ML20235A808
| ML20235A808 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 09/18/1987 |
| From: | Bailey J WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| ET-87-0297, ET-87-297, GL-87-12, NUDOCS 8709230426 | |
| Download: ML20235A808 (14) | |
Text
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LF CREEK W8) NUCLEAR OPERATING John A. Bailey Vice Proeident En0ineering and Technk:al Services September 18, 1987 ET 87-0297 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.C.
20555
Subject:
Docket No. 50-482:
Response to Generic Letter 87-12 Loss of RHR While the RCS is Partially Filled Gentlemen:
This letter is provided as requested in Generic Letter 87-12,
" Loss of Residual Heat Removal (RHR) While the Reactor Coolant System (RCS) is Partially Filled", dated July 9, 1987.
Attached is Wolf Creek Nuclear Operating Corporation's response to items 1 through 9 in the Generic Letter.
Additionally, this response encompasses the topics contained in Enclosure 1 to the Generic Letter.
The information attached provides a description of the operation of the Wolf Creek Generating Station, during the approach to and during operation with a partially filled Reactor Coolant System.
If you have any questions concerning this submittal, contact me or Mr. D. L.
Maynard of my staff.
Very truly yours, l
g l
John A. Bailey Vice-President Engineering & Technical Services j
JAB /jad Attachment
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cc: P. O'Connor (2)
R. Martin i
J. Cummins 8709230426 870918 h[/
PDR ADOCK 05000482 P
PDR P.O. Box 411/ Burlington, KS 66839 / Phone: (316) 364-8831 1
g An Equal opportunity Ernployer MF/HC/ VET 4
STATE OF KANSAS
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) SS COUNTY OF COFFEY
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John A.. Bailey, of lawful age, being first duly sworn upon oath says that he
-is Vice-President Engineering and Technical Services of Wolf Creek Nuclear Operating Corporation;- that he has read the -foregoing document and knows the content-thereof;.
that he has' executed that same for and.on behalf of said Corporation with full power and authority to do so; and that the facts therein stated are. true and correct to -the best of his knowledge, information and belief.
Bk f' John A. Bailey
/
Vice-President Engineering and Technical Services SUBSCRIBED and sworn to before me this /[
day of 1987.
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WOLF CREEK NUCLEAR OPERATING CORPORATION.
WOLF CREEK GENERATING STATION RESPONSE TO GENERIC LETTER 87-12 LOSS OF RESIDUAL HEAT REMOVAL.(RHR) WHILE THE REACTOR COOLANT SYSTEM (RCS) IS PARTIALLY FILLED i
Prepared by:
Mike Estes Jerry Houghton
u
Attachment:
to ET 87-0297 September 18, ~ 1987...
Page ;1 of L11 Response to Generic Letter 87-12 Loss of Residual Heat Removal While RCS Partially Filled-This attachment provides a description of the operation: of the Wolf. Creek Generating Station during the approach to'a partially filled' Reactor Coolant System (RCS) condition and during operation with a partially ~ filled RCS.
The following-items re-state specific Generic Letter requests for
-information and are followed.by the corresponding Wolf Creek Nuclear Operating Corporation responses.-
i Generic Letter Item 1
..A detailed description of the circumstances and conditions under which your plant would be-entered into and brought through a draindown process and operated with the RCS partially filled, including any interlocks that could' cause a disturbance to the system.
Examples of the type of information required are the. time between full-power operation and reaching a partially filled condition (used to determine decay heat loads); requirements for minimum steam generator (SG) levels;' changes in the status of-equipment for maintenance and testing and coordination of such operations while the RCS is i
partially filled; restrictions regarding testing, operations, and maintenance that could perturb the nuclear steam supply system (NSSS);
ability of the RCS to withstand pressurization if the reactor vessel head and-steam generator manway are in place; requirements pertaining to isolation of containment; the time required to replace the equipment hatch should replacement be-necessary; and requirements pertinent to reestablishing the integrity of the RCS pressure boundary.
1
Response
The Reactor Coolant System (RCS) is placed in a partially drained condition
(" half-loop") for certain maintenance and inspection activities (i.e., Steam Generator Tube Inspection, RCP seal replacement) where isolation at other conditions is impractical or impossible.
While other methods of isolation such as freeze seals and relying on check valves are plausible, the
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potential for increased equipment damage and/or personnel harm would seem to outweigh any " half-loop" concern.
