ML20044F796
| ML20044F796 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 05/24/1993 |
| From: | Burdick T, Pegg W, Jacqwan Walker NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML18058B833 | List: |
| References | |
| 50-255-93-10, NUDOCS 9306010060 | |
| Download: ML20044F796 (11) | |
See also: IR 05000255/1993010
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U. S. NUCLEAR REGULATORY COMMISSION
REGION III
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Report No. 50-255/93010(DRS)
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Docket No. 50-255
License No. DPR-20
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ticensee:
Consumers Power Company
Palisades Nuclear Generating Plant
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27780 Blue Star Memorial Highway
Covert, MI 49043-9530
Facility Name:
Palisades Nuclear Plant
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Inspection At:
Covert, Michigan
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Inspection Conducted: April 6 through May 4, 1993
Inspectors:
fcrt
f/2#/f7
J. Walker
Date
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T
W
DDilf3
W. Pegg
Date /
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Approved By:
[
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T. Burdick, Chief
Date
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Operator Licensing Section 2
Inspection Summarv
Lnstection on April 6 throuah May 4.1993 (Report No. 50-255/93010(DRS))
Areas Inspected:
Special inspection to assess the implementation of
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commitments made in response to the 1986 SER and to assess the operability and
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availability of the MSIVs.
Results: Within the areas inspected, one deviation was identified regarding a
conflict between the E0Ps dealing with a main steam line break (MSLB)
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coincident with a single failure of the intact steam generator's MSIV to close
and commitments made to the NRC (Section 2.4.6.1).
Open items were identified
regarding differences between the training of operators and the
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recommendations of the SER on the possible loss of instrumentation inside of
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containment during this event (Section 2.4.6.2); a commitment to include as-
found stroke time testing of the MSIVs in the operational procedures (Section
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2.4.4); operator training on loss of instrumentation and procedural changes
related to this event were not developed; graphs to account for errors in
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pressurizer and SG narrow range level were referred to in footnotes in the
safety status, but no guidance is given as to when these graphs are required
(Section 2.4.7.1); and discrepancies between the containment pressure and
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temperature values observed during simulator runs of the MSLB event and the
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expected values (Section 2.4.7.3).
No problems with the maintenance or
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operability of the MSIVs were noted.
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9306010060 930524:
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ADOCK 05000255
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DETAILS
1.0
Exit Meetina Attendees
Consumers Power Company
T. Palmisano, Operations Manager
R. Orosz, Manager, Nuclear Engineering and Construction
J. Hanson, Operations Superintendent
R. Heimsath, Training Administrator
J. Kuemin, Licensing Administrator
R. Kasper, Maintenance Manager
R. Gerling, Reactor and Safety Analysis Manager
D. Malone, Operations Staff Support Supervisor
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F. Yanik, Risk Assessment Supervisor
B. Kubacki, System Engineering Section Head
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E. Letke, Main Steam System Engineer
U. S. Nuclear Reaulatory Commission (NRC)
T. Burdick, Chief, Operator Licensing Section 2
J. Heller, Senior Resident Inspector
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The NRC inspectors also contacted and interviewed other licensee
personnel during the inspection.
2.0
Backaround Information
2.1
Purpose of the Inspection
The purpose of the inspection was to assess implementation of commitments made
in response to the 1986 SER and to assess the operability and availability of
the MSIVs through an examination of the MSIV maintenance and operational
history.
2.2
History
In the early 1980's, it was discovered that the design of the main steam
system did not preclude the possibility that a MSLB inside containment
concurrent with a failure of the intact steam generator's (SG) MSIV could
occur.
The steam system at Palisades relies on reverse mounted check valves
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as MSIVs. There are no non-return check valves downstream of the MSIVs to
prevent blow-down of both steam generators in the event of a steam line break
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inside containment concurrent with a passive failure of the opposite steam
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generator's MSIV to close.
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In 1983, the licensee committed to make modifications to either replace the
MSIVs with motor-operated valves, install non-return check valves downstream
of the MSIVs to preclude backflow from one steam generator into the other on a
MSLB, or qualify the instrumentation and equipment inside of containment for
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the bounding conditions of the dual steam generator blowdown.
