ML042120423
| ML042120423 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 07/30/2004 |
| From: | Blough A Division Reactor Projects I |
| To: | Peschel J, Warner M Florida Power & Light Energy Seabrook |
| bellamy r r | |
| References | |
| EA-04-139 IR-04-003 | |
| Download: ML042120423 (33) | |
See also: IR 05000443/2004003
Text
July 30, 2004
Mr. Mark E. Warner
Site Vice President
c/o Mr. James M. Peschel
Seabrook Station
P.O. Box 300
Seabrook, NH 03874
SUBJECT:
SEABROOK STATION - NRC INTEGRATED INSPECTION REPORT
Dear Mr. Warner:
On June 30, 2004, the NRC completed an inspection at the Seabrook Nuclear Power Station.
The enclosed report documents the inspection findings which were discussed on July 22, 2004,
with yourself and other members of your staff.
This inspection examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
This report documents an apparent violation which is being considered for escalated
enforcement action in accordance with the General Statement of Policy and Procedure for
NRC Enforcement Actions (Enforcement Policy), NUREG-1600. The apparent violation
involved a failure to properly implement 10 CFR 50.59 for a change to the Seabrook Updated
Final Safety Analysis Report (FSAR). Since the change involved more than a minimal increase
in the frequency of occurrence and the consequences of an accident previously evaluated in
the Seabrook FSAR, the change required prior NRC approval, but approval was not requested.
The NRC has not made a final determination in this matter, therefore, no Notice of Violation is
being issued for this apparent violation, at this time.
Before the NRC makes its enforcement decision, we are providing you an opportunity to either:
(1) respond to the apparent violation addressed in this inspection report within 30 days of the
date of this letter or (2) request a predecisional enforcement conference. If a conference is
held, it will be open for public observation. The NRC will also issue a press release to
announce the conference. Please contact Dr. Ronald Bellamy at (610) 337-5200 within 7 days
of the date of this letter to notify the NRC of your intended response.
If you choose to provide a written response, it should be clearly marked as a "Response to An
Apparent Violation in Inspection Report No. 05000443/2004003; EA-04-139" and should include
for the apparent violation: (1) the reason for the apparent violation, or, if contested, the basis
Mr. Mark E. Warner
2
for disputing the apparent violation, (2) the corrective steps that have been taken and the
results achieved, (3) the corrective steps that will be taken to avoid further violations, and (4)
the date when full compliance will be achieved. Your response may reference or include
previous docketed correspondence, if the correspondence adequately addresses the required
response. If an adequate response is not received within the time specified or an extension of
time has not been granted by the NRC, the NRC will proceed with its enforcement decision or
schedule a predecisional enforcement conference.
In addition, please be advised that the number and characterization of apparent violations
described in the enclosed inspection report may change as a result of further NRC review. You
will be advised by separate correspondence of the results of our deliberations on this matter.
In accordance with 10 CFR 2.390 of the NRCs "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 NRCs document system
(ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm.html
Also, our current Enforcement Policy is included on the NRCs Web site at www.nrc.gov; select
What We Do, Enforcement, then Enforcement Policy.
Sincerely,
/RA/
A. Randolph Blough, Director
Division of Reactor Projects
Docket No. 50-443
License No: NPF-86
Enclosure:
Inspection Report No. 05000443/2004003
w/ Attachment: Supplemental Information
Mr. Mark E. Warner
3
cc w/encl:
J. A. Stall, FPL Senior Vice President, Nuclear & CNO
J. M. Peschel, Manager - Licensing
G. F. St. Pierre, Station Director - Seabrook Station
R. S. Kundalkar, FPL Vice President - Nuclear Engineering
D. G. Roy, Nuclear Training Manager - Seabrook Station
Office of the Attorney General
P. Brann, Assistant Attorney General
M. S. Ross, Attorney, Florida Power & Light Company
R. Walker, Director, Dept. of Public Health, Commonwealth of Massachusetts
B. Cheney, Director, Bureau of Emergency Management
C. McCombs, Acting Director, MEMA
Administrator, Bureau of Radiological Health, State of New Hampshire
W. Meinert, Nuclear Engineer, Massachusetts Municipal Wholesale Electric company
J. Devine, Polestar Applied Technology
R. Backus, Esquire, Backus, Meyer and Solomon, New Hampshire
Town of Exeter
Board of Selectmen
S. Comley, Executive Director, We the People of the United States
R. Shadis, New England Coalition Staff
M. Metcalf, Seacoast Anti-Pollution League
Mr. Mark E. Warner
4
Distribution w/encl: (VIA E-MAIL)
Region I Docket Room (with concurrences)
H. Miller, RA
J. Wiggins, DRA
R. Bellamy, DRP
K. Jenison, DRP
G. Dentel, SRI - Seabrook
W. Lanning, DRS
R. Crlenjak, DRS
J. Rogge, DRS
L. Doerflein, DRS
R. Conte, DRS
R. Lorson, DRS
J. White, DRS
C. Miller, EDO Coordinator
J. Clifford, NRR
L. Licata, Backup
DOCUMENT NAME: C:\\ORPCheckout\\FileNET\\ML042120423.wpd
After declaring this document An Official Agency Record it will be released to the Public.
To receive a copy of this document, indicate in the box: "C" = Copy without attachment/enclosure "E" = Copy with
attachment/enclosure "N" = No copy
OFFICE
RI/DRP
RI/DRS
RI/DRP
RI/ORA
NAME
GDentel/KMJ for
WSchmidt/WLS
RBellamy/RRB
RUrban/JLN for
DATE
07/29/04
07/29/04
07/29/04
07/29/04
OFFICIAL RECORD COPY
Enclosure
i
U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No.:
05000443
License No.:
Report No.:
Licensee:
Florida Power & Light Energy Seabrook, LLC (FPL)
Facility:
Seabrook Station, Unit 1
Location:
Post Office Box 300
Seabrook, New Hampshire 03874
Dates:
April 1, 2004 to June 30, 2004
Inspectors:
Glenn Dentel, Senior Resident Inspector
Steve Shaffer, Resident Inspector
George Malone, Salem Resident Inspector
Thomas Moslak, Health Physicist
Jamie Benjamin, Reactor Inspector
Kenneth Jenison, Senior Project Engineer
Shani Lewis, Project Engineer
Approved by:
Dr. Ronald Bellamy, Chief
Projects Branch 6
Division of Reactor Projects
Enclosure
ii
TABLE OF CONTENTS
SUMMARY OF FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii
REACTOR SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1R04
Equipment Alignment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1R05
Fire Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
1R06
Flood Protection Measures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
1R07
Heat Sink Performance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
1R11
Licensed Operator Requalification Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
1R12
Maintenance Effectiveness . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
1R13
Maintenance Risk Assessments and Emergent Work Evaluation . . . . . . . . . . 10
1R14
Personnel Performance Related to Non-Routine Plant Evolutions and Events
11
1R15
Operability Evaluations
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
1R16
Operator Workarounds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
1R19
Post-Maintenance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
1R22
Surveillance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
1R23
Temporary Plant Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
1EP6
Drill Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
RADIATION SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
2OS1 Access Control to Radiologically Significant Areas
. . . . . . . . . . . . . . . . . . . . . 16
OTHER ACTIVITIES [OA] . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18
4OA1 Performance Indicator Verification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18
4OA2 Identification and Resolution of Problems . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
4OA5 Other Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
4OA6 Meetings, including Exit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
ATTACHMENT: SUPPLEMENTAL INFORMATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
KEY POINTS OF CONTACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . A-1
LIST OF DOCUMENTS REVIEWED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-2
LIST OF ACRONYMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-5
Enclosure
iii
SUMMARY OF FINDINGS
IR 05000443/2004003; 04/01/2004-06/30/2004; Seabrook Station, Unit 1; Flood Protection
Measures.
