ML19344E477
| ML19344E477 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 06/26/1980 |
| From: | William Jones OMAHA PUBLIC POWER DISTRICT |
| To: | Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| References | |
| IEB-80-12, NUDOCS 8009020106 | |
| Download: ML19344E477 (6) | |
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g June 26, 1980 Mr. K. V. Seyfrit, Director U. S. Nuclear Regulatory Commission Office of Inspection and Enforcement Region IV 611 Ryan Plaza Drive Suite 1000 Arlington, Texas 76011
Reference:
Docket No. 50-285, Fort Calhoun Station 4
Dear Mr. Seyfrit:
The Omaha Public Power District received IE Bulletin 80-12, dated May 9, 1980, regarding decay heat removal system operability.
Attached hereto is the District's response to the bulletin.
Sincerely,
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/, /V.s. Jones Division Manager Production Operations, WCJ/KJM/BJH:Jmm Attach.
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Director Division of Reactor Operations Inspection NRC Office of Inspection and Enforcement Washington, D. C.
20555 LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N. W.
Washington, D. C.
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1.
Review the circumstances and sequence of events at Davis-Besse as described in Enclosure 1.
The events as described in Enclosure 1 have been reviewed. Al-though no electrical schematics were provided, it appears from the description that fundamental design differences at Fort Calhoun Station would make a similar event unlikely.
2.
Review your facility (f es) for all DHR degradation events experi-enced, especially for events similar to the Davis-Besse incident.
Operations Incidents from 1974 through the present were reviewed.
No degradations were noted which rendered the shutdown cooling system inoperable.
3.
Review the hardware capability of your facility (ies) to prevent DHR loss events, including equipment redundancy, diversity, power source reliability, instrumentation and control reliability, and overall reliability during the refueling and cold shutdown modes of operation.
A review (attached) of the shutdown cooling system at the Fort Calhoun Station has been completed.
This review included the shut-down cooling system and other auxiliary systems which are used for decay heat removal.
Based on this review, criteria and guidelines have been developed for equipment operability and status to ensure decay heat removal system operability.
4.
Analyze your procedures for adequacy of safeguarding against loss of redundancy and diversity of DHR capability.
The station maintenance and operating procedures will be reviewed against the criteria and guidelines established in Item 3's review.
Based on this review, the procedures will be updated to ensure operability of the decay heat removal system.
5.
Analyze your procedures for adequacy of responding to DHR loss events.
Special emphasis should be placed upon responses when maintenance or refueling activities degrade the DHR capability.
At present, the District has emergency procedures which deal with the loss of certair of the shutdown cooling auxiliary systems (i.e., raw wats ani; component cooling).
Based on the guidelines and criteria of Iten 3's review, an emergency procedure will be developed to provide operator guidance for loss of shutdown cool-ing.
6.
Until further notice or until Technical Specifications are revised to resolve the issues of this bulletin, you should:
a.
Inplement as soon as practicable, administrative controls to assure that redundant or diverse CHR methods are available during all modes of plant operation.
(NOTE: When in a
~ refueling mode with water in the refueling cavity and the head removed, an acceptable means could include one DHR train and a readily accessible source of borated water to replenish any loss of inventory that might occur subsequent to the loss of the available DHR train.)
b.
Implement administrative controls as soon as practicable, for those cases where single failures or other actions can result in only one DHR train being available, requiring an alternate means of DHR or expediting the restoration of the lost train or method.
Operations Memorandum 80-02 has been issued to provide interim guidance for operator action until Items 4 and 5 are completed.
7.
'leport to the NRC within 30 days of the date of this bulletin the results of the above reviews and analyses, describing:
a.
Changes to procedures (e.g., emergency, operational, adminis-trative, maintenance, refueling) made or initiated as a result of your reviews and analyses, including tne schedules or actual dates of accomplishment.
(NOTE:
NRC suggests that you consider the following:
(1) limiting maintenance activities to assure redundancy or diversity and integrity of DHR cap-ability, and (2) bypassing or disabling, where applicable, automatic actuation of ECCS recirculation in addition to disabling high pressure injection and containment spray preparatory to the cold shutdown or ref'ueling mode.)
