ML20236C649
| ML20236C649 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 10/16/1987 |
| From: | Burger W, Rogger F, Schepens R, Sinkule M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20236C608 | List: |
| References | |
| 50-424-87-56, NUDOCS 8710270281 | |
| Download: ML20236C649 (12) | |
See also: IR 05000424/1987056
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UNITED STATES
/pn AtogD
NUCLEAR REGULATORY COMMISSION
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REGION 11
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101 MARIETTA STREET, N.W.
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ATLANT A, GEORGI A 30323
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'+4 * * * . * s
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Report No.:
50-424/87-56
Licensee:
Georgia Power Company
P. O. Box 4545
Atlanta, GA 30302
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Docket No.:
50-424
License No.:
Facility Name: Vogtle 1
Inspection Conducted: September 5 - October 7, 1987
4; A N
/0 N6
Inspecto
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. F. Rogge, Senior Resident Inspector
Date Signed
W4 A hal4
to/16 /27
b M . J. Schepens, Resident Inspector
Date Signed
f.edre: A8&
lO//6 /2 7
% W. Burg r, Repident Inspector
Date Signed
Approved by: 7
2
a/C-
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7
u-
M. V; Sinkule, Section Chief
Date' Sighed
Division of Reactor Projects
SUMMARY
Scope:
This routine, unannounced inspection entailed resident inspection in
the following areas:
plant operations, radiological controls, maintenance,
surveillance, fire protection, security, quality programs and administrative
controls affecting quality, and follow-up on previous inspector identified
items.
Results:
Two violations were identified - failure to maintain the residual
heat removal system operable and failure to adequately report the event.
8710270281 871016
DR
ADOCK 05000424
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REPORT DETAILS
1.-
Persons Contacted
Licensee Employees
- G. Bockhold,- Jr., General Manager, Nuclear Operations
- E. M. Dannemiller, Technical Assistant to General Manager
- T. V. Greene, Plant Support Manager
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- R. M. Bellamy, Plant Manager
- W. F. Kitchens, Manager Operations
J. E. Swartzwelder, Deputy Manager - Operations
- R. E. Lide, Engineering Support Supervisor
C. W. Hayes, Vogtle Quality Assurance Manager
G. R. Frederick, Quality Assurance Site Manager - Operations
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C. E. Belflower, Quality Assurance Assessment Manager
-W. E. Mundy, Quality Assurance Audit Supervisor
M. A. Griffis, Maintenance Superintendent
- J. F. D'Amico, Nuclear Safety and Compliance Manager
R. M. Odom, Plant Engineering Supervisor
- C. L. Cross, Senior Regulatory Specialist
W. C. Gabbard, Regulatory Specialist
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S. F. Goff, Regulatory Specialist
- A. L. Mosbaugh, Assistant Plant Support Manager
H. M. Handfinger, Assistant Plant Support Manager
M. P. Craven, Nuclear Security Manager
R. E. Spinnatu, ISEG Supervisor
- C. C. Echert, Technical Assistant to Plant Manager
- P. A. Herrmann, Senior Nuclear Engineer, Corporate
Other licensee employees contacted included craftsmen, technicians,
supervision, engineers, operations, maintenance, chemistry, inspectors,
and office personnel.
- Attended Exit Interview
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2.
Exit Interviews (30703)
The inspection scope and findings were summarized on October 7, 1987, with
those persons indicated in paragraph 1 above. The inspector described the
areas inspected and discussed in detail the inspection results.
No
dissenting comments were received from the licensee. The licensee did not
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identify as proprietary any of the materials provided to or reviewed by
the inspector during this inspection.
Region based NRC exit interviews
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were-attended during the inspection period by a resident inspector.
The
inspection closed one unresolved item.
The items identified during this
inspection were:
Violation 50-424/87-56-01, " Failure To Maintain The Residual Heat
Removal System Operable per TS 3.5.2" - Paragraph 6.b.(1)
Violation 50-424/87-56-02, " Failure To Adequately Report the Residual
Heat Removal System Inoperable Event" - Paragraph 6.b.(2)
3.
Licensee Action on Previous Enforcement Matters (92702)
This area was not inspected.
4.
Unresolved Items (92701)
Unresolved items are matters about which more information is required to
determine whether they are acceptable or may involve violations or
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deviations.
No new unresolved items we*e identified.
One previous item
is discussed in Paragraph 6.b.(1).
