ML030570444
| ML030570444 | |
| Person / Time | |
|---|---|
| Site: | Kewaunee  |
| Issue date: | 02/21/2003 |
| From: | Pederson C Division of Reactor Safety III |
| To: | Coutu T Nuclear Management Co |
| References | |
| IR-02-007 | |
| Download: ML030570444 (37) | |
See also: IR 05000305/2002007
Text
February 21, 2003
Mr. T. Coutu
Site Vice President
Kewaunee Nuclear Power Plant
N490 Hwy 42
Kewaunee, WI 54216
SUBJECT: KEWAUNEE NUCLEAR POWER PLANT
NRC INSPECTION REPORT 50-305/02-07(DRS)
Dear Mr. Coutu:
On November 8, 2002, the NRC completed an inspection at your Kewaunee Nuclear Power
Plant. The enclosed report documents the inspection findings, which were discussed on
November 8, 2002, with you and other members of your staff. Follow-up telephone exits were
held with you and members of licensee management, on December 19, 2002, and January 21,
2003.
The inspection examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel. Specifically, this inspection focused on the design and performance capability of the
component cooling water system to ensure that it was capable of performing its required
safety-related functions. In addition, the inspection reviewed a sample of permanent plant
modifications and changes made under 10 CFR 50.59.
Based on the results of this inspection, the inspectors identified two issues of very low safety
significance (Green) that were determined to involve violations of NRC requirements. However,
because of their very low safety significance and because they were entered into your
corrective action program, the NRC is treating the issues as Non-Cited Violations in accordance
with Section VI.A.1 of the NRCs Enforcement Policy. If you deny these Non-Cited Violations,
in whole or in part, you should provide a response with a basis for your denial, within 30 days of
the date of this inspection report, to the Nuclear Regulatory Commission, ATTN: Document
Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, Region
III; the Director, Office of Enforcement, United States Nuclear Regulatory Commission,
Washington, DC 20555-0001; and the NRC Resident Inspector at the Kewaunee Nuclear
Power Plant.
T. Coutu -2-
In accordance with 10 CFR 2.790 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/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA by RCaniano Acting For/
Cynthia D. Pederson, Director
Division of Reactor Safety
Docket No. 50-305
License No. DPR-43
Enclosure: Inspection Report 50-305/02-07(DRS)
cc w/encl: D. Graham, Director, Bureau of Field Operations
Chairman, Wisconsin Public Service Commission
State Liaison Officer
T. Coutu -2-
In accordance with 10 CFR 2.790 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/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA by RCaniano Acting For/
Cynthia D. Pederson, Director
Division of Reactor Safety
Docket No. 50-305
License No. DPR-43
Enclosure: Inspection Report 50-305/02-07(DRS)
cc w/encl: D. Graham, Director, Bureau of Field Operations
Chairman, Wisconsin Public Service Commission
State Liaison Officer
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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 RIII E RIII E NRR** RIII RIII
NAME ZFalevits DHills ZFalevits for MKunowski for RCaniano for
ADunlop:sd JLamb KReimer CPederson
DATE 02/13/03 02/20/03 02/18/03 02/14/03 02/21/03
OFFICIAL RECORD COPY
- NRR concurrence for section 1R17.b of inspection report
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No: 50-305
License No: DPR-43
Report No: 50-305/02-07
Licensee: Nuclear Management Company, LLC
Facility: Kewaunee Nuclear Power Plant
Location: N490 State Highway 42
Kewaunee, WI 54216
Dates: October 21, 2002, through November 8, 2002
Re-exit Dates: December 19, 2002
January 21, 2003
Inspectors: A. Dunlop, Reactor Engineer
Z. Falevits, Reactor Engineer
J. Neurauter, Reactor Engineer
S. Sheldon, Reactor Engineer
T. Bilik, Reactor Engineer, Trainee
J. Panchison, Mechanical Contractor
H. Anderson, Mechanical Contractor
C. Baron, Mechanical Contractor
Approved by: David E. Hills, Chief
Mechanical Engineering Branch
Division of Reactor Safety
SUMMARY OF FINDINGS
IR 05000305/02-07(DRS); Nuclear Management Company, LLC; on 10/21-11/8/2002,
Kewaunee Nuclear Power Plant. Safety System Design and Performance Capability Inspection.
The inspection was a three-week baseline inspection of the design and performance capability
of the component cooling water system. In addition, the biennial reviews of permanent plant
modifications and 10 CFR 50.59 evaluations were concurrently performed. The inspection was
conducted by regional engineering specialists with mechanical consultants assistance. The
inspection identified two issues of very low significance.
The significance of most findings is indicated by their color (Green, White, Yellow, Red) using
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
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. Inspection Findings
Cornerstone: Mitigating Systems
- Green. A finding of very low safety significance associated with a Non-Cited Violation of
10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified that pertained
to improper application and use of a common non-safety related power supply to feed
two redundant safety related circuits. This was not in accordance with the plant
engineering specification procedure, the Updated Safety Analysis Report and the
applicable Electrical and Electronics Engineers Standards.
This finding was more than minor because this finding was associated with design
control attributes which affected the Mitigating Systems Cornerstone objective to ensure
the reliability and capability of the component cooling water (CCW) system to respond to
initiating events to prevent undesirable consequences. The use of a common balance
of plant (non-safety) power supply to feed redundant safeguard electrical circuits, the
lack of adequate electrical separation, and evaluation of seismic qualifications of some
of these redundant circuits and components have the potential to upset plant stability,
challenge critical safety functions during shutdown as well as power operations, and
could potentially affect the reliability and capability of the CCW system to respond to
This design deficiency finding is assessed as Green because it did not result an actual
loss of the CCW systems safety function. A review of the system design identified a
number of electrical separation issues, but did not result in any immediate operability
concerns. This provides reasonable assurance that there has not been an actual loss of
system function due to this condition. Therefore, this issue was screened out of the
significance determination process as Green (Section 1R17).
- Green. A finding of very low safety significance associated with a Non-Cited Violation of
10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified that related to
2
the control and quality of design basis engineering calculations. Specifically, a number
of concerns were identified related to the indexing and control of existing calculations,
the lack of available calculations to support some aspects of the current design basis,
and errors in existing calculations. As a result of these issues, the current design basis
calculations, as well as the existing calculation control processes, may not be adequate
to ensure that the design basis will continue to be maintained. Although none of the
specific deficiencies identified during the inspection resulted in immediate operability
concerns, it was concluded that the component cooling water system design basis was
not being adequately controlled by the existing calculations.
This finding was more than minor based on the potential that the lack of adequate
control and quality of design basis calculations could result in the ability of the
component cooling water system to perform its safety functions to be degraded. Design
basis calculations were routinely used in support of design changes, operating
procedures, test acceptance criteria, and operability determinations. This finding was of
very low safety significance (Green) because it did not represent an actual loss of the
component cooling water systems safety function. (Section 1R21.2)
C. Licensee-identified Violations
No findings of significance were identified.
3
REPORT DETAILS
1. REACTOR SAFETY
Cornerstone: Initiating Events, Mitigating Systems, and Barrier Integrity
1R02 Evaluations of Changes, Tests, or Experiments (71111.02)
Review of Evaluations and Screenings for Changes, Tests, or Experiments
a. Inspection Scope
The inspectors reviewed nine 10 CFR 50.59 evaluations and twelve screenings. These
documents were reviewed to ensure consistency with the requirements of 10 CFR
50.59. The inspectors used Nuclear Energy Institute (NEI) 96-07, Guidelines of 50.59
Evaluations, Revision 1, to determine acceptability of the completed evaluations and
screenings. The NEI document was endorsed by the NRC in Regulatory Guide 1.187,
Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments,
November 2000. The inspectors also consulted Inspection Manual, Part 9900, 10 CFR
GUIDANCE: 50.59. Documents reviewed during the inspection are listed at the end of
the report.
b. Findings
No findings of significance were identified.
1R17 Permanent Plant Modifications (71111.17B)
Review of Recent Permanent Plant Modifications
a. Inspection Scope
The inspectors reviewed 17 permanent plant modifications that were performed by the
licensee's engineering staff during the last two years, 10 of which were commercial
grade dedications. Three of the modifications affected the component cooling water
system and therefore, review of these modifications counted for completion of activities
under both NRC Inspection Procedures 71111, Attachments 17 and 21. The
modifications were reviewed to verify that the completed design changes were in
accordance with specified design requirements and the licensing bases and to confirm
that the changes did not affect the modified system or other systems' safety function.
Calculations which were performed or revised to support the modifications were also
reviewed. As applicable to the status of the modification, post-modification testing was
reviewed to verify that the system, and associated support systems, functioned properly
and that the modification accomplished its intended function. The inspectors also
verified that the completed modifications did not place the plant in an increased risk
configuration. The inspectors evaluated the modifications against the licensee's design
basis documents and the Updated Safety Analysis Report (USAR). The inspectors also
used applicable industry standards, such as the American Society of Mechanical
4
Engineers (ASME) Code and the Institute of Electrical and Electronics Engineers (IEEE)
Standards, to evaluate acceptability of the modifications.
b. Findings
Introduction: Green. The inspectors identified a Non-Cited Violation (NCV) of 10 CFR
Part 50, Appendix B, Criterion III, Design Control, that pertained to improper
application and use of a common balance-of-plant (BOP) non-safety power supply to
feed two redundant safety related control valve circuits.
Discussion: Design Change Request (DCR) 3163 was initiated on January 30, 2000, to
align the service water (SW) system on a safety injection (SI) signal to maximize flow to
the containment fan coil units early in the event of an accident. Specifically, the design
change modified the control circuits for SW to component cooling water (CCW) heat
exchangers temperature control valves CV-31406/SW-1306A (Train A) and
CV-31407/SW-1306B (Train B). The design change modified the control logic and
added control switches, relays, and solenoid valves, which would cause the
SW-1306A/B valves to open on a SI signal and on loss of the non-safety control power.
The valves were designed to modulate and control SW flow to the CCW heat
exchangers, thereby controlling CCW temperature during normal plant operation. If the
valves were fully open, the CCW temperature at the heat exchanger outlet would be
cooled to approximately the SW temperature. This would then result in a subsequent
cooldown of the letdown flow temperature. The valves were designed to fail open on a
SI signal, loss of air, or loss of electrical power.
