ML061950308
| ML061950308 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 07/14/2006 |
| From: | William Jones NRC/RGN-IV/DRP/RPB-B |
| To: | Naslund C Union Electric Co |
| References | |
| IR-06-011 | |
| Download: ML061950308 (35) | |
See also: IR 05000483/2006011
Text
July 14, 2006
Charles D. Naslund, Senior Vice
President and Chief Nuclear Officer
Union Electric Company
P.O. Box 620
Fulton, MO 65251
SUBJECT: CALLAWAY PLANT - NRC SPECIAL INSPECTION REPORT 05000483/2006011
Dear Mr. Naslund:
On April 11-14, 2006, the U.S. Nuclear Regulatory Commission (NRC) conducted a special
inspection at your Callaway Plant. The inspection effort continued with in-office and additional
on-site reviews through June 16, 2006. The purpose of the inspection was to evaluate the
impact of the discovery that component cooling water would not be established to the residual
heat removal heat exchangers until after the postloss-of-coolant accident recirculation phase
was initiated. The enclosed report documents the inspection findings, which were discussed on
June 26, 2006, with Mr. Tim Herrmann and members of your staff.
The inspection was conducted as a result of your staffs identification, during a plant simulator
exercise, that component cooling water to the residual heat removal heat exchangers would not
have been established until the containment recirculation phase of emergency core cooling
system injection had been initiated. The failure to establish procedures that were consistent
with the safety analysis could have challenged the ability of the emergency core cooling system
in performing its safety functions during the containment recirculation phase. As discussed in
detail in the enclosed report, because the underlying safety concern was corrected on
March 30, 2006, and does not represent a current safety concern, the inspection focused on
the circumstances that lead up to your staff identifying this condition, AmerenUEs response,
including the root cause and extent of condition reviews, and the identification of any generic
issues related to the design and operating practices that resulted in this condition.
This inspection report documents several opportunities prior to March 27, 2006, including
operating experience and review of other emergency operating procedure deficiencies, to
identify that the established emergency operating procedures did not ensure that the facility
would be operated in accordance with the safety analysis. In addition, the inspection team
identified that, after the condition was identified, the immediate actions that were taken to place
the plant in a configuration to meet the safety analysis did not adequately consider the
component cooling water system response to a loss of offsite power. The plant was
subsequently placed in a configuration that supports the design basis component cooling water
system requirements.
Union Electric Company -2-
Based on the results of this inspection, the NRC identified two findings, each evaluated under
the risk significance determination process as having very low safety significance (Green). The
NRC also determined that there was a violation associated with each of the findings. These
violations are being treated as noncited violations, consistent with Section VI.A of the
Enforcement Policy. These noncited violations are described in the subject inspection report.
In addition, a licensee-identified violation, which was determined to be of very low safety
significance, is listed in the report. If you contest these violations or the significance of the
violations, 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 copies to the Regional Administrator, U.S.
Nuclear Regulatory Commission, Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington,
Texas 76011-4005; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001; and the NRC Resident Inspector at the Callaway Plant facility. 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 component of NRC's document system (ADAMS).
ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the
Public Electronic Reading Room).
Sincerely,
/RA/
William B. Jones, Chief
Project Branch B
Division of Reactor Projects
Docket: 50-483
License: NPF-30
Enclosure:
Inspection Report 05000483/2006011
w/attachments: Supplemental Information
Timeline Describing CCW to RHR Heat Exchangers Problem
Charter Memorandum dated April 10, 2006
cc w/enclosure:
Professional Nuclear Consulting, Inc.
19041 Raines Drive
Derwood, MD 20855
John ONeill, Esq.
Pillsbury Winthrop Shaw Pittman LLP
2300 N. Street, N.W.
Washington, DC 20037
Union Electric Company -3-
Keith A. Mills, Supervising Engineer,
Regional Regulatory Affairs/
Safety Analysis
AmerenUE
P.O. Box 620
Fulton, MO 65251
Missouri Public Service Commission
Governors Office Building
200 Madison Street
P.O. Box 360
Jefferson City, MO 65102
H. Floyd Gilzow
Deputy Director for Policy
Missouri Department of Natural Resources
P. O. Box 176
Jefferson City, MO 65102-0176
Rick A. Muench, President and
Chief Executive Officer
Wolf Creek Nuclear Operating Corporation
P.O. Box 411
Burlington, KS 66839
Dan I. Bolef, President
Kay Drey, Representative
Board of Directors Coalition
for the Environment
6267 Delmar Boulevard
University City, MO 63130
Les H. Kanuckel, Manager
Quality Assurance
AmerenUE
P.O. Box 620
Fulton, MO 65251
Director, Missouri State Emergency
Management Agency
P.O. Box 116
Jefferson City, MO 65102-0116
Keith D. Young, Manager
Regulatory Affairs
AmerenUE
P.O. Box 620
Fulton, MO 65251
Union Electric Company -4-
David E. Shafer
Superintendent, Licensing
Regulatory Affairs
AmerenUE
P.O. Box 66149, MC 470
St. Louis, MO 63166-6149
Certrec Corporation
4200 South Hulen, Suite 630
Fort Worth, TX 76109
Keith G. Henke, Planner
Division of Community and Public Health
Office of Emergency Coordination
930 Wildwood, P.O. Box 570
Jefferson City, MO 65102
Chief, Radiological Emergency
Preparedness Section
Kansas City Field Office
Chemical and Nuclear Preparedness
and Protection Division
Dept. of Homeland Security
9221 Ward Parkway
Suite 300
Kansas City, MO 64114-3372
Union Electric Company -5-
Electronic distribution by RIV:
Regional Administrator (BSM1)
DRP Director (ATH)
DRS Director (DDC)
DRS Deputy Director (RJC1)
Senior Resident Inspector (MSP)
Branch Chief, DRP/B (WBJ)
Senior Project Engineer, DRP/B (RAK1)
Team Leader, DRP/TSS (RLN1)
RITS Coordinator (KEG)
J. Lamb, OEDO RIV Coordinator (JGL1)
ROPreports
CWY Site Secretary (DVY)
W. A. Maier, RSLO (WAM)
SUNSI Review Completed: ___WBJ_ ADAMS: / Yes G No Initials: __WBJ_
/ Publicly Available G Non-Publicly Available G Sensitive / Non-Sensitive
R:\_REACTORS\_CW\2006\CW2006011RP-DED.wpd
RIV:RI:DRP/B RI:DRS/EB2 SRA:DRS C:DRS/EB2 C:DRP/B
DEDumbacher GAPick DPLoveless LJSmith WBJones
E-WBJones E- WBJones /RA/ /RA/ /RA/
7/10/06 7/10/06 7/14/06 7/13/06 7/13/06
OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket: 50-483
License: NPF-30
Report: 05000483/2006011
Licensee: AmerenUE
Facility: Callaway Plant
Location: Junction Highway CC and Highway O
Fulton, Missouri
Dates: April 11-14, 2006, with additional on-site in-office inspection through
June 16, 2006
Team Leader: D. Dumbacher, Senior Resident Inspector, Project Branch B
Inspectors: G. Pick, Senior Reactor Inspector, Engineering Branch
D. Loveless, Senior Reactor Analyst
Approved By: W. B. Jones, Chief, Project Branch B, Division Reactor Projects
-1- Enclosure
SUMMARY OF FINDINGS
IR 05000483/2006011; 04/11-06/16/06; Callaway Plant: Special Inspection to evaluate
AmerenUEs discovery that component cooling water flow to the residual heat removal heat
exchangers would not have been established until after the postloss-of-coolant accident
recirculation phase was initiated.
This report covered the initial on-site inspection conducted April 11-14, 2006, with in-office
review and additional on-site inspection conducted through June 16, 2006, by a special
inspection team consisting of one resident inspector, one region-based reactor inspector, and
one region-based senior reactor analyst. Two noncited violations were identified. The
significance of most findings is indicated by their color (Green, White, Yellow, Red) using
Inspection Manual Chapter 0609, "Significance Determination Process." 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 3, dated July 2000.