The District will complete the review, update, and issuance of the procedures described in Items 4 and 5 by October 1, 1980.
Training will be completed in approximately 6 weeks following the October 1, 1980, date.
Since the Fort Calhoun Station has just returned (June of 1980) to operation from a refueling outage, no use of the shutdown cooling system is anticipated until the refueling outage of 1981.
Based on this, the District feels that the October 1,1980, implementation date will not present a safety concern.
The Operations Memorandum will ensure adequate safety measures for the interim period.
b.
The safeguards at your facility (ies) against DHR degradation, including your assessment of their adequacy.
Based on the attached review criteria and guidelines, the District feels that upon completion of the procedure update of Item 7 a.,
the safeguards to ensure decay heat removal operability are ade-quate.
Adequate time will be available to institute one of the alternate cooling methods should some SDC degradation occur.
DECAY HEAT REMOVAL CAPABILITY REVIEW The decay heat removal system (designated shutdown cooling - SDC) at the Fort Calhoun Station provides a means of core cooling independent of the reactor coolant pump / steam generator system.
This system is designed to be placed in service at 265 psia and will remain in service at pressures down to atmospheric.
As designed, a line from the number 2 steam generator hot leg is opened permitting either of the two low pressure safety injection (LPSI) pumps to take suction from the reactor coolant system (RCS).
The re-actor coolant is then pumped through the shutdown cooling heat exchangers and is then pumped into the reactor cold legs via the low pressure safety injection header.
The Missouri River provides the ultimate heat sink via the closed component cooling water system and the raw water system.
The following review divides the SDC system and its auxiliaries into related subsystems.
1)
Shutdown cooling suction line and isolation valves HCV-347 and HCV-348.
Reference drawings E-23866-210-130, sheets 1 and 2, E-23866-210-110 and Figure 8.1.1.
A single suction line is installed to the suction of the two low pressure safety injection pumps.
Since the line is used only for shutdown cooling, a single line is considered ade-quate.
The single line also minimizes exposure to LOCA. The line is isolated during operation above 265 psia by HCV-347 and HCV-348.
The isolation valves are Limitorque motor operated valves powered from the emergency 480v buses.
The valves fail as is (HCV-347 is powered from DG2 and HCV-348 is powered from DG1).
Two redundant channels are provided which will close HCV-347 l
and HCV-348 should pressure exceed 265 psia.
The pressure protection channel P105 and P115 will not close the isolation
<alves on loss of power (P105 will be modified to accomplish t
this as a result of the review of IEB 79-27).
If the protection channel should fail while on shutdown cooling, the isolation valves can be manually opened via a hand wheel.
1 2)
Shutdown cooling pumps.
Reference E-23866-210-130 and Figure 8.1.1.
Under normal operation, either of the LPSI pumps SI-1A (powered from engineered safety features train A) or SI-1B (powered from engineered safety features train B) may be used to provide shutdown cooling flow.
Since shutdown cooling i
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During any time that only a single LPSI pump is available, one of the three containment spray pumps may be used if the LPSI pump is inoperable. The use of the spray pumps (SI-3A, SI-3B, or SI-3C) will be restricted to times when the RCS is below 150 psig.
During the times when the refueling pool is flooded, if required, the pumps (the entire low pressure safety injection header) may be taken out of service.
At this time, several alternate means (i.e., charging pumps, spray pumps, HPSI pumps which are powered from an engineered safety feature bus) may be used to make-up to the pool.
Pool temperature is monitored to ensure the core is remaining cooled.
3)
Shutdown cooling heat exchangers.
Reference E-23866-210-130, 11405-M-10, 11405-M-100.
Two shutdown cooling heat exchangers are available.
These are manually aligned for operation with a comon remote manual inlet and outlet control valve (HCV-335 inlet and HCV-341 outlet).
In addition, FCV-326, which bypasses the SDC heat exchangers, is installed for temperature control purposes.
If control power (or instrumentation) fails, the remote valves can be manually repositioned to provide cooling.
Normally, component cooling water is used as the heat removal medium in the SDC heat exchangers.
If the component cooling water system should fail, cooling water from the raw water system can be diverted directly through the SDC heat exchangers.
4)
Loop injection system - SDC supply.
Reference E-23866-210-130, sheet 2.
