ML050560348
| ML050560348 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 08/31/2003 |
| From: | NRC/RGN-II |
| To: | |
| References | |
| FOIA/PA-2004-0277, IR-03-006 | |
| Download: ML050560348 (11) | |
See also: IR 05000321/2003006
Text
- 210
August xx, 2003 S
SUBJECT: EDWIN I. HATCH NUCLEAR POWER PLANT - NRC TRIENNIAL FIRE
PROTE TION INSPECTION REPORT &o 82t0e-e& AND 60 3664*46-
Dear Mr. Sumn
On July 25, 2003, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at
your Hatch Nuclear Plant Units 1 and 2. The enclosed inspection report documents the
inspection findings, which were discussed on that date with Mr. R. Dedrickson and other
members of your staff.
The inspection examined activities conducted under your license as they relate to safety and
compliance with the Commission's 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 four findings that have potential safety significance greater than very low
significance, however a safety significance determination has not been completed. One issue Avo//Mf
o d did present an immediate safety c ae .
The-
-4-rovn the tcaionieI proeuetslv
Oirc h oor ietfie yt;6~o
sro-odu- iv cio roowe i evthisedton. The other three issues i
present ediatesafety concern
In adison, the report documents three NRC-identified findings of very low safety significance\
(Green), all of which were determined to involve violations of NRC requirements. However,
because of the very low safety significance and because they are entered into your corrective
action program, the NRC is treating these three findings as non-cited violations (NCVs)
consistent with Section VL.A of the NRC Enforcement Policy. If you contest any NCV in this
report, you should provide a response within 30 days of the date of this inspection report, with
the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control
Desk, Washington DC 20555-0001; with copies to the Regional Administrator Region 1I; the'
Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington,
DC 20555-0001; and the NRC Resident Inspector at Hatch Nuclear Power Plant.
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O '.pAD
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In accordance with 10 CFR 2.790 of the NRC's 'Rules of Practice," a copy of this letter and its
enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publically Available Records (PARS) component of NRC's document system
(ADAMS). ADAMS is accessible from the NRC Website at
http://www.nrc.gov/readina-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
IRAI
Charles R. Ogle, Chief
Engineering Branch 1
Division of Reactor Safety
Docket Nos.: 50-321, 50-366
Enclosure: Inspection Report 50-321, 366/03-06
ice A: U.S. NUCLEAR REGULATORY COMMISSION
551- Li
REGION II
Docket Nos.: 50-321, 50-366
Report No.: 5O82I0308~
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Licensee: Southern Nuclear Operating Company
Facility: E. I. Hatch Nuclear Plant
-Location: P. O. Box 2010
Baxley, GA. 31513
Dates: July 7-11, 2003 (Week 1)
July 21-25, 2003 (Week 2)
Inspectors: C. S E., Senior Reactor Inspector, (Lead Inspector)
R. Sc nior Reactor Inspector
G. Wiseman, Fire Protection Inspector
K. Sullivan, Consultant, Brookhaven National Laboratory
Accompanying
Personnel: S. Belcher, Nuclear Safety Intern, (Week
by: Charles R. Ogle, Chief
Engineering Branch I
Division of Reactor Safety
wiv'
%.abe
SUMMARY OF FINDINGS _ _. __ - -
REPORT DETAILS _ _ _ _ -
REACTOR SAFETY
FIRE PROTECTION _ . __ _ _
Systems Required to Achieve and Maintain Safe Shutdown - - - - -
re Protection of Safe Shutdown Capability - --
Post Fi 7 town Capability
Operational Implementation of Alternative Shutdown Capability - -
C munications
mergency Lighting
Cold Shutdown Repairs
Fire Barriers and Fire Area/Zone/Room Penetration Seals
ire Protection Systems, Features, and Equipment . -
SAFETY SYSTEM DESIGN AND PERFORMANCE CAPABILITY
DCR 91-134, SRV Backup Actuation via Pressure Transmitter Signals
OTHER ACTIVITIES
Identification and Resolution of Problems
Meetings Including Exit
ulemental Information