The report covers a 13-week period of inspection by resident inspectors and an announced
inspection by a regional senior health physics inspector. The significance of most findings is
indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process" (SDP). 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 3, dated July 2000.
A.
NRC-Identified and Self-Revealing Findings
TBD. The inspectors identified an apparent violation of 10 CFR 50.59 for implementing
a change in the facility that resulted in a more than minimal increase in the frequency of
occurrence and consequence of an accident previously evaluated, without obtaining
NRC approval pursuant to 10 CFR 50.90. In 1997, Seabrook identified that turbine
building flood diversion devices (scuppers) had not been installed in the plant as
described in Seabrooks final safety analysis report (FSAR). Between 1997 and 2000,
Seabrook implemented a design change which removed the turbine building scuppers
from the FSAR without prior NRC approval as required by 10 CFR 50.59 and 50.90.
The inspectors determined that traditional enforcement applied because this issue
impacted the NRC's ability to perform its regulatory function. The turbine building
scuppers were designed to mitigate the consequences of a circulating water system
failure. The system failure could create a turbine building flood, which if not addressed,
could impact onsite and offsite power sources. The design change resulted in a more
than minimal increase in the frequency of occurrence and consequences of a loss of
offsite power accident. (Section 1R06)
B.
Licensee-Identified Violations
None.
Enclosure
REPORT DETAILS
Summary of Plant Status
The plant began the period at full rated thermal power and operated at or near full power for the
entire report period.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R04
Equipment Alignment (71111.04)
a.
Inspection Scope
Full System Walkdown - Chemical Volume and Control System (CVCS) (71111.04S - 1
Sample)
The inspectors conducted a detailed review of the alignment and conditions of the
CVCS. The inspectors performed a walkdown to verify the system alignment was
maintained in accordance with system drawings and procedures. Control room
indications were verified to be appropriate and consistent with technical specification
requirements and the Updated Final Safety Analysis Report (UFSAR). The inspectors
reviewed and evaluated the potential impact on system operation from open work
orders, condition reports and tagged equipment. The system health report was
reviewed, verified during the walkdown and discussed with the system engineer.
The inspectors reviewed the following documents to support the walkdown and to verify
proper system alignment:
Piping and instrumentation drawings (P&IDs) for the CVCS;
A sample of historical condition reports (CRs) relative to the CVCS and its
support systems (CRs 04-01665, 03-10736, and 03-08317);
MA 4.8, "Control of Scaffolding," Rev. 7;
MS0599.47, "Erection of Scaffolding," Rev. 0.
Partial System Walkdowns. (71111.04Q - 4 Samples)
The inspectors performed the following partial system walkdowns:
On April 16, the inspectors performed a walkdown of the service water system
while maintenance was being performed on a service water valve (SW-V-4)
which provides isolation to the non-safety loads (Work Order 0216668).
On May 12, the inspectors performed a walkdown of the service air system after
the failure of the Centec service air compressor.
2
Enclosure
On May 10 through 14, the inspectors performed a walkdown of the "B"
emergency diesel generator (EDG) while the "A" EDG was out of service for
scheduled maintenance.
On May 10 through 14, the inspectors performed a walkdown of the "A" and "B"
residual heat removal systems.
The inspectors conducted a walkdown of each system to verify that the critical portions
of selected systems, such as valve positions, switches, and breakers, were correctly
aligned in accordance with Seabrooks procedures and to identify any discrepancies that
may have had an effect on operability.
The inspectors reviewed applicable piping and instrumentation drawings and applicable
operational lineup procedures to support the walkdowns and to verify proper system
alignment.
b.
Findings
No findings of significance were identified.
1R05
Fire Protection (71111.05)
a.
Inspection Scope (71111.05Q - 9 Samples)
The inspectors examined several areas of the plant to assess: 1) the control of transient
combustibles and ignition sources; 2) the operational status and material condition of
the fire detection, fire suppression, and manual fire fighting equipment; 3) the material
condition of the passive fire protection features (fire doors, fire dampers and fire
penetration seals); and 4) the compensatory measures for out-of-service or degraded
fire protection equipment. The following areas were inspected:
B Charging Pump Room, Elevation 7'-0"
Train A Residual Heat Removal (RHR), Containment Building Spray (CBS),
Safety Injection (SI) Equipment Vault, elevation (-) 61'
Train B RHR, CBS, SI Equipment Vault, elevation (-) 61'
Train A RHR, CBS, SI Equipment Vault, elevation (-) 50'
Train B RHR, CBS, SI Equipment Vault, elevation (-) 50'
Diesel Generator Building Train A Generator Room, elevation 21'6" & 51'6"
Diesel Generator Building Oil Tank Rooms Train A, elevation (-)16'
Turbine Floor, North & South Ends, elevation 75'
Turbine Building Ground Floor, North & South Ends, elevation 21'
The inspectors reviewed the following documents:
Fire Protection Pre-Fire Strategies and Fire Hazard Analysis;
3
Enclosure
Compensatory List of Fire Protection Equipment out-of-service;
Fire Protection Equipment Layout Drawings.
b.
Findings
No findings of significance were identified.
1R06
Flood Protection Measures (71111.06 - 3 Samples)
a.
Inspection Scope
The inspectors reviewed three samples of flood protection measures. These reviews
were conducted to evaluate the licensee's protection of safety-related systems from
internal and external flooding conditions. The inspectors performed a walkdown of two
specific internal areas of interest where licensee documents indicated increased risk
significance from internal flooding events. In addition, an external walkdown was
conducted. The two internal areas and the external area consisted of:
A and B RHR pump rooms
A and B vital switchgear rooms and turbine building condenser bays.
Turbine building and other building exteriors
The inspectors determined whether internal and external flooding conditions were
adequately addressed by Seabrook. The inspectors reviewed the Seabrook Final
Safety Analysis Report (FSAR) and other design basis documents, including several
flooding calculations. The inspectors compared the as-found equipment and conditions
to ensure they remained consistent with those indicated in the design basis
documentation, flooding mitigation documents, and risk analysis assumptions.
Documents reviewed during the inspection are listed in the Attachment.
b.
Findings
Introduction
The inspectors identified an apparent violation of 10 CFR 50.59 for making a change to
the facility that resulted in a more than minimal increase in the frequency of occurrence
and consequence of an accident previously evaluated in the Final Safety Analysis
Report (FSAR) without prior NRC approval. In 1997, Seabrook determined that turbine
building flood diversion devices (scuppers) had not been installed in the plant as
described in the FSAR. Subsequent to this determination, Seabrook failed to properly
evaluate the change to the facility in accordance with 10 CFR 50.59.