The DCR documented that actuators for SW-1306A/B, the SI relay contacts, the new
switches, relays, and the cabling from the existing relays to the new relays were all
classified QA1 (safety related) and were to be separated per plant Engineering
Specification ES-9010, Cable Installation and Separation Criteria, and IEEE Standard
308-1971, Criteria for Class 1E Electric Systems for Nuclear Power Generating
Stations. The inspectors noted that separation criteria in ES-9010 included the
following:
- Section 4.1, Safeguard Separation stated, The objective of the following
criteria is to achieve independent electrical systems compatible with and for
redundant equipment. Cable separation shall provide sufficient isolation
between redundant systems so that no single failure or credible incident can
render both systems inoperable or remove them from service.
- Section 4.1.2 stated, There are two trains provided for the Redundant
Safeguard System and four channels provided for the Reactor Protection
System. Separation of these trains or channels must be maintained to preclude
the possibility of any single incident causing both trains or more than one
channel from becoming inoperative. The power, control, and instrumentation
cables and trays for the Safeguard System and Reactor Protection System shall
be separated as follows: Train A, Train B...
- Section 4.1.3 stated, The power cables for each Redundant Safeguard System
may be placed in the cable trays only of the same train.
5
- Section 4.1.14 stated, Where the wiring for redundant engineered safety
features is within a single panel or panel section, this wiring shall be separated,
one group from the other by six-inch (6") air space or fireproof barrier..., wiring
not associated with either train" may be grouped with one train but may not
cross from one train bundle to the other train.
The inspectors also noted that USAR Section 8.2-2, Separation Criteria, Revision 17,
contained similar separation requirements to the one specified in ES-9010. The
separation criteria in the USAR included the following:
- Cable separation provides sufficient isolation between redundant systems so that
no single failure or electrical incident can render both redundant systems
inoperable or remove them from service.
- Non-safety related power, control or instrumentation cable shall not be permitted
to cross over from one safeguard tray to another.
- Where the wiring for redundant engineering safety features is within a single
panel or panel section, the wiring is separated one group from another, by a
6-inch air space or a fireproof barrier. The barriers are steel metal or flexible
metallic conduit. Wiring not associated with either train may be grouped with
one train but may not cross from one train bundle to the other train.
IEEE Standard 308-1971, Section 5.4, Vital Instrumentation and Control Power
Systems, stated in part,
Dependable power supplies are required for the vital instrumentation and control
systems of the unit(s) including the engineering safety feature instrumentation
and control systems.
Power must be supplied to these systems in such a manner as to preserve their
reliability, independence and redundancy. Typically one or more of the following
may be required: (3) two or more independent alternating current power
supplies having a degree of reliability and availability, compatible with systems
they serve.
The inspectors concluded that use of a common non-safety related power supply to feed
both trains of safety related circuits was not in accordance with the requirements stated
above. The non-safety related power supply was not considered quality power that was
free from adverse voltage and current transients, which can disturb component
operation.
IEEE Standard 279-1968, Proposed IEEE Criteria for Nuclear Power Plant Protection
Systems, required that protection systems that generate reactor trip or engineered
safeguards actuation meet the single failure criterion specified in the IEEE Standard.
Section 4.2 states under Single Failure Criterion, any single failure within the protection
system shall not prevent proper protection system action when required. Valves
SW-1306A and B were designed as redundant safeguard components/systems and
were therefore required to meet the single failure criterion of IEEE Standard 279.
6
Section 3, Design Basis, states in part, a specific protection system design basis shall
be provided for each nuclear power plant and shall document as a minimum the
following: (h) the malfunction, accidents, or other unusual events (e.g., fire, explosion,
missiles, lightening, flood, earth-quake, etc.) which could physically damage protection
system components or could cause environmental changes leading to functional
degradation of system performance and for which provisions must be incorporated to
retain necessary protection system action.
The inspectors reviewed the safety evaluation for this DCR. In response to question
No. 1, the safety evaluation for this DCR stated that the power supply for the control
circuit remained the same and that the new valves were powered from separate power
supplies, separated by Engineering Specification ES-9010. However, the inspectors
determined that the 120VAC power supply for valves SW-1306A and SW-1306B
redundant control circuit logic was not being provided from separate safeguards power
supplies (as it should have been for redundant circuits) and was not separated per the
separation requirements delineated in Engineering Specification ES-9010. The DCR
design implemented in the field indicated that the redundant safeguards valves were
powered from the same BOP (non-safeguard) power feed supplied by fuse panel
RR172 (circuits ACNI-9 and ACNI-10), as shown on schematic diagram E-2492,
Revision G. The licensee, however, considered it separate power supplies based on the
use of a separate fuse from the same BOP source to feed each of the redundant valves
control circuits. As such, the licensee considered that the installed modification was in
agreement with the statements in the safety evaluation. On February 4, 2003, the
licensee initiated CAP014584 which documented the difference between the licensees
and inspectors positions with respect to the statements in the safety evaluation. The
CAP stated that this was not an operability issue and that there was no failure potential
that can impact the operability of the CCW system from fulfilling its safeguards function.
However, the inspectors noted that there was no detailed engineering analysis to
evaluate all potential failures that could result from feeding both redundant circuits from
the same BOP feed.
The inspectors also determined that while the DCR stated that the SW-1306A/B valve
actuators (CV-31406 and CV-31407) were QA 1 components, they were supplied and
installed as non-safety (QA-2) components (reference CAP013501, dated October 30,
2002). In addition, the inspectors noted that an evaluation was not performed for
DCR 3163 to ensure that SW-1306A/B control switches 19904 and 19905 were
seismically qualified. CAP014389 was initiated on January 20, 2003, to address this
issue. The inspectors also noted that temperature controllers TC-26309 and TC-26310
used for controlling CCW temperature by modulating opening positions of valves 1306A
and 1306B had been designated as non-safety components and were also fed from the
same common non-safety power supply.
The DCR stated that normal (non-safeguards) power will be used to power the new
solenoid valves consistent with the remainder of the SW 1306A/B valves and that the
valves will be powered from two existing separate circuits. However, the inspectors
noted that the remainder of the SW-1306A/B control circuits were designed and
installed as safeguard systems but were fed from a common BOP feed.
7
The inspectors reviewed the electrical schematic and wiring diagrams for SW-1306A/B
and noted that terminal box (TB)1371, shown on wiring diagram E-2112, Revision V,
contained field wiring for both SW-1306A and SW-1306B valve circuits. Electrical
conductors coded ACN1-9L1 and ACN1-9L2 (designated as Train A wires), electrical
conductors coded ACN1-10L1 and ACN1-10L2 (designated as Train B wires), and BOP
conductors ACN1-42L1 and ACN1-42L2 were all terminated to terminal blocks inside
TB1371. In addition, a conduit containing the cables feeding control circuits for
SW-1306A and SW-1306B valves was routed from Train A section to Train B section of
TB2771. This conduit contained wire codes ACN1-42L1(power supply to BOP lights
and controllers for both 1306A and 1306B valves), ACN1-9L1 and ACN1-9L2 (power
supply to SW-1306A control circuit), and ACN1-10L1 and ACN1-10L2 (power to
SW-1306B control circuit).
The inspectors also conducted a field inspection of SW-1306A/B and its associated
components. Wiring diagram E-I531, Revision AJ, showed TB2771 wiring which
included the new relays and switches. TB2771 was divided into two sections, which
were separated horizontally by a fireproof metal barrier to separate SW-1306A (Train A)
electrical components from SW-1306B (Train B) electrical components. The BOP feeds
from common fuse panel RR172 were routed via the same conduit into TB2771. Train
A related (9L1) 120VAC BOP feed was routed to the Train A section of TB2771 and
Train B related (10L1) 120VAC BOP feed was routed via the same conduit to the Train
B portion of TB2771. A short conduit was routed from Train A section to Train B section
of TB2771. This conduit contained the BOP feed cables conductors. The inspectors
determined that the present installed configuration of the 120VAC BOP feeds to
SW-1306A/B resulted in electrically connecting Train A and Train B circuitry through the
120VAC BOP power supplies. Each of the SW-1306A/B control circuits was protected
by one fuse and one slug located in RR172. The inspectors determined that the
installed electrical configuration was contrary to the electrical separation requirements
delineated in ES-9010, USAR 8.2.2, and IEEE-308-1971.
During review of condition reports, the inspectors identified that since May 2000, the
SW-1306A and/or the SW-1306B valve(s) inadvertently opened on at least nine
separate occasions. These following events occurred during normal plant operation due
to random grid disturbances, lightning strikes, and/or surveillance testing activities.
- May 10, 2000, (Kewaunee Assessment Process (KAP) 00-001414) SW-1306A/B
failed open when grid perturbation caused short lived loss of voltage. The KAP
stated that this condition has been experienced in the past.
- September 2, 2000, (KAP 00-003120) an electrical disturbance caused by a
lightning induced spike resulted in reactivity problems when SW-1306A and B
had failed open.
- November 24, 2001, (KAP 01-018732) SW-1306B failed open during
performance of SP-33-110, Diesel Generator Automatic Test, as a result of
load shedding and restarting of large loads. The KAP stated that the apparent
cause for the identified problem appears to be that the system design is subject
to this type of event because a momentary loss of power which occurs when
switching 120VAC QA2 power will result in valves SW-1306A and B failing open.
8
- November 20, 2001, (KAP 01-18695) valves SW-1306A and B failed open during
performance of surveillance testing SOP-ELV-40-8, after losing power during a
power switching activity.
- June 24, 2002, (CAP012001) a transient where both SW-1306A and B valves
opened due to an electrical transient. This caused the CCW temp to decrease,
which could have had a positive reactivity affect on the reactor had the operators
not taken actions. The CAP documented that operator workaround 01-22 and
abnormal procedure A-CC-31A, Abnormal Conditions in the Component Cooling
System, were implemented to bypass the letdown demin and an auxiliary
operator was dispatched to regain control of the system. Reactivity effects were
monitored, although no changes were seen due to early recognition of the
problem. The inspectors determined that loss of the common non-safety power
supply resulted in both valves opening unexpectedly, challenging the operators
by use of an operator workaround to expeditiously bypass letdown demin and
prevent a potential positive reactivity effect.
- July 9, 2002, (CAP012174) a misalignment of substation capacitor bank opening
and closing resulted in a voltage dip that caused SW-1306B to fail open.
Operator workaround 01-22 and abnormal procedure A-CC-31A were
implemented to bypass the letdown demin and an auxiliary operator was
dispatched to regain control of the system.