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In 1985, the licensee submitted a PRA to the NP,C to show that the proposed
modifications were not cost effective considering the probability of the event
occurring and the cost of the improvements relative to the decrease in overall
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risk for each of the proposed modifications.
On February 28, 1986, the NRC issued an SER on this issue concurring with the
licensee's evaluation provided alternative measures were implemented. These
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alternative measures, the licensee's response to them and the NRC's evaluation
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of the response are provided in Section 2.4.6 of this report.
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2.3
Valve Desion
The Palisades MSIVs are 30-inch Atwood-Morrill swing disc check valves that
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are reverse mounted and utilize an external air cylinder and piston connected
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to the valve disc shaft to hold the valves open. To close the MSIVs, a
solenoid is actuated which vents the air from under the piston.
This allows a
compressed spring above the piston to expand and force the piston down,
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assisting the valve in closing.
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2.4
Inspection Findinos
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2.4.1
Evaluation of Industry Exoerience
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On September 29, 1991, while in the prccess of shutting down, the MSIVs at the
Point Beach Nuclear Plant, Unit 1, fail =J to close due to corrosion problems
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in the pneumatic operators. On November 15, 1991, the licensee began a review
of this industry experience and determined that the Palisades plant had eight
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Atwood-Morrill check valves of a similar design, including the MSIVs.
Prior
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to evaluation of the MSIV failure's applicability to Palisades, the licensee's
document tracking the industry experience was lost due a computer entry error
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and a failure of the responsible review group to return the evaluation to the
originator. The review was started again on March 11, 1993, after NRC
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operations inspectors made inquiries about the MSIVs and the event at Point
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Beach. Consequently, the licensee did not complete an evaluation of the
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applicability of the Point Beach experience to Palisades in a time frame
commensurate with the safety significance. The control exhibited over this
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review is considered a weakness.
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2.4.2
MSIV Operational History
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The inspectors reviewed licensee event reports, abnormal occurrence reports,
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and interviewed operations personnel to determine if the MSIVs had ever failed
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to close on demand. The inspectors determined that the MSIVs had failed to
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close three times in the period between 1972 and 1973. There have been no
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failures to close since 1973. The actions taken in response to the initial
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failures were successful in preventing recurrence.
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2.4.3
MSIV Maintenance Historv
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The inspectors reviewed work orders from the past ten years, with emphasis on
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maintenance activities over the past three years.
In Work Order 24104695, the
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pneumatic operator of CV-0510 was disassembled and inspected per Procedure
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MSS-M-19, " Disassembly, Inspection and Reassembly of Main Steam Isolation
Valves CV-0501 and CV-0510", Revision 5.
The inspection of the actuator
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internals revealed some spots of corrosion, scratches and other normal wear.
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The degree of wear and degradation discovered would not have affected the
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ability of the actuator to perform it's function. The licensee will
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disassemble and inspect the pneumatic actuator for CV-0501 during the next
cold shutdown. Although provisions are made in this procedure to perform
disassembly and inspection of the pneumatic actuator (step 5.3.9), step 3.8.4
allows the system engineer to bypass this step of the procedure. This step in
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the procedure had been bypassed since the installation of the actuators in
1979. The length of time between the installation of the pneumatic actuators
in 1979 and the first disassembly and inspection of the actuators in 1992.is
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excessive considering the importance of the actuators and is a weakness.
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The inspectors were concerned about the use of Q Sealant, Furmanite, or other
temporary means to repair stuffing box and packing leaks without the benefit
of as-found testing to determine if such activities unduly affected the
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performance of the HSIVs. As-found testing would provide necessary feedback
on the impact of such repairs on the shaft friction and the ability of the
actuator to overcome the increased friction.
The inspectors found the maintenance procedures to be clear and complete.
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inspectors interviewed maintenance personnel and determined that the
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maintenance personnel were qualified and cognizant of the maintenance
requirements of the MSIVs. No other problems were noted.
2.4.4
MSIV Testina
The inspectors reviewed the technical specification (TS) surveillance history
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of the MSIVs for the past five years. The MSIVs closed within the TS
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allowable times and conditions.