4
Enclosure
Description
When the plant was licensed in 1983, the Seabrook Safety Evaluation Report (SER),
NUREG 0896, stated that Seabrook had provided an analysis of the effect of flooding on
safety related equipment as a result of a postulated failure in the Circulating Water
system (CW). The SER stated that the CW system had the potential for flooding the
turbine building if a CW line ruptured and the CW pumps were not stopped. The SER
further stated that continued operation of the pumps would cause water to flow out of
the turbine building through scuppers and doors to the yard and away from plant
buildings. Shutdown of the pumps would eventually stop the flow. The SER concludes
that a total failure in the CW system would not result in flooding which would
compromise plant safety. The Seabrook SER further stated that water level alarms
were installed in the CW pits in the turbine building that would alert the control room
operators in the event of a CW system rupture. The NRC staff concluded that the
circulating water system [met] the requirement of GDC 4 with respect to protection of
safety related systems, and that it met the acceptance criteria of SRP 10.4.5. The
FSAR mirrored the wording in the SER. In 1997, the FSAR stated that the "scuppers ...
[in] the turbine building will allow the water to flow out of the building, preventing
excessive water build up on the building floor. ... the [CW] pit would fill up ... unless
prompt action by the operator is taken. No safety-related equipment is affected by a
failure of this equipment."
In 1997, Seabrook identified that components in the turbine building and the CW system
that were part of the original design basis, as described in the FSAR, Section 10.4.5,
were never installed. These components include turbine building flood diversion devices
(scuppers) and CW pit level switches. In 1997, the inspectors evaluated the non-
conformance to the FSAR and issued a non-cited violation (see NRC Inspection Report
05000433/97-06). In response to this issue, Seabrook completed a design change
which installed the level switches and removed the description of the scuppers in the
FSAR through a 10 CFR 50.59 evaluation. Seabrook determined that the change did
not need NRC approval.
The inspectors reviewed the 1997 design change (DRC 97-0033) and its associated 10 CFR 50.59 evaluation and determined that the change to eliminate the scuppers
resulted in a more than minimal increase in the frequency of occurrence and the
consequences of an accident previously evaluated in the FSAR and thus needed prior
NRC approval. Therefore, the change was a violation of 10 CFR 50.59. The inspectors
performed this evaluation using both the 10 CFR 50.59 regulation in existence in 1997
and the subsequent revision to the regulation in 1999. NEI Guidance 96-07, Guidelines
for 10 CFR 50.59 Implementation, November 2000, Revision 1, defines and provides
examples of what conditions constitute more than minimal increases of risk. Based on
the guidance, the inspectors concluded that the removal of the turbine building scuppers
used for mitigating a turbine building flood constituted a more than minimal increase in
risk. The flooding of the turbine building potentially impacts the turbine building relay
room (offsite power sources) and the emergency switchgear rooms (onsite power
5
Enclosure
sources). The inspectors also identified that Seabrook had two opportunities to identify
the violation during closeout review of design change notices (DCN) and during their
reexamination of the 10 CFR 50.59 evaluation associated with the design change.
Analysis.
Screening for Old Design Issues
NRC Inspection Manual Chapter (IMC) 0305 allows credit to be given to the
licensee for self identification of certain old design issues, such as engineering
calculations, engineering analyses, associated operating procedures or plant
equipment installations, if all four of the criteria in IMC 0305 are met. The
inspectors determined that two of the four old design criteria were not met in
the case of the 1997 facility change: (1) The licensees failure to meet 10 CFR 50.59 requirements in 1997 and in 2000 were identified by the NRC and the
NRC review of the facility change has not yet been performed; (2) The
inspectors determined that there were performance deficiencies associated with
the 1997, 10 CFR 50.59 and with the 2000 design change notice (DCN 1-3) in
that the 2000 DCN affirmed, incorrectly, that the 1997 review was adequate.
Screening for Traditional Enforcement
In accordance with Inspection Manual Chapter (IMC) 0612, Appendix B, Issue
Disposition Screening, the inspectors determined that traditional enforcement
applied because this issue potentially has impacted the NRCs ability to perform
its regulatory function. Specifically, the licensee failed to perform an adequate
10 CFR 50.59 analyses and failed to obtain a license amendment pursuant to
10 CFR 50.90, in 1997 as required by 10 CFR 50.59(c), when it was discovered
that the turbine building, as-built, had been changed from the facility as
described in the final safety analysis report (FSAR) [as updated]. The licensee
also departed from the method of evaluation described in FSAR Section 10.4.5
regarding the impact of non-safety-related systems on safety-related systems
used to establish that the design basis of the circulating water (CW) system met
GDC 4. NRC IMC 0612 provides for a risk assessment to this traditional
enforcement process when NRC considers it appropriate. The change increased
the likelihood of loss of offsite power and loss of vital buses initiating events.
The elimination of the turbine building internal flood mitigating scuppers resulted
in a more than minimal increase in the frequency of occurrence
[10 CFR 50.59(c)(2)(i)] of the loss of offsite power accident previously evaluated
in FSAR Section 15.2.6.
6
Enclosure
Phase III of the NRC Significance Determination Process
The Regional Risk analyst performed a Phase 3 SDP using the Seabrook,
Rev 3, Standardized Plant Analysis Risk (SPAR) Model, Level 1, Change 3.01,
created January 2004, that was updated using NUREG-5496 data regarding loss
of offsite power (LOOP) initiating event (IE) frequency, recovery probabilities and
emergency diesel generator (EDG) mission times. Common cause failure alpha
factors were updated using information that the Idaho National Engineering and
Environmental Laboratory (INEEL) staff and the Nuclear Regulatory
Commission's (NRC) Office of Nuclear Regulatory Research have developed
from data collection and analysis of common cause failure events from 1980
through 2001.
The analysis assumed that if the plant had been constructed in accordance with
the design and licensing basis as described in FSAR section 10.4.5 (before
update in 1997), there would be a minimal contribution to plant risk, core
damage probability or loss of offsite power frequency due to a CW system
rupture and a turbine building flood.
Without the internal flood diversion scuppers installed, the failure of a CW
expansion joint and flooding of the turbine building could result in three potential
conditions: (1) a non-recoverable loss of offsite power (LOOP); (2) a non-
recoverable LOOP combined with the loss of one vital 4kV AC bus; (3) a non-
recoverable LOOP combined with the loss of both vital 4kV AC buses.