A. NRC-Identified and Self Revealing Findings
Cornerstone: Mitigating Systems
- Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B,
Criterion XVI, for the failure to take adequate corrective action to prevent recurrence of
a significant condition adverse to quality. Specifically, AmerenUE failed to correct the
Emergency Operating Procedure deficiencies associated with Final Safety Analysis
Report requirements following an April 15, 1998 notification of the same deficiencies at
another standardized nuclear unit power plant system plant. At that time AmerenUE did
not identify and correct similar deficiencies involving the component cooling water
system support function for residual heat removal heat exchangers. The Emergency
Operating Procedure deficiencies were discovered by plant personnel on March 27,
2006, during a simulator exercise involving the transition to the emergency core cooling
system recirculation phase. Problem identification and resolution crosscutting aspects
were identified for the failure to adequately identify and correct Emergency Operating
Procedures deficiencies to ensure operation within the design basis.
This issue was more than minor because it affected the Mitigating Systems cornerstone
objective of equipment reliability. The failure to provide for component cooling water
system flow through the residual heat removal heat exchangers for initial containment
recirculation could result in a loss of the component cooling water system and thus
become a much more significant safety concern. AmerenUEs evaluation of the
condition was considered for the time allowable to establish component cooling water
flow before a loss of the component cooling water system would occur. AmerenUE
provided an evaluation that demonstrated a loss of component cooling water would not
occur based on the timing of operator actions. Because the timing did affect the
probabilistic risk assessment for human reliability, a Phase 3 risk assessment was
performed by an NRC senior reactor analyst. The analyst determined that the finding
-2- Enclosure
was of very low safety significance, Green. AmerenUE entered this issue into their
corrective action program as Callaway Action Request 200602565 (Section 03).
- Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B,
Criterion XVI, for AmerenUEs failure to implement appropriate corrective actions for
maintaining component cooling water flow consistent with design basis requirements.
On April 11 and 12, 2006, AmerenUE placed the Train A component cooling water
system in a configuration which could result in component cooling water pump runout in
the event of a loss-of-coolant accident coincident with a loss of offsite power.
Crosscutting aspects associated with problem identification and resolution were
identified for the failure to implement appropriate corrective actions to ensure the
component cooling water system remained operable for other design basis events.
This issue was more than minor because it affected the Mitigating Systems cornerstone
objective of equipment reliability in that a loss of one train of the component cooling
water system could cause other mitigating equipment (i.e., pumps and heat exchangers)
to fail and thus become a much more significant safety concern. Using the NRC
Inspection Manual Chapter 0609, Significance Determination Process, Phase 1
Screening Worksheet, the finding was determined to be of very low safety significance
because it did not result in a loss of safety function for a single train for greater than its
Technical Specification allowed outage time. AmerenUE entered this issue into its
corrective action program as Callaway Action Request 200602995 (Section 04.02).
B. Licensee-Identified Finding
A violation of very low significance, which was identified by AmerenUE, has been
reviewed by the inspectors. Corrective actions taken or planned by AmerenUE have
been entered into AmerenUE's corrective action program. This violation and the
corrective action tracking number are listed in Section 4OA7 of this report.
-3- Enclosure
REPORT DETAILS
01 Background
01.1 Summary of Discovery and Immediate Response to Component Cooling Water (CCW)
System Operability for Emergency Core Cooling System (ECCS) Containment
Recirculation
On March 27, 2006, operations personnel were conducting emergency operating
procedure (EOP) validations on the plant simulator to verify time critical manual
operator actions. During this activity a senior reactor operator identified a concern with
the timing of CCW initiation during ECCS containment recirculation. Although the
validation actions were not specifically being conducted to validate the time at which
CCW would be initiated, the operator noted that CCW may not be established to the
residual heat removal (RHR) heat exchangers until after the postloss-of-coolant accident
(post-LOCA) recirculation phase was automatically initiated. Subsequently, the
Callaway Training Department requested that Wolf Creek Generating Station provide
information on CCW initiation for ECCS recirculation and a calculation for the allowed
maximum design basis CCW temperatures from a previous NRC violation (50-
482/9812-01).
On March 29, 2006, Callaway received the requested information and the review was
completed on March 30, 2006. Corrective Action Request 200602565 was initiated the
same morning. The concern with the timing of CCW initiation during ECCS containment
recirculation was then relayed to the Operations shift crew who aligned CCW to the
RHR heat exchangers to provide continuous flow during power operation.
In accordance with Management Directive 8.3, NRC Incident Investigation Program,
the NRC determined that a special inspection was warranted, in part, on the basis of the
potential safety significance of a loss of CCW. AmerenUE established a root cause
team and a past operability determination team on April 6, 2006. The NRC chartered a
special inspection which began on April 11, 2006. The inspection team completed all
aspects identified in the charter on June 16, 2006. The team used NRC Inspection
Procedure 93812, Special Inspection, to perform the scope identified in the inspection
charter, dated April 10, 2006. The charter may also be found in the NRC Public
Document Room or from the Publicly Available Records component of NRC's document
system (ADAMS) under Accession Number ML061010217. ADAMS is accessible from
the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Reading
Room).
01.2 Impact of CCW Initiation to RHR Heat Exchangers Following Post-LOCA Recirculation
Phase
The Callaway Final Safety Analysis Report (FSAR), Section 9.1.3.2.3 and Table 6.3-8,
specified that the operators initiate CCW flow to the RHR heat exchangers as the
refueling water storage tank (RWST) level neared the automatic transfer setpoint and
prior to the recirculation phase. This was significant because the automatic transfer of
RHR pump suction from the RWST to the containment recirculation sump would
-4- Enclosure
introduce hot water, approximately 265EF, to the RHR heat exchangers. Containment
ECCS recirculation, without CCW cooling flow to the RHR heat exchangers, would heat
up the shell side of the RHR heat exchangers to temperatures in excess of the design
bases CCW system temperature and possibly cause boiling of the CCW water.
02 Prior Opportunities to Address Emergency Operating Procedure Deficiencies
02.01 Generic Communications Related to Containment Recirculation
The following provides a summary of selected generic communications applicable to
Callaway ECCS containment recirculation and CCW initiation.
10/14/76 Westinghouse issued Letter SLBE 6-803 recommending automatic CCW
initiation to the RHR heat exchangers prior to the swapover point.
Callaway plant, owned by Union Electric Company, was part of the
Standardized Nuclear Unit Power Plant System (SNUPPS) group.
SNUPPS documented that the manual action was acceptable as
operators were expected, with training, to safely perform the requirement
and because automatic action would result in additional unnecessary
surveillances. The letter stated that automatic function could be
backfitted by the NRC at the FSAR stage.
5/29/80 Westinghouse issued SNUPPS Letter SNP-3346. The letter stated that
CCW must be aligned to the RHR heat exchanger prior to swapover in
the recirculation mode.
12/82 Generic Letter 82-33, SUPPLEMENT 1 TO NUREG-0737-
REQUIREMENTS FOR EMERGENCY RESPONSE CAPABILITY,
Section 7.1, established requirements for licensees to reanalyze
transients and accidents and prepare technical guidelines. These
analyses were to identify critical operator tasks and were to be the bases
for upgraded EOPs. AmerenUEs commitments were to provide a
procedures generation package, including a program for validating EOPs.
Callaway had several opportunities to validate that CCW is established to
RHR heat exchangers prior to transfer to the cold leg recirculation phase.
02.02 Licensee Documents Addressing Callaway Containment ECCS Recirculation
The following provides a summary of selected corrective action and licensing documents
involving the Callaway ECCS containment recirculation and CCW initiation.
1980 to 1982 Callaway FSAR was issued. FSAR Section 9.1.3.2.3 and Table 6.3-8
stated that the CCW initiation must be initiated prior to ECCS
recirculation mode swapover.
1984 Callaway EOPs were initiated and required, in Procedure ES 1.3,
Transfer to Cold Leg Recirculation, that the CCW to the RHR heat
exchangers be initiated. The Westinghouse emergency response
-5- Enclosure
guideline (ERG), for Procedure E-1, Loss of Reactor or Secondary
Coolant, did not have a step to open the CCW inlet valves to the RHR
heat exchangers. The ERG basis to Procedure ES 1.3, step 2, specifies
that the step to align CCW was a "verify step that assumed previous
CCW flow initiation to each RHR heat exchanger.
4/15/1998 Callaway initiated a corrective action document, SOS 98-1577, noting
that the NRC had issued Wolf Creek Generating Station a 10 CFR 50.59
violation highlighting that late initiation of the CCW to the RHR heat
exchangers could result in 270EF recirculation sump water being
introduced to the RHR heat exchangers. Without cooling this could result
in exceeding the design temperature of the CCW system and cause
boiling to occur (Wolf Creek Generating Station PIR 973483).