t of Items Opened, Closed, and Discussed
ist of Documents Reviewed
REPORT DETAILS
- wL,
1. 'REACTOR
171-
SAFETY41"_. A ,' f 4 A f__ 4----
Cornerstones: Initiating Events, Mitigating Systems and Barrier Integrity
1R05 FIRE PROTECTION
The purpose of this triennial fire protection inspection was to perform a risk-infcormed
inspection of defense-in-depth mitigating elements provided to ensure the succ essful
accomplishment of safe shutdown conditions in the event of fire at the Hatch Nuclear
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systems, equipment and operating procedures. The evaluation did not include a / (gYL
comprehensive review of the potential impact of fire-induced failures in associated
circuits of concern to post-fire safe shutdown. The inspection was performed in
accordance with the nj Nuclear Regulatory Commission (NRC) reactor oversight
process using a risk i formed approach for selecting the fire areas and attributes to be
inspected. The team used Plant Hatch Individual Plant Examination of External Events,
to choose several risk significant areas for detailed inspection and review. The fire
areas chosen for review during this inspection were:
Fire Area 2016, West 600 V Switchgear Room, Control Building, Elevation 130
. /. feet.
.T Fire Area 2104, East Cableway, Turbine Building, Elevation 130 feet.
A) Fire Area 2404, Switchgear Room 2E, Diesel Generator Building, Elevation 130
feet.
A(0 Fire Area 2408, Switchgear Room 2F, Diesel Generator Building, Elevation 130
feet.
From a r iew of licens documents and obs ations noted du 4ng observations of
facility con ons (i.e., plawalk-downs), the injection team det mined th t a fire in
the selected fl areas prese d a significant cont ion to overall t risk _d
conditional core e probab
Docu rviewed by t m are'd in t attachment.
A01A Systems Required to Achieve and Maintain Post-Fire Safe Shutdown] _
Inspection Scope
The licensee's Safe Shutdown Analysis Report (SSAR) was reviewed to determine the
components and systems necessary to achieve and maintain safe shutdown conditions
in the event of fire in each of the selected fire areas. The objectives of this evaluation
were as follows:
Use (30
2
(a) Verify that the licensee's shutdown methodo gy has correctly identified
the components and systems necessary t achieve and maintain a safe
shutdown condition.
(b) Confirm the adequacy of the systems lected for reactivity control,
reactor coolant makeup, reactor heat emoval, process monitoring and
support system functions.
(c) Verify that a safe shutdown can be hieved and maintained without off-
site pow r, when it can be confirm that a postulated fire in any of the
selected Sire Areas could cause loss of off-site power.
(d) Verify that local manual operator actions are consistent with the plant's
fire protection licensing basis.
b. Jssuec-aid Findings
Clicensing Basis for Repair Activities (Opening/Closing of Links) Performed to Achieve
Safe Shutdown Condition.
(Introduction: The licensee's SSAR is based on as ring that a minimum set of
systems and equipment, that are capable of satisf ing the'
of Appendix R would be available inthe event of/ire in an) / at. &
This minimum set of systems and equipment ireferred to
shutdown path. Three specific paths for safe hutdown of'
Paths 1 or 2 would be used in the event of fi in areas thv
requirements of Appendix R Section III.G.2. Path 3 is an a E -'
and is used in the event of a significant fire in the control X
cable spreading room which forces operators to abandon i
fire damage or environmental (i.e., control room habitabilit
shutdown panels would be utilized for Path 3 shutdown.'Ni
for review during this inspection required this capability, ar
-reviewed during this inspection.
Q>Descrirtion: Systems required to perform the shutdown functions of reactor shutdown,
over pressure protection, maintenance of coolant inventory, and decay heat removal
have been identified for each path. The reactpr shutdown function is provided by the
reactor protection system (RPS) for all s.