The analyst calculated the initiating event frequency for each of these three
flooding related events using the method described in Section 12.1, Internal
Flooding, of the 2002 Seabrook Probabalistic Safety Study (SSPSS-2002). The
analyst accepted some of the licensees inputs for this calculation method
including the large and very large internal flooding initiating event frequencies,
the probability values for the flood to propagate into each of the two vital
switchgear rooms, and the probability that the switchgear rooms (75 gpm max)
floor drains would mitigate and prevent the flooding from affecting the two vital
The analyst independently calculated the human error probability inputs using
the NRC SPAR Model Human Error (HE) Worksheets. SSPSS-2002
documented that the times to the loss of offsite power due to flooding of the
offsite power relay room in the turbine building were 8 and 40 minutes for very
large and large floods, respectively. However, in response to this issue, the
licensee recalculated the flooding rate and determined that the times to the loss
of offsite power due to large and very large floods are 18 and 92 minutes. The
staff used these more recently-calculated times to complete the SPAR HE
Worksheets. Use of these times resulted in the performance shaping factors
(PSFs)for the available diagnosis times to be considered barely adequate and
7
Enclosure
extra for the very large and large floods, respectively. Although the inspectors
raised issues regarding the adequacy of the flooding alarms, adequacy of the
alarm response and emergency operating procedures, and the stress and
complexity associated with the required actions, the staff analysis set all other
PSFs to nominal values. The human error probabilities (HEP) were dominated
by the very large flood and were calculated to be in the low E-1 range using the
SPAR HE Worksheets. Alternatively, if the SPSS-2002 time (8 minutes) to loss
of offsite power for a very large flood was used in the HRA calculation, the HEP
value would default to 1.0 (one order of magnitude higher) due to inadequate
time diagnoses and response to the event. Using the SPAR risk assessment
tools and the 18 and 92 minute assumptions, the frequency per year (and the
increase in frequency above the initial minimal value) of the non-recoverable loss
of offsite power due to a large or very large flood was calculated in the low E-4
range. Therefore, the staff determined that the change to the facility from that
described in the pre-1997 FSAR resulted in a more than minimal increase in the
frequency of occurrence of the loss of offsite power accident previously
evaluated in the FSAR.
The licensee used the HCR\\ORE\\THERP method from the EPRI HRA calculator
and calculated an HEP value, in the low to mid E-2 range, for the operator action
to open the turbine building roll-up door to provide a flood drainage path and
mitigate the very large flood event. This calculation assumed time pressure
and skill-based response performance shaping factors. The licensees HRA
analysis also assumed simple response with very good cues and indication,
operations at control room panels, and low stress. Considering the possibility to
recover from the inability to open the turbine building door by stopping the
circulating water pump as the recovery method, the licensees calculated HEP
value would drop (an order of magnitude) to the low to mid E-3 range.
For the first event (a non-recoverable loss of offsite power due to a large or very
large flood), an initiating event assessment was performed by setting the LOOP
initiating event (IE) and the failure to recover offsite power probabilities to true
(or 1.0). All other initiating event probabilities were set to 0 and a conditional
core damage probability (CCDP) was calculated. The product of the CCDP and
the frequency (per year) of the flood induced non-recoverable LOOP event was
calculated to determine the change in core damage frequency (CDF). A similar
assessment was performed for the two remaining events and the sum of the
CDF values for the three turbine building flooding events was determined to be
in the low E-6 range. Therefore, the change to the facility from that described in
the pre-1997 FSAR represented a more than minimal increase in the
consequences of the loss of offsite power accident previously evaluated in the
FSAR.
8
Enclosure
Enforcement
10 CFR 50.59 (c) states in part, that the licensee may make a change in a facility
as described in the final safety analysis report (FSAR) without prior Commission
approval, provided the proposed change does not involve more than a minimal
increase in the frequency of occurrence and the consequences of an accident
previously evaluated in the FSAR. NRC Regulatory Guide 1.187, Guidance for
Implementation of 10 CFR 59, provides guidance on the implementation of
10 CFR 50.59 and endorses (with exceptions) Nuclear Energy Institute
(NEI) 96-07,Guidelines for 10 CFR 50.59 Implementation, November 2000,
Revision 1. NEI 96-07 defines and provides examples of conditions that
constitute more than minimal increases of risk.
Contrary to the above, Seabrook completed a change to the facility that
represented more than a minimal increase in the frequency of occurrence and
the consequences of a previously evaluated accident without prior NRC
approval. Between 1997 and 2000, Seabrook implemented DCR 97-0033 and
design change notices 01 through 03 which removed the turbine building
scuppers from Seabrooks FSAR.
This violation of requirements is being treated as an apparent violation of
10 CFR 50.59 (c), 05000443/2004-003-001, Failure to Obtain Prior NRC
Approval for a Change to the Facility.
1R07
Heat Sink Performance (71111.07A - 2 Samples)
a.
Inspection Scope
The inspectors reviewed two samples of safety related heat exchangers to identify any
degraded performance or potential for common cause problems that could increase
plant risk. The inspectors reviewed the results of recent residual heat removal (RHR)
system and component cooling water (CCW) system health reports and ensured that
associated performance data were documented and met the design performance
criteria. In addition, the inspectors compared the most recent performance data of the
RHR and CCW heat exchangers with the trend and system data in the system health
report. The inspectors also reviewed the Final Safety Analysis Report to ensure that
RHR heat exchanger performance criteria were consistent with the Seabrook design
basis. The inspectors verified that adverse conditions and work orders documented in
the RHR system health report were appropriately entered into Seabrooks corrective
action program and adequately addressed.
b.
Findings
No findings of significance were identified.
9
Enclosure
1R11
Licensed Operator Requalification Program (71111.11)
Quarterly Resident Inspector Review (71111.11Q - 1 Sample)
a.
Inspection Scope
On June 14, the inspectors observed an operator training session focusing on human
performance of time critical tasks. The inspectors reviewed the operators abilities to
correctly evaluate the training scenario and implement the emergency plan. Operator
actions were reviewed against Seabrook's procedural requirements. The inspectors
also evaluated whether deficiencies were identified and discussed during critiques.
b.
Findings
No findings of significance were identified
1R12
Maintenance Effectiveness (71111.12)
a.
Inspection Scope (71111.12Q - 2 Samples)
The inspectors completed two maintenance rule samples including one system review
and one specific issue review.
System Review
The inspectors evaluated Maintenance Rule (MR) implementation for the diesel air
handling system. The system was categorized in 10 CFR 50.65(a)(1) due to repetitive
failures of the starters for the diesel air handling fans on the 480 VAC motor control
centers. The inspectors interviewed engineers, reviewed specific maintenance rule
criteria for the 480 VAC motor control centers, and examined the apparent cause
determination and corrective actions of CR 03-07222. The inspectors reviewed the
(a)(1) improvement plan and system monitoring plan and evaluated the activities against
Maintenance Rule Functional Failure (MRFF) Review
The inspectors reviewed the application of the maintenance rule for a temporary loss of
seal injection flow during the performance of procedure OS1003.03. The inspectors
conducted interviews, reviewed the Updated Final Safety Analysis Report (UFSAR),
specific maintenance rule criteria and the system health report for the CVCS system.
Additionally, the inspectors reviewed the associated apparent cause for condition report
(CR 03-08317) and assigned corrective actions. The inspectors compared the
maintenance rule functional failure evaluation against 10 CFR 50.65 requirements and
against the guidance in NUMARC 93-01, "Industry Guideline for Monitoring the
effectiveness of Maintenance at Nuclear Power Plants," Rev. 2. Based on the
10
Enclosure
inspection, CR 04-04903 was generated to reevaluate the maintenance rule functional
failure determination.
b.
Findings
No findings of significance were identified.