Procedure RI-17, " Main Steam Isolation Velve
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Circuits Test and Valve Closure Timing", Revision 12, documented the required
TS closure time testing. The required testing, however, did not accurately
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reflect the ability of the MSIVs to perform their safety function during the
previous operational cycle, as the MSIVs were tested on power ascension after
they had already been exercised, repaired or conditioned.
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Prior to this inspection, the licensee had documented closure of the MSIVs on
a shut down check list, but no measurement of the closure time was taken.
The
licensee has since committed to include as-found stroke time testing of the
MSIVs in the operational procedures. The results of this as-found testing
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will indicate if there was any degradation in the ability of the MSIVs to
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perform their function over the previous operational cycle. The commitment to
include as-found stroke time testing of the MSIVs in the operation procedures
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will be tracked as an open item (255/93010-Ol(DRS)).
2.4.5
Dual SG Blowdown
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The following is a brief synopsis of what would happen in a dual SG blowdown,
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based on the PRA and simulator runs of this scenario on the Palisades
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certified plant specific simulator. The use of the designators "A"
and "B"
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are arbitrary and are provided for clarity.
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A guillotine break occurs on the B Main Steam line, upstream of
the MSIV.
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The MSIVs receive a low steam generator pressure signal to close.
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The MSIV for the A steam generator fails to close and remains in
the full open position.
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Backflow from the A steam generator pushes the MSIV for the B
steam generator open and allows the A steam generator to blow down
through the break.
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Both steam generators blow down into containment.
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The primary coolant system (PCS) cools down to approximately 350*F
and depressurizes to approximately 1000 psia. All primary coolant
pumps (PCPs) are stopped.
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Upon dry out of the "Most Affected" steam generator the PCS begins
to reheat and repressurize. The saturation temperature in the
"Least Affected" steam generator is initially above PCS
temperature preventing it from acting as a heat sink and
inhibiting natural circulation.
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Voids are formed in the upper head region of the reactor vessel,
again inhibiting natural circulation flow.
2.4.6
Licensee Response to SER
2.4.6.1
Operator Actions
The NRC concern, documented in the SER issued in 1986, was as follows:
"The emergency procedures dealing with secondary line breaks, E0P-6 and
E0P-7, and the procedure for normal reactor trip, E0P-1, do not provide
definitive guidance regarding maintenance of a heat sink (use of steam
generator level or other secondary parameter for feedback) even if the
operator perceives that both steam generators may be faulted. In fact,
in E0P-1 it is noted that if dry out occurs, the affected steam
generator is to be considered inoperable. Additionally, the absence in
the procedures of any recognition of overcooling expected to occur in
these events enhances the potent 51 for inappropriate spontaneous
operator action. Procedures based on systems analysis should be
developed to enable the operator to cope with these events and to
achieve a controlled cooldown. The staff considers that analyses are
needed to identify a control strategy that considers the decay heat
generation available to the operator and how the response changes over
the course of the cooldown."
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The licensee's response to this concern, documented in a letter to the NRC
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dated April 28, 1986, was as follows
"All new E0Ps provide definitive guidance regarding maintenance of the
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heat sink.
This includes monitoring of specific parameters including
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wide range steam generator levels, core temperature indications, steam
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generator pressures and steam generator feed flow. All new E0Ps have
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safety function status checks which include the safety function of core
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heat removal and PCS heat removal."
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"In all new E0Ps, emphasis is placed on continuous feed to at least one
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steam generator, even if both SGs are " dried out", until shutdown
cooling or long term-post PCS break-core cooling is sufficient for decay
heat removal ."
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"It should also be emphasized that in all new E0Ps, confirmation of
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adequate core heat removal requires the confirmatory indication of at
least in-core temperature.
In the Functional Recovery Procedure, if in-
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containment indications are considered lost, the continuing actions for
PCS Heat Removal and Core Heat Removal include maximizing feed flow to
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available steam generators."
"The new E0Ps are structured such that, if in-containment vital
instrumentation is considered lost then the functional Recovery
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Procedure is implemented.