5,
Operational Safety Verification (71707) (93702)
The plant began this inspection period in Power Operation (Mode 1) at 100%
power until October 2 when power was reduced to 75% due to secondary
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chemistry problems. The unit returned to 100% power the same day following
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repairs to a leaking condenser tube.
The plant experienced one ESF
actuation of the Control Room Emergency Ventilation System on
September 21, 1987.
a.
Control Room Activities
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Control Room tours and observations were performed to verify that
facility operations were being safely conducted within regulatory
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requirements.
These inspections consisted of one or more of the
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following attributes as appropriate at the time'of the inspection.
Proper Control Room staffing
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Control Room access and operator behavior
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Adherence to approved procedures for activities in progress
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Adherence to Technical Specification (TS) Limiting Conditions
for Operations (LCO)
Observance of instruments and recorder traces of safety related
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and important to safety systems for abnormalities
Review of annunciators alarmed and action in progress to correct
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Control Board walkdowns
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Safety parameter display and the plant safety monitoring system
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-operability status
Discu',sions and interviews with the On-Shift Operations
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Supervisor, Shif t Supervisor, Reactor Operators, and the Shift
Technical Advisor to determine the plant status, plans and
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assess operator knowledge
Review of the operator logs, unit log and shift turnover sheets
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No violations or deviations were identified,
b.-
Facility Activities
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Facility tours and observations were performed to assess the
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effectiveness of the administrative controls established by direct
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observation of plant activities, interviews and discussions with
licensee personnel, independent verification of safety systems status
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and LCOs, licensee meetings and facility records.
During these
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inspections.the following objectives are achieved:
(1) Safety System Status (71710)- Confirmation of system
operability was obtained by verification that flowpath valve
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alignment, control and power supply alignments, component
conditions, and support systems for the accessible portions of
the ESF trains were proper.
The inaccessible portions are
confirmed as availability permits.
Additional indepth
inspection of the Auxiliary Feedwater System (AFW) was performed
to review the system lineup procedure with the plant drawings
and as-built configurations, compare valve remote and local
indications, and walkdown of hangers, supports, snubbers and
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electrical equipment interiors. The inspector verified that the
lineup was in accordance with license requirements for system
operability.
The AFW system received additional inspection when
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one pump could not pass the flow requirements during
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surveillance and again when the building sump overflowed and
backed up into all three pump rooms.
Corrective action for the
first event involved replacing the orifice in the pump miniflow
line and conducting the surveillance with a portion of the flow
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into the Steam Generators.
The second event required the
cleaning of all three floor drain lines and the removal of
foreign debris from the check valves.
(2) Plant Housekeeping Conditions - Storage of material and
components and cleanliness conditions of various areas
throughout the facility were observed to determine whether
safety and/or fire hazards existed.
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(3) Fire Protection - Fire protection activities, staffing and
equipment were observed to verify that fire, brigade staffing was
appropriate ' and that fire alarms, extinguishing equipment,
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actuating controls,
fire fighting equipment, emergency
-equipment, and fire barriers were operable.
(4) Radiation Protection (71709) - Radiation' protection activities,
staffing and equipment .wert. observed = to verify proper program
implementation.- The inspection included review of the plant
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program effectiveness.
Radiation work permits and personnel
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cunpliance were reviewed during' the daily
tours.
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Observations of Radiation Control Areas (RCAs) plant
were conducted to.
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verify proper identification and implementation.
The inspector
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. discussed with the licensee previous issues which have resulted
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in an NRC view that this programmatic- area needs additional
management at'tention (staffing, experience, high rad area
control,etc.).
(5) Security (71881) - Security controls -were observed to verify
that security barriers were intact, guard forces were on duty,
and access to the Protected Area (PA) was controlled in
accordance with the facility security plan.
Personnel within
the PA were observed to verify that badges were properly
displayed and that personnel requiring escort were properly
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escorted.
Personnel within vital areas were observed to ensure
proper authorization for the area.
Equipment operability and
proper compensatory activities were verified on a periodic
basis.
(6) Surveillance (61726) (61700) - Surveillance tests were observed
to verify that approved procedures were being used; qualified
personnel were conducting the tests; tests were adequate to
verify equipment operability; calibrated equipment was utilized;
and TS requirements were followed.
The inspectors observed
portions of the surveillance
and reviewed completed data
against acceptarce criteria.