1R13
Maintenance Risk Assessments and Emergent Work Evaluation (71111.13 - 5 Samples)
a.
Inspection Scope
The inspectors reviewed the scheduling and control for two planned maintenance
activities and three emergent work troubleshooting activities in order to evaluate the
effect on plant risk. The inspectors conducted interviews with operators, risk analysts,
maintenance technicians, and engineers to assess their knowledge of the risk
associated with the work, and to ensure that other equipment was properly protected.
The inspectors evaluated the compensatory measures against Seabrook procedures,
Maintenance Manual 4.14, "Troubleshooting, and Work Management Manual 10.1,
"On-Line Maintenance." Specific risk assessments were conducted using Seabrook's
"Safety Monitor." The inspectors reviewed the following items:
On May 19 and 20, the inspectors reviewed the plant risk configuration for
maintenance on the "A" emergency feedwater pump and one switchyard
breaker;
On April 7 and 8, the inspectors reviewed the on-line maintenance assessment
for troubleshooting work on the slow flow solenoid for MS-V-88. The inspectors
observed portions of the work activity, examined the work order (WO) 0415025
and associated documents, and interviewed the maintenance technicians. The
work documents were evaluated against various Seabrook procedures including
Work Management Manual (WM) 8.4, Work Control Practices, Rev. 2;
On May 19 and 20, the inspectors reviewed the on-line maintenance assessment
for troubleshooting work for the steam driven emergency feedwater pump due to
higher than expected temperatures on the outboard bearing of the pump. The
inspectors observed portions of the work activity, reviewed WOs 0340705,
0419680, and 0419691. The inspectors also interviewed engineers,
maintenance technicians and operators;
On May 10 to 14, the inspectors reviewed the plant risk configuration during the
"A" EDG maintenance outage. The inspectors also evaluated the emergent
activities associated with high jacket water temperature instrument failure and an
inadvertent auxiliary fuel oil pump auto start.
11
Enclosure
On June 22 to 24, the inspectors reviewed the on-line maintenance assessment
for troubleshooting work for the Containment Enclosure Ventilation Area (CEVA)
to atmosphere differential pressure instrumentation. The instruments required
rescaling to be able to measure the required differential pressure due to a
change in calculation methodology. The inspectors observed portions of the
work activities and reviewed WOs 0423283, 0423284, 0423344, and Design
Change Request (DCR) 04DCR008. The inspectors interviewed engineers, I&C
technicians, and operators involved in the operation.
b.
Findings
No findings of significance were identified.
1R14
Personnel Performance Related to Non-Routine Plant Evolutions and Events (71111.14
- 2 Samples)
a.
Inspection Scope
The inspectors reviewed operator response to two non-routine evolutions.
Main Feedwater Pump Issue
The inspectors reviewed operator performance in response to increasing vibration levels
on the B main feedwater pump FW-P-32-B. The inspectors verified that operators
evaluated the increasing vibration and took appropriate actions to address the condition
in accordance with procedures. The A main feedwater pump was biased under
WO 0412384 using procedure ON1035.10, Main Feed Pump Standby and Start Up
Operation, Rev. 7.
Circulating Pump Trip
On May 18, the "C" circulating water pump tripped due to a human performance error
during a maintenance activity. The unit remained at full power as the two remaining
circulating water pumps maintained sufficient flow to the plant. The inspectors reviewed
operator performance in response to the loss and subsequent restart of the pump. The
inspectors examined operator response against alarm response procedures, "CW Pump
C Breaker Trip and L/O," and ON 1238.01, "Circulating Water Screens Fouled
Abnormal," Rev. 5. The inspectors reviewed operator actions to restart the pump
against operating procedures, ON 1038.01, "Circulating Water System Pump Startup,"
Rev. 7 and ON 1017.02, "Circulating Water Screen Wash Operation," Rev. 5.
b.
Findings
No findings of significance were identified.
12
Enclosure
1R15
Operability Evaluations (71111.15 - 4 Samples)
a.
Inspection Scope
The inspectors reviewed operability evaluations and/or condition reports in order to
verify that the identified conditions did not adversely affect safety system operability or
plant safety. The evaluations were reviewed using criteria specified in Generic Letter 91-18, "Resolution of Degraded and Nonconforming Conditions" and Inspection Manual
Part 9900, "Operable/Operability - Ensuring the Function Capability of a System or
Component." In addition, where a component was determined to be inoperable, the
inspectors verified that Technical Specifications (TS) limiting condition for operation
implications were properly addressed. The inspectors performed field walkdowns,
interviewed personnel, and reviewed the following items:
CR 04-03519, which evaluated degraded or non-conforming conditions on the
safety related 4160 volt breakers to ensure that operability was justified and that
mitigating systems or those affecting barrier integrity remained available. The
inspectors reviewed licensee performance to ensure all related TS and FSAR
requirements were met.
CR 04-04438, which evaluated the impact of a failed containment temperature
input which was removed from alarm but remained in the computer calculation
for average containment temperature. This calculation is used to satisfy the TS
surveillance requirement 4.6.1.5. The inspectors reviewed the apparent cause
and corrective actions, interviewed operators and system engineers, examined
procedure ON 1090.06, "Use and Control of Deleted Analog and Digital Points,"
Rev. 3, and reviewed associated CRs (03-04884 and 04-05778) and WOs
(0319738, 0319743, and 0422593). The inspectors also reviewed past
containment temperature data to determine whether the TS maximum average
temperature had been exceeded.
CRs 04-04086, 04-04002 and 04-04010, which evaluated the impact of the
residual heat removal (RHR) breaker not capable of being racked out to the test
position. Operators attempted to rack out the breaker but encountered
interference with a metal raceway in the breaker cubicle. The inspectors
reviewed the operability evaluation, conducted independent walkdowns of the
RHR and other 4kV breakers, and interviewed several engineers.
CR 04-04780, which evaluated the impact of increasing emergency feedwater
pump outboard bearing temperature. The inspectors observed portions of
multiple surveillance tests, reviewed the operability evaluation, and examined the
detailed temperature trend data.
13
Enclosure
b.
Findings
No findings of significance were identified.
1R16
Operator Workarounds (71111.16 - 1 Sample)
a.
Inspection Scope
The inspectors completed a review of one specific operator workaround.
The inspectors reviewed the staging of flashlights at the control panels as a
compensatory measure due to Inverter ED-I-9 being inoperable. In a station blackout
with ED-I-9 being inoperable, the control room would lose overhead lighting. The
inspectors reviewed CR 04-05505 and Standing Operating Order 04-016 in order to
evaluate the impact on operators.
b.
Findings
No findings of significance were identified.
1R19
Post-Maintenance Testing (71111.19 - 5 Samples)
a.
Inspection Scope
The inspectors reviewed post-maintenance testing (PMT) activities to ensure: 1) the
PMT was appropriate for the scope of the maintenance work completed; 2) the
acceptance criteria were clear and demonstrated operability of the component; and 3)
the PMT was performed in accordance with procedures. The following PMTs were
reviewed:
On April 27, the activities associated with the repair of the "B" Charging Pump
discharge vent piping. The inspectors observed portions of the maintenance
activity, interviewed maintenance technicians and operators, and reviewed
On May 21 and 24, the retests described in WO 0418722 were performed
following replacement of the A Electrohydraulic Control Pump. The inspectors
reviewed the test results and the work order.