If a decision is made to not rely on any in-
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containment instrumentation, the result is that (1) the PORVs are opened
and all available safety injection (SI) pumps started and SI valves
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open; (2) auxiliary feed flow is maximized and directed to at least one
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S/G; (3) containment air coolers are placed in emergency configuration
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and containment spray manually maximized; and (4) all available service
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water and component cooling water pumps operated."
In developing a strategy to recover from this event, the licensee made the
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assumption that all. in-containment instrumentation would be lost (Palisades
Plant Emergency Operating Basis Document, E0P 6.0 page 38 of 41, E0P 9.0 page
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25 of 133 and PRA, Appendix 6, page 2). This would- require that the strategy
explained above be performed.
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Initially, the operators would complete E0P 1.0, " Standard Post Trip Actions."
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Upon completion of these actions, the operator would either initially enter
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E0P 6.0, " Excessive Steam Demand Event," or go directly to E0P 9.0,
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" Functional Recovery Procedure," if the operator determines that more than one
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event is in progress.
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Upon entering E0P 9.0, the operator is directed to Success Path HR-3, "S/G
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heat sink and safety injection pump operation," based on actual S/G level-
being less than - 84% of wide range steam generator levels. The following
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actions are required'by HR-3:
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1)
When PCS pressure falls below 1300 psia, trip all PCPs. [PCS
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pressure will decrease to below 1300 psia due to the rapid
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cooldown of the PCS caused by the steam break. The functional
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recovery procedure, not assuming any specific single event is in
progress, has the PCPs tripped in the event a LOCA could be
occurring];
2)
Operate all safety injection and charging pumps;
3)
Isolate the MOST affected steam generator, [this requires the
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closing of both MSIVs which cannot be accomplished);
4)
Verify feedwater criteria by either steam generator level
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[ steam generator level is one of the instruments assumed lost by
this event), or feedwater flow, in conjunction with steam
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generator levels [not available];
5)
Check for acceptance criteria of HR-3 by.SG level [not available),
core exit thermocouples or loop cold leg temperature, [not
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available), and adequate safety injection flow; and
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If HR-3 acceptance criteria is not met, then go to.HR-4, "Once-
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Through-Cooling."
[By using instruments inside containment which
are not reliable, the operator may not be implementing once-
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through-cooling at this point in the procedure.]
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Decisions made in using the E0Ps as written rely upon instrumentation that has
already been assumed to be inoperable and/or unreliable. The PRA agrees that-
in-containment instrumentation will be lost along with in-containment
electrical components.
" Violation of EEQ is assumed to lead to failure of
instrumentation required by the operator to monitor systems and equipment-
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within containment important to maintaining adequate core cooling."
(PRA
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Appendix 6, Page 2)
The licensee continues to take credit for the in-containment instrumentation
and electrical components when using E0P 6.0, " Excessive Steam Demand Event,"
and E0P 9.0, " Functional Recovery Procedure," for recovery.
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The licensee maintains the PORVs closed with their block valves isolated.
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There is no guarantee that the block valves would be able to be opened in the
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harsh environment inside containment.
" Currently to comply with 10 CFR 50, Appendix R, PORV breakers52-196 and 52-224 are normally left open.
This is
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to prevent the possibility of " hot shorts" simultaneously in control circuits
for each PORV and its associated block valve, resulting in a LOCA."
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(Supplement 1 to NUREG-0737-Response to Draft Safety Evaluation - Procedures
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Generation Package, Attachment 1, page 3).
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The licensee has failed to meet their commitment with regard to the following:
(1) go to once-through-cooling; (2) ensure maximum feed flow to at least one
steam generator; (3) maximize containment spray-flow and place containment
coolers in emergency alignment; and (4) maximize service. water and component
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cooling water flow. Due to the facility relying on inoperable or unreliable
instrumentation, there is no guarantee that the steps committed to would be
performed. With the potential loss of instrumentation and other compcnents
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upon diagnosis of the dual steam generator blowdown with failed open MSIV and
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upon exiting E0P 1.0,-the operator should immediately go to once through
cooling to ensure the core remains cooled, then make attempts to regain SGs as
a viable heat sink. This is considered a deviation from the specified
commitment (255/93010-02(DRS)).