(7) Maintenance Activities (62703) - The inspector observed
maintenance activities to verify that correct equipment
clearances were in effect; work requests and fire prevention
work permits, as required, were issued and being followed;
quality control personnel were available for inspection
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activities as required; retesting and return of systems to
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service was prompt and correct; and TS requirements were being
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followed. Maintenance backlog was reviewed.
No violations or deviations were identified.
6.
Review of Licensee Reports (90712) (90713) (92700)
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a.
In-Office Review of Periodic and Special Reports
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This inspection consists of reviewing the below listed reports to
determine whether the information reported by the licensee is
technically adequate and consistent with the inspector knowledge of
the material contained within the report.
Selected material within
the report is questioned randomly to verify accuracy to provide a
reasonable assurance that NRC personnel and/or other users have an
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appropriate document for their activities.
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Monthly Operating Reports - The report dated September 11, 1987, was
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reviewed.
The' inspector had no significant comments regarding these
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reports.
b.
Licensee Event Reports (LERs) and Deficiency Cards (DCs)
Licensee Event Reports (LERs) and Deficiency Cards (DCs) were
reviewed for potential generic impact, to detect trends, and to
determine whether corrective actions appeared appropriate.
Events
which were reported pursuant to 10 CFR 50.72, were reviewed as they
occurred to determine if the technical specifications and other
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regulatory requirements were satisfied.
In-office review of LERs may
result in further followup to verify that the stated corrective
actions have been completed, or to identify violations in addition to
those described in the LER.
Each LER is reviewed for enforcement
action in accordance with 10 CFR Part 2, Appendix C.
Review of DCs
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was performed to maintain a realtime status of deficiencies,
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determine regulatory compliance, follow the licensee corrective
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actions, and assist as a basis for closure of the LER when reviewed.
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Due to the numerous DCs processed only those DCs which result in
enforcement action or further inspector followup with the licensee at
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the end of the inspection are discussed as listed below.
The items
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marked with an asterisk indicated that a reactive inspection occ'urred
at the time of the event prior to receipt of the written report.
(1) Deficiency Card reviews:
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DC 1-87-2018,1-87-2057 and 1-87-2241 - These deficiency cards
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documented a problem where the plant was placed in an unanalyzed
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condition
outside the established design basis when the RHR
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crossover isolation valves (1HV-8716 A & B) were closed on
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several occasions for surveillance testing.
With either of
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these valves closed the RHR subsystems are not capable of
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injection into all four RCS loops per the ECCS analysis.
This
condition was described in NRC IE Intermation Notice 87-01 a'ated
January 6,1987, which the licensee had received on site on
January 12, 1987.
The licensee formed an event critique task
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force on August 31, 1987, per their administrative procedures to
conduct a root cause determination of why this information was
not effectively processed to prevent the condition.
The task
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force reviewed plant procedures to determine a complete account
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of when and for how long the plant was in an unanalyzed
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condition. The task force identified six (6) times when the RHR
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crossover isolation valves were closed.
Three times were for.
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valve stroking on 4/3/87, 7/3/87, and 8/8/87 per the quarterly
inservice valve test surveillance. procedure 14825-l' and the
other three times were for RHR pump testing on'2/24/87, 4/16/87,
and. 7/20/87_ per the RHR pump and check valve inservice test
surveillance procedure 14805-1. The licensee estimates that the
valves were closed for a maximum time of 3 minutes during the
performance of the valve stroking surveillance and review of the
maximum ~ time which they were closed during the RHR -pump
inservice testing surveillance indicates that one surveillance
on'7/20/87 degraded the RHR system 72 minutes.
The task force
also identified that the RHR cold leg isolation valves
(1HV-8809A & B) were closed on 3/18/87 during the performance of
the RCS pressure isolation valve leak test surveillance
procedure 14450-1 which was performed in Mode 3 for a 12 minute-
duration. The licensee's task force has completed the review of
these events to determine the duration for which these valves
were closed. The task force also conducted a complete review of-
all maintenance work orders and clearances since initial fuel
which did not identify any other instances were these valves
were closed. The licensee's immediate corrective action has
consisted of but was not limited to the following:
1) surveillance procedures 14805-1 and 14825-1 have been revised
to no longer require closing of the RHR crossover isolation
valves in Modes 1, 2, or 3; 2) a night order has been issued to
the shifts describing the concern with closing either the RHR
crossover or cold leg isolation valves and a copy of NRC IE
Information Notice 87-01 has been placed in the . operations
required reading book; and 3) Administrative' Controls have been
placed on the RHR crossover and cold leg isolation valves by
hanging a "for information" tag on the handswitches on the main
control board -instructing operators not to close these valves
when in Modes 1, 2, or 3 and that if any of these valves are
found to be closed then they should be opened immediately or if
they can not be opened then comply with Technical Specification 3.0.3.