5/5/98 In response to SOS 98-1577, Callaway recognized that Procedure E-1
did not have a step prior to entry to Procedure ES 1.3 and added Step 14
to open the CCW inlet valve to each RHR heat exchanger. The change
was made as a temporary change notice (TCN 98-0427). The added
step was not validated to ensure it would address the concern.
9/5/2002 Callaway corrective action document Callaway Action Request
(CAR) 200205499 stated that the Callaway EOP validation process had
validated Westinghouse recommendations in regard to EOP steps to
enter cold leg recirculation. The CAR stated that Callaway Plant had no
interim configuration issues and that FSAR Table 6.3.2 commitments for
timing actions during the swapover were met.
1/2/2004 Callaway corrective action document CAR 200400017 noted that Wolf
Creek Generating Station required that the CCW inlets to each RHR heat
exchanger be opened in 90 seconds or less following the automatic sump
swapover. The CAR initiator asked if Callaway had any similar concerns.
The Callaway accident analysis group identified no concerns.
1/27/2005 Callaway corrective action document CAR 200500564 stated that FSAR
Table 6.3-8 assumed that CCW flow is aligned to the RHR heat
exchangers before the RWST low-low-1 swapover point is reached. The
initiator questioned why the RWST outflow analysis did not explicitly
include times to align CCW flow to the RHR heat exchangers. The
response to the CAR was that steps not directly associated with the
swapover were not appropriate.
03 Corrective Actions to Address CCW Initiation on Containment Recirculation
a. Inspection Scope
The inspectors reviewed AmerenUEs actions to evaluate EOP deficiencies prior to
identifying the concern with the timing of CCW initiation during ECCS containment
recirculation on March 27, 2006. The team considered whether AmerenUEs corrective
-6- Enclosure
action program had opportunities to identify and prevent the EOP deficiencies
associated with ECCS recirculation cooling. Specifically, the inspectors reviewed
whether AmerenUEs past reviews adequately considered:
1) System safety function - classification and prioritization of the problem
commensurate with its safety significance
2) EOP validation - identification of corrective actions which are appropriately
focused to correct the problem
3) Licensing bases requirements including 10 CFR 50.59 reviews
4) Operability/reportability issues
5) Review of operational experience
b. Findings
Failure to Identify and Correct Inadequate EOPs
Introduction: The inspectors identified a Green noncited violation (NCV) of 10 CFR
Part 50, Appendix B, Criterion XVI, for the failure to identify and implement appropriate
corrective actions for EOP deficiencies associated with CCW cooling to RHR heat
exchangers as required to respond to a large-break LOCA.
Description: On March 27, 2006, during performance of EOP validations on the plant
simulator, AmerenUE recognized that CCW would not be established to the RHR heat
exchangers until after the post-LOCA recirculation phase was automatically initiated.
The automatic transfer of each RHR pump suction path from the RWST to the
containment recirculation sump would introduce hot water, approximately 265EF, to
each RHR heat exchanger prior to CCW flow being established. This could result in the
CCW system exceeding its design basis maximum temperature.
Callaway FSAR, Section 9.1.3.2.3 and Table 6.3-8, required that operators initiate CCW
to the RHR heat exchangers as the RWST level neared the automatic transfer setpoint
and prior to the recirculation phase of a LOCA. The hot water, without CCW cooling
flow, would heat up the shell side of the RHR heat exchangers to temperatures in
excess of the design bases CCW system temperature and possibly create boiling of the
CCW water in the RHR heat exchangers. Procedure E-1 Loss of Reactor or Secondary
Coolant, as written, had manual operator actions to align cooled CCW water to the
RHR heat exchangers which could not be performed prior to reaching the RWST lo-lo-1
level setpoint. This would cause a delay in cooling hot containment recirculation sump
water.
AmerenUE reviewed the simulator data and initiated a CAR on March 30, 2006.
AmerenUE established a plant lineup that provided continuous CCW flow through each
RHR heat exchanger until a permanent resolution could be established. This addressed
the immediate safety concern. The team verified that the failure to meet assumptions in
-7- Enclosure
the accident analyses had no impact on peak containment temperatures and pressures
for the LOCA accident sequences as peak conditions are mostly a function of the
containment spray system function and not the time of initiation of CCW into the RHR
heat exchanger.
AmerenUE performed heat transfer calculations and EOP validations to ensure that no
boiling of the stagnant CCW water would have occurred prior to initiating CCW water in
step 2 of ES 1.3. The heat transfer calculations determined that, over the range of
performance of different operating crews, 8 to 37 seconds of margin existed between
the initiation of opening the CCW valves and the onset of boiling. Based on plant
inservice testing, the valves were fully opened in 50 to 51 seconds. As a result of these
very low margins of time to boil, AmerenUE performed impact studies associated with
the collapse of steam bubble formation and steam slug flow analyses for the tube region
of the RHR heat exchangers. These studies resulted in approximately 60 seconds to
boil off the volume (approximately 12 percent of the total volume) of CCW water above
the heat exchanger tubes. The conclusion was there would be no significant water
hammer or steam slug flow forces created by the collapse of steam that would have
been formed. The team independently reviewed the calculations and supporting
documentation for these conclusions. This review included EPRI-NP-6766, Water
Hammer Prevention, Mitigation and Accommodation, and NRC NUREG-CR-6519,
Screening Reactor Steam/Water Piping Systems for Water Hammer.
AmerenUE documented the following opportunities to have identified and implemented
appropriate corrective action to address the inadequate EOP and safety system design
aspect:
- Westinghouse Letters SLBE 6-803 and SNP-3346
- Callaway initial FSAR reviews
- Callaway corrective action documents directly associated with the issue
(SOS 98-1577, CAR 200106536, CAR 200400017, CAR 200202808,
CAR 200503084, and CAR 200507150)
- Operational experience associated with SNUPPs plant (Wolf Creek), NRC Safety
System Engineering inspection finding (05000482/1998-012)
- Two EOP change requests and associated 10 CFR 50.59 screening reviews
associated with Procedures E-1 and ES 1.3
- Wolf Creek corrective action document problem identification Report PIR 973483
In addition, the team considered the following documents in their assessment of the
overall corrective action effectiveness to address the EOP deficiencies associated with
containment ECCS recirculation and impact on the supporting safety system design
aspect.
-8- Enclosure
- NRC Generic Letter 82-33 response and inclusion into Technical
Specification (TS) 5.4.1.b
- Reviews associated with the initiation of Callaway EOPs versus Westinghouse
documented critical operator EOP response times being exceeded. The
deficiency resulted in critical operator response times taking longer than
assumed in the accident analysis. AmerenUE review identified three similar
extent of condition reviews but missed the noncompliance with FSAR
assumptions described in this finding.
- Callaway corrective action documents directly associated with the issue,
CAR 200205499 (Callaway EOP validation process had responded to
Westinghouse OE regarding EOP steps to enter cold leg recirculation) and
CAR 200500564 (FSAR Table 6.3-8 assumed that CCW flow is aligned to the
RHR heat exchangers before the RWST low-low-1 swapover point is reached)
Analysis: In accordance with NRC Inspection Manual Chapter 0612, Section 05.01,
Screen for Performance Deficiencies, the team determined that this issue constituted a
performance deficiency because AmerenUE repeatedly failed to identify and correct the
issues related to a previous NRC finding (05000483/200306-02), CAR 200500564 and
other identified significant conditions adverse to quality. Consequently, AmerenUE had
operated the plant for years with the potential for boiling in the shell side of each RHR
heat exchanger following a postulated a large-break LOCA. Each missed opportunity to
correct inadequate emergency operating procedures was a result of ineffective
corrective action reviews, a lack of understanding of the accident analysis and licensing
bases, and poor interface between AmerenUEs accident analysis and emergency
operating procedures writers groups.
Phase 1 Screening Logic, Results, and Assumptions
In accordance with NRC Inspection Manual Chapter 0612, Section 05.03, Screen for
Minor Issues, the inspectors determined that the finding was more than minor. This
finding was associated with the equipment performance, reliability, attribute of the
mitigating systems cornerstone and was determined to affect the objective of that
cornerstone. Specifically, the finding could have resulted in the loss of CCW following a
postulated large-break LOCA.
The inspectors evaluated the issue using the Significance Determination Process (SDP)
Phase 1 Screening Worksheet for the Initiating Events, Mitigating Systems, and Barriers
Cornerstones provided in NRC Inspection Manual Chapter 0609, Appendix A,
"Significance Determination of Reactor Inspection Findings for At-Power Situations."