Path 1 utilizes reactor core iso i ooling (RCIC), SRVs: WA'
removal (RHR) system in alternate shutdown cooling rr
inventory makeup, deca eat removal, and depressurizati
approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> nto the event, at which time the rea
the low-pressure coolant injection (LPCI) operability range
mitigate the impact of a spurious actuation of the automatic - _
(ADS) at a time when RHR system may not be available dul
licensee has assured that Core Spray (CS) would be availa' .
Path 2 utilizes the High Pressure Coolant Injection (HPCI),'
in the alternate shutdown cooling mode of operation. The HFui system and one SRV
are utilized during the first 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of a fire event to maintain the reactor water level and
14
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3
pressure within acceptable limits. After approximately 4 hoofs, the RHR system is
started in the alternate shutdown cooling mode of operotibn.
For the fire areas evaluated, the licensee idenfidthe
it ctures, systems and
components needed to achieve and mainta safe utdown conditions in the event of
fire. The team evaluated required manu opera r actions in order to verify that they
were consistent with the plant's fire pro ction icensing basis. Based on this evaluation
the team determined that the licensee es (manualoperator actions to open
terminal board links as a means of preventing an undesirid Qt -'
SRVs. (see section 1R21.01).
(:Analvsis: This finding is greater than min
reliability objectives and the equipment pel
cornerstone. Additionally, human factors 1 L/ Dr
I
to not successfully complete the task. The 7
1ic-V opening of terminal board links are consid*
potential safety significance greater than IC)/ 6 HeS 5 *
Enforcement: The licensee's current lice. -
Power request for exemption dated May 16 -- ri.'.t.ty cvaluation
Report (SER) dated January 2, 1987) characterized the opening of links as 'a repair
activity that is not permitted as a means of complying with Section 1Il.G of Appendix R.
Based on these documents the opening of links was consis ^ - ' -.
licensee and the NRC staff in 1987. The licensee could n
justify why these actions are not characterized as a repair'
/ . o ,, asC
In response to this inspection finding ,the licensee initiated
2003800152, dated 7/24/03) to evaluate actions to open Iir ,4CL d3 7
they are necessary to achieve hot shutdown, and if an exe! 4
required. This issue is identified as URI 50-366/03-06-01, 1
Activities (Opening/Closing-of-Linkspo-Achieve Safe ShutW
rerffains open pending review and acceptance 6f additional
documentation which demonstrates that actions necessary
dered a repair necessary to achieve and maintain hot'
.02 Fire Protection of Safe Shutdown Capability
a. Inspection Scope
For the selected fire areas, the team evaluated the frequency of fires or the potential for
fires, the combustible fire load characteristics and potential fire severity, the separation
of systems necessary to achieve safe shutdown (SSD), and the separation of electrical
components and circuits located within the' same fire area to ensure that at least one
SSD path was free of fire damage. The team also inspected the fire protection features
to confirm they were installed in accordance with the codes of record to satisfy the
applicable separation and design requirements of 10 CAR 50, Appendix R, Section IIL.G,
and Appendix A of BTP APCSB 9.5-1. The team reviewed the following documents,
which established the controls and practices to prevent fires and to control combustible
fire loads and ignition sources, to verify that the objectives established by the
NRC-approved fire protection program (FPP) were satisfied:
- A
2
SUMMA OF FINDINGS
IR 05000321/2003-00 000366/2003-006; _ U O t C;
- and 7/21-25/203 E. l. Hatch Nuclear Plant, Units 1 and 2; Triennial Fire
Protection
The report covered a two-week period of inspection by three regional inspectors and a
contractor from Brookhaven National Laboratory. Three Green non-cited violations (NCVs) and
four unresolved items with potential safety significance greater than Green were identified. 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-1 649, "Reactor Oversight Process," Revision 3, dated July 2000.
A. NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
ucZ The team identified an unresolved item in that a local manual operator action, to
prevent spurious opening of all eleven safety relief valves (SRVs) during a fire event,
-1
-V would not be performed in sufficient tiriine to be effective. Also, licensee reliance on this
manual action for hot shutdown during a fire, instead of physically protecting cables from
fire damage, had not been approved by the NRC.
This finding is unresolved pending completion of a significance determination. In
response to this pot tial issue, the licensee promptly moved the manual action step to
the front of the fire rocedure to enable operators to accomplish the action much
sooner during a fire'vent. This finding was determined to have potential safety
significance greater than very low significance because of the use of manual actions in
lieu of physical protection as required by 10 CFR 50 Appendix R, Section Ill.G.2.
(Section 1R05.05.b.1)
URI. The team identified an unresolved item in
cause all eleven SRVs to open at a time when ru
may not be available. To mitigate this event, the
report (SSAR) credits the use of Core Spray Lo
However, the licensee did not provide any object
or analysis) which demonstrated that, assumingl
2104, the limited set of equipment available wot
a manner that satisfies the shutdown performani
section L.1.e to 1OCFR 50.
This finding is unresolved pending completion ofl.... . _,.. Uf
I
record which demonstrates the capability of the Core Spray system to mitigate the
above event. This finding was determined to have potential safety significance greater
than very low significance because of a lack of a calculation of record and
3
documentation of the limited set of equipment that would be credited for safe shutdown
under these conditions. (Section 1R.05.03.b)
-rbO"d: The team identified an unresolved item in that the licensee's current fire'
protection licensing basis characterizes the opening of terminal board links in control
panels as a repair activity which is not permitted to achieve and maintain hot shutdown
conditions. The licensee could not provide any evidence to justify why these actions
were not characterized as a repair activity in its current SSAR. In response to this
inspection finding, the licensee initiated a Condition Report (CR 2003800152, dated
7/24/03) to evaluate actions to open links, in order to determine if they are necessary to
achieve hot shutdown, and if an exemption from Appendix R is required.
This finding is unresolved pending completion of a significance determination. This
finding is greater than minor because it impacts the mitigating system cornerstone and
has the potential for the operator not successfully completing the action because of
adverse human factor conditions. (Section 1R.05.01 .b)
< -Z -URI: The team identified an unresolved item in connection with the implementation of
design change request (DCR)91-134, SRV Backup Actuation via Pressure Transmitter
Signals. The installed plant modification failed to implement the one-out-of-two taken
twice logic that was specified as design input requirements in the design change
package. Additionally, implementation of a two-out-of-two coincident taken twice logic,
has introduced a potential common cause failure of all eleven SRVs because of fire
induced damage to two instrumentation circuit cables in close proximity to each other.
This finding is unresolved pending completion of a significance determination. This
finding is greater than minor because it impacts the mitigating system cornerstone. This
finding has the potential for defeating manual control of Group WK SRVs that are
required for ensuring that the suppression pool temperature will not exceed the heat
capacity temperature limit (HCTL) for the suppression pool. (Section 1R21.01)
6 O Green. The team identified a finding with very low safety significance in that a local
manual operator action to operate safe shutdown equipment was too difficult and was
also unsafe. The licensee had relied on this action instead of providing physical
protection of cables from fire damage or preplanning cold shutdown repairs. However,
the team judged that some operators would not be able to perform the action.
This finding involved a violation of 10 CFR 50, Appendix R,Section II.G.1 and
Technical Specification 5.4.1. The finding is greater than minor because it affected the
availability and reliability objectives and the equipment performance attribute of the
mitigating systems cornerstone. Since the licensee could have time to develop and
implement cold shutdown repairs to facilitate accomplishment of the action, this finding
did not have potential safety significance greater than very low safety significance.
(Section 1R05.05.b.2)
4
(2Green. The team identified a finding with very low safety sigrificance in that the
licensee relied on some manual operator actions to operate safe shutdown equipment,
instead of providing the required physical protection of cables from fire damage, and
awithout NRC approval.