The inspection included the attendance at the PRB
Meeting where the event critique team presented the event review
report and root cause determination.
The inspection determined
that the two most-important corrective actions are improvements
to the Operating Experience Program (0EP) and to the Inservice
Test program (IST).
The OEP program will be revised to include
an operational impact review and immediate compensatory action
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prior to normal program changes and the IST program changes will
include procedures for making 10 CFR 50.59 changes.
The method utilized for testing the RHR pumps and check valves
as specified in Revision 1 of surveillance procedure 14805-1
requires the closure of the RHR crossover isolation valve for
the opposite train being tested as well as the closure of the
RHR heat exchanger outlet valve for the train being tested. The
RHR pump and check valve performance is then demonstrated by
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opening 'the- RHR heat exchanger bypass valve and recirculating
flow ba'ck to the RWST via two no.rmally locked closed manual
isolation valves ~(1-1205-U4-027 - and 1-1205-U4-226). . The
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inspector' determined that this test method-in effect renders the.
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low head safety injection flow path inoperable as defined in TS
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3.5.2 in that the train being . tested would have its flow
diverted to the RWST and the other train would only be. capable
of-injecting into two of the four loops.
This test method was
conducted during the following times.
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Date'
RHR-Pump
Mode
Power
Duration
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02/24/87*
A
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53 Minutes
02/24/87*
B
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29 Minutes
04/16/87**
A
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48%
46 Minutes
04/16/87**
B'
1
48%
39 Minutes
07/20/87
A
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100%
72 Minutes
07/20/87
A
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100%
34 Minutes
- Initial Criticality Achieved on 3/9/87
- The NRC I'nspection determined that the worst case f(r RHR System
unavailability was 85 Minutes since the RWST manual isolation
valves were not closed between pump tests.
During the' inspection the inspector also noted that Westinghouse
letter No. GAE-4704 dated' February 27, 1987, communicated
Westinghouse's position regar#ng IE Information Notice 87-01 -
RHR Valve Misalignment Causes Degradation of ECCS in PWRs.
This
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letter informed the licensee that every Westinghouse plant ECCS
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Analysis assumes low head safety injection. into all cold legs.
Therefore, the isolation of RHR flow to any of the reactor
coolant system cold legs to allow surveillance testing wil.1
place a plant in an unanalyzed condition outside its established
design basis.
Furthermore, the inspector noted that FSAR
paragraph 6.3.2.2.4 describes the RHR pumps and states that the
minimum flow bypass. lines prevent deadheading of the pumps an_d
permit pump testing during normal operation.
FSAR paragraph
6.3.2.7 describes the provisions for performance testing and
states that all pumps have miniflow lines for us.e in testing
operability.
FSAR 6.3,4.2 describes reliability tests and
inspections and states:
1) that the design measures ensure that
active components may be tested periodically for operability
(e.g., pumps on miniflow); and 2) that the design features which
ensure this test capability are met.
Specifically, RHR pumps
can also be tested periodically when the plant is at power using
the miniflow recirculation lines.
FSAR figure 6.3.2-2 sheet 4
of '21 provides a valve alignment chart for the three modes of
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ECCS and in part shows the following valve positions.
__-__a
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. Cold Leg-
Hot Leg
. Valve
Injection _
Recirculation
Recirculation
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HV-8809A-
Open
Open
' Closed
HV-8809B
Open
Open
Closed
D
HV-8716A
Open
. Closed.
Open
HV-87168-
.0 pen
Closed
Closed
HV-0606
'Open
.0 pen
Open
t:
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HV-0607'
0 pen
'Open
Open
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. Contrary to. the above information, the RHR. system was rendered
inoperable by the surveillance procedure 14805-1 in that both
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. trains of RHR were impacted due -to the valve lineup and test
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. method which resulted in eliminating the four loop injection
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that each RHR subsystem must have to be considered operable.
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This surveillance was performed on several occasions subsequent
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to:the licensee receiving the IE Information Notice and the
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Westinghouse letter which serves as notification of the
potential problem and that the plant would be placed outside the
design-basis.