Following a postulated large-break LOCA, the component cooling water system would
not have functioned without quick operator action because of boiling in the RHR system
heat exchanger. This represents a loss of the system safety function. Therefore, the
screening indicated that a Phase 2 estimation was required.
-9- Enclosure
Phase 2 Estimation for Internal Events
In accordance with NRC Inspection Manual Chapter 0609, Appendix A, Attachment 1,
"User Guidance for Determining the Significance of Reactor Inspection Findings for
At-Power Situations," the inspectors estimated the risk of the subject finding using the
Risk-Informed Inspection Notebook for Callaway, Revision 2. The inspectors made the
following assumptions:
1) The performance deficiency that resulted in inadequate EOPs (failure to
establish CCW cooling to the RHR heat exchangers until after the post-LOCA
recirculation phase was automatically initiated) existed from April 15, 1998, until
March 30, 2006, when licensee personnel revised the procedure to correct the
deficiency. Therefore, this deficiency affected plant risk for an extended period
of time and the Phase 2 exposure window of greater than 30 days was used to
estimate the risk impact of the deficiency over a 1-year assessment period.
2) The failure to establish CCW cooling to the RHR heat exchangers prior to the
recirculation phase, following a postulated large-break LOCA, would have
resulted in a complete loss of the CCW system without operator intervention.
3) Table 2 of the Risk-Informed Inspection Notebook identified that worksheets for
all initiating events except the total loss of service water were applicable when a
finding affected the CCW system. However, the senior reactor analyst
determined that this performance deficiency only impacted the plant during a
large-break LOCA. Therefore, none of the sequences on any other worksheet
were applicable or quantified.
4) Table 1 of the Risk-Informed Inspection Notebook identified that the initiating
event likelihood for a large-break LOCA having an exposure time window of
greater than 30 days was 5. The inspectors noted that the performance
deficiency did not increase the likelihood of a large-break LOCA.
5) Given Assumption 2, the inspectors adjusted the low pressure recirculation
mitigation in the large-break LOCA worksheet from a credit of 3 to a credit of 0
because cooling would have been lost to the sump without CCW.
6) Despite not being required before the recirculation phase began, the actions to
establish CCW cooling to the RHR heat exchangers were proceduralized in the
emergency operating procedures. Additionally, operating crews being tested in
the plant simulator were able to establish CCW prior to the postulated failure of
the CCW system as defined by licensee calculations. Therefore, the inspectors
gave operator recovery credit in the worksheet indicating that sufficient time was
available to implement the actions, operators had been trained in the procedures
that could be implemented entirely from the main control room, and that no
special equipment was necessary to complete the actions. Therefore, as
-10- Enclosure
defined in NRC Inspection Manual Chapter 0609, Appendix A, Attachment 1,
Table 4, Remaining Mitigation Capability Credit, the inspectors gave a
Recovery of Failed Train Credit (PCREDIT) of 1.
Based on the above assumptions, only Sequence 1 of the large-break LOCA worksheet
was applicable. The resulting sequences are provided in Table 1 below:
Table 1
Phase 2 Worksheet Results
Initiator Sequence Initiating Event Likelihood Mitigating Functions Result
Large-Break LOCA 1 5 Operator Recovery of CCW 6
By application of the counting rule, the internal event risk contribution of this finding to
the change in delta core damage frequency (CDF) was of low to moderate risk
significance (White). The approximate value of this frequency (CDFPHASE 2) was
calculated by the senior reactor analyst to be 3.3 x 10-6.
Phase 3 Analysis
Assumption 6 made during the Phase 2 estimation process was overly conservative
and did not completely represent the actual probability that operators would fail to
establish CCW cooling to the RHR heat exchangers prior to the time that the CCW
system would no longer be capable of performing its intended safety function following a
postulated large-break LOCA. Therefore, the senior reactor analyst performed a
modified Phase 2 estimation to better indicate the risk of the subject performance
deficiency.
Internal Initiating Events:
The analyst utilized the simplified plant analysis Risk H (SPAR-H) method used by Idaho
National Engineering and Environmental Laboratories (INEEL) during the development
of the SPAR models and published in NUREG/CR-6883, INEEL/EXT-02-10307, The
SPAR-H Human Reliability Analysis Method, as an appropriate tool for evaluating the
probability that operators would establish CCW cooling to the RHR heat exchangers in a
timely manner following a postulated large-break LOCA.
The probability (PRECOVERY) that operators failed to properly perform the EOPs and/or
failed to perform them prior to the failure of the CCW system upon demand was
calculated to be 2.0 x 10-2. In calculating this failure probability, the analyst assumed
that the nominal action failure rate of 0.001 should be adjusted by multiplying this
nominal rate with the following performance shaping factors:
' Available Time: 10
The available time was barely adequate to complete the action. Licensee
operating crews in the plant simulator took up to 2-1/2 minutes after the
-11- Enclosure
switchover to recirculation to establish CCW flow to both RHR heat exchangers.
By licensee calculations, this action would have occurred approximately 1 minute
prior to boil off of the CCW water in the isolated RHR heat exchangers.
' Stress: 2
Stress under the conditions postulated would be high. Multiple alarms would be
initiated, causing loud, continuous noise in the main control room. Additionally,
the operators would readily identify that a large break had occurred in the reactor
coolant system and would understand that the consequences of their actions
would represent a threat to plant safety.
' All remaining performance shaping factors were considered to be nominal under
the subject conditions.
Using this more realistic operator recovery credit, the analyst recalculated the change in
core damage frequency as follows:
CDF = CDFPHASE 2 ÷ PCREDIT * PRECOVERY
= 3.3 x 10-6 ÷ 0.1 * 2.0 x 10-2
= 6.6 x 10-7
By modification of the Phase 2 estimation and in accordance with NRC Inspection
Manual Chapter 0609, Appendix A, Attachment 1, Phase 3, Risk Evaluation Using Any
Risk Basis that Departs from the Phase 1 or Phase 2 Process, the analyst determined
that the internal event risk contribution of the subject finding to the CDF was of very
low risk significance (Green). The best estimate value of this frequency was calculated
by the senior reactor analyst to be 6.6 x 10-7.
External Events
The plant-specific SDP worksheets do not currently include initiating events related to
fire, flooding, severe weather, seismic, or other external initiating events. In accordance
with Manual Chapter 0609, Appendix A, Attachment 1, step 2.5, "Screen for the
Potential Risk Contribution Due to External Initiating Events," experience with using the
site-specific Risk-Informed Inspection Notebook has indicated that accounting for
external initiators could result in increasing the risk significance attributed to an
inspection finding by as much as one order of magnitude. Therefore, the analyst
assessed the impact of external initiators because the Phase 2 SDP result provided a
risk significance estimation of 7 or greater. However, the analyst determined that the
likelihood that an external event could result in a large-break LOCA was so small as to
be negligible to the quantification of the risk of the subject performance deficiency.
-12- Enclosure
Potential Risk Contribution from Large Early Release Frequency (LERF)
In accordance with Manual Chapter 0609, Appendix A, Attachment 1, step 2.6, "Screen
for the Potential Risk Contribution Due to Large Early Release Frequency (LERF)," the
analyst determined that the finding needed to be screened for its potential risk
contribution to LERF using Manual Chapter 0609, Appendix H, Containment Integrity
Significance Determination Process, because the estimated CDF result provided a
risk significance estimation of greater than 1 x 10-7.
According to Appendix H, Section 4.1, the subject performance deficiency represented a
Type A finding because the finding influenced the likelihood of accidents leading to core
damage. As documented in Appendix H, Table 5.1, the only accident sequences that
would lead to LERF for a pressurized water reactor with a large-dry containment like
Callaways would be steam generator tube ruptures and intersystem LOCAs. The
analyst noted that the only affected core damage sequence involved a large-break
LOCA initiator. These sequences do not typically result in containment bypass
accidents.
Based on the above, and in accordance with Appendix H, the analyst screened out all
accident sequences related to the finding as not significant to LERF.
Conclusion
The performance deficiency resulted in a finding that was of very low risk significance
(Green). The best estimate change in core damage frequency was 6.6 x 10-7,
representing the risk related to internal initiators. The change in risk related to external
events, as well as the change in LERF, was determined to provide only negligible
increase in risk.
The inspection team found that this finding has crosscutting implications in the problem
identification and resolution performance area. AmerenUEs inadequate evaluations
resulted in not correcting a licensing basis safety issue.