Typically, from full power operation to " half-loop" operation is 4-5 days.
j
- This, in turn, is subject to the amount of outage work planned during l
cooldown, completion of required surveillance tests and system lineups to j
support " half-loop" operation and other operational concerns. If any change i
in this time estimate were to be postulated, it would be in a more lengthy
{
direction vice less lengthy.
J
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Attachment to'ET. 87-0297-l:
September '18, J 1987 Page 2 of 11 During all modes of operation at Wolf Creek Generating Station (WCGS), an Equipment-Out-Of-Service Log,. Action Statement Summary Log, and Tech Spec-Surveillance Schedules ensure that' applicable equipment is kept operable'or the appropriate Limiting Condition For Operation is met.
Work. activities-receive initiel-review in a pre-outage planning. stage'for.any. conflicts or problems associated with " half-loop" operation.
The Operations. Coordinator
- -Operations,.
a ~ presently licensed
' Senior. Reactory Operator- (SRO),
participates lin'this review.
During the actual' outage, daily. meetings are held. where any concerns-.are addressed.
In-addition, the Operations Coordinator -Operations plays an important part in. assisting-the. Control Room in maintaining safe." half-loop" operation through his input.to various work groups had-the Outage Manager.
Finally,.before any work starts, a
central. work authority (CWA)- who is also a presently licensed SRO, must review plant status and approve work to start..
.The. CWA - maintains close communications 'with. Control-Room. personnel.
Thus,- control'of-work is.
maintained through' pre-outage stage, daily outage meetings,
'and: CWA approval.
To. protect. the EHR. system from excessive pressure and inadvertent flow paths,.certain valves are interlocked. The'following discussion pertains to Train "A" of the, system. The Train "B" discussion would be identical..
RCS hot leg loop 1 to RHR pump "A" suction valves PV-8702A and HV-8701A are used to isolate the RCS from the RHR system.
To open valves PV-8702A and HV-8701A, the following conditions must be met:
Valve HV-8811A, containment sump to RHR pump "A" suction, must be closed.
Valve HV-8812A, RWST to RHR pump "A" suction, must be closed.
Valve HV-8804A, RHR pump "A" discharge to the CCP's suction and SI pump "A" suction must be closed.
RCS pressure sensed on PT-403 (RCS press.)
RCS pressure sensed on PT-405 (RCS press.)
This interlock prevents the operator from inadvertently connecting the RCS to the Refueling Water Storage tank (RWST) the containment recirculation sump, or the Chemical and Volume Control System.
Valve HV-8701A will close automatically if RCS pressure increases to 682 psig.
While some level in all steam generators will exist, it may vary from partially drained to wet layup.
The controlling procedure for draining the i
RCS, GEN 00-007,
" MODE 5-RCS Drain Down",
specifies as initial conditions i
that " Steam Generator Narrow Range levels are at a level necessary to support the outage schedule." This initial condition is intended to allow some flexibility in steam generator secondary side outage activities while draining the RCS.
During the period prior to draining the RCS, Technical Specification 3.4.1.4.1 requires one RHR loop shall be operable and in
Attachm:nt to ET 87-0297 i
September 18, 1987 Page 3 of 11 operation and either another RHR' loop is operable aor water level in two steam generators is greater than 10% wide range.
Upon completion of draining'the RCS,
" half-loop" operation is established and Technical Specification 3.4.1.4.2 requires two RHR loops shall be operable and one shall be_in operation.
This technical specification, correctly recognizes that with " half-loop" operation established, any water level in the steam generators will not be as effective for decay heat removal because the primary side of the steam generator U-tubes are drained.
]
The traditional approach at WCGS has.been that minimal maintenance or testing goes on during draining of the RCS.
The approach used is to estimate the amount of time required to drain the RCS and give the operating j
crews that amount of time to establish stable " half-loop" operation.
At that time full maintanence activities will again commence.
Therefore, l
distraction of the operating crew while establishing " half-loop" operation is minimized.
.Although these restrictions are not formalized in any 9rocedure, management adopted this approach in the 1986 refueling outage.
There are several mechanisms which protect the RCS from over pressurization during draining of the RCS and " half-loop" operation. First, Power Operated Relief Valves (PORV) on the pressurizer are in the cold-overpressure mode (i.e.,
their setpoint is automatically lowered based on RCS temperature).