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The NRC inspection determined that the licensee's task force did
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not -identify the significance of the inadequate test method
utilized in the abnve pump test events which recirculated RHR
flow back to the RWST in lieu of recirculating the' flow thru'the
miniflow line as specified in the FSAR.
In addition, plant
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personne1' utilized a surveillance procedure from another utility
as the basis during their procedure development.
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The quarterly inservice valve test procedure No. 14825-1
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requires the stroking of the :RHR crossover isolation valves
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(1-HV-8716A & B) to demonstrate operability in Modes 1, 2, 3 and
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This surveillance was conducted on the following times.
Date
Mode
Power
Duration
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4/3/87
1
35%
3 Minutes
7/3/87
1
100%
3 Minutes
8/8/87
1
90%
3 Minutes
The inspector determined that valve stroking would not be
allowable by the technical specification because they cause each
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RHR subsystem to be, incapable of injection into all four cold
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were to occur (failure of an RHR pump)g as no active failures
legs.
The inspector notes that as lon
that this would have no
affect on the ability to perform the safety function.
In fact,
this stroke test is a desirable test since this valve is
required to be repositioned during the course of an accident and
quarterly stroking would provide reasonable assurance of
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operability.
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The ' RCS pressure isolation valve leak test surveillance
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procedure 14450-1 Rev. O required the closing of the RHR cold
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' leg- isolation valves (1-HV-8809A & B). to . demonstrate that
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leakage of the RCS pressure isolation valves is within limits of
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TS - 4.4.6.2.2.
The. licensee estimated 'that. these valves were
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closed for _ approximately 12 minutes each during the performance -
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of the . surveillance test on March 18, 1987.
The inspector
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. determined that this test placed the plant _ outside the design
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basis in that it' prevented either RHR pump from being able to
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inject into all four loops.
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The inspector ' determined that' the failure to implement prompt
corrective . action. following notification by the NRC resulted in
-three situations of inoperability of the Residual Heat Removal'
System. (TS 3.5.2) for various periods of time ranging from 3
minutes to as long as 85 minutes.
Each occurrence places the
plant in the condition whereby TS 3.0.3 is applicable. TS 3.0.3
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cannot be voluntarily utilized for maintenance or testing
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purposes and had it .been entered it would- have necessitated
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action within one hour to place the unit in hot standby within
six. hours. - The licensee did not take any' action for the two
events on .4/16/87 and 7/20/87 which exceed the one hour time
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frame. This violation is identified as 50-424/87-56-01 " Failure
To Maintain The Residual Heat Removal System Operable per TS 3.5.2".
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(Closed) Unresolved Item 50-424/87-49-01, " Review Licensee
Evaluation.and Corrective Action Regarding the Closure of either
the RHR Cold Leg Isolation (1HV-8809A & B) or Crossover
Isolation (1HV-8716A & B) Valves which Renders the System
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This item is closed based on the above
identification of a violation.
(2) The following LERs were reviewed and are ready for closure
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pending'. verification that the licensee's stated corrective
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actions have been completed.
- 50-424/87-44, Rev.1, " Control Room HVAC Design Violates Single
Failure Criteria".
This event involves two conditions which
could have potentially prevented the control room pressure from
. reaching the emergency mode design value since the introduction
rate of outside air would have been lower than postulated in the
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analysis.
Backdraft dampers were installed to correct one
condition and the inlet dampers were deenergized to prevent the
second condition.
Final corrective action is under review and
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will be the subject of a supplemental LER.
No enforcement
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action is warranted since Region II granted discretionary
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enforcement action.
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- 50-424/87-50, Rev. O,
" Reactor Trip Caused By Instrument
Technician's Error".
This event involves a reactor trip from
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100% power on a turbine trip signal which was a result of a low
hydraulic fluid pressure.
The low pressure resulted when a
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technician improperly connected the test equipment across the
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generator current transformer and produced a false momentary
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power load unbalance signal, which initiated closure of the
turbine control and intermediate stop valves.
The licensee
corrective action was to counsel technicians and engineers
regarding the need to accurately assess the potential impact of
maintenance activities that are performed on their system.
No
violations were identified during or as a result of this event.
- 50-424/87-48, Rev. O, " Local Leak Rate Test Not Performed Within
TS Time Limit".
This event regards the failure to perform the
appropriate surveillance after maintenance to demonstrate
containment integrity.