On June 2, the torque checks performed on the manifold bolts on feedwater
transmitters (1-FW-FT-543, 1-FW-FT-513, 1-FW-FT-533, 1-FW-FT-512, and
1-FW-FT-532). The inspectors observed the torque checks and reviewed WOs
0420265 through 0420269.
14
Enclosure
On June 16 and 17, the torque checks performed on the manifold bolts on nine
feedwater transmitters (1-FW-LT-501, 1-FW-LT-4310, 1-FW-LT-502, 1-FW-LT-
4320, 1-FW-FT-523, 1-FW-LT-503, 1-FW-LT-4330, 1-FW-LT-504, and 1-FW-
LT-4340). The inspectors observed the torque checks and reviewed WOs
0421228 through 0421236.
On June 11, fuse 19, associated with a shutdown control rod, and the blown fuse
indicator for this fuse were replaced. The inspectors observed the fuse
replacement, interviewed maintenance technicians, and observed the initial
thermography of the replacement fuse and the blown fuse indicator. The
inspectors also reviewed WO 0421227.
b.
Findings
No findings of significance were identified.
1R22
Surveillance Testing (71111.22 - 5 Samples)
a.
Inspection Scope
The inspectors observed portions of surveillance testing activities of safety-related
systems to verify that the system and components were capable of performing their
intended safety function, to verify operational readiness, and to ensure compliance with
required Technical Specifications and surveillance procedures.
The inspectors attended some of the pre-evolution briefings, performed system and
control room walkdowns, observed operators and technicians perform test evolutions,
reviewed system parameters, and interviewed the system engineers and field operators.
The test data recorded was compared to procedural and technical specification
requirements, and to prior tests to identify any adverse trends. The following
surveillance procedures were reviewed.
On May 19, OX1436.02, Turbine Driven Emergency Feedwater Pump Quarterly
and Monthly Valve Test, Rev. 8;
On June 10, OX1416.05, "Service Water Cooling Tower Pumps Quarterly and 2
Year Comprehensive Test," Rev. 7. The inspectors conducted an in-office
review of the completed surveillance test;
On June 16, OX1423.07, Containment Enclosure Emergency Exhaust Filter
System 31 Day, Rev. 6;
On June 18, OX1426.17, DG 1B Tech Spec Action Statement Surveillance,
Rev. 4; and
On June 24, OX1430.04, "Main Steam System Valve Operability Tests," Rev. 3.
The inspectors conducted an in-office review of the completed surveillance test.
b.
Findings
15
Enclosure
No findings of significance were identified.
1R23
Temporary Plant Modifications (71111.23 - 1 Sample)
a.
Inspection Scope
The inspectors reviewed a plant modification to determine whether it met the criteria of a
temporary modification or temporary alteration. The modification involved the
installation of service air tubing with isolation and drain valves to provide service air
system pressure at the inlet of the portable air compressor to facilitate auto-start
capabilities of the temporary air compressors.
The inspectors interviewed engineers and operators, completed field walkdowns, and
reviewed the Temporary Modification Request, 2004-004, Rev. 01 and WO 0418708.
The inspectors verified that the equipment was installed in accordance with NRC
requirements and plant procedures. The inspectors also examined the combined effect
of the modification with the outstanding temporary modifications.
b.
Findings
No findings of significance were identified.
1EP6
Drill Evaluation (71114.06 - 1 Sample)
a.
Inspection Scope
The inspectors reviewed emergency classification and notification completed by
operators during requalification training on June 14 (See Section 1R11). The inspectors
evaluated the results against Seabrooks Emergency Response Manual 1.1,
"Classification of Emergencies" and NEI 99-02, "Regulatory Assessment Performance
Indicator Guideline," Rev. 2.
b.
Findings
No findings of significance were identified.
16
Enclosure
1. RADIATION SAFETY
Occupational Radiation Safety [OS]
2OS1 Access Control to Radiologically Significant Areas (71121.01 - 21 Samples)
a.
Inspection Scope
On May 17 to 20, the inspectors verified that Seabrook was properly implementing
physical, administrative and engineering controls for access to locked high radiation
areas, and other radiologically controlled areas during power operations, and that
workers were adhering to these controls when working in these areas. Implementation
of these controls was reviewed against the criteria contained in 10 CFR 20, applicable
industry standards, and Seabrooks procedures.
Plant Walkdown and RWP Reviews
The inspectors identified exposure significant work areas including areas in the Waste
Handling Building, Containment Building, Primary Auxiliary Building, and Fuel Handling
Building. Tasks in the Waste Handling Building included transfer of a spent resin liner
from the storage area into a shipping cask and preparation of the cask for shipment.
Tasks in the Containment Building included accumulator sampling, boric acid cleaning,
and ECCS valve verification. Tasks in the Primary Auxiliary Building included removal
and transfer to storage of a spent fuel pool filter (SFP-F-33). Tasks in the Fuel Handling
Building included inspection of wall surfaces in the fuel transfer canal and testing of fuel
transfer equipment. The inspectors reviewed the radiation work permits (RWP) and the
radiation survey maps associated with these work areas to determine whether the
radiological controls were acceptable.
The inspectors toured accessible radiological controlled areas, and with the assistance
of a radiation protection technician, performed independent radiation surveys of selected
areas to confirm the accuracy of survey data and adequacy of postings.
In reviewing RWPs, the inspectors reviewed electronic dosimeter dose/dose rate alarm
set points to determine if the set points were consistent with the survey indications and
plant policy. The inspectors verified that the workers were knowledgeable of the actions
to be taken when the electronic dosimeter alarms or malfunctions for tasks being
conducted under selected RWPs. Work activities reviewed included spent resin liner
handling (RWP 04-R-00020, Task 1), various tasks performed in the Containment
Building during power operations (RWP 04-R-00010, Tasks 1, 2, 3, 4), removal/transfer
of a spent fuel pool filter (RWP 04-R-00013, Task 2), and fuel transfer canal inspection
(RWP 04-R-00026, Tasks 1, 2).
The inspectors reviewed various RWP and associated instrumentation, respiratory
protection, and engineering controls for potential airborne radioactivity areas. Through
17
Enclosure
review of relevant documentation and discussions with cognizant plant staff, the
inspectors confirmed that no worker received an internal dose in excess of 50 mrem due
to airborne radioactivity since the last inspection.
The inspectors reviewed the physical and administrative controls for highly
contaminated materials stored in the spent fuel pool.
Problem Identification and Resolution
The inspectors reviewed elements of Seabrooks Corrective Action Program related to
controlling access to radiologically controlled areas, to determine if problems were being
entered into the program for resolution. Details of this review are contained in Section
4OA2 of this report.