Enforcement: Title 10 of the Code of Federal Regulations, Part 50, Appendix B,
Criterion XVI, "Corrective Action," required that conditions adverse to quality are
promptly identified and corrected. Further, the requirement states that, in the case of
significant conditions adverse to quality, measures shall be taken to ensure that the
cause of the condition is determined and corrective action taken to preclude repetition.
Contrary to the above, the corrective actions taken for a previous NRC finding
(05000483/200306-02) and CAR 200500564, as well as other identified opportunities to
correct the deficient EOP, were a significant condition adverse to quality where the
measures taken to ensure that the cause of the condition is determined and corrective
action taken to preclude repetition were not effective. This finding is a noncited violation
(NCV 05000483/2006011-01) consistent with Section VI.A of the NRC Enforcement
Policy. AmerenUE entered this issue into its corrective action program as
-13- Enclosure
04 Adequacy of Planned or Completed Corrective Actions
a. Inspection Scope
The team reviewed AmerenUEs immediate corrective actions needed to ensure the
function of the RHR heat exchangers in a large-break LOCA event and those actions to
prevent recurrence of a failure of the EOPs to meet licensing bases accident analyses.
The corrective actions implemented by AmerenUE involved three sets of actions. The
first set of actions was to establish continuous CCW flow through the RHR heat
exchangers. This would preserve the FSAR described licensing basis and ensure that
the hot containment recirculation sump water does not boil the CCW in the RHR heat
exchangers.
The second set of actions was to have a root cause team formed to determine the
extent of condition of missed licensing bases related requirements as they apply to the
EOPs. This team was to provide immediate evaluation and communication of
requirements not clearly met and provide input to the root cause determination for
The third set of actions was to ensure that plant procedures and planned maintenance
associated with the CCW system were reviewed to ensure compliance with TSs and
other aspects of the current licensing bases.
The inspectors independently reviewed the adequacy of AmerenUEs initial and planned
corrective actions.
b. Observations and Findings
.1 Upon discovery of the EOP deficiency, AmerenUE placed the plant in a configuration
that ensured adequate ECCS flow to the RHR heat exchangers during a large-break
LOCA and for containment ECCS recirculation. AmerenUEs root cause team had
sufficient resources allocated and performed a thorough extent of condition review of
licensing bases requirements pertaining to the EOPs. Three minor licensing bases
conflicts with the EOPs were identified and actions were assigned to immediately
resolve the conflicts. These are discussed in Section 04.3 of this report. Initial reviews
were performed on plant procedures and planned maintenance but they did not
effectively prevent a potential CCW pump runout scenario associated with a LOCA with
a loss of offsite power event as discussed in Section 04.2.
.2 Corrective Action to Establish Continuous CCW Flow to RHR Results in Possible CCW
Runout Conditions
Introduction: On April 12, 2006, the inspectors identified a Green, noncited violation of
10 CFR Part 50, Appendix B, Criterion XVI, due to AmerenUEs failure to implement and
adequately communicate the corrective actions identified to permanently establish CCW
to RHR heat exchangers on a large-break LOCA. This failure resulted in potential CCW
pump runout conditions for a LOCA with loss of offsite power.
-14- Enclosure
Description: On March 30, 2006, CAR 200602565 (CCW alignment for large-break
LOCAs), operability determination, operator night orders, and the acceptance criteria for
Engineering Procedure ETP-EG-0002, Component Cooling Water System Flow
Verification, Revision 4, had identified that CCW train flow should not be run above
7250 klbm/hr due to potential pump runout concerns. On April 11, 2006, CCW Pump A
was started to augment Train A CCW Pump C flow to address a Train A charging pump
oil cooler low flow alarm. The operating shift then aligned CCW flow to the spent fuel
pool to balance Train A CCW system flows. The combined flow for the two CCW
pumps was in excess of 8400 klbm/hr. On April 12, 2006, the inspectors informed
AmerenUE operating personnel that this alignment was in conflict with the operability
determination and may not ensure CCW pump operability in a postulated LOCA with a
loss of offsite power. The design of the sequencer for the essential Train A 4160 volt
bus would only start CCW Pump A. Pump A would experience a runout condition,
causing damage to the pump and possible pump failure. On failure of Pump A, Pump C
would automatically start and subject Pump C to the same conditions. Pump Cs
reliable operation would then be challenged.
The 10 CFR 50.59 safety evaluation screening for changes to Procedure
OTN-EG-00001, Component Cooling Water System, Revision 25, on April 7, 2006,
discussed potential runout system alignment but did not result in a change to the
annunciator response procedure describing what maximum flow for a single CCW pump
would be. CAR 200602995 was written to address the inspectors concerns. This CAR
identified that on two occasions, each less than 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> duration on April 11 and 12,
2006, flow in Train A was in excess of 7250 klbm/hr flow. CAR 200602995 also stated
that TS 3.7.7, Limiting Condition for Operation, Action A, for an inoperable CCW pump
was missed but the 72-hour allowed outage time was not exceeded.
Analysis: The team determined that this finding constituted a performance deficiency.
AmerenUEs corrective action for the EOP deficiency resulted in subsequent plant
configurations that did not ensure that a single CCW pump would not be subjected to
pump runout conditions. This issue was more than minor because it affected the
Mitigating Systems cornerstone objective of equipment reliability and capability of
systems that respond to initiating events in that established CCW pump operating flow
conditions would not have ensured an operable Train A CCW pump. Using the NRC
Inspection Manual Chapter 0609, Phase 1 Screening Worksheet, the finding was
determined to be of very low safety significance since it did not result in a loss of safety
function for a single train for greater than its TS allowed outage time.
The inspection team found that this finding has crosscutting implications in the problem
identification and resolution performance area. The inadequate CCW system flow was
a result of inadequate corrective action described in CAR 200602565.
Enforcement: Corrective actions for operability of the RHR system prescribed in
CAR 200602565 failed to ensure that each train of the CCW system was available to
provide RHR heat exchanger cooling. Title 10 of the Code of Federal Regulations,
Part 50, Appendix B, Criterion XVI, "Corrective Action," required that conditions adverse
to quality are promptly identified and corrected. Contrary to 10 CFR Part 50,
-15- Enclosure
Appendix B, Criterion XVI, on both April 11 and 12, 2006, AmerenUE corrective actions
lead to improper CCW system conditions that challenged the CCW train to RHR heat
exchanger safety function.
This finding is an NCV (NCV 05000483/2006011-02, Inadequate Corrective Actions
Result in Possible CCW Runout Conditions) consistent with Section VI.A of the NRC
Enforcement Policy. AmerenUE entered this issue into its corrective action program as
.3 Comprehensiveness of the Licensee's Determination of the Extent of Condition
a. Inspection Scope
Through interviews and documentation reviews, the team evaluated the
comprehensiveness of AmerenUEs extent of condition review for the failure to
implement FSAR design bases requirements. Specifically, the team assessed whether
licensee personnel had adequately reviewed procedures and engineering issues
associated with the newly established CCW to RHR heat exchangers alignment. Also
the inspectors independently reviewed the FSAR and EOPs to assess whether other
associated licensing bases requirements were met.
The inspectors reviewed the licensing documents identified by AmerenUE that were not
directly met by the current revision of the Callaway EOPs. These were:
- FSAR, Chapter 6, Engineering Safety Features, Section 6.2.2, stated that
containment spray cannot be terminated until completion of the injection phase.
Procedure E-1, step 7, allowed both containment spray pumps to be turned off,
prior to completing the injection phase, provided containment pressure had been
reduced below 5.5 psig. Having containment pressure less than 5.5 psig
ensured that no adverse impact on dose consequences would occur.
- FSAR, Chapter 9, Auxiliary Systems, Section 9.2.1.2.2.3, stated that auxiliary
feedwater low suction pressure signal opened the essential service water system
isolation valves to ensure essential service water supply to the auxiliary
feedwater system. The FSAR provided only a percent margin to the ultimate
heat sink total volume for essential service water system use and not a specific
time requirement for the operators to secure the auxiliary feedwater use. The
license calculated the time based on the percent margin and determined the
operators have 65 minutes to complete the task. This was validated as having
sufficient time to perform the task to realign the auxiliary feedwater system
suction away from the ultimate heat sink.