The licensee discovered the June 18,
1987 event on July 14, 1987, during a review of work orders.
The test was performed and successfully completed.
Following
the maintenance a seal leakage test was performed; however, the
TS requires an overall air lock test at 45 psig.
The licensee
counseled the supervisors and is revising the work order process
to ensure early identification of maintenance work orders which
effect TS surveillance requirements. This item is identified as
a licensee identified violation (LIV) which meets the criteria
for not issuing a Notice of Violation and will be identified as:
50-424/ LIV 8756-01, "LER-87-48 - Failure to Perform TS 4.6.1.3b
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Surveillance on the Containment Air Lock Following Maintenance".
- 50-424/87-51, Rev. 0, "AFW Flow Transmitters Inoperable Due To
Inadequate Instructions and Personnel Errors".
This issue was
reviewed in NRC report 50-424/87-49 and resulted in identifying
a Licensee Identified Violation. The LER was reviewed and found
to be consistent with the finding in the previous inspection.
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Implementation of the corrective actions remains to be
completed. This item will receive further followup.
- 50-424/87-55, Rev. O, " Closure of RHR System Valves Causes Loss
of Availability of One RHR Pump". This LER was submitted to the
NRC on 10/5/87 entitled " Closure of RHR System Valves Causes
Plant to Operate Outside its Design Bases". The LER states the
report to be voluntary and only discusses the closure of
1-HV-8809A & B events of March 18, 1987.
The inspector
questioned why the event was not submitted pursuant to
50.73(a)(2)(ii)(A) or (B) criteria which requires a report for
any event or condition that resulted in the nuclear power plant
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being:
(a) in an unanalyzed condition that significantly
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compromised plant safety; or (b) in a condition that was outside
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the design basis of the plant.
The licensee response was that
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they had spent many hours on this issue and felt that it did not
fit the criteria.
The other events were not included because
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they were not as significant or interesting.
Based on the
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conclusions as stated in arriving at Violation 50-424/87-56-01
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above the inspector has concluded that a failure to report all
the. events;where the plant was in a unanalyzed condition
constituted..a separate violation.
This violation is identified
- as 50-424/r -56-02, " Failure To Adequately Report the Residual
Heat Removal = System Inoperable Event".
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7.
Management Meetirs (303028)
On September 10,_1987, a meeting was conducted at the site-.to discuss the
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plant status and served as a followup meeting to previous meetings. The
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focus of the . meeting was directed at the licensee's progress regarding
reactor . trip reduction, security improvements, and other areas regarding
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operational performance.
The meeting included a tour of the security
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barriers that had recently been found deficient.
The NRC requested that
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another meeting be planned for November.
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On September 15, 1987, a meeting was conducted at the GPC Corporate office
-in Atlanta, GA to discuss .the status of licensing and inspection items on
the'Vogtle and Hatch facilities.
8.
IE Bulletin Followup _ (92703)
~(Closed) 50-424 & 50-425/86-BU-03, " Potential Failure of Multiple -
Emergency Core ~ Cooling System (ECCS) Pumps Due to Single Failure of Air
Operated ' Valve in Minimum Flow' Recirculation Line".
The licensee
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responded to this bulletin in a letter dated November 11, 1986, and stated
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that they had evaluated the bulletin to determine whether a single-failure
vulnerability exists in the minimum flow recirculation line of any ECCS
pump that could cause a failure of more than one ECCS train.
Their
- evaluation has concluded that the concerns-'of the bulletin are not
applicable to this facility.
The only ECCS components that share common
minimum flow recirculation lines and/or valves are the Safety Injection
System (SIS) and the Chemical Volume and Control System (CVCS).
The SIS
has miniflow return lines to the Refueling Water Storage Tank (RWST).
The
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lines become common after the pressure breakdown orifices for each pump.
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The individual portions of the miniflow lines contain an isolation valve
each, and the common portion contains a single valve. All of these valves
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are motor operated and as such are fail-as-is upon loss of control
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circuitry or power.
The CVCS minimum recirculation normal flow paths
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connect to a common line that contains a single motor operated isolation
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valve that is also-fail-as-is upon loss of power.
On a safety injection
signal, the CVCS pump miniflow paths automatically align to alternate
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- miniflow paths which have no common valves.
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Based on an in-office review by the Region II inspectors, this item is
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closed. Closure for Unit 2 is via report 50-424/87-39.
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