The first three items above were determined by the licensee to be maintenance rule
functional failures in maintenance rule evaluation MRE000082, dated November 21,
2001. The fourth item above was classified as a maintenance preventible functional
failure in KAP 01-18695. Condition Evaluation CE002373, dated February 12, 2002,
and apparent cause evaluation ACE001828, dated June 21, 2002, concluded that as a
result of the numerous instances where valves SW-1306A and B have failed open,
System 38 Function 04 (supplies 120VAC QA2 power) has had a repetitive MPFF and
was considered (a)(2) degraded. ACE001828 documented three more instances where
SW-1306A or B valves failed open on June 23, July 21, and July 22, 2002, during
substation breaker manipulation and lightening strikes. Licensees investigation
(ACE001828) revealed the following three distinct concerns related to the SW-1306A
and B valve events: (1) The effects of random grid disturbances while at full power
should not result in these valves fully opening at times when plant power is not lost or
interrupted and a SI signal in not present, (2) train separation (should the power supply
for these valves be separated instead of tied to the same source), and (3) the controllers
are obsolete.
To identify the correct cause of the SW-1306A/B valves inadvertent openings and to
determine if Design Change 3205 (initiated to modify the power supplies to the
electronic controllers) will address the concern of the undesired opening of these valves
under certain conditions, the licensee issued temporary change TC 02-01 on July 2,
2002, to install monitoring equipment on the SW-1306B train. This has not yet been
implemented in the field. Therefore, the inspectors noted that actual cause of
SW-1306A/B failing open during normal plant operations has yet to be determined.
In a related matter, the licensee documented in OTH002449, dated August 30, 2001,
that CC water temperature could reach 390F during an event where a SI signal was
generated (SW-1306A and B open). The licensee stated in the OTH that this
9
temperature was not considered in the piping analysis and that the issue needed to be
examined by Westinghouse.
Analysis: Evaluation of this issue concluded that it was a design control issue resulting
in a finding of very low safety significance (Green). The design control issue was due to
a licensee performance deficiency in that the licensee failed to adequately control the
design modification process for modification DCR 3163 as required by established plant
and industry design standards.
In accordance with Manual Chapter 0612, the inspectors determined the issue was
more than minor because this finding was associated with design control attributes
which affected the Mitigating Systems Cornerstone objective to ensure the reliability and
capability of the CCW system to respond to initiating events to prevent undesirable
consequences. The use of a common BOP (non-safety) power supply to feed
redundant safeguard electrical circuits, the lack of adequate electrical separation, and
evaluation of seismic qualifications of some of these redundant circuits and components
have the potential to upset plant stability, challenge critical safety functions during
shutdown as well as power operations, and could potentially affect the reliability and
capability of the CCW system to respond to initiating events.
This design deficiency finding is assessed as Green because it did not result in an
actual loss of the CCW systems safety function. A review of the system design
identified a number of electrical separation issues, but did not result in any immediate
operability concerns. This provides reasonable assurance that there has not been an
actual loss of system function due to this condition. Therefore, this issue was screened
out of the significance determination process as Green.
Enforcement: 10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in
part, that measures be established to assure that applicable regulatory requirements
and the design basis are correctly translated into specifications, drawings, procedures,
and instructions. It further states that design changes shall be subject to design control
measures commensurate with those applied to the original design. Section 4.1.2 of
ES-9010 states in part that cable separation shall provide sufficient isolation between
redundant systems and that the power and control cables for the safeguard system shall
be separated.
Contrary to the above, on June 30, 2000, the installed electrical configuration was not
in accordance with plant and industry established electrical separation design
requirements as specified in IEEE Standard 308-1971, and in ES-9010 for the control
circuits for temperature control valves SW-1306A/CV-31406 and SW-1306B/CV-31407.
The licensee used non-safety related 120VAC power supplies from a common fuse
cabinet to feed the redundant safeguard system control circuits for these valves in lieu
of separate safety related power supplies, which would provide sufficient isolation
between these safeguard redundant systems.
Because of the low safety significance of this issue and because it was entered in the
licensee's corrective action program (CAP013801), the issue is being treated as a
Non-Cited Violation, consistent with Section VI.A.1 of the NRC Enforcement Policy
(NCV 50-305/02-07-01).
10
1R21 Safety System Design and Performance Capability (71111.21)
Introduction
Inspection of safety system design and performance verifies the initial design and
subsequent modifications and provides monitoring of the capability of the selected
systems to perform design bases functions. As plants age, the design bases may be
lost and important design features may be altered or disabled. The plant risk
assessment model is based on the capability of the as-built safety system to perform the
intended safety functions successfully. This inspectable area verifies aspects of the
mitigating systems cornerstone for which there are no indicators to measure
performance.
The objective of the safety system design and performance capability inspection is to
assess the adequacy of calculations, analyses, other engineering documents, and
operational and testing practices that were used to support the performance of the
selected systems during normal, abnormal, and accident conditions. The inspection
was performed by a team of inspectors that consisted of a team leader, three Region III
inspectors, and three mechanical consultants.
The component cooling system was selected for review during this inspection based
upon:
- having a high probabilistic risk analysis ranking;
- having had recent significant issues; and
- not having received recent NRC review.
The criteria used to determine the systems performance included:
- applicable technical specifications;
- applicable USAR sections; and
- the systems design documents.
The following system and component attributes were reviewed in detail:
System Requirements
Process Medium - water, electricity
Energy Source - electrical power, air
Control Systems - initiation, control, and shutdown actions
System Condition and Capability
Installed Configuration - elevation and flow path operation
Operation - system alignments and operator actions
Design - calculations and procedures
Testing - flow rate, pressure, temperature, voltage, and levels
11
Components
The component cooling water pumps and heat exchanger were selected for detailed
review during the inspection. These components were specifically reviewed for
component degradation due to the impact that its failure would have on the plant.
.1 System Requirements
a. Inspection Scope
The inspectors reviewed the updated safety analysis report, technical specifications,
system descriptions, drawings and available design basis information to determine the
performance requirements of the component cooling water system. The reviewed
system attributes included process medium, energy sources, and control systems. The
rationale for reviewing each of the attributes was:
Process Medium: This attribute required review to ensure that the component cooling
water pumps would supply the required flow to the safety related components following
design basis events. To achieve this function, the inspectors verified that the
component cooling water system would be able to accept the design heat loads from the
applicable safety related components through the residual heat removal heat exchanger
and transfer sufficient heat to the service water system through the component cooling
water heat exchanger to maintain system operability.
Energy Sources: This attribute required review to ensure that the component cooling
water pumps would start when called upon, and that appropriate valves would have
sufficient power to change state when so required. To achieve this function, the
inspectors verified that the interactions between the component cooling water pumps
and their support systems were appropriate such that all components would start when
needed under normal or standby electrical power.
Controls: This attribute required review to ensure that the automatic controls for
starting the component cooling water pumps, and associated system components, were
properly established. Additionally, review of alarms and indicators was necessary to
ensure that operator actions would be accomplished in accordance with the design.
b. Findings
No findings of significance were identified.
.2 System Condition and Capability
a. Inspection Scope
The inspectors reviewed design basis documents and plant drawings, abnormal and
emergency operating procedures, requirements, and commitments identified in the
updated safety analysis report and technical specifications. The inspectors compared
the information in these documents to applicable electrical, instrumentation and control,
12
and mechanical calculations, setpoint changes, and plant modifications. The inspectors
also reviewed operational procedures to verify that instructions to operators were
consistent with design assumptions.
The inspectors reviewed information to verify that the actual system condition and tested
capability was consistent with the identified design bases. Specifically, the inspectors
reviewed the installed configuration, the system operation, the detailed design, and the
system testing, as described below.
Installed Configuration: The inspectors confirmed that the installed configuration of
the component cooling water system met the design basis by performing detailed
system walkdowns. The walkdowns focused on the installation and configuration of
piping, components, and instruments; the placement of protective barriers and systems;
the susceptibility to flooding, fire, or other environmental concerns; physical separation;
provisions for seismic and other pressure transient concerns; and the conformance of
the currently installed configuration of the systems with the design and licensing bases.
Design: The inspectors reviewed the mechanical, electrical, and instrumentation design
of the component cooling water system to verify that the system and subsystems would
function as required under accident conditions. This included a review of the design
basis, design changes, design assumptions, calculations, boundary conditions, and
models as well as a review of selected modification packages. Instrumentation was
reviewed to verify appropriateness of applications and set-points based on the required
equipment function. Additionally, the inspectors performed limited analyses in several
areas to verify the appropriateness of the design values.
Testing: The inspectors reviewed records of selected periodic testing and calibration
procedures and results to verify that the design requirements of calculations, drawings,
and procedures were incorporated in the system and were adequately demonstrated by
test results. Test results were also reviewed to ensure automatic initiations occurred
within required times and that testing was consistent with design basis information.
Pre-operational test data was also reviewed to confirm initial design parameters that
could not be tested under normal operations.
b. Findings
Design Basis Information
Based on the inability or difficulties in retrieving design information requested by the
inspectors, licensee personnel documented that, in many cases, design basis
information for the CCW system was difficult if not impossible to locate. Licensee
personnel initiated CAP013087 and CAP013119 to enter the problem in the corrective
action program. This issue was also identified during the previous NRC Safety System
Design and Performance Capability Inspection for the service water system and entered
into the corrective action program as KAP 00-002566. The licensee in response to this
issue has been developing Design Basis System Functional Matrixes for a number of
systems including the component cooling water system. These documents were still in
13
draft at the time of the inspection, although it appears that some progress has been
made in identifying and controlling design basis information.
Calculation Control and Quality Issues
Introduction: Green. The inspectors identified a Non-Cited Violation of 10 CFR Part 50,
Appendix B, Criterion III, Design Control, that related to the control and quality of
design basis engineering calculations. Specifically, the inspectors identified a number of
concerns related to the indexing and control of existing calculations (including the failure
to use appropriate and/or current calculation inputs and assumptions), the lack of
available calculations to support some aspects of the current design basis, and errors in
existing calculations. As a result of these issues, the inspectors determined that the
current design basis calculations, as well as the existing calculation control processes,
may not be adequate to ensure that the design basis will continue to be maintained.
Although none of the specific deficiencies identified during the inspection resulted in
immediate operability concerns, the inspectors concluded that the CCW system design
basis was not being adequately controlled by the existing calculations.
Discussion: During the inspection the inspectors noted a number of calculation
deficiencies. The licensee initiated individual CAPs, as appropriate, to ensure that each
of these conditions will be addressed by the corrective action system. In addition, the
licensee initiated two high level CAPs, CAP013531 and CAP013532, to address
calculation indexing and calculation errors, respectively. The following discussion
includes examples of calculation deficiencies identified during the inspection.