In the temperature / pressure range where draining of.the RCS is started (i.e., < 200 F, < 425 psig), the PORVs setpoint is approximately 600 psig by Technical Specification 3.4.9.3..
Second, with RHR in operation, the RHR suction relief valves which discharge to the Pressurizer Relief Tank and are set at 450 psig.
- Finally, if neither of the above Cold Overpressure Protection System is. operable, an RCS vent of > 2 square inches is required to be established per Technical Specification 3.4.9.3 This vent would likely be a removed pressurizer code safety valve or PORV to obtain the required discharge area.
Procedure GEN 00-007, requires the verification of one Cold Overpressure Protection System operable as an initial condition.
Normal operating practice at WCGS is with the PORVs and the RHR suction relief valves operable to provide overpressure protection.
Containment integrity, with the exception of one containment air-lock door, the equipment hatch, and any containment isolation valves undergoing local leak rate tests (LLRT) or maintenance is maintained.
In the 1986 refueling outage, it was demonstrated that the equipment hatch could be set in four hours.
The air-lock door could be shut immediately and the valve (s) undergoing LLRT isolated immediately.
There are no technical specification requirements to maintain containment integrity during Mode 5.
As statti previously, maintenance and inspection hetivities at " half-locp" operation are carefully reviewed prior to and during the outage to ensure only essential activities are done.
Since only essential maintenance and inspection activities are allowed, threats to the RCS pressure boundary integrity are minimal.
No technical specification requirements exist for reestablishing the RCS pressure boundary integrity during Modes 5 and 6.
This would be difficult to do since postulation of all possible situations which might occur is impractical.
[
Attachment to ET 87-0279 September 18, 1987' Page 4 of 11
~ Generic Letter Item 2 A detailed description of the instrumentation and alarms provided to the
. operators. for controlling thermal and hydraulic aspects of the NSSS during operation with the RCS partially filled.
You should describe temporary-connections, piping, and instrumentation used for this RCS condition and the quality control process to ensure proper functioning of such connections, piping, and instrumentation, including assurance that they do not contribute to loss of RCS inventory or otherwise lead to perturbation of the NSSS while the RCS is partially filled.
You should also provide a description of your ability to monitor RCS pressure, temperature, and level after the RHR function may be lost.
Response
l Temperature indication comes from one wide range hot leg resistance temperature detector (RTD) and one wide range cold leg RTD per RCS loop.
These are immersed in thermowells which protrude into the reactor coolant flow path.
These RTD's feed a subcooling monitor, the main control board chart recorders, and are used in the Cold Overpressure Protection System to develop PORV setpoints discussed in the response to Item 1.
After establishment of " half-loop" operation, changes in temperature indication will be delayed due to the physical location of the RTDs in the RCS piping.
There is a temperature element at the inlet of each RHR heat exchanger which indicates on the main control board chart recorder.
There is also local temperature indication at the discharge of each RHR heat exchanger. Another temperature element on the discharge of each RHR heat exchanger indicates on the same main control board chart recorder as the RHR heat exchanger inlet temperature. Hence, a highly visible trend of RCS inlet / outlet temperature is available as long as the RHR loop is aligned to the RCS.
Pressure indicators exist on the inlet of each RHR suction line from the RCS, upstream of the first isolation valve to the RHR loop.
These are local indicators and read inside containment.
RCS pressure transmitters are provided which are safety grade with remote indications in the Control Room.
They tap off the bottom of the reactor vessel via the seal table and the top of the vessel and, hence, as long as the vessel head is installed, provide RCS pressure whether or not the RHR System is aligned to the RCS.
It is the pressure transmitters which control automatic RHR isolation on increasing RCS pressure ( 682 psig).
They also prevent opening the RHR loop suction isolation valves-at RCS pressure above 383 psig.
They are routed to a main control room chart recorder.
A local pressure indicator is provided on the suction of each RHR pump downstream of the RHR loop isolation valves.
Attachment to ET 87-0279 September 18, 1987 Page 5 of 11 k
A cold-calibrated pressurizer level indicator (which reads out on the main control board (MCB) ) is used to monitor draining of the RCS until level is off-scale. A non-pressure compensated loop level indicator installed'on the Loop 1 cross-over indicates and alarms on the MCB at a high-and low -level.
It is used in conjunction with tygon tubing installed between BB V311 (also on loop 1 cross-over piping) and BB V295 on the pressurizer vapor space.