Jobs-In-Progress
The inspectors observed aspects of various maintenance and operational activities
being performed during the inspection period to verify that radiological controls, such as
required surveys, area postings, job coverage, and pre-job RWP briefings were
conducted; personnel dosimetry was properly worn; and that workers were
knowledgeable of work area radiological conditions. Tasks observed were selected
aspects of transferring a spent resin liner to a shipping cask, a containment entry for
accumulator sampling, removal/transfer of a spent fuel pool filter, and fuel transfer canal
inspections.
High Risk Significant, High Dose Rate HRA and VHRA Controls
The inspectors discussed with the Health Physics Supervisor the controls and
procedures pertaining to High Dose Rate (HDR) areas and Very High Radiation Areas
(VHRA). The inspectors verified that any changes to relevant Seabrook procedures did
not substantially reduce the effectiveness and level of worker protection. Controls for
significant high risk areas that were reviewed included an entry into the containment
building during power operations and inspections in the fuel transfer canal.
The inspectors discussed with senior radiation protection technicians the controls in
place for special areas that have the potential to become VHRA during certain plant
operations. These special areas include the fuel transfer canal and Containment
Building during power operations. The inspectors verified the prerequisite radiation
protection departments communications and controls were in place to allow completion
of timely actions, such as properly posting and controlling access to affected areas.
Keys to Locked High Radiation Areas (LHRA) and Very High Radiation Areas (VHRA),
maintained at the health physics control point and in the control room, were inventoried,
and accessible LHRAs were verified to be properly secured and posted during plant
tours.
18
Enclosure
Radiation Worker/Radiation Protection Technician Performance
The inspectors observed radiation worker and radiation protection technician
performance by attending various pre-job RWP briefings and morning staff meetings.
The inspectors reviewed condition reports related to radiation worker and radiation
protection errors to determine whether an observable pattern traceable to a similar
cause was evident.
b.
Findings
No findings of significance were identified.
4. OTHER ACTIVITIES [OA]
4OA1 Performance Indicator Verification (71151 - 3 Samples)
The inspectors sampled licensee submittals for the performance indicators (PIs) listed
below for the period from April 2003 through March 2004. PI definitions and guidance
contained in NEI 99-02, "Regulatory Assessment Indicator Guideline," Rev. 2 were used
to verify the accuracy of the PI data reported during that period and the basis in
reporting for each data element.
Mitigating Systems Cornerstone
Safety System Unavailability, High Pressure Safety Injection Systems
Safety System Unavailability, Emergency AC Power
Safety System Unavailability, Heat Removal System (Emergency Feedwater)
The inspectors reviewed operator logs, surveillance tests, condition reports, system
health reports and other relevant documents, and interviewed applicable licensee
personnel to verify the accuracy and completeness of Seabrook's PI data. The
inspectors also reviewed the accuracy of the number of required/critical hours reported.
19
Enclosure
4OA2 Identification and Resolution of Problems (71121.01)
1.
Access Control to Radiologically Significant Areas
a.
Inspection Scope
The inspectors reviewed twelve CRs, recent Radiation Safety Committee meeting
minutes, a Nuclear Oversight Audit Report (SBK-04-01), Daily Quality Summary
Reports, and materials used in presenting the As Low As Reasonably Achievable
(ALARA) Plan for the next refueling outage, to evaluate Seabrooks threshold for
identifying, evaluating, and resolving occupational radiation safety problems. This
review included a check of possible repetitive issues such as radiation worker and
radiation protection technician errors.
The review was conducted against the criteria contained in 10 CFR 20, Technical
Specifications, and Seabrooks procedures.
b.
Findings
No findings of significance were identified.
2.
Problem Identification and Resolution Trend Review (71152 - 1 sample)
a.
Inspection Scope
The inspectors reviewed Seabrook's corrective action program to identify trends that
may indicate existence of more safety significant issues. The inspectors reviewed the
corrective action database through the review of individual components to identify
equipment degradation trends. Additionally, the inspectors reviewed Seabrook's
programs for identifying trends through their performance improvement group, the
individual departments, and the condition report oversight group. The inspectors also
reviewed several trend condition reports.
b.
Findings
No findings of significance were identified.
20
Enclosure
4OA5 Other Activities
1.
TI 2515/156, Offsite Power System Operational Readiness
Cornerstones: Initiating Events, Mitigating Systems
a.
Inspection Scope
The inspectors performed Temporary Instruction 2515/156, Offsite Power System
Operational Readiness. The inspectors collected and reviewed information pertaining
to the offsite power system specifically relating to the areas of the maintenance rule (10 CFR 50.65), the station blackout rule (10 CFR 50.63), offsite power operability, and
corrective actions. The inspectors reviewed this data against the requirements of 10 CFR 50 Appendix A General Design Criterion 17, Electric Power Systems, 10 CFR 50.65 (a)(4), and Plant Technical Specifications.
b.
Findings
No findings of significance were identified.
4OA6 Meetings, including Exit
Exit Meeting Summary
The inspectors presented the inspection results to Mr. M. Warner on July 22, 2004,
following the conclusion of the period. The licensee acknowledged the findings
presented. The licensee did not indicate that any of the information presented at the
exit meeting was proprietary.
Site Management Visit
On June 18, Mr. Hubert Miller, Regional Administrator, US NRC Region I, and
Mr. Richard Crlenjak, Deputy Division Director, Division of Reactor Safety, toured the
site and met with Mr. Mark Warner and other members of Seabrook's management.
ATTACHMENT: SUPPLEMENTAL INFORMATION
A-1
Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
P. Allen
Senior Health Physics Technician
M. Bianco
Supervisor, Radiological Waste Services
L. Bladow
Manager, Nuclear Oversight
R. Campion
Nuclear Oversight Auditor
W. Cash
Health Physics Department Manager
D. Cormier
Senior Health Physics Technician
T. Date
Senior Health Physics Technician
P. Dundin
Shift Operations Manager
D. Flahardy
Senior Health Physicist
D. Hampton
Supervisor, Health Physics
L. Johnson III
Senior Health Physics Technician
M. Kiley
Operations Manager
P. Nardone
Reactor Engineer
M. OKeefe
Regulatory Compliance Supervisor
J. M. Peschel
Manager - Licensing
M. Scannell
Supervisor, Health Physics
R. Sterritt
Senior Nuclear Analyst, ALARA
M. Sullivan
Senior Health Physics Technician
R. Thurlow
Health Physics Technical Supervisor
J. Watts
Nuclear Oversight Auditor
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened:
Failure to Obtain Prior NRC Approval for a
Change to the Facility
Closed:
None.
Opened and Closed
None.
Discussed
None.