Indexing and Control of Existing Calculations - The inspectors identified concerns
related to the indexing and control of existing calculations. As shown in the following
examples, conditions were identified where design basis calculations were not based on
current input data, were based on assumed inputs in lieu of calculated values, were not
consistent with other design basis calculations, or were not revised when appropriate to
reflect a change in input data. Review of calculation indexes and discussions with
licensee personnel indicated that these issues were related to inadequate indexing and
control of design basis calculations. The inspectors also found that it was difficult to
identify the status of calculations, and to determine if a calculation was a current design
basis calculation. In response to these concerns, the licensee initiated CAP013531 that
concluded that Kewaunee was not up to industry standards with regard to calculation
controls, and addressed in the generation of system functional matrixes the need to use
the Currator database for the indexing of calculations.
- Calculation C11353, Determination of CCW Pump delta-P Acceptance Criteria
for use in SP 31-168, concluded that an acceptable pump degradation for the
CCW pumps was 10 percent, which was consistent with the permissible
degradation established in ASME OM-6. Subsequent to the issuance of the
referenced calculation, a CCW system hydraulic flow model was developed and
depicted in calculation C11409. Interpolating the flow model results from the
calculation results indicated that the CCW pumps were limited to approximately a
5 percent degradation based on the required flows during post LOCA [loss-of-
coolant-accident] recirculation. Using the results of calculation C11353 would
permit degradation of the CCW pumps to less than design basis flow
14
requirements. At the time of the inspection no operability issues were associated
with this condition since new CCW pumps had been installed and were exhibiting
very little degradation. Additionally, subsequent to the development of the
hydraulic model, a sensitivity analysis was performed by the licensee to
demonstrate that a reduced CCW flow requirement would be adequate during
post LOCA recirculation.
Although there were no operability concerns, design basis documents existed
that were not consistent as to inputs and assumptions and were not properly
linked together. This particular example was identified by the licensee just prior
to the inspection and was documented in CAP013269, however this is an
example of the inspectorss concern found in other design basis calculations.
- Calculation 611.1128.M3, Determine the Highest Relieving Pressure in the CC
System, determined the maximum pressure the CCW system could experience
as a result of a tube rupture in one of the major heat exchangers. The
calculation concluded that the low point in the system, the residual heat removal
(RHR) pump seal water heat exchangers, could exceed their design pressure by
approximately 15.7 percent. The calculation concluded that this condition was
acceptable, and the results of the calculation were reflected in USAR Section
9.3.3.
One of the inputs to this calculation was the maximum (shutoff) head of the
CCW pumps. A maximum pump head value of 265 feet was used based on the
original CCW pump curves. The inspectors noted that the new CCW pumps
(DCR 3128) were provided with a maximum head of greater than 270 feet.
Calculation 611.1128.M3 had not been revised to reflect this more limiting input.
In addition, the inspectors noted a slight difference between the calculation
results and the values presented in USAR Section 9.3.3. In response to these
concerns, the licensee initiated CAP013567. The licensee evaluated the
condition and concluded that there were no operability concerns based on the
margins associated with the ASME code, the operating history of this equipment,
and the fact that the system was originally pressure tested to 225 psig.
- Calculation C11396, Effect of Sleeving and 50 Equivalent Plugged Tubes in the
Component Cooling Water Heat Exchangers, assumed a 2500 gpm CCW heat
exchanger flow value, which appeared to be non-conservative. The licensee
stated that the flow value was based on calculation C11376, Determine
Acceptable SW Flow to Component Cooling Water Heat Exchanger, and that
the flow value in C11376 was based on test data from surveillance procedure
SP31-168. The licensee also stated that a concern with this flow value had been
identified shortly before the inspection, and initiated CAP013220.
As discussed in CAP013220, the assumed CCW flow of 2500 gpm would not be
bounding for a single failure scenario resulting in one CCW pump providing flow
to two CCW heat exchangers. In response to this issue, Addendum A to
calculation C11376 was issued to verify that the actual flow rate would be
sufficient for the required heat removal. Addendum A to calculation C11376
included the assumption that 50 equivalent CCW heat exchanger tubes were
15
plugged to be consistent with calculation C11396. As a result of Addendum A to
calculation C11376, it was concluded that the results of calculation C11396 are
bounding.
- Calculation C11053, Evaluate the Acceptability of the Throttled Positions of
Valves CC-402A and CC-402B, assumed CCW system alignment and CCW
flow rates to provide the necessary heat removal to support maintaining the
reactor coolant system temperature at 140F during refueling mode activities.
These unverified assumptions were included in the thermal performance
calculation concerning the alignment and flow through each CCW heat
exchanger, and the flow either through a single RHR heat exchanger or through
other available flow paths that would be in parallel to the flowpath through the
RHR heat exchanger.
The licensee modified existing CAP 008661 and CAP 013259/OTH 008995 to
include verification, using the system hydraulic model, of the assumed CCW flow
through the CCW heat exchangers and through the other downstream parallel
flow paths in subsequent revision of calculation C11053 and associated
operating procedures.
The inspectors also identified that calculation C11053 did not address instrument
accuracy in determining SW system temperature limitations to support
maintaining refueling mode temperatures at 140F. The licensee initiated
CAP013477 to revise calculation C11053 to account for instrument accuracy
(+/- 2F) in determining limitations on the main SW header local temperature
indicators, which would be used to monitor SW inlet temperature.
- Calculations C10510, Voltage Ratings of Safeguard DC Operated Devices,
C-038-003, 125 VDC Safeguard Distribution Network Cable Voltage Drops, and
ESR 90-104, Evaluate DC Distribution to Diesel Generators, each addressed
an aspect of design adequacy for the safeguard 125VDC distribution system.
Calculation C10510, referenced calculation C-038-003, Revision 3, while
Revision 5 had already been issued. Licensee personnel previously determined
that calculation C10510 should have been revised. The inspectors also noted
that ESR 90-104 results were not reflected back into calculation C10510. The
licensee initiated CAP 013368 to address this issue. Due to the fact that the
125VDC safeguard batteries were sized for an 8-hour mission time, and the
licensee was licensed for a 4-hour mission time, there did not appear to be any
operability concerns associated with this issue.
Lack of Available Calculations to Support Aspects of the Current Design Basis - The
inspectors identified the following examples of design basis requirements that were not
supported by available calculations. These conditions also appear to be related to the
deficiencies in calculation control. Because an index of available design basis
calculations was not available, the inspectors found that it was difficult to identify those
design basis requirements that were not supported by calculations.
- The inspectors requested the supporting calculations for the performance of the
CCW system during an 10 CFR 50, Appendix R safe shutdown. The licensee
16
responded that there was not an analysis to address the CCW systems
capability to reach cold shutdown conditions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with a single train of
CCW available as required by Appendix R and initiated CAP013454. The
licensee stated that there was a high level of confidence that this safe shutdown
requirement could be met based on an existing analysis of the Point Beach CCW
system for an Appendix R safe shutdown. The Point Beach CCW system was
capable of achieving cold shutdown conditions for two units within the required
time with only one CCW pump available. The Kewaunee CCW system would be
required to achieve cold shutdown conditions for only one unit with one similar
CCW pump available. Therefore, the licensee concluded that the Point Beach
analysis would be bounding.
- The inspectors requested the supporting documentation to verify the capability of
relief valves CC-611A and CC-611B to pass sufficient flow in the event of a
postulated reactor coolant pump thermal barrier rupture. The Westinghouse
specification sheet indicated that the valves were sized to pass 570 gpm of
water. However, in the event of a thermal barrier rupture these valves would be
required to pass a mixture of steam and water to prevent overpressurization of
the associated CCW piping. CAP 013574 was initiated to address this issue.
The licensee stated that the Point Beach relief valves, which were similar in
design but smaller in size, were sized to pass 380 gpm of steam/water mixture at
25 percent quality. The latest available information from Westinghouse indicated
that a Kewaunee thermal barrier rupture would result in a leakage equivalent to
260 gpm. The licensee stated that a Kewaunee thermal barrier rupture would
also result in a steam/water mixture of 25 percent quality. Therefore, the larger
Kewaunee relief valves appear to be adequate to prevent overpressurization of
the CCW piping system.
Errors in Calculations - The inspectors identified the following examples of a variety of
errors in the calculations reviewed during the inspection. In response to this concern,
the licensee initiated CAP013532 to address the overall issue of calculation errors and
discrepancies.
- Calculation C11400, NEP 4.10 Evaluation of Piping Changes Associated with
DCR 3413, evaluated the effect of adding small vent lines to the CCW system.
A calculation assumption stated, a stress intensification factor for calculating
stress in the main pipe header does not need to be considered for the addition of
the vent line assemblies since the diameter of the vent line branch is less than
D/4 (item 6, Form NEP 4.16-3), where D is the nominal diameter of the header
pipe. NEP 4.16, Piping Configuration Reconciliation to Comply with IEB 79-14,
did not provide any justification to omit stress intensification factors for branch
lines with a diameter less than D/4, which was required by USA Standard Code
for Pressure Piping B31.1.0-1967, Power Piping. CAP013456 was initiated to
address this issue. The licensee, however, did review header pipe stress reports
at the vent line locations and documented that the stresses were very low such
that there were no immediate operability concerns for calculation C11400.
- Calculation C10659, Maximum Working Pressure of RHR Pump Seal Heat
Exchanger, applied the rules of ASME Section VIII to conservatively calculate
17
maximum allowable internal pressure of the RHR pump seal heat exchanger,
even though it was not an ASME stamped vessel. The calculation did not
adequately address the requirements of ASME Section VIII Part UCI,
Requirements for Pressure Vessels Constructed of Cast Iron. Maximum
allowable vessel internal pressure was calculated using UG-22 formulas, but
UCI-3 imposed more conservative pressure-temperature limits that were not
considered. Also, this calculation did not evaluate all applicable loadings of
UG-22 as required by UCI-23 as only internal pressure was evaluated. The
licensee initiated CAP 013592 and demonstrated RHR pump seal cooler
operability for maximum temperature-internal pressure.
Analysis: Evaluation of this issue concluded that it is a design control deficiency
resulting in a finding of very low safety significance (Green). The design control
deficiency was due to a licensee performance deficiency in that design calculations
either did not exist or contained errors. The Mitigating Systems Cornerstone was
affected due to the potential for the CCW system providing long term heat removal
function being degraded by this condition. No other cornerstones were degraded as a
result of this issue.
The inspectors determined that this finding was associated with design control attributes
and affected the objective of the Mitigating Systems Cornerstone to ensure the
capability of the CCW system to respond to initiating events to prevent undesirable
consequences, and is therefore greater than minor. The lack of adequate control and
quality of design basis calculations had the potential to result in the ability of the CCW
system to perform its safety functions to be degraded. Design basis calculations were
routinely used in support of design changes, operating procedures, test acceptance
criteria, and operability determinations.