Since this tubing is connected to the vapor space where a positive pressure is maintained during draining of the RCS, the tubing is representative of actual reactor vessel level at any time. Also used in an indirect manner to indicate " half-loop" level is the amount of water added. to the Recycle Holdup Tanks.
The comparison between the installed " half-loop" level indicator, the tygon tubing, and Recycle Holdup Tank level provides three independent checks of " half-loop" level.
The following alarms on the MCB are provided to the operators for controlling thermal and hydraulic aspects of the. NSSS during " half-loop" operation:
RHR - RCS ISO VLV OPEN RHR LOOP 1 FLOW LOW RHR A DISCH PRESS HI RHR PUMP TROUBLE RCS LOOP LEV LO RHR LOOP 2 FLOW LOW
-RHR B DISCH PRESS HIGH RHR HX A CCW FLOW HI LO RHR HX B CCW FLOW HI L0 RCSSATUgATE
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MONITOR ALARM In addition, since component cooling -water (CCW) cools the RHR heat exchanger, the CCW alarms may indicate problems with shutdown cooling.
The indications mentioned were designed and are being maintained to appropriate quality standards.
Installation of the tygon tubing is in accordance with plant procedures and therefore is bounded by the Quality Control program.
The tubing is the most likely source of RCS inventory loss if it were to break or separate at the connection to the RCS loop.
During " half-loop" operation, the tubing is checked every 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift and during draining of the RCS a watch is continually stationed to monitor the tubing.
The ability to monitor RCS pressure after RHR loss is reliant on pressure transmitters connected to the vessel previously mentioned and local pressure indicators on the RHR pump suctions upstream of isolation valves.
Loop level via tygon hose and installed loop level indicator would also remain available.
RCS temperature would be reliant on the wide range RTD's with some delay time.
Attachment to ET 87-0279.
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September 18, 1987 Page 6 of 11 Generic Letter item 3 Identification of all pumps that can be used to control NSSS inventory.
Include:.(a) pumps you require be operable or capable of operation (include
'information about such1 pumps that may be temporarily removed from service
'for testing or' maintenance); (b) other pumps not included in item a (above);
and (c) an evaluation of items-a and b (above) with respect to applicable TS requirements.
Response
The following pumps are available for RCS inventory control:
2 Centrifugal Charging Pumps 2 Intermediate Head Safety Injection (SI) Pumps 2 Residual Heat Removal Pumps 1 Positive Displacement Pump With the reactor vessel head on in Mode 5, only one centrifugal charging pump is kept operable to comply with technical specification concerning cold overpressure protection.and boron injection cabability.
All other pumps except RHR pumps are " Danger Tagged" at their breakers.
If a pump is needed (e.g.,an SI pump to fill the accumulators) a valve between its discharge and the RCS is required to be shut.
In a matter of 15 to 30 minutes any of these tagged pumps could be made ready for operation unless they were undergoing maintenance.
Generic Letter Item 4 A description of the containment closure condition you require for the conduct of operations while the RCS is partially filled.
Examples of areas of consideration are the equipment hatch, personnel hatches, containment purge valves, SG secondary-side condition upstream of the isolation valves (including the valves), piping penetrations, and electrical conditions.
Response
l During " half-loop" operations, WCGS normally maintains one containment air-lock door operable and capable of being shut and normal containment integrity operability of automatic isolation valves except for local leak i
rate testing.
Personnel are also on call to secure the open equipment
- hatch, if necessary, with a completion time of four hours expected.
Occasionally, both air-lock doors may be open for equipment / personnel transit as necessary.
Normally the Containment Shutdown Purge System and/or Containment Minipurge System is lined up and running or at least available for operation.
Steam generators may/or may not be in wet layup as described in the response to Item 1 above.
Attachment to ET 87-0279 September 18.-1987 Page;7.of 11 Generic Letter Item 5 Reference to and a summary description of procedures in the control room of your plant which describe operation while the RCS is partially filled. Your response should include the analytic basis you used for procedures development.
We are particularly interested -in your treatment of draindown to the condition where the RCS is partially filled, treatment of minor variations from expected behavior such as caused by air entrainment and de-entrainment, treatment of boiling in the core with and without RCS pressure boundary integrity, calculations of approximate time from loss of RHR to core damage,
. level differences in the RCS and the effect upon instrumentation indications, treatment of air in the RCS/RHR system, including the impact of air upon NSSS and instrumentation response, and treatment of vortexing at the connection of the RHR suction line(s) to the BCS.