- FSAR, Chapter 15, Safety Analysis, Section 15.6.3.2.2, and Table 15.6.1 stated
that, following a steam generator rupture accident, a cooldown to RHR
conditions using the intact steam generator atmospheric steam dumps must be
initiated at approximately 60 minutes from the start of the event. Since this
occurs after the leak from the ruptured steam generator tubes is stopped, no
-16- Enclosure
additional radioactivity is released. This requirement had no logic bases
documented and is being reviewed by AmerenUE.
b. Observations and Findings
The inspectors found that the corrective actions, in response to normal and off-normal
procedures associated with the CCW system were generally being correctly applied to
the Callaway Plant. CCW annunciator response procedures associated with CCW
loads were not appropriately addressed as discussed in the finding in Section 04.02.
The inspectors also found AmerenUE had not fully evaluated smaller and medium break
LOCA scenarios or possible CCW pump loss of net positive suction head. AmerenUE
provided data and calculations to show that these cases with elevated initial CCW
temperatures still resulted in no significant impact.
.4 Evaluation of Licensees Initial Root Cause Determination
a. Inspection Scope
The team reviewed AmerenUEs preliminary root cause determination of the failure to
implement FSAR design bases requirements for independence, completeness, and
accuracy.
b. Observations and Findings
The inspectors found AmerenUEs direct cause determination to be accurate. This
initial licensee report, however, did not emphasize why so many opportunities to identify
the issue were missed. Organizational interface ineffectiveness and an inadequate
corrective action program allowed AmerenUEs organization to repeatedly miss
opportunities to understand an unanalyzed safety issue. Specifically the bases and
sequence of FSAR requirements were not researched when questions arose. This lead
to inaccurate and incomplete initial reviews by CAR lead responders and prevented
licensee management from becoming appropriately engaged.
The team noted that AmerenUE had identified three preliminary causes of not having
established procedures that would ensure RHR heat exchanger cooling in a LOCA prior
to automatic introduction of hot containment recirculation sump water. These were
discussed in significant condition adverse to quality CAR 200602565.
wording. The FSAR wording had remained unchanged since initial issue.
- AmerenUE had at least three opportunities to identify and correct the problem.
Reviews of corrective action documents and the emergency procedures were
narrowly focused.
- Ineffective communication between Callaway and Wolf Creek contributed to the
narrow focus of the reviews and corrective action evaluations.
-17- Enclosure
.5 Discussion of the Potential Impact Associated with Boiling in the RHR Heat Exchanger
CCW Side
The team reviewed the engineering calculations to evaluate whether the safety function
of heat removal from the containment sump following a LOCA could be achieved.
Specifically, the concern related to aligning CCW to the shell side of the RHR heat
exchangers after the RHR suction path was realigned to the containment sump from
the RWST. Because the containment recirculation sump water would be at a saturation
temperature of approximately 265EF, boiling of CCW on the shell side of the RHR heat
exchangers would occur with no shell side fluid flow.
AmerenUE performed four separate calculations that determined: (1) the temperature
rise in the shell side of the heat exchangers, (2) the heat exchanger voiding rate, (3) the
magnitude and impact of any resulting water hammer, and (4) the CCW inlet
impingement plate response to a water hammer.
The actions to ensure CCW flow is delivered to the shell side of the heat exchangers
were specified in the EOPs; consequently, the delivery of CCW to the shell side of the
heat exchangers was a function of operator time.
The team determined that AmerenUE used appropriate design inputs, calculation
methodologies, and conservative assumptions. The calculations determined that the
time required to boil and subsequently void the shell side of the heat exchangers would
have been prevented by prior operator action. Further, although it was anticipated that
CCW would be delivered to the heat exchangers prior to voiding of the heat exchangers
shell, AmerenUE demonstrated the collapse of a steam bubble that would void the
space above the heat exchanger tubes would create a water hammer of small
magnitude. The maximum force predicted for this case was equivalent to approximately
115 psig.
.6 Generic Implications
a. Inspection Scope
The team reviewed AmerenUEs design bases to determine whether generic issues
related to the design and operating practices existed with other Callaway systems or
other nuclear plants.
b. Observations and Findings
The team, with assistance from AmerenUE, and the NRCs Office of Nuclear Reactor
Regulation, determined that other Westinghouse primary water reactor plants without
automatic CCW initiation may not be in full compliance with their licensing bases. This
issue is being reviewed by the NRC.
-18- Enclosure
4OA6 Meetings, Including Exit
On April 14, 2006, the team presented the status of the inspection, to date, to
Mr. Tod Moser, Manager, Plant Engineering.
On May 2, 2006, the team leader conducted an exit meeting with Mr. Tim Herrmann,
Vice President, Engineering, and other members of his staff.
On June 26, 2006, the team leader conducted a supplemental exit meeting with
Mr. Tim Herrmann, Vice President, Engineering, and other members of his staff.
While proprietary information was reviewed, no proprietary information is being retained
or is included in this report.
4OA7 Licensee-Identified Violations
The following violation of very low safety significance (Green) was identified by
AmerenUE and is a violation of NRC requirements which meet the criteria of Section VI
of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as NCVs.
TS 5.4.1.b, EOP Program," required that requirements of NUREG 0737 as described in
Generic Letter 82-33 be adhered to, ensuring that applicable accidental analysis
licensing bases are correctly translated into emergency procedures. Contrary to this,
AmerenUE had not translated the FSAR described licensing basis into EOPs.
Callaways requirements to initiate CCW flow to the RHR heat exchangers prior to the
opening of the containment recirculation sump valves on a postulated large-break LOCA
were not met. Specifically FSAR, Section 9.1.3.2.3, and Table 6.3-8 required that
operators initiate CCW to the RHR heat exchangers as the RWST level neared the
automatic transfer setpoint prior to the recirculation phase of a LOCA.
The particular function to prevent boiling in the RHR heat exchangers on a large-break
LOCA required subsequent analysis to ensure the RHR and CCW functions were not
unrecoverable. Through calculations, AmerenUE was able to demonstrate that using
the actual EOP step location, operator action occurred in time to prevent boiling.
This issue is more than minor because it was similar to Example 3.I of Appendix E of
Manual Chapter 0612. It was necessary for AmerenUE to perform a calculation to
determine whether the existing EOPs were acceptable. Because there was available
margin in the time to boil and time to RHR heat exchanger tube uncovery calculations,
this issue was confirmed not to involve a loss of function of the system in accordance
with Part 9900, Technical Guidance, Operability Determination Process for Operability
and Functional Assessment. Therefore, this issue screens as Green during Phase 1 of
the SDP as described in Manual Chapter 0609, Appendix A, Attachment 1.
This issue was identified in AmerenUEs corrective action program as CAR 200602565.