This finding was assessed as Green because it did not represent an actual loss of the
CCW systems safety function. A review of the system calculations identified a number
of deficiencies, but did not result in any immediate operability concerns. This provided
reasonable assurance that there was not an actual loss of system function due to this
condition. Therefore, this issue was screened out of the significance determination
process as Green.
Enforcement: 10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in
part, that measures be established to assure that applicable regulatory requirements
and the design basis are correctly translated into specifications, drawings, procedures,
and instructions.
Contrary to the above, as of November 8, 2002, the design basis of the component
cooling water system were not correctly translated into plant documents, in that design
calculations contained errors or were not available to verify that the CCW system design
basis capability was maintained.
Because of the low safety significance of this issue and because it is in the licensee's
corrective action program, the issue is being treated as a Non-Cited Violation, consistent
with Section VI.A.1 of the NRC Enforcement Policy (NCV 50-305/02-07-02). The
18
licensee initiated CAP 013531 to address calculation indexing and CAP 013532 to
address calculation errors.
.3 Components
a. Inspection Scope
The inspectors examined the component cooling water pumps and component cooling
heat exchangers to ensure that component level attributes were satisfied. The attribute
selected for review was component degradation.
Component Degradation: This attribute was verified through review of component
repair histories and review of corrective action documents. The inspectors reviewed the
attribute to verify the licensee was appropriately maintaining components in the
component cooling water system
b. Findings
No findings of significance were identified.
4. OTHER ACTIVITIES (OA)
4OA2 Identification and Resolution of Problems
a. Inspection Scope
The inspectors reviewed a sample of component cooling water system, permanent plant
modifications, and 10 CFR 50.59 program problems that were identified by the licensee
and entered into the corrective action program. The inspectors reviewed these issues to
verify an appropriate threshold for identifying issues and to evaluate the effectiveness of
corrective actions related to design issues. In addition, condition reports initiated on
issues identified during the inspection were reviewed to verify adequate problem
identification and incorporation of the problem into the corrective action system. The
specific corrective action documents that were sampled and reviewed by the inspectors
are listed in the attachment to this report.
b. Findings
No findings of significance were identified.
4OA6 Meetings, Including Exits
Exit Meeting
The inspectors presented the inspection results to Mr. T. Coutu, and other members of
licensee management, on November 8, 2002. The licensee acknowledged the findings
presented. The inspectors asked the licensee whether any materials examined during
the inspection should be considered proprietary. Two documents were determined to be
19
proprietary information and both were returned to the licensee at the end of the
inspection. Two follow-up telephone exits were held with Mr. Coutu and other members
of licensee management, on December 19, 2002, and January 21, 2003. The licensee
indicated they did not agree with the NCV 50-305/02-07-01 documented in section 1R17
of this report and may submit an appeal based on different interpretation of which
requirements were applicable for this modification.
20
KEY POINTS OF CONTACT
Licensee Management
M. Aulik, Supervisor Engineering (Modifications)
L. Armstrong, Engineering Director
T. Coutu, Site Vice President, Kewaunee Site
G. Harrington, Compliance Supervisor
K. Hull, Supervisor Engineering (Mechanical)
J. McCarthy, Operations Manager
M. Reddemann, Vice President Engineering
P. Rescheske, Senior Engineer (50.59s)
K. Schommer, Supervisor Engineering (Electrical)
T. Webb, Regulatory Affairs Manager
E. Weinkam, Director Regulatory Services (Hudson)
NRC
A. Gill, Acting Section Chief, Electrical Engineering Branch, NRR
D. Hills, Chief, Mechanical Engineering Branch, Division of Reactor Safety, RIII
J. Lamb, Kewaunee Project Manager, NRR
J. Lara, Senior Resident Inspector
T. Narinder, Electrical Engineering Branch, NRR
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
50-305/02-07-01 NCV Failure to maintain adequate separation of safety related circuits
50-305/02-07-02 NCV Design basis calculations contained errors or did not exist
Discussed
None
21
LIST OF ACRONYMS USED
ADAMS Agency-wide Document Access and Management System
ASME American Society of Mechanical Engineers
BOP Balance-of-Plant
CAP Corrective Action Process
CFR Code of Federal Regulations
CC/CCW Component Cooling Water
DCR Design Change Request
DRS Division of Reactor Safety
F Fahrenheit
gpm Gallons per Minute
IEB Inspection and Enforcement Bulletin
IEEE Institute of Electrical and Electronics Engineers
KAP Kewaunee Assessment Process
LOCA Loss-of-coolant-accident
NCV Non-Cited Violation
NEI Nuclear Energy Institute
NRC Nuclear Regulatory Commission
NRR Office of Nuclear Reactor Regulation
PARS Publicly Available Records System
SDP Significance Determination Process
SI Safety Injection
TB Terminal Board
USAR Updated Safety Analysis Report
VAC Volts Alternating Current
22
LIST OF DOCUMENTS REVIEWED
Number Title Revision/Date
Calculations
1179.M7 Effect of Increased Service Water Temperature on the Revision 0
Component Cooling Heat Exchanger Post-LOCA
Performance
1179.M8 Service Water Elevated Temperature Report Revision 0
8814-05-EPED1 CAPTOR Data Loading for Breaker Coordination Addendum A
611.1098.M2 Component Cooling System Heat Loss Calculations Revision 0
611.1128.M1 Investigation of Component Cooling System Revision 0
Overpressurization
611.1128.M2 Component Cooling Surge Tank Overflow Line Head Revision 1
Calculations
611.1128.M3 Determine the Highest Relieving Pressure in the CC Revision 0
System
611.1147.M1 Containment Integrity Technical Review - Component Revision 0
Cooling Water, Excess Letdown Heat Exchanger Collapse
Pressure
812.1179.P1 Component Cooling Surge Tank Shell and Nozzle Stress Revision 0
Analysis
812.1179.P4 Calculation of Minimum Thickness of Shell and Cover Revision 0
Plates on the Component Cooling Heat Exchangers
C10014 Fuse Tripping Time (Ref. RE-39-023) February 15, 1992
C10030 Electrical Overcurrent Coordination 12/208VAC Distribution March 20, 1992
Cabinets BRA-127 and BRB-127
C10510 Voltage Ratings of Safeguard DC Operated Devices Revision ORG
C10650 DCR 1236 Part 2-Boric Acid Heat Tracing (BAHT) BAHT May 8, 1984
Transformer (BAHT) Evaluation
C10659 Maximum Working Pressure of RHR Pump Seal Heat Revision 0
Exchanger
C10678 1992 Service Water Flow Test Analysis Revision 0
C10809 KNPP Containment Pressure and Temperature Transients Revision 0
Following a Design Basis LOCA or a 3 ft2 Pump Suction
Break
C10915 Safeguard Diesel Generator Loading Adjustments for September 10, 2002
Operation at Frequencies Other than 60 Hertz
C10920 Component Cooling Water System Margin in Post-LOCA Revision 0
Containment Sump Recirculation Mode
C10952 Performance Evaluation of Component Cooling Heat Revision 0
Exchanger under Off-Design Conditions
C10972 Evaluation of Component Cooling in Support of Revision 0
Component Cooling Pump A Replacement
C11053 Evaluate the Acceptability of the Throttled Positions of Revision 1
Valves CC-402A and CC-402B
C11247 480V Safety Related Circuit Breakers; Control Voltage March 21, 2002
Calculation
23
Number Title Revision/Date
C11344 Document the Results of SW System Testing Performed Revision 0
during the Fall 2001 Outage under SOP-SW-02-16 and
SOP-SW-02-17
C11352 Evaluation of As-Found Piston Settings for Snubbers Inside Revision 0
Containment
C11353 Determination of CCW Pump delta-P Acceptance Criteria Revision 0
for use in SP 31-168
C11355 Minimum Desired Component Cooling Flow to the Letdown Revision 0
Heat Exchanger
C11356 CC Pump Motor Operation at 280HP January 25, 2002
C11357 Evaluate the Ability of a Lower Lake Temperature to Revision 0
Compensate for Reduced CC Flow to an RHR Hx
C11359 Component Cooling Flow Evaluation of 02-1932 Revision 0
C11359 Determine Minimum RHR Hx CC Flow Rate and Refine Addendum A
SW Temperature Restriction Based on SOP-CC-31-18
C11376 Determine Acceptable SW Flow to Component Cooling Revision 0
Water Heat Exchanger
C11176 Determination of Available SW to the CCW Hx A with Revision 0
SW-1300A throttled Further Closed
C11195 Maximum Pressure Drop on Component Cooling Heat Revision 0
Exchangers
C11321 Instrument Allowable Value for Foxboro 63U Alarm Relays Revision A
C11357 Evaluate the Ability of a Lower Lake Temperature to Revision 0
Compensate for Reduced CC Flows to RHR Hx
C11357 Identify Approved Reference Documents for References Addendum A
5.1, 5.2, and 5.3