2.4.6.2
Loss of Instrumentation
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The NRC concern, documented in the SER issued in 1986, was as follows:
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"Because some instrumentation inside containment may be affected by the
two steam generator blowdown environment, operator training / procedures
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to cope with possible loss of information or misinformation is needed."
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The licensee's response to this concern, documented in a letter to the NRC
dated April 28, 1986, was as follows:
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" Caution notes will be added to the beginning of operator actions and
Safety function Status checks of the optimal recovery procedures for
LOCA and Excess Steam Demand Event (ESDE-the events in which degraded
containment atmospheric conditions are expected). These notes will
state that due to the potential adverse effects of high containment
temperature / pressure on in-containment instrumentation the operator-
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should not rely on any single instrument, but rather observe
confirmatory indications; and significance should be placed heavily on
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trending and less significance on specific instrument readings.
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"In light of these notes, the existing guidance in the LOCA and
ESDE draft procedures for maintaining SG feed flow would prevent
deliberately stopping auxiliary feedwater flow. Additionally, the
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SI pump throttling criteria, which require a high operator confidence
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level in primary temperature and pressure indications, would correctly
prevent any operator action limiting SI flow.
"The acceptance criteria for containment atmospheric conditions in the
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LOCA and ESDE procedures will be changed to have an upper limit of
design pressure and temperature. Therefore, t'iese conditions in safety
function status check should " flag" operators into using the Functional
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Recovery Procedure".
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The licensee, in stating that a caution note would be added to the E0Ps, took
credit for an existing caution.
"During degraded Containment conditions, the
operator should not rely on any single instrument indication due to large
instrument errors. Alternate / additional instrumentation should be used to
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confirm trending of PCS conditions." This caution does not give the operator
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any guidance as to what instrumentation is lost or what alternate
instrumentation is available.
In addition, the basis documents for E0P 6.0
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and E0P 9.0 state that all containment instrumentation is assumed lost for the
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MSLB with single MSIV failure scenario (EOP 6.0 Basis, Revision 4, page 38 of
41 and E0P 9.0 Basis, Revision 5, page 25 of 133).
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The licensee met their commitment in this case, but failed to fully address
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the concerns expressed by the NRC.
In particular, adequate procedure actions
to cope with the loss of information or misinformation caused by the loss of
instrumentation were not established. The inspectors reviewed the training
records related to this event and found no records cf training on loss of
instrumentation.
Some guidance is given in E0P 6.0 for narrow range steam
generator level and pressurizer level, but this would only be applicable for a
single uncomplicated steam line break (no other events in progress concurrent
with the steam line break).
Operations personnel stated that they did not know what instrumentation would
be affected, and that they would have to wait for the engineering staff to
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evaluate containment conditions and determine which instrumentation would be
available for use. This evaluation would take place during the event.
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The licensee failed to fully address the concerns raised by the NRC in
response to item 2.
Specifically, adequate changes to the procedures and
operator training on loss of instrumentation related to this event were not
developed. This is considered an open item (255/93010-03(DRS)).
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2.4.7
Additional Areas Examined
2.4.7.1
Emeroency Operatina Procedures
The caution prior to . step 12 ir E0P 9.0, Success Path 9.0, states "If
maintaining an SG heat sink is immediately needed to protect the Public
Health / Safety and both SGs are required to be isolated, then the Shift
Supervisor may direct departing from the SG isolation steps (reference
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10CFR50.54X)." The basis for this caution states "If both SGs are affected,
isolating both SGs will result in a violation of HR-3 safety function
acceptance criteria. Only HR-4, "Once-Through-Cooling," would be available".
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This caution should direct the operators to the proper procedure, HR-4.
The
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caution, stating that the Shift Supervisor may deviate from the procedure via
10 CFR 50.54x, provides no useful information.
In this case, actions that can
provide adequate protection are not specified even though they are available.