The low pressure injection system provides the means by which the SDC is injected into the cold leg.
The incoming LPSI line branches to four cold leg injection points, ensuring flow to i
core.
The injection valves are motor operated valves which fail as is.
The valves are powered from diverse safety buses and can be manually repositioned if required.
5) 480 3 p 60 HZ AC electrical system.
Reference Figure 8.1.1.
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Under normal operating conditions with both trains of ESF equipment available, either train can be used for shutdown cooling (one raw water pump (RW), one LPSI pump, and one component cooling water (CCW) pump).
These components may be carried by either an offsi?e power (161 KV or 345 KV) source I
or by the onsite emergency diesel generators.
If maintenance is required on one of the 4.16 KV ESF train buses, the remaining bus coupled with switching on the 480 l
distribution level 'will provide adequate equipment avail-l ability to ensure the shutdown cooling will remain operable.
With one 4.16 KV ESF bus out of service, one LPSI pump, two RW pumps, three CCW pumps and all motor operated valves remain l
1
.. This bus alignment uses one single 4.16 KV bus for distribu-tion. However, the likelihood of a bus fault is small and the exposure time for single bus operation is relatively short.
For testing. purposes, it is felt that the operable diesel generator does not have to be in an automatic operating mode.
If offsite power is lost, the diesel can be manually placed in service.
During periods when the pool is full, with the pool providing an adequate heat sink for the core for a limited period of time, the energency 13.8 KV supply can be used to directly supply the 183C/1B3C-4C/184C 480 buses which would permit makeup to the pool independently of the single supply of a 4.16 KV bus, via charging pump CH-1B.
The supply valves for CH-1B could be manually aligned.
6) 130VDC and 120 1460HZ AC control power.
Reference Figure 8.1.1 The control power systems at Fort Calhoun Station consist of two emergency batteries, one for each train, which feed the 130VDC control system and.four 120 1460HZ solid state in-verters -2 to each train.
The batteries are normally float and supply power only if the 480 V safety buses are de-ener-gized. Manual transfer switches are installed on the DC system to transfer the DC loads if a battery should fail.
If an inverter fails, its load may be transferred to an inverter on the same train.
If control power is lost to valves, they will go to their fail-safe position or may be manually aligned, for shutdown cooling operation.
In addition, 480V breakers can be manually closed at the breaker.
During certain battery maintenance, the entire station control load is handled by one battery.
This failure is considered extremely unlikely.
- However, during periods when the pool is flooded, the 13.8V emergency supply and manual equipment operation should permit adequate core cooling.
7)
Spurious Actuation of ESF If recirculation actuation signal (RAS) were to occur, the operating LPSI pump would be tripped, but can be manually restarted.
The system, as designed, requires a two out of four logic low tank level (safety tank low signal) and an accident signal to initiate RAS (loss of power will not give RAS).
Since the system makes use of level switches to initiate STLS, a spurious actuation is unlikely.
The only initiation on RAS will occur if the pool is drained or instrument air is lost.
It is not felt that special precautions are required to dis-able RAS.
-~ 8)
Auxiliary system raw water and component cooling.
Reference 11405-M-10, 11405-M-100.
Both the raw water (RW) and component cooling water (CCW) systems are required for heat removal from the reactor core to the ultimate heat sink - the Missouri River.
One CCW pump and one RW pump are required for decay heat removal.
The system is designed such that the RW system can provide cooling water directly to the SDC heat exchangers in the event the CCW system or control power fails (also loss of instrument air).
During periods when the pool is flooded, both the RW and CCW systems may be taken out of service.
During this time, pool temperature is monitored.
Also, during periods when the head is on the reactor, the CCW provides an adequate heat sink for decay heat for limited outage of the RW system.
During this time the RCS temperature is monitored.
PROCEDURE CRITERIA AND GUIDELINES 1)
Review redundant equipment and establish administrative procedures as required to ensure shutdown cooling operability.
2)
Issue an EP for loss of shutdown cooling, providing alter-native methods on action based on equipment availability.
3)
Ensure I&C calibrations have caution on accidental initiation of STLS, PPLS, CPHS.
4)
Ensure surveillance testing of initiating signal addresses loss of SDC.
,