1C14 AOP 1; Loss of Component Cooling Water; Revision 16
- CAP 1178236; No High Energy Line Break Flooding Calculation for the Turbine Building;
April 15, 2009
- Reactor Coolant Pump Seal Loss of Coolant Accident Break Sizes and Loss of Coolant
Accident Success Criteria; February 10, 2009
- Prairie Island Nuclear Generating Plant CC/HELB Summary Report; June 5, 2009
3 Attachment
- Calculation PI-996-61-M01; Evaluation of the Effects of Jet Impingement on Component
Cooling Water Piping in the Turbine Building; Revision 0
- Letter from A. Washburn, Sulzer Pumps, to S. Skoyen, Prairie Island Nuclear Plant;
June 4, 2009
- MPR Report - Prairie Island Nuclear Generating Plant Safety Injection Pump Operability
Evaluation; May 12, 2009
- Engineering Change 13762; Prairie Island Turbine Building HELB Sensitivity Study;
February 11, 2009
- NF-39246-1; Unit 2 Component Cooling Water System Flow Diagram; Revision 76
4 Attachment
LIST OF ACRONYMS USED
ACE Apparent Cause Evaluation
ADAMS Agencywide Document Access Management System
AV Apparent Violation
CAP Corrective Action Program Document
CCDP Conditional Core Damage Probability
CDF Core Damage Frequency
CFR Code of Federal Regulations
CCW Component Cooling Water
DC Direct Current
DRP Division of Reactor Projects
ECCS Emergency Core Cooling System
IMC Inspection Manual Chapter
LERF Large Early Release Frequency
LOCA Loss of Coolant Accident
NCV Non-Cited Violation
NRC U.S. Nuclear Regulatory Commission
PARS Publicly Available Records
RASP Risk Assessment of Operational Events
RCP Reactor Coolant Pump
SDP Significance Determination Process
SRA Senior Reactor Analyst
TS Technical Specifications
USAR Updated Safety Analysis Report
5 Attachment