-19- Enclosure
ATTACHMENTS: SUPPLEMENTAL INFORMATION
TIMELINE DESCRIBING CCW TO RHR HEAT EXCHANGERS PROBLEM
CHARTER MEMORANDUM DATED APRIL 10, 2006
-20- Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
D. Fuller, System Engineer
B. Huhmann, Supervising Engineer, Nuclear Engineering Systems, Mechanical
M. Jennings, Operating Supervisor
S. Maglio, Superintendent, Systems Engineering
J. Milligan, Shift Manager, Operations
K. Mills, Supervising Engineer, Regional Regulatory Affairs/Safety Analysis
T. Moser, Manager, Plant Engineering
S. Petzel, Engineer, Regional Regulatory Affairs
T. Herrmann, Vice President, Engineering
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
05000483/2006011-01 NCV Failure to Recognize and Correct Inadequate Emergency
Procedures (Section 03)05000483/2006011-02 NCV Inadequate Corrective Actions Result in Possible CCW
Runout Conditions (Section 04.2)
LIST OF DOCUMENTS REVIEWED
Calculations
Number Title Revision
EJ-M 18 RHR Pump Recirculation Operation versus Time of 1
(Wolf Creek) Initiation of CCW Flow to RHR Heat Exchangers
C-4176-00-01 Callaway RHR Heat Exchanger Shell Side Temperature 0
Rise
C-4176-00-02 Callaway RHR Heat Exchanger Transient Voiding Rate 0
C-4176-00-03 Likelihood and Magnitude of Water Hammers in the 0
Callaway RHR Heat Exchanger
M-EG-05, Calculate the NPSH Available to the CCW Pumps With the 0
ADD 2 Surge Tank Empty
BN-16 Maximum RWST Transfer Volumes and Swapover Times 0
A1-1 Attachment 1
Callaway Action Requests
199100746 19860054 199801577 200106536 200202808
200205499 200400017 200500564 200503084 200507150
200602565 200602908 200602992 200602995
Drawings
Number Title Revision
FSAR Piping and Instrumentation Diagram - CCW System NA
Figure 9.2-3,
Sheet 1
FSAR Piping and Instrumentation Diagram - CCW System NA
Figure 9.2-3,
Sheet 2
M-23EG01(Q) Piping Isometric CCW System, Auxiliary Building, Train A 6
M-23EG03(Q) Piping Isometric CCW System, Auxiliary Building, Train B 7
M-23EG04(Q) Piping Isometric CCW System, Auxiliary Building, Train B 3
M-23EG05(Q) Piping Isometric CCW System, to Fuel Building, Train B 2
5736 Vertical Residual Heat Exchanger Outline Drawing 5
5739 Vertical Residual Heat Exchanger Details 3
5740 Vertical Residual Heat Exchanger Details 4
Miscellaneous Documents
Number Title Revision/Date
Memorandum Summary of Screening Criteria for the Evaluation of April 12, 2006
4176-00-01 Steam Water Hammer at Power Plants
50.59 Screen Evaluate the Use of Gothic Software 38204
RFR 23374
Safety Kewaunee Nuclear Power Plant - Review for Kewaunee Revision 3,
Evaluation Reload Safety Evaluation Methods Topical September 10,
Report WPSRSEM-NP 2001
A1-2 Attachment 1
Miscellaneous Documents
Number Title Revision/Date
Inspection Wolf Creek Generating Station Design Inspection, February 23,
Report 05000 1998
482/97-201
EPRI Water Hammer Prevention, Mitigation and July 1992
NP-6766 Accommodation
PRA Risk Assessment for CCW Flow to RHR Heat 38819
Evaluation Exchanger Issue
Request
06-269
NUREG/ Screening Reactor Steam/Water Piping Systems for September
CR-6519 Water Hammer 1997
PIR 973483 Wolf Creek Performance Improvement Request 35731
Describing USAR and EMG ES-12 Conflicts
Draft Event & Causal Factors Chart, CAR 200602565 CCW 38818
Revision 2 Flow Requirements to RHR Heat Exchangers Not
Meeting FSAR
Formal Safety Callaway FSE for Evaluating FSAR Chapters 6 and 15 1998
Evaluation for as a Result of Calculation BN-16, Revision 0 Which
RFR 19025 Determined the Maximum Times for Swapover of
emergency core cooling system and CS Pumps from
Injection Phase to Recirculation Phase for 5 Cases
FSAR Fuel Pool Cooling System 5/97
Section 9.1
FSAR Component Cooling System 5/97
Section 9.2
FSAR Sequence of Changeover Operation from Injection to 5/97
Table 6.3-8 Recirculation
Night Order Callaway Night Order, CCW Alignment Requirements 38813
based on CAR 200602565
A1-3 Attachment 1
Miscellaneous Documents
Number Title Revision/Date
Night Order Callaway Night Order, CCW Alignment Requirements 38806
based on CAR 200602565
Computer History Printouts
Type Period
RHR Pump B 19 minutes on 2/11/2004
Current
RWST 3 years beginning 1/1/2001
Temperature
CCW Flow & 50 seconds of operation to show flow versus valve position
Valve
Operation
Procedures
Number Revision Subject
E-1 5 Loss of Reactor or Secondary Coolant
E-1 6 Loss of Reactor or Secondary Coolant
OTA-RK-00020 1 Annunciator Response Window 52B for CCW Pump A
or C Pressure Low
OTA-RK-00020 1 Annunciator Response Window 54B for CCW Pump B
or D Pressure Low
OTN-EG-00001 25 CCW System
OTN-EG-00001 26 CCW System
OTO-BB-00002 23 Rcp Off-Normal
OTO-EG-00001 9 CCW System Malfunction
EOP E-1 1B2 Loss of Reactor or Secondary Coolant
A1-4 Attachment 1
Procedures
Number Revision Subject
ES-1.3 ERG NA Westinghouse Owner Group ERG Background for ES-1.3
(background)
ES-1.3 ERG NA Westinghouse Owner Group ERG for ES-1.3
ES-1.3 0 Transfer to Cold Leg Recirculation
ES-1.3 5 Transfer to Cold Leg Recirculation
ES-1.3 6 Transfer to Cold Leg Recirculation
CAR Callaway Action Request
CCW component cooling water
CDF delta core damage frequency
CFR Code of Federal Regulations
ECCS emergency core cooling system
EOP emergency operating procedure
ERG emergency response guideline
ESW essential service water
FSAR Final Safety Analysis Report
INEEL Idaho National Engineering and Environmental Laboratories
LERF Large Early Release Frequency
LOCA loss-of-coolant accident
NCV noncited violation
RWST refueling water storage tank
SDP significance determination process
SNUPPS Standardized Nuclear Unit Power Plant System
SPAR simplified plant analysis risk
TS Technical Specification
A1-5 Attachment 1
TIMELINE DESCRIBING CCW TO RHR HEAT EXCHANGERS PROBLEM
October 14, Westinghouse issued letter SLBE 6-803 recommending automatic CCW
1976 initiation to the RHR heat exchangers prior to the swapover point. Callaway
Plant, owned by Union Electric Company, was part of the SNUPPS group.
SNUPPS felt that manual action was acceptable as operators are expected
to be trained and felt that automatic action would result in additional
unnecessary surveillances. The letter stated that automatic function could
be backfitted by the NRC at the FSAR stage.
May 29, Westinghouse issued SNUPPS Letter SNP-3346. It stated that CCW must
1980 be aligned to the RHR heat exchangers prior to swapover in the recirculation
mode.
1980 to Callaway FSAR issued. In two locations it was stated that the CCW initiation
1982 must be prior to recirculation mode swapover. (Section 9.1.3.2.3 and
Table 6.3-8.
December Generic Letter 82-33, Section 7.1, established requirements for licensees to
1982 reanalyze transients and accidents and prepare technical guidelines. These
analyses were to identify critical operator tasks and were to be the bases for
upgraded EOPs. AmerenUEs EOPs were to provide a procedures
generation package, including a program for validating EOPs. Callaway had
several opportunities to validate that CCW is established to RHR heat
exchangers prior to the transfer to the cold leg recirculation phase.
June 7, Callaway EOPs were initiated and required, only in Procedure ES 1.3, that
1905 the CCW to the RHR heat exchangers be initiated. This was contrary to the
FSAR sections requiring prior initiation. The Westinghouse ERG, for the
Procedure E-1 Loss of Reactor or Secondary Coolant, response, also did
not have a step to open the CCW inlet valves to the RHR heat exchangers.
The ERG clearly identified, in the basis to Procedure ES 1.3, step 2, that the
step to align CCW was a "verify step that assumed previous attempts to
initiate CCW flow to the RHR heat exchangers.
April 15, Callaway initiated a corrective action document, SOS (previous CAR
1998 name) 98-1577, noting that the NRC had issued Wolf Creek a 50.59 violation
highlighting that late initiation of the CCW to the RHR heat exchangers could
result in 270EF recirculation sump water being introduced to the RHR heat
exchangers. Without cooling, this could result in exceeding the design
temperature of the CCW system and cause boiling to occur. (Wolf Creek
PIR 973483).
A2-1 Attachment 2
May 5, 1998 Callaway recognized that Procedure E-1 did not have a step prior to entry to
Procedure ES 1.3 and added a step to open the CCW inlet valve to each
RHR heat exchanger. However the change was made as a temporary
change notice (TCN 98-0427) and the 50.59 screening question addressing
whether the change was to a procedure as described in the FSAR was
answered "NO."
CAR Callaway CAR 200205499 stated that the Callaway EOP procedure
200205499 validation process had validated OE14159 in regard to EOP steps to enter
cold leg recirculation. The CAR stated that Callaway had no interim
configuration issues and that FSAR 6.3.2 commitments for timing actions
during the swapover were met.
January 2, Callaway CAR 200400017 noted that Wolf Creek nuclear power plant
2004 required that the CCW inlets to each RHR heat exchanger be opened in
90 seconds or less following the automatic sump swapover. The CAR
initiator asked if Callaway had any similar concerns and the accident analysis
group replied "No."
January 27, CAR 200500564 stated that FSAR Table 6.3-8 assumed that CCW flow is
2005 aligned to the RHR heat exchangers before RWST low-low-1 swapover point
is reached. The initiator questioned why the RWST outflow analysis did not
explicitly include times to align CCW flow to the RHR heat exchangers. The
response to the CAR was that steps not directly associated with the
swapover were not appropriate.