C11359 Component Cooling Flow Evaluation of 02-1932 Revision 0
C11359 Determine Minimum RHR Hx Flow Rate and Refine SW Addendum A
Temperature Restriction Based on SOP-CC-31-18
C11376 Determine Acceptable SW Flow to Component Cooling Revision 0
Water Heat Exchanger
C11376 Evaluate CCW Flow to CCW Hx during Cont. Sump Addendum A
Recirculation
C11380 Condensate and Feedwater Model Revision 0
C11396 Effect of Sleeving and 50 Equivalent Rugged Tubes in the Revision 0
Component Cooling Heat Exchangers
C11396 Address 120F Maximum Service Water Outlet Addendum A
Temperature during Post-LOCA also identified in C11376
and Add Revision Number to a Referenced Calculation
C11396 Document Acceptability of CCW Hx Plugging/ Sleeving Addendum B
Configuration
C11398 CC Hx Tube Sleeve DP Revision 1
C11399 Evaluation of Partial Tube and Sleeve Repair for the CC Revision 0
Hx
C11400 NEP4.10 Evaluation of Piping Changes Associated with Revision 0
DCR 3413
24
Number Title Revision/Date
C11401 CC Hx Tube Sleeve Leakage Revision 0
C11402 CC Hx Analysis and Evaluation by EFCO Revision 0
C11409 CC System Flow Model Development Revision 0
C11432 Emergency Makeup to CCS Revision 0
C11443 Containment Thermal Hydraulic Response to Design Basis Revision 0
Analysis (DBA) Loss of Coolant Analysis (LOCA) with
Reduced Component Cooling System (CCS) Flow
C-038-003 125 VDC Safeguard Distribution Network Cable Voltage Revision 5
Drops
C-042-001 Safeguard Diesel Generator Loading (Addendum A) April 22, 2002
ER 31-003 Throttle Valve Position Control December 18, 1991
ESR 90-104 Evaluate DC Distribution to Diesel Generators September 27, 1990
ESR 93-100 Component Cooling Surge Tank Level Revision 0
GMP-238 MOV Thrust and Torque Evaluations - FW-12A August 31, 2001
September 24, 2001
GMP-238 MOV Thrust and Torque Evaluations - CC-612A November 1, 2002
M-1052-1 Kewaunee Plant Outside Shield Building Safeguard and Revision 1
Important to Safety Equipment Area Temperatures
Following a LOCA Inside Containment
Minimum Flow Study for Pump Graham Seal Cooler July 10, 1992
Performance
Condition Report Documents Reviewed During the Inspection
KAP 97-622 IST Acceptance Criteria for ESF Pumps February 7, 1997
KAP 00-001414 Ground Pertabation on 5/10/00 Causes Short Lived Loss of May 10, 2000
Voltage
KAP 00-002566 Design Basis Information for SW and SW Components July 13, 2000
Difficult to Locate
KAP 00-003120 Various Alarms-SW-1306A and B Failed Open September 3, 2000
KAP 01-018695 SW-1306A/B Failures November 20, 2001
KAP 01-018732 SW-1306B Fails Open During SP-33-10 November 24, 2001
CE 009496 Calculation C10920- CCW Pump Analyses & Assumptions March 10, 2000
CAP002818 PS 26018, (CC Pmps 1A/1B Low Dish Press Backup Pump May 9, 2000
Start/Low Alarm) Drift
CAP002706 CC-4A Handwheel Broke Free When Valve Closed October 16, 2000
CAP002684 AC/DC Load Forms Misplaced November 17, 2000
CAP002008 Documentation on the Results of the Kewaunee Flooding June 21, 2001
Study is Incomplete
CAP008327 Pressure Setting of Some CCW Thermal Relief Valves do August 3, 2001
no Account for Sufficient Backpressure
OTH002449 CCW Temperature Could Reach 39F During an Event or August 30, 2001
an SI Signal
CAP007760 Relief Valves Disassembled Prior to As-found Tests October 31, 2001
CAP000844 CFC Tube Life Unknown Due to Material Change November 7, 2001
25
Number Title Revision/Date
CAP000761 SW-1306A Actuator Adjustment November 15, 2001
CAP000646 SW-1306A/B Failures November 21, 2001
ACE000100 SW-1306A/B Failed Open During performance of SOP- November 21, 2001
ELV-40-8
MRE000082 Maintenance Rule Evaluation for Valves SW-1306A/B November 21, 2001
Failures
CAP000656 SW-1306B Fails Open During SP-33-110 Testing November 24, 2001
ACE000103 Apparent Cause Evaluation for Valve SW-1306B Failure November 24, 2001
During Test SP-33-110
MRE000084 Maintenance Rule Evaluation for Valve SW-1306B Failure November 24, 2001
During Test SP-33-110
CAP000587 Complete 1993 Calculations C-038-009 & 010 November 29, 2001
CAP000588 Calculation C-038-011 125VDC Battery Duty Cycle for November 29, 2001
Battery C & D Has No Acceptance Criteria
CAP007663 FW-12B Fails Timing Test Following DCR 3325 December 3, 2001
CAP000503 Lack of Operations Administrative Guidance December 6, 2001
CAP000145 Design Change versus Systematic Approach to Training January 15, 2002
CAP000074 Possible CC Pump Runout on LOOP and Single Failure January 23,2002
CE 000061 Condition Evaluation per CAP000074 January 23, 2002
CA 000070 Install Valve travel Limiter on Valve CC-302 January 24, 2002
CA 000071 Install Ultrasonic Flow Meters on RHR HX A & B CC Piping January 24, 2002
CAP002927 Snubber FDW-H114 Appears to Be Bottomed Out and January 28, 2002
Carrying Load
CA 000073 CC-302 AOV Program Scoping and Categorization February 7, 2002
Process
CAP003114 Perform Maintenance Rule (a)(1) Evaluation for SW- February 12, 2002
1306A/B Valves
CAP003191 Inadequate Procedure February 20, 2002
CAP011530 CC System Leak Developed following Flush of CC Hx May 2, 2002
CAP011556 Evaluate B CCW Hx Condition After finding Tube Cracks in May 5, 2002
A CCW Hx
CAP011560 USAR Changes Involving Plant Design Load Change May 6, 2002
Capability Require 50.59 Review
CAP011582 CCW Hx A Tube Leaks May 7, 2002
CAP011828 System 31 Maintenance Rule (a)(1) Evaluation Required June 7, 2002
CAP011972 CCW Accident Flow Rate for CC-3A(B) not Specified in June 20, 2002
Test Procedure
CAP012001 SW-1306A and SW-1306B Opened due to Electrical June 24, 2002
CE010129 SW-1306A and SW-1306B Opened due to Electrical June 25, 2002
CAP012029 ECP Concern June 25, 2002
CAP012174 Shoto Substation Capacitor Bank Problem July 9, 2002
26
Number Title Revision/Date
MRE001523 Maintenance Rule Evaluation for Valves SW-1306A/B July 9, 2002
Failures
CE010241 Shoto Substation Capacitor Bank Problem July 10, 2002
MRE001526 Shoto Substation Capacitor Bank Problem July 10, 2002
CAP012211 Perform Maintenance Rule Evaluation on Failure - 6/23/02 July 12, 2002
CAP012212 NAO Discovered that Hose House #2 South of Main July 12, 2002
Transformers Was Not Sealed
CE10063 Perform a Condition Evaluation Per CAP011928" July 17, 2002
CAP012631 Evaluate CC Pump Check Valve Slam and Method to August 19, 2002
Avoid it
RCE-576 Root Cause Evaluation - Tube Leaks Identified in August 27, 2002
Component Cooling Water Heat Exchanger
CAP012749 Predicted CC Flow for Components Supplied is Less than August 28, 2002
Documented Requirements
CAP012800 Calculation Error September 3, 2002
CAP012942 Seal Water Heat Exchanger Component Cooling Outlet September 15, 2002
Flow Indicator Pegged High
CAP012993 Final Configuration of CCW Hxs Not Specifically September 18, 2002
Addressed in DCR Package
CA 00861 Ensure That Affected Operations Procedures Will Be September 20, 2002
Revised as Stated in CA 007724
CAP013087 QA Vault Missing 5 CCW Records September 25, 2002
CAP013094 Insufficient Description in Calculation C10266 September 25, 2002
CAP013104 CC Surge Tank Level XMTR 24041 Drift and Nonlinearity September 26, 2002
CAP013119 SSDI CCW Missing Documents September 27, 2002
CAP013137 Adequacy of Heat Exchanger Configuration Control September 30, 2002
CAP013177 Current Status of Numerous System 31 Calculation not October 3, 2002
Readily Apparent
CAP013209 Potential USAR Discrepancy Regarding Hi Flow Alarm on October 7, 2002
CC Return Flow From a RXCP
CAP013212 Set Point Discrepancy for SI Pump Low Flow Alarm October 8, 2002
CAP013220 Post Accident Analysis Flow Rate Assumption for CCW HX October 8, 2002
CAP013269 Component Cooling Water System IST Acceptance Criteria October 11, 2002
CAP013259 Revise calculation C11053 October 14, 2002
CAP013368 A Review of Calc. # C10510 Orig. Identified Several Issues October 18, 2002
CE 10920 Condition Evaluation per CAP13269 October 15, 2002
CA 009118 Corrective Action for CE10920 October 29, 2002
ACE 001828 SW1306A/B Opens During Disturbances on External
Electrical Grid
Condition Reports Written as a Result of the Inspection
CAP013427 Axial IST Vibration Reading Not Taken on CCW Pumps October 23, 2002
CAP013430 Apparent Lack of a UFSAR Discussion of the Maximum October 23, 2002
Allowable SW Temperature
27
Number Title Revision/Date
CAP013448 CC Heat Exchanger Performance Testing October 24, 2002
CAP013454 Lack of Cooldown Analysis for Appendix R October 24, 2002
CAP013456 Stress Intensification Factor October 24, 2002
CAP013457 AS-Built Drawing Discrepancy for DCR 3413 October 25, 2002
CAP013471 Determine Adequacy of Scope of NEP 14.17 Evaluation for October 28, 2002
CC-612A & B
CAP013475 MOV Drawing Irregularities October 28, 2002
CAP013477 Revise Calculation C11053 to Include Instrument Accuracy October 28, 2002
CAP013500 Discrepancies in CII356 (CC Motor at 280 HP) October 30, 2002
CAP013501 Design Description 3163 Discrepancy Found During SSDI October 30, 2002
CAP013504 RHR Hx Transfer Surface Area October 30, 2002
CAP013515 Revise Calculation No. 812.1179.P4 October 30, 2002
CAP013517 Component Cooling Surge Tank Level Loop 618 is not October 31, 2002
Calibrated Properly