A-2
Attachment
LIST OF DOCUMENTS REVIEWED
Section 1R06 - Flood Protection Measures
Documents
NRC Safety Evaluation Report (SER) NUREG 0896
Individual Plant Examination for Seabrook Station, dated March 1991
FPL Seabrook Engineering Evaluation 90-05
FPL Seabrook Memo (OKeefe/Robertson) dated 5/10/04
FPL Seabrook Report SB2002X Model SSPSS 2002 dated 2/5/03
General Rubber memo (Aanonsen/Heckscher) dated 1/18/00
General Rubber Technical data sheets and test reports, dated 18/08/00
Circulating Water System Description dated 6/14/93
ACR 97-21283, No Scuppers are Install in the Turbine Building
ACR 00-1898, dated 2/16/2001
DCR 97 -0033 - Condenser Pit Flood Level Switch, 10/15/97
DCN 01, 02 and 03 to DCR 97-0033, closed out on 10/16/00
Calculation C-X-1-21802, Expansion Joint Rupture in the Circulating Water System Located in
the Turbine Building, dated October 1997
Calculation United Engineers Study of a Rupture of Circulating Water Expansion Joints. dated
February 1974
10 CFR 50.59 Evaluation , Condenser Pit Flood Level Switch, dated October 1997
NRC Safety Evaluation Report (SER) NUREG 0896
FPL Seabrook email (OKeefe/Robertson) dated 5/10/04
United Engineers and Constructors Specification for GE Installation for Package 21
Alarm Response Procedures and Level Instrumentation Data
Level Switch 1-DF-LSHH-5985/computer point D6688
Level Switch 1-DF-LSHH-598xx/computer point D8433
Alarm Response D6688 for Level Switch 1-DF-LSHH-5985
Alarm Response D8433 for Level Switch 1-DF-LSHH-598x
Work Orders
0219738
Condenser Pit Level
0233612
Main Condenser Water box
A-3
Attachment
Condition Reports (CR)
00-01898
Contrary to UFSAR Section 10.4.5.3, There are No Scuppers in the East Wall of
the Circulating Water Pump House
00-04854
Contrary to UFSAR Section 10.4.5.3, There are No Scuppers in the East Wall of
the Circulating Water Pump House
00-09273
Corrective Actions for UFSAR update
Sections 1R13 and 1R15 - Operability Determination and Maintenance Activity
Documents
On-Line Maintenance Plan, Diesel Generator (DG) "A" LCO - May 10 and 12, 2004
Work Orders
0339314
Current Injection Testing of DG_P-122A
0339249
Switchgear Breakers Trip Check
0339292
Line Breaker Differential Relay - SA-1 Inoperable
0339330
Lube Oil Inlet Temperature
0339329
EDG Prelube Oil Temperature Switch
0319915
Rocker Arm Lube Pumps
0339312
EDG Coolant Backup Pump
0231219
EDG Relief Valve 1-DG-V-62-A
0336213
EDG 4160 Volt Breaker Repair
0413751
EDG 4160 Volt Breaker Inspection and Repair
91D0239
EDG 1-DG-TT-7-A2 Repair
00C5708
EDG DG-1A Engine Air Cooler CCW Temperature
0417332
EDG Fuel Oil Receipt
0413126
1-DG-1-A 4160 Volt Breaker Inspection and Repair
0336216
1-DG-1-A 4160 Volt Breaker Inspection and Repair
0339292
1-DG-1-A 4160 Volt Breaker Inspection and Repair
Condition Reports
CR 03-05177 Standby Conditions for EDG
CR 04-01942 EDG Exhaust Corrosion
CR 04-02327 EDG Engine Signature Analysis
CR 04-04504 EDG Biobor out of specification high
CR 04-03519 EDG ED-X-3B Breaker Arching Contact
A-4
Attachment
Other References
Colt Pielstick PC-2V Drawing 004-020
Alarm Response Procedure D6560.pro DG "A" AUX FUEL Oil Pump Running
FSAR Section 9.5
EDG Fuel Oil Storage and Transfer System
Repetitive Activity
95RM1136500 RAT XFMR 1-ED-X-3B
Repetitive Activity
98RM44843001 XFMR 1-ED-X-3B
Repetitive Activity
99RM17414001 XFMR 1-ED-X-3B
Repetitive Activity
97RM44594001 XFMR 1-ED-X-3B
Repetitive Activity
97RM41281001 4160 Breaker control circuit inspection
Repetitive Activity
97RM11365001 4160 Breaker control circuit inspection
Repetitive Activity
98RM44842001 4160 Breaker control circuit inspection
Repetitive Activity
99RM44907001 4160 Breaker control circuit inspection
Repetitive Activity
97RM44718001 4160 Breaker control circuit inspection
Repetitive Activity
97RM44719001 4160 Breaker control circuit inspection
Repetitive Activity
97RM17605001 4160 Breaker control circuit inspection
Section 2OS1: Access Control to Radiologically Significant Areas
Procedures
HD0958.03, Rev 23
Personnel Survey and Decontamination Techniques
HD0958.17, Rev 12
Performance of Routine Radiological Surveys
HD0958.30, Rev 23
Inventory and Control of Locked or Very High Radiation Area Keys
and Locksets
HD0963.02, Rev 13
Administrative Guidelines for Health Physics Instrumentation
HD0992.02, Rev 28
Issuance and Control of Personnel Monitoring Devices
HN0951.04, Rev 06
Health Physics Repetitive Tasks
HN0958.13, Rev 25
Generation and Control of Radiation Work Permits
HN0958.25, Rev 25
High Radiation Area Controls
HN0958.30, Rev 23
Inventory and Control of Locked or Very High Radiation Area Keys
and Locksets
HN0958.39, Rev 04
Multi-Badge Control & Exposure Tracking
JD0999.910, Rev 0
Reporting Key Performance Indicators
RP 2.1, Rev 18
General Radiation Worker Instruction and Responsibilities
RP 9.1, Rev 17
RCA Access/Egress Requirements
RP 9.2, Rev 8
Radiological Access Requirements to Containment Area
RP 13.2, Rev 4
Storage of Highly Radioactive Material in the Reactor Cavity or
Spent Fuel Pool
RP 15.1, Rev 17
Job Pre-Planning and Review for Radiation Exposure Control
RP 15.2, Rev 09
ALARA Recommendations
RP 15.4, Rev 10
Use and Control of Temporary Shielding
RP 15.5. Rev 03
Exposure Goals
OE 3.6, Rev 5
Condition Reports
ON1090.04, Rev 3
Containment Entry
WN0598.076, Rev 0 Moving High Dose Rate Containers (>1R/Hr)
A-5
Attachment
Quality Assurance Reports
Radiation Protection/Process Control/Radwaste Programs Audit, SBK-04-01
Condition Reports
04-01552, 04-03996, 04-03051, 04-03711, 04-01517, 04-01832, 04-03767, 04-01508, 03-11055,
03-09695, 04-01198, 04-01505
Radiation Safety Committee Meeting Minutes
Meeting No. 03-05 dated December 2, 2003
Meeting No. 04-01, dated March 18, 2004
Health Physics Study/Technical Information Document (HPDTID)
Radiological Response to Repair RC-FT-434, (HPSTID 04-001)
LIST OF ACRONYMS
CEVA
Containment Enclosure Ventilation Area
CR
Condition Reports
Containment Building Spray
Component Cooling Water
Chemical Volume and Control System
Final Safety Analysis Report
High Dose Rate
MA
Maintenance Manual
Maintenance Rule Functional Failure
Non-cited Violation
NRC
Nuclear Regulatory Commission
Publicly Available Records
Performance Indicator
Piping and instrumentation drawings
Post Maintenance Testing
Radiation Work Permit
Spent Fuel Pool
Safety Injection
TS
Technical Specification
WM
Work Management Manual
Work Order
A-6
Attachment
Very High Radiation Area