March 20, Licensed operator retraining to perform EOP validations questioned whether
2006 CCW initiation to RHR heat exchangers was time critical on a large-break
LOCA.
March 30, CAR 200602565 was initiated describing the discovery of the simulator EOP
2006 validation. The Operations department placed the RHR heat exchanger
CCW alignment in a safe condition.
April 7, 2006 Operability determination and 50.59 screening for changes to
Procedure OTN-EG-00001 (CCW system) describe the extent of condition
for the current CCW to RHR heat exchangers alignment. Each describes a
maximum 7250 klbm/hr flow rate for a single CCW train due to pump runout
concerns during a large-break LOCA scenario with loss of offsite power to an
engineered safety features bus. Concern is that only a single CCW pump
will be sequenced onto the bus with a CCW system alignment for two pump
operation.
April 10, Callaway forms root cause and engineering teams to address the EOP/CCW
2006 issue.
A2-2 Attachment 2
April 11, NRC charters a Special Inspection Team to respond to the discovery that
2006 CCW would not be initiated to the RHR heat exchangers prior to auto
swapover to the recirculation phase on a large-break LOCA.
April 11, Low flow on the Train A charging pump oil cooler occurs. Shift operators
2006 review annunciator response Procedure OTA-RK-00020 guidance, start a
second Train A CCW pump, and increase Train A flow to 8400 klbm/hr.
April 12, NRC inspector questions the conflict with the operability determination and
2006 the actions by the operating crew in response to the 4/11/06 low CCW flow
on the charging pump.
April 13, CAR 200602995 describes two times when the 7250 klb m/hr CCW pump
2006 limit was exceeded. One was approximately 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> on 4/11/06 and again
for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> on 4/12/06.
April 14, Initial onsite inspection completed by NRC team
2006
A2-3 Attachment 2
April 10, 2006
MEMORANDUM TO: David Dumbacher, Resident Inspector, Callaway Station
Project Branch B, Division of Reactor Projects
Greg Pick, Senior Reactor Inspector
Engineering Branch 2, Division of Reactor Safety
FROM: Arthur T. Howell III, Director, Division of Reactor Projects /RA/
SUBJECT: SPECIAL INSPECTION CHARTER TO EVALUATE CALLAWAY PLANT
COMPONENT COOLING WATER INITIATION TO THE RESIDUAL
HEAT REMOVAL HEAT EXCHANGERS DURING THE INITIAL POST-
LOCA RECIRCULATION PHASE
A Special Inspection Team is being chartered in response to the discovery that component
cooling water (CCW) would not be established to the residual heat removal (RHR) heat
exchangers until after the postloss of coolant accident (LOCA) recirculation phase was initiated.
This could lead to a failure of the CCW system and a loss of safety injection and other essential
loads (such as spent fuel pool cooling). The licensee implemented prompt actions to establish
flow to the RHR heat exchangers to restore the safety systems and essential loads to an
operable status. You are hereby designated as the Special Inspection Team members. Mr.
Dumbacher is designated as the team leader.
A. Basis
On March 30, 2006, the Callaway Plant reported (CAR 200602565) that, during a
simulator exercise on March 20, 2006, an operator raised a concern regarding the
timeliness of initiation of the CCW flow to the RHR heat exchangers during post-LOCA
(large break) recirculation from the containment safety injection sumps. The licensee
identified that the sequence of establishing CCW flow, and the delays in its initiation
because of the sequence in the emergency operating procedures, could result in the
potential to exceed the CCW design temperature during a large LOCA when
containment recirculation is first initiated. The licensee found during a simulator
exercise that CCW flow to the RHR heat exchangers was not initiated until 4-6 minutes
after containment recirculation flow was first established through the RHR heat
exchangers. The Final Safety Analysis Report describes that CCW is placed in service
prior to refueling water storage tank lo-lo 1 level being reached and the swapover
occurring. The licensee had previously established, through the emergency operating
A3-1 Attachment 3
Multiple Addressees -2-
procedures, that CCW would be initiated through the RHR heat exchanger following the
swapover to containment recirculation. The licensees identification that the CCW
system may not actually be aligned in sufficient time to ensure adequate cooling of the
RHR heat exchanger resulted in the licensee questioning their ability to meet design
basis requirements. The licenses immediate corrective action included aligning and
running the CCW system continuously to ensure that adequate cooling water was
available to the RHR heat exchanger in the event of a design basis LOCA event.
This Special Inspection Team is chartered to compare the as-found conditions to the
licensing basis for containment recirculation; determine if there are generic safety
implications associated with the timing of CCW initiation post-LOCA through the RHR
heat exchangers; review the identification, evaluation, and determination whether the
CCW system and associated safety injection systems were inoperable for the
postrecirculation phase; review the licensees compensatory measures following
discovery of the condition; and review the licensees calculations regarding the impact of
the timing of CCW initiation to the RHR heat exchangers as provided in their emergency
operating procedures.
B. Scope
The team is expected to address the following:
1. Develop a complete sequence of events related to the discovery of the CCW
timing concern for post-LOCA safety injection and the followup actions taken by
the licensee.
2. Compare operating experience involving post-LOCA emergency core cooling
system (ECCS) cooling requirements to actions implemented at the Callaway
Plant. Review prior opportunities to have addressed EOP and/or design
considerations associated with ECCS recirculation cooling requirements,
including the effectiveness of those actions. Determine if there are any generic
issues related to the design and operating practices associated with post-LOCA
recirculation and ECCS cooling. Promptly communicate any potential generic
issues to regional management.
3. Review the extent of condition determination for this condition and whether the
licensees actions are comprehensive. This should include potential for other
EOP validation issues as well as potential ECCS recirculation timing issues.
4. Review the licensees determination of the cause of any procedural design
deficiencies and/or operating practices that allowed the potential for CCW
system design temperature to be exceeded. Independently verify key
assumptions and facts. If available, determine if the licensees root cause
analysis and corrective actions have addressed the extent of condition for
problems with CCW cooling to the safety systems.
A3-2 Attachment 3
Multiple Addressees -3-
5. Determine if the Technical Specifications were met for the ECCS and CCW
systems following the implementation of compensatory measures.
6. Determine if the supporting analyses for the licensees compensatory measures
were made in accordance with 10 CFR 50.59.
7. Review the calculations the licensee is developing to evaluate the CCW initiation
sequence for post-LOCA ECCS and CCW operability.
8. Collect data necessary to support a risk analysis. Specifically obtain information
associated with the degree to which the ECCS and CCW systems would be
affected during post-LOCA recirculation, the break sizes that are affected, the
containment response, the ability to recover failed pumps and other components,
and the dominant accident sequences.
C. Guidance
Inspection Procedure 93812, "Special Inspection," provides additional guidance to be
used by the Special Inspection Team. Your duties will be as described in Inspection
Procedure 93812. The inspection should emphasize fact-finding in its review of the
circumstances surrounding the event. It is not the responsibility of the team to examine
the regulatory process. Safety concerns identified that are not directly related to the
event should be reported to the Region IV office for appropriate action.
The Team will report to the site, conduct an entrance, and begin inspection no later than
April 11, 2006. While on site, you will provide daily status briefings to Region IV
management, who will coordinate with the Office of Nuclear Reactor Regulation, to
ensure that all other parties are kept informed. A report documenting the results of the
inspection should be issued within 30 days of the completion of the inspection.
This Charter may be modified should the team develop significant new information that
warrants review. Should you have any questions concerning this Charter, contact me at
(817) 860-8248.
cc via E-mail:
B. Mallett M. Peck
T. Gwynn R. Kopriva
A. Vegel D. Overland
D. Chamberlain W. Jones
R. Caniano S. O'Connor
L. Smith D. Terao
J. Clark J. Donohew
V. Dricks M. King
W. Maier
A3-3 Attachment 3
Multiple Addressees -4-
SUNSI Review Completed: _WBJ_____ ADAMS: / Yes G No Initials: _WBJ
/ Publicly Available G Non-Publicly Available G Sensitive / Non-Sensitive
S:\DRP\DRPDIR\CHARTER\Callaway April 2006.wpd ML061010217
RIV:C:DRP/B DD:DRP D:DRS D:DRP
WBJones;df:lao AVegel DDChamberlain ATHowell
/RA/ /RA/ /RA/ /RA/
4/10/06 4/10/06 4/10/06 4/10/06
OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
A3-4 Attachment 3