CAP013523 50.59 (SE #02-06) Conclusion for TCR 02-02 October 31, 2002
CAP013530 Errors in NEP 4.9 Evaluations Discovered During SSDI October 31, 2002
CAP013531 Lack of Calculation Indexing October 31, 2002
CAP013532 Review Recent Calculation Errors and Discrepancies for October 31, 2002
Common Issues
CAP013561 No Guidance on the Effect of Low Temp CCW to the RXCP November 5, 2002
Thermal Barrier Post-SI
CAP013564 IPEOP Improvement November 5, 2002
CAP013566 Full Flow Test Requirements for CC-3A(B) November 5, 2002
CAP013567 Replacement Component Cooling Pump, Maximum Pump November 5, 2002
Head Determination
CAP013572 CCW SSDI - Update Calculation 611.1128.M1, as November 5, 2002
Reference for Relief Flow Rate
CAP013574 Relief Valve CC-611A(B) Required Capacity Discrepancy November 5, 2002
CAP013575 Evaluate Manual Valve Maintenance in the CCW System November 5, 2002
CAP013580 Problems Identified Under ICPs May not be CAPd November 6, 2002
CAP013581 Review to Determine If Any DG Loads May be Removed November 6, 2002
CAP013582 NEP 4.9 Recommendations Needs to Be Formalized November 6, 2002
CAP013584 CC System Flow Balancing November 6, 2002
CAP013588 Post Installation Vibration Data for CCW Pumps November 6, 2002
CAP013592 Inadequacy of RHR Pump Seal Cooler Max Operating November 6, 2002
Pressure Calculation
CAP013593 Discovered Drawing Discrepancy November 7, 2002
CAP013594 Z Dimensions for Snubbers November 7, 2002
CAP013607 Accuracy of UFM Measurements November 7, 2002
CAP013608 Areas of DGs SP testing for TS4.6.a that need revision November 7, 2002
CAP013805 Questions on SW-1306A/B Separation and Failure Impact November 25, 2002
on Reactivity
CAP014584 Inadequate 50.59 Evaluation February 4, 2003
28
Number Title Revision/Date
Design Change Requests
DCR 955 Modify the Logic to CC-610A and B Valve June 17, 1980
DCR 1560 Remove Relief Valve on Comp. Cooling Tnk. & Vent Tnk. February 15, 1988
Using Existing Vent Line
DCR 2283 Replace Instruments 26309, 26310. 25015 and 35016 January 2, 1988
DCR 2603-1 Evaluate Connection of Static Trip II Replacement Relays May 17, 1993
for Six Breakers
DCR 2728 Replacement of Westinghouse BFD Relays October 7, 1994
DCR 3055 Replace RCS Flow Transmitters May 28, 2002
DCR 3128 Replace Component Cooling Pumps March 30, 2001
DCR 3163 Modify Controls for Valves SW 1300 A(B) and SW October 3, 2000
1306A(B) on an SI Signal
DCR 3325 Upgrade FW12A & FW12B Actuators August 26, 2001
DCR 3331 Interposing Relays for Control Room Status Lights March 21, 2002
DCR 3355 Reactor Coolant Pump (RXCP) Overcurrent Relay Setting September 11, 2002
DCR 3412 CCW Heat Exchanger Tube Sleeving May 11, 2002
TCR 02-15 Install Data Acquisition Equipment to CC-302 August 29, 2002
Drawings
110E001, Sh. 3 Auxiliary Coolant System Engineering Flow Diagram Revision 12
206C927, Sh. 6 Line List Auxiliary Coolant System Revision 1
237127A-E2492 Schematic Diagram-Control Valves CV-31406,31407 Revision D
237127A-M326 Auxiliary Coolant Piping Sheet 1 of 3 Revision AC
237127A-M360 Reactor Bldg. Piping - Chem. & Vol. Control, Auxiliary Revision R
Coolant & Safety Injection
237127A-M361 Reactor Bldg. Piping - Chem. & Vol. Control, Auxiliary Revision X
Coolant & Safety Injection
834823-M-1423 2" SW Emergency Makeup to CC Isometric Revision 1
E-204 Integrated Logic Diagram Component Cooling System Revision AK
E-235 Circuit Diagram 480V SWGR.-Safeguard Buses Revision AJ
E-240 Circuit Diagram 4160V & 480V Power Sources Revision AQ
E-567 Motor Control Center 1-52B and 1-62B Revision T
E-604 W/D Motor Control Center 1-52B (Sh.2) Revision AS
E-614 W/D Motor Control Center 1-62E Revision AK
E-615 W/D Motor Control Center 1-62E (Sh.2) Revision BG
E-625 External Connection Motor Operated Valves Sh.4 Revision AC
E-627 External Connection-Motor Operated Valves Sh.6 Revision BJ
E-778 W/D Sequence Loading Panel DR116 Train B Revision AV
E-799 W/D Technical Cabinet TC 1956 Revision DY
E-1082 Control Schematic 480V Breakers 15108 & 15109 Revision T
E-1089 Control Schematic 480V Breakers 166108 & 16109 Revision S
E-1345 Schematic Diagram M.C.C. 1-52B Motors 1-102 Revision M
29
Number Title Revision/Date
E-1349 S/D MCC 1-52B Motors 1-399 Revision AC
E-1351 W/D External Connection Sol. Valves TBs 1357, 1358, Revision AJ
1360 and 1391
E-1404 S/D MCC 1-62B Motors 1-584 and 1-763 Revision U
E-1425 S/D MCC 1-62E Motor 1-364 MCC 1-62B Motor 1-364 Revision S
E-1427 Schematic Diagram MCC 1-62E Motors 1-446 Revision F
E-1531 W/D Ext. Conn. Sol. & Cont. Valves TBs 1357, 1358, Revision AJ
1360, & 1391
E-1540 S/D Solenoid Valves SV33077, 78, 80, 81, 82 and 83 Revision L
E-1621 Integrated Logic Diagram DG Mechanical System Revision AH
E-1632 Integrated Logic Diagram Service Water System Revision AH
E-1637 Integrated Logic Diagram Diesel Generator Electric Revision W
E-1638 Integrated Logic Diagram Diesel Generator Electric Revision V
E-1639 Integrated Logic Diagram Diesel Generator Electric Revision M
E-1815 W/D Mechanical Control Console C View B CR104 Revision AJ
E-1816 W/D Mechanical Control Console C View C CR104 Revision AQ
E-1828 W/D Mechanical Vertical Panel A View B CR106 Revision AJ
E-1830 W/D Mechanical Vertical Panel A View C Lower CR106 Revision BM
E-1832 W/D Mechanical Vertical Panel A View D Lower CR10 Revision BN
E-1876 Schematic Diagram Load Shedding Train A Revision P
E-1877 Schematic Diagram Load Shedding Train A Revision T
E-1881 Schematic Diagram Sequence Loading Bus 1-5 Revision P
E-1912 Schematic Diagram Solenoid Valves SV 33074 and 75 Revision G
E-2026 Integrated Logic Diagram Chemical and Volume Control Revision Q
System
E-2045 Integrated Logic Diagram Component Cooling System Revision AC
E-2055 Integrated Logic Diagram Component Cooling System Revision M
E-2105 External Connection Sol. and Cont Valves TB 1351 and Revision CK
1363
E-2116 W/D Sol. and Cont. Valves TB 1372 Revision S
E-2112 W/D Sol V/VS Cont. V/VS and Dampers TB-1371, 1377, Revision V
1378, 1458
E-2198 S/D Sol. Valves 3343301, 2, 4, 33769 and 33770 Revision F
E-2233 Relay Settings Sh.33 Revision F
E-2234 Relay Settings Sh.34 Revision G
E-2243 Relay Settings Revision L
E-2244 Relay Settings Revision L
E-2358 W/D Fuse Panel RR172 AC Normal 1 Dist. Revision AQ
E-2359 W/D Fuse Panel RR173 AC Normal 2 Dist. Revision BE
E-2492 Schematic Diagram - Control Valves CV-31406, 31407 Revision G
E-2545 Instrument W/D Component Cooling Flow Return & Revision E
Component Cooling Heat Exgrs 1A/1B Outlet Temp.
E-2551 Instrument W/D Reactor Coolant Sys Flow-Loop A Revision C
30
Number Title Revision/Date
E-2565 Instrument W/D Reactor Coolant Sys Flow-Loop B Revision C
E-2566 Instrumentation W/D Reactor Coolant Sys Flow- Loop B Revision C
E-2567 Instrumentation W/D Reactor Coolant Sys- Flow Loop B Revision D
E-2568 Instrumentation W/D Reactor Coolant Sys Flow - Loop A Revision C
E-2569 Instrumentation W/D Reactor Coolant Sys Flow - Loop B Revision E
E-3106 Integrated Logic Diagram Component Cooling System Revision G
M-328 Auxiliary Coolant Piping Revision AF
XK-100-18 Flow Diagram Auxiliary Coolant System Revision AL
XK-100-19 Flow Diagram Auxiliary Coolant System Revision AE
XK-100-20 Flow Diagram Auxiliary Coolant System Revision T
XK-100-61-1 KNPP Component Cooling Surge Tank Revision 4A
XK-100-622 Interconnection Wiring Diagram Rack R2 WPS Nuclear Revision 2L
Power Plant Reactor Protection System
X-K100-717 Relief Valve Revision A1
X-K100-731 Relief Valve Revision A1
Miscellaneous
031-014 SSFI Documentation Sheet - Normal Heat Load Calcs for September 26, 1990
CC Water Hx Review
D-31-047 SSFI Documentation Sheet - Review of Calculation October 4, 1990
Component Cooling Heat Exchanger
D-31-081 SSFI Documentation Sheet - CCW Low Pressure Alarm October 19, 1990
and CCW Pump Auto-Start Setpoint
D-31-099 SSFI Documentation Sheet - CC Hx Capacity October 12, 1990
D-31-099 SSFI Documentation Sheet - CCW System Relief Valve October 18, 1990
ER-031-012 Evaluation - Component Cooling Water Heat Exchanger March 13, 1993
ER-031-013 CCW Pump - Low Pressure Alarm Setpoint April 15, 1991
ER-031-023 SSFI Evaluation Sheet - CCW System Relief Valve Testing April 19, 1991
Form U-1 Manufacturers Data Report - Item AH-CC550 (CCW HX) N/A
FSD/SS-M-3357 Westinghouse Letter - Component Cooling Water System July 13, 1984
Safety Review Committee Finding
K100-2557 Westinghouse Instruction Manual - Auxiliary Heat N/A
Exchangers
KP-S-2213 Pioneer Letter - Auxiliary Coolant Valves March 1, 1972
KP-W-1455 Pioneer Letter - Component Cooling System - Component May 11, 1972
Pressure Drops
Pioneer Memorandum - Berzins to Hickey - Component March 28, 1973
Cooling System Set-point Change (FIA-26602)
NSC-KP-M-SLR Pipe Rupture Analysis - Component Cooling System April 17, 1972
-83
OEA 93-204 Recirculation Phase Design Issue January 31, 1996
OEA 97-056 NRC Information Notice 1996-031 July 10, 1997
R-31-012 Request for Information - CC Heat Exchanger October 25, 1990
31
Number Title Revision/Date
R-31-013 Request for Information - CCW Low Pressure Alarm and October 18, 1990
CCW Pump Auto-Start Setpoint