Explain how your analytic basis supports the following as pertaining to your facility:
(a) procedural guidance pertinent to timing of operations, required instrumentation, cautions, and critical parameters; (b) operations control and communications requirements regarding operations that may perturb the NSSS, including restrictions upon testing, maintenance, and coordination ' of operations that.could upset the condition of the NSSS; and (c) response to loss of RHR, including regaining control of RCS heat removal, operations involving the NSSS if RHR cannot be restored, control of effluent from the conteirvent if containment was not in an isolated condition. at the time of loss of RHR, and operations to provide containment isolation if containment was not isolated at the time. of loss of RHR (guidance pertinent to tiding-of operations, cautions and warnings, critical parameters, and notifications is to be clearly described).
Response
The following procedures exist for use during " half-loop" operation:
a) GEN 00-007 - MODE 5 - RCS Drain Down b) 0FN 00-015 - Loss of Shutdown Cooling (RUR) c) Alarm Responses - Exist for all alarms discussed in the response to Item 2 above d) GEN 00-001 - MODE 5 Fill and Vent of the RCS e) SYS EJ-120 - Startup of a Residual Heat Removal Trsin f) SYS EJ-110 - RHR System Fill and Vent Including Initial RCS Fill g) SYS EJ-321 - Shutdown of a Residual Heat Removal Train h) SYS BB-110 - Reactor Coolant System Fill and Vent
Attachment to ET 87-0297 September 18, 1987 Page 8 of 11 In all cases except fer OFN 00-015 the analytic basis for the procedures include WCGS-system descriptions, vendor technical manuals, piping and electrical drawings, technical specifications, and good operating practice coupled with. lessons learned during the startup phase.
These procedures were written by experienced operators (either utility personnel or consultants) with extensive licensed / power plant experience.
In addition, these procedures have been used on the actual plant and modified as necessary to be more effective.
Alarm procedures and 0FN 00-015 have been tested on the simulator and necessary modifications made.
- Hence, the l
analytic basis for these procedures is a strong one.
Analysis have been done at WCGS to provide a curve for operators of RCS loop level versus RHR flow to indicate flow rates at which vortexing/ air entrainment of RHR pump suctions will most likely occur.
Initial work on this aspect is already complete, but it has not been finalized for incorporation into operating procedures.
OFN 00-015 initially checks obvious items such as if power to the RHR pumps is available.
If power is not available to each train, then alternate cooling methods are established by such means as steam generators, safety injection accumulator float, RWST alignment to RHR pumps or RWST to operable centrifugal charging pump.
If the RCS is above " half-loop" level fuel pool cooling to refueling pool is aligned.
As a last resort, and with Plant Manager concurrence, the fire header in conjunction with the diesel fire pump or motor driven pump can supply water to the core. Steps are provided to isolate containment if loss of shutdown cooling occurs with the reactor vessel head removed.
Procedure GEN 00-007 provides specific steps to monitor cavitation of RHR pumps and to refer to 0FN 00-015, as necessary, if RHR is lost.
There are notes addressing the fact that draining should be terminated if level anomalies occur.
Specific notes are provided to address monitoring of tygon tubing and the loop level indication. A step is provided to pre-arrange RHR suction venting so it is readily available if needed.
Precautions and limitations address excessive RHR flow, cavitation, and fluctuations in RCS level as steam generator U-tubes are drained.
Generic Letter Item 6 A britf description of training provided to operators and other affected personnel that is specific to the issue of operation while the RCS is partially filled.
We are particularly interested in such areas as maintenance personnel training regarding avoidance of perturbing the NSSS and response to loss of decay heat removal while the RCS is partially filled.
s 1
g Attachment to ET 87-0297 l
_ September.18, 1987 Page.9 of 11-l
Response
The ' following is. a brief description of. the present training activities related to loss'of shutdown cooling.
1.
An annual review of 0FN 00-015,
" Loss of Shutdown Cooling-(RHR)",-
is required of all license holders.
2.
In addition to.the above annual review, on a biennial frequency, all license holders are required to respond to a loss of shutdown cooling as part of the similator training program.
3.
The RHR system is taught to Hot License candidates and covers decay heat history, loss of RHR due to air binding and WCGS' response to prevent air binding.