This finding involved a violation of 10 CFR 50, Appendix R, Section lll.G.2. The finding
is greater than minor because it affected the availability and reliability objectives and thE
equipment performance attribute of the mitigating systems cornerstone. Since the
actions could reasonably be accomplished by operators in a timely manner, this finding
did not have potential safety significance greater than very low safety significance.
(Section 1R05.05.b.3)
) Green. The team identified a finding with very low safety significance in that emergency
a
lighting was not adequate for some manual operator actions that were needed to
support post-fire operation of safe shutdown equipment.
This finding involved a violation of 10 CFR 50, Appendix R, Section III.J. The finding is
greater than minor because it affected the reliability objective and the equipment
performance attribute of the mitigating systems cornerstone. Since operators would be
able to accomplish the actions with the use of flashlights, this finding did not have
potential safety significance greater than very low safety significance. (Section
1R05.07.b)
B. Licensee-Identified Violations
None-- --- _ _ -
_A
. r . 6
21
relays. The total of 12 relays described above, (6 in ATTS cabinet 2H1 1-P927 and 6 in
ATTS cabinet 2H1 1-P928), were intended to be wired to provide "one-out-of-two taken
twice logic" for actuation of the SRVs. The design objective was to assure that a single
relay failure in either Division would not cause an inadvertent SRV actuation.
Coincident logic input is required from both Division instrument loops in order to initiate a
SRV backup actuation via the pressure transmitter signals.
Analysis: The licensee in their SSAR takes credit for manual control of Group WA" SRVs
in order to achieve and maintain safe shutdown conditions. Manual control of Group "A"
SRVs are required for a fire in the fire areas selected for review.
The team performed a circuit analysis of SRV 2B21-F1 3F ( Path 1) and SRV 2B21-
F01 3G (Path 2) in order to verify that the design objectives of implementing a one-out-
of-two taken twice logic had been achieved. Based on this review the team determined
that the design objective of implementing a one-out-of-two taken twice logic had not
been installed for the SRVs. The logic installed for the SRVs was a two-out-of-two
coincident taken twice logic in addition to a one-out-of-two coincident taken twice logic.
The logic implemented results in spurious actuation of group "A" SRVs for a fire in fire
area 2104 and defeats the capability to manually control these SRVs as is required per
the SSAR.
Enforcement 10 CFR 50, Appendix B, Criterion ll, requires that design control
measures shall provide for verifying or checking the adequacy of design. The accepted
industry standard, ANSI N45.2.11-1974, section 4, requires design activities to provide
for relating the final design back to the source of design input.
The logic implemented by the licensee for DCR 91-134 was different from the specified
design input requirements. The plant installation failed to correctly implement the one-
out-of-two taken twice logic that was specified for the SRV backup actuation via
pressure transmitter signals design change package. This failure has created a
condition where fire induced failures of two instrument circuit cables, (within close
proximity to each other), could result in spurious actuation of all eleven SRVs with the
eleven SRVs assuming a stuck open mode of operation, based on the logic input from
trip master unit relays K31OD, and K335D and their associated trip unit slave relays.
The 10 CFR 50.59 Evaluation performed for the plant modification failed to identify this
failure mode. Additionally, the 10 CFR 50.59 Evaluation was inadequate in that it did
not provide an adequate technical basis that an Unreviewed Safety Question (USQ) had
not been created by implementation of the plant modification. Pending additional review
by the NRC, this item is identified as URI 50-366/03-06-06, Implementation of DCR 91-
134 Results in Spurious Actuation of Eleven SRVs because of Fire Induced Faults.
This inspection finding may be a" Potentially Generic Issue" by having implications for
other licensees who have implemented a plant modification similar to DCR 91-134 for a
BWR having a Mark 1 containment.
4. OTHER ACTIVITIES