R-31-023 Request for Information - CCW System Relief Valve Test October 23, 1990
RESP-031-012 Request for Information Response April 22, 1993
RESP-031-013 Request for Information Response April 15, 1991
Section 4.6 PRA Component Cooling Water System Notebook August 29, 2002
SER Section 8.3 On-site Power System-AC Power System July 14, 1972
SER Section Component Cooling System July 24, 1972
9.3.2
SSEP-13-1 Operating Conditions Evaluation - OPS Valve No. Revision Original
CC-612A
SSEP-13-1 Operating Conditions Evaluation - OPS Valve No. Revision Original
CC-601B
SSEP-13-1 Operating Conditions Evaluation - OPS Valve No. Revision Original
CC-601A
SSEP-13-1 Operating Conditions Evaluation - OPS Valve No. Revision Original
CC-612B
System No. 31 KNPP System Description - Component Cooling Water Revision 2
System (CC)
UCR # R18-006 USAR Change Request (UCR) No. R18-006 / Pending October 1, 2002
SW-1306A and SW-1306B Control Circuit Evaluation January 6, 2003
EQ Temperature Profile for Containment N/A
Component Cooling Water System (CC) Design Basis Draft
System Functional Matrix
System Health Report - Component Cooling September 2002
Kewaunee Inservice Testing Program
AOV Ranking Worksheets CC610A/B Revision 0
CCW Pumps Inservice Testing Hydraulic and Vibration
Data -2000 through 2002
Preoperational Tests
CC-1 Component Cooling Water Initial Fill & Operation October 16, 1973
CC-2 Component Cooling Water Cold Functional Testing November 5, 1973
CC-3 Component Cooling Water Hot Functional Testing November 1, 1973
Procedures
A-CC-31 Abnormal Component Cooling System Operations Revision A
A-CC-31A Abnormal Conditions in the Component Cooling System Revision 0 (deleted)
A-CC-31B Leakage Into Component Cooling System Revision I (deleted)
A-SW-02 Abnormal Service Water System Operation Revision S
32
Number Title Revision/Date
DC/PM 3128-2 CC Pump A Installation - Retest 08-28-01
DC/PM 3128-4 CC Pump B Installation - Retest 08-28-01
E-CC-31 Loss of Component Cooling Revision L (deleted)
E-0 Reactor Trip or Safety Injection May 30, 2002
E-0-06 Fire in Alternate Fire Zone Revision O
E-1 Loss of Reactor or Secondary Coolant Revision N
ECA-0.0 Loss of All AC Power Revision Y
ECA-0.1 Loss of All AC Power Recovery Without SI Required Revision M
ECA-0.2 Loss of All AC Power Recovery With SI Required Revision L
ECA-1.2 LOCA Outside Containment Revision I
ES-1.2 Post LOCA Cooldown and Depressurization Revision M
ES-1.3 Transfer to Containment Sump Recirculation Revision S
GNP-04.03.01 Guide to Safety Review, Safety Evaluations and Second Revision A
Level Reviews
GNP-04.03.02 Plant Physical Change Screening Revision C
GNP-04.03.03 Plant Physical Change Control Revision D
GNP-04.03.04 Calculation - Preparation, Review, and Approval Revision D
GNP-04.04.01 50.59 Applicability Review and Pre-Screening Revision B
GNP-04.04.02 50.59 Screening and Evaluation Revision A
GNP-06.02.01 Procurement Technical Evaluation Administration Revision B
GNP-06.02.02 Procurement Technical Evaluation Procedure Revision C
ICP 31-02 CC - Flow Indicators Calibration Revision K
N-CC-31 Component Cooling System Operation Revision X
N-CC-31-CL Component Cooling System Prestartup Checklist Revision W
N-RC-36A Reactor Coolant Pump Operation Revision AA
N-SW-02 Service Water System Revision W
NAD-04.03 Plant Physical Changes Revision D
NAD-04.04 Changes Tests and Experiments (10CFR50.59) Revision B
NAD-06.02 Procurement Technical Evaluations Program Revision D
NEP-04.16 Piping Configuration Reconciliation to Comply with IEB 79- Revision B
14
SP-33-110 Diesel Generator Automate Test Revision AC
SP-42-04A Diesel Generator A Operated Test Revision S
SP-42-047B Diesel Generator B Operated Test Revision U
SP-42-312A Diesel Generator A Availability Test Revision R
SP-55-177 Inservice Testing of Pumps Vibration Measurements Revision Y
SP-168A Train A Component Cooling Pump and Valve Test - IST Revision Original
SP-168B Train B Component Cooling Pump and Valve Test - IST Revision Original
47024-H CC Surge Tank Level High/Low Revision C
Procurement Technical Evaluation
PTE 92-0154 Diode for Diesel Generator February 19, 2002
33
Number Title Revision/Date
PTE 92-0186 Service Water Pump Parts Revision 24
PTE 92-0196 Brass Whitey Valves Revision 7
PTE 93-0031 Oils, Greases and Lubricants Revision 29
PTE 94-0009 Crosby 3/4 Inch JMAK Spec Type B Relief Valves Revision 3
PTE 00-0023 Component Cooling Heat Exchanger Parts Revision 3
PTE 01-0058 Anchor Darling 3" 150 lb. Flex-Wedge Gate Valves Revision 1
PTE 02-0016 Upgraded Gaskets for the RCP Revision 0
PTE 02-0022 Lever Pin for Anchor-Darling 6" Swing Check Valve Revision 0
PTE 02-0025 ASCO Solenoid Valves - Model 8342 Revision 0
Specifications
S3397-1 Main Feedwater Flow Control Valves Trim Replacement June 5, 2002
NEP 4.9 Electrical Load Addition DCR-3190 November 4, 2002
No. 2003 Specification for Piping Design Revision 9
ES-9010 Cable Installation and Separation Criteria March 21, 1987
E-Spec 676257 Westinghouse- CC System Relief Valves, Sheet 19 Revision 3
E-Spec 676257 Westinghouse- CC System Relief Valves, Sheet 2 Revision 4
Surveillances (Date Shown Is Date Surveillance Was Completed)
CMP-31-02, (CC) Component Cooling Water Heat Exchanger Cleaning March 18, 1993
GMP 137 (QA-1) - Performed for CC Hx A March 17, 1994
April 12, 1994
April 14, 1995
October 31, 1995
November 14, 1995
September 30, 1996
April 30, 2000
September 13, 2001
CMP-31-02, (CC) Component Cooling Water Heat Exchanger Cleaning March 12, 1993
GMP 137 (QA-1) - Performed for CC Hx B April 17, 1994
April 10, 1995
September 25, 1995
October 2, 1995
October 2, 1996
May 4, 2000
September 13, 2001
DCR 2468 Service Water Flow Test Train A April 8, 1992
ICP 31-01 CC - Surge Tank Level Loop 618 Calibration October 29, 1999
February 1, 2001
September 25, 2002
ICP 31-04 CC - Heat Exchanger 1A/1B Flow Loop 619 Calibration March 5, 1999
September 6, 2000
February 26, 2002
ICP 31-05 CC - Pumps 1A/1B Discharge Header Pressure Indicator May 5, 2000
Controller 26018 Calibration March 19, 2001
June 25, 2002
34
Number Title Revision/Date
ICP 31-11 CC - Heat Exchanger 1A/1B Outlet Temperature Loop 621 August 14, 1998
Calibration January 19, 2000
July 16, 2001
SOP-CC-31-1 Component Cooling Flow Test January 5, 1993
SOP-CC-31-17 CC Flow Measurement - Post CC-302 Limiter Installation January 26, 2002
SOP-CC-31-18 CC Flow Measurement thru Both RHR Hxs with CC-302 March 21, 2002
Full Open
10 CFR 50.59 Evaluations
DCR 3128 Replace Component Cooling Pumps March 30, 2001
DCR 3163 Modify Controls for Valves SW 1300 A(B) and SW Revision 1
1306A(B) on an SI Signal
SE 01-56 Upgrade the ICCMS train A and B Modems to Eliminate November 15, 2001
Self Coupling
SE 01-64 DCR 3260 November 28, 2001
SE 02-01 CC Pump Operation with Two Pumps Running January 11, 2002
SE 02-02 SOP-CC-31-16 & TCR 0201 January 25, 2002
SE 02-03 SOP-CC-31-16, Rev ORIG January 24, 2002
SE 02-04 SOP-CC-31-17, Rev ORIG January 26, 2002
SE 02-06 TCR 02-02, Bypass Forebay Low-Low Level CW Pump March 1, 2002
Trip
SE 02-07 SOP-CC-31-18, Rev ORIG January 25, 2002
10 CFR 50.59 Screenings
DCR-3128 Replace Component Cooling Pumps March 30, 2001
SCRN 02-003 DG Loading Calculation Revisions April 22, 2002
SCRN 02-012 ES-1.3 / Revision S April 30,2002
SCRN 02-033 DCR 3413 May 9, 2002
SCRN 02-034 DCR 3412, CCW Heat Exchanger Tube Sleeving May 8, 2002
SCRN 02-040 Revision to C10915 Rev 3 September 9, 2002
SCRN 02-061 Replace Trim in Main Feedwater Regulating Valves July 17, 2002
SCRN 02-069 DCR 3350 Replace AFW Check Valves July 22, 2002
SCRN 02-075 PTE 02-0016, Revision 0 June 20, 2002
SCRN 02-115 DCR/PM 3394 September 13,2002
SCRN 02-120 Revision to C10032 Rev 1 September 16, 2002
SCRN 02-121 Revision to C-038-003 Rev 5 September 16, 2002
Technical Specifications Section 3.6 Containment Systems Amendment # 155
Section 3.3.d Component Cooling System Amendment # 116
Section 4.6 Periodic Tests of Emergency Power System Amendment # 119
Updated Safety Analysis Report
Section 5.3 Reactor Containment Vessel Isolation Systems Revision 16
35
Number Title Revision/Date
Section 8.2 Electrical Systems Revision 17
Section 9.3 Auxiliary Coolant System Revision 17
Section 9.6 Facility Services Revision 17
Table 5.2-2 Reactor Containment Vessel Penetrations Revision 16
Table 6.2-7 Residual Heat Exchangers Design Parameters Revision 16
Table 6.2-9 Shared Functions Evaluation Revision 16
Table 8.2-1 Diesel Generator Load (Max.) for DBA Revision 17
Table 9.3-1 Component Cooling System Component Data Revision 17
Table 9.3-2 Residual Heat Removal System Component Data Revision 16
Table 9.3-3 Spent Fuel Pool Cooling System Component Data Revision 17
Table 9.3-4 Auxiliary Coolant System Code Requirements Revision 16
Table 9.3-5 Auxiliary Coolant System Failure Analyses Revision 17
Table 11.2-7 Radiation Monitoring System Channel Data Revision 17
Table 14.3.4-19 LOCA Containment Response Analysis Parameters Revision 17
Vendor Manuals
Ingersoll-Dresser Component Cooling Water Pumps March 2001
91456 Controlotron Field Manual System 1010P Uniflow Universal
Portable Flowmeter
36