4.- A Lesson Plan entitled " Refueling Concerns", is included in' the License Requalification Program and encompasses decay heat history, industry events concerning loss of shutdown cooling problems and procedure OFN 00-015 Required crew briefings are conducted during shift turnover in which potential operational concerns / problems are emphasized.
The briefing are conducted by the shift supervisor / supervising operator.. The Diablo Canyon event was placed in operator required reading for operations personnel to read.
The following is a brief description of present training activities related to " half-loop" operations for maintenance personnel.
Principles of PWR operation are provided-to non-operator personnel as part of the maintenance and professional-technical training programs.
Included in these programs will be training on the principles and hazards of " half-loop" operations.
Training will be provided to selected Maintenance and I&C supervisory personnel prior to " half-loop" operations.
The training will discuss the principles and hazards of " half-loop" operation.
Attachment to ET-87-0297-September.18, 1987
-Page 10 of 11 Generic Letter Item 7 1
Identification of additional resources provided to the operators while the i
RCS is partially filled, such as assignment of additional personnel with specialized knowledge involving the phenomena and instrumentation.
j
Response
During " half-loop" operations, no formal requirements exist. for augmenting the operating crew. An outage coordinator for Operations with a SRO license l
or previous license is on site and available to the Control Room to discuss planned activities in regard to " half-loop" operations.
During refueling i
outages, a Superintendent is available to provide management support.
Additionally, Maintenance, Instrumentation and Control, and Engineering support are either available on shift or by call-out.
Generic Letter Item 8 Comparison of the requirements implemented while the RCS is partially filled and requirements used in other Mode 5 operations.
Some requirements and procedures followed while the RCS is partially filled may not appear in the other modes. An example of such differences is operation with a reduced RHR flow rate to the minimize the liklihood of vortexing and air ingestion.
Response
Procedure GEN 00-007 (draining of the RCS) specifically addresses reduced RHR flow as " half-loop" level is approached.
This is a significant difference from other Mode 5 operations. Operations at " half-loop" consists of additional level monitoring as discussed in the response to Item 2 above whereas other Mode 5 operations do not.
More attention is dedicated in GEN 00-007 to RHR cavitation /vortexing than in other Mode 5 operations.
More emphasis is placed on being ready to the shut equipment hatch.
Technical specification requirements at " half-loop" operation regarding operable RHR systems are different than with RCS loops filled.
(See Item 1 above)
On shift personnel are used to monitor local level indication (tygon tube) when
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approaching " half-loop" level.
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Generic Letter Item 9 As a result of your consideration of these issues, you may have made changes to your current program related to these issues.
If such changes have strengthened your ability to operate safely during a partially filled situation, describe those changes and tell when they were made or are scheduled to be made.
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Attachment to ET 87-0297' I
i September. 18, 1987 l
Page 11 of 11 1
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Response
L As a result of loss of shutdown cooling events.in the industry, WCGS L
implemented the following' changes during the 1986 refueling outage:
I
- 1) RHR suction' vents prearranged in' GEN 00-007 so they can be rapidly used' if needed.
- 2) Cautions / notes in procedures regarding RHR cavitation; reduced RHR flow rates approaching " half-loop" operation, and air binding of RHR pumps.
- 3) Detailed loss of shutdown cooling procedure (OFN 00-015) is in place'and training for licensed personnel has been implemented.
- 4) Industry events concerning loss of shutdown cooling problems.has been provided to operations personnel as part of the required reading program.
Additional refresher training on loss of shutdown cooling will occur for all crews prior to performing license duties for " half-loop" operations for the 1987 refueling outage.
Several enhancements to the existing " half-loop" level' monitoring capabilities are being evaluated.
As a result of loss of RHR while the RCS is partially filled events, the following changes to existing programs are being pursued:
Complete the development of " half-loop" level versus RHR flow curves mentioned in the response to Item 5 above and incorporate into plant operating procedures.
This is scheduled to be completed prior to 1987 refueling outage.
1 CONCLUSION Based on the response to this letter, WCNOC has done much to address concerns on " half-loop" ' operation.
Significant changes were implemented during the 1986 cutage based on 3 events of RHR pump cavitation.
Our present program has been proven in actual practice to be effective in preventing RHR pump cavitation.
Additional items are being considered to improve already existing items.
We believe Wolf Creek is and will continue to safely operate at " half-loop" configuration without undue risk to the public.
_ _ _ _ _ _ _ _ _ _ _ _.