In addition,10 CFR 50.54x is invoked in an emergency when reasonable actions
that depart from a license condition or a technical specification are needed
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to protect the public health and safety and no actions consistent with license
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conditions and technical specifications that can provide adequate or
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equivalent protection are immediately apparent.
In this case, appropriate
actions can be taken through HR-4.
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E0P 6.0, Attachment 2, contains graphs to account for errors-in pressurizer
and SG narrow range level. This attachment in not referred to in the body of
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the procedure. It is referred to in footnotes in the Safety Status, but no
guidance is given as to when these graphs are required. This was brought to
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the licensee's attention with no immediate response. This is considered an
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open item (255/93010-04(DRS)).
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2.4.7.2
Trainina
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The inspectors reviewed the training for this event and determined that the
only training provided was on E0P 6.0.
This training primarily consisted of
being able to identify the "Most Affected" and "Least Affected" steam
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generators.
It did not cover the background to this event or the commitments
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made to the NRC as to what actions were necessary to recover from this event.
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No training was conducted on the potential loss of instrumentation or actions
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needed to be taken in response to the loss of instrumentation. No training
was conducted on E0P 9.0, which is the facility identified primary procedure
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for recovering from this event.
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During the 1989 Emergency Operating Procedure Team Inspection, it was
identified that, "there were no lesson plans for individual success path
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procedures of E0P 9.0, either in the simulator or the classroom phases of
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instruction."
In the current training material, the lesson plan for E0P 9.0
covered the generic steps for E0P 9.0 and addressed the basis for each success
path in an overview format.
Individual training for each success path was not
performed.
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In discussions with the training staff, the inspectors noted that during the
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previous training cycle, a scenario had been run to cover this event. The
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following items were noted in a review of the Simulator Exercise Guide.
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Plant trip was initiated by the inadvertent closure of the MSIVs
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with failure of one MSIV to fully close.
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The training staff inserted leaks on both steam generators to
simulate a MSLB in containment concurrent with failure of one
MSIV. This was done in conjunction with operator over rides on
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the MSIV indications to give the appearance of the failed open
MSIV. The dynamics of this event differed from those obtained by
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running a dual steam generator blowdown with one MSIV failed open.
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The criteria for ending this scenario was, at a minimum, for the
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operators to enter E0P 9.0 and determine which success path to
use. The scenario could be allowed to continue until the most
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affected SG is isolated or until the crew had entered HR-4,
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"Once-Through-Cooling".
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No training was conducted on the potential loss of instrumentation
inside containment.
2.4.7.3
Observations of Simulator Exercise
The following parameters did not appear to be properly modeled on the
Palisades Plant specific simulator.
Containment Pressure - During the first scenario with all ECCS
equipment available, the containment pressure peaked at 12 psig. During
the second scenario with a loss of off site power and only one train of
ECCS equipment available, maximum containment pressure was 22 psig.
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Previous analyses indicated that these pressures are lower than what
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would be present during an actual event.'
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Containment Temperature - The maximum containment temperature
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identified during either scenario was 150*F.
Previous analyses
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indicated that this temperature is lower than what would be present
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during an actual event.
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As a result of the containment temperature and pressure discrepancies, it is
believed tnat the training on this event had a negative effect. The operators
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were not presented with realistic values to enable them to make the required
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decisions. This tends to lead to a false sense of security and reliance on
procedures that do not appear adequate to recover from this event. This is
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considered an open item (255/93010-05(DRS)).
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3.0
Open Items
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Open items are matters which have been discussed with the licensee, which will
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be reviewed further by the inspectors, and which involve some action on the
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part of the NRC or licensee or both. Open items disclosed during the
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inspection are discussed in Sections 2.4.4, 2.4.6.2, 2.4.7.1, and 2.4.7.3 of
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this report.
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4.0
Exit Meetina
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The inspectors met with the licensee staff (denoted in Section 1) at the site
on May 4, 1993, for an exit meeting to summarize the purpose, scope, and
findings of the inspection. A verbal summary of the inspection findings was
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provided to the licensee at that time. The inspectors discussed the likely
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informational content of the inspection report with regard to documents or
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processes reviewed by the inspectors during the inspection. The licensee did
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not identify any such documents or processes as proprietary.
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