ML20245A536
| ML20245A536 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 04/17/1989 |
| From: | Morris K OMAHA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LIC-89-331, NUDOCS 8904250260 | |
| Download: ML20245A536 (6) | |
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9 Omaha Public Power District I
1623 Harney Omaha, Nebraska 68102 2247 402/536 4000 April 17, 1989 LIC-89-331 1
l U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station P1-137 Washington, DC 20555
Reference:
1.
Docket No. 50-285 i
2.
10CFR50.63 - Station Blackout Rule i
Gentlemen:
SUBJECT:
OPPD Response to the Station Blackout Rule As It Applies To Fort Calhoun Station The Omaha Public Power District (0 PPD) has conducted a detailed study in response to the Station Blackout (SB0) Rule.
The study consisted of detailed engineering analyses which demonstrates the Fort Calhoun Station's capability to cope with a station blackout event. These analyses were performed using the guidelines and recommendations described in Regulatory Guide (RG) 1.155
" Station Blackout," and NUMARC 87-00 " Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors." The Attachment provides the OPPD detailed response to the Station Blackout Rule.
The NUMARC 87-00 Generic Response Format was followed in developing this response.
In summary, the results of the engineering analyses document that the probable duration of a station blackout at the Fort Calhoun Station is four hours.
During the four hour period the plant is maintained in a hot shutdown condition.
Decay heat is removed in natural circulation by a steam driven auxiliary feedwater pump and the following conditions exist:
1.
There is sufficient condensate in the emergency feedwater tank to supply the required amount of feedwater to the steam generators for the removal of decay heat.
2.
There is sufficient DC battery capacity to supply the required power to the Station Blackout equipment.
3.
Operability of Station Blackout equipment is not adversely affected by the loss of HVAC.
4.
Appropriate containment integrity is ensured by the containment isolation valves.
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. U. S. Nuclear Regulatory Commission LIC-89-331 Page 2 5.
Sufficient core coolant inventory is maintained to prevent core uncovery.
During the review process of the results of the engineering analyses, a concern was identified regarding possible leakage through the Control Element Drive Mechanism (CEDM) seals.
This condition was not addressed in the reactor coolant inventory analysis as described in the Attachment, Item C.6.
The possible CEDM seal leakage problem was identified in the late stages of the SB0 Study, therefore OPPD will initiate actions to analyze this condition and a schedule will be submitted to the Commission within sixty (60) days to indicate when this analysis will be completed.
Sincerely, M
K. J. Morris
' Division Manager Nuclear Operations KJM/jak Attachment c:
LeBoeuf, Lamb, Leiby & MacRae R. D. Martin, NRC Regional Administrator P. D. Milano, NRC Project Manager P. H. Harrell, NRC Senior Resident Inspector
Attachment OPPD RESPONSE TO STATION BLACK 0UT RULE FOR FORT CALHOUN STATION On July 21, 1988, the Nuclear Regulatory Commission (NRC) amended its regulations in 10CFR Part 50. A new section, 50.63, was added which requires that each light-water-cooled nuclear power plant be able to withstand and recover from a SB0 of a specified duration.
Utilities are expected to have the baseline assumptions, analyses, and related information used in their coping evaluation available for NRC review.
It also identifies the factors that must be considered in specifying the SB0 duration.
Section 50.63 requires that, for the SB0 duration, the plant be capable of maintaining core cooling and appropriate containment integrity.
Section 50.63 further requires that each licensee submit the following information:
1.
A proposed SB0 duration including a justification for the selection based on the redundancy and reliability of the onsite emergency AC power sources, the expected frequency of loss of offsite power, and the probable time needed to restore offsite power; 2.
A description of the procedures that will be implemented for SB0 events for the duration (as determined in 1 above) and for recovery there from; and 3.
A list and proposed schedule for any needed modifications to equipment and associated procedures necessary for the specified SB0 duration.
The NRC has issued RG 1.155 " Station Blackout" which describes a means acceptable to the NRC Staff for meeting the requirements of 10 CFR 50.63.
RG 1.155 states that the NRC Staff has determined that NUMARC 87-00 " Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout At Light Water Reactors" also provides guidance that is in large part identical to the RG 1.155 guidance and is acceptable to the NRC Staff for meeting these requirements.
Table 1 to RG 1.155 provides a cross-reference between RG 1.155 and NUMARC 87-00 and notes where the RG takes precedence.
OPPD has evaluated the Fort Calhoun Station against the requirements of the SB0 rule using guidance from NUMARC 87-00 except where RG 1.155 takes precedence.
The results of this evaluation are detailed below.
A.
Proposed Station Blackout Duration l
NUMARC 87-00, Section 3 was used to determine a proposed SB0 duration of four hours. No modifications were required to attain this proposed coping duration category.
The following plant factors were identified in determining the proposed SB0 duration:
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- 1. -
AC Power Design Characteristic Group is P1 based on:
I -
a.
Expected frequency of grid-related LOOP events does not exceed once per 20 years; b.
Estimated frequency of LOOP events due to extremely severe weather places the plant in ESW group 2; c.
Estimated frequency of LOOP events due to extremely severe weather places the plant in SW group 2; d.
The offsite power system is in the 11/2 Group.
2.
The emergency AC power configuration group is C based on:
a.
There are two emergency AC power supplies not credited as alternate AC power sources.
b.
One emergency AC power supply is necessary to operate safe shutdown equipment following a loss of offsite power.
3.
The target EDG reliability is 0.95.
A target EDG reliability of 0.95 was selected based on having a nuclear unit average EDG reliability for the last 20 demands equal to 0.95.
B.
Procedure Description Plant procedures have been reviewed and changes necessary to meet NUMARC 87-00 will be implemented in the following areas:
1.
Station blackout response per NUMARC 87-00, Section 4.2.1; 2.
AC power restoration per NUMARC 87-00, Section 4.2.2; 3.
Severe weather per NUMARC 87-00, Section 4.2.3.
4.
Procedure changes associated with any modifications required after assessing coping capability per NUMARC 87-00, Section 7.
l C.
Proposed Modifications and Schedule The ability of the Fort Calhoun Station to cope with a SB0 for four hours in accordance with NUMARC 87-00, Section 3.2.5 and as determined in Section "A" above; was assessed using NUMARC 87-00, Section 7 with the following results:
1.
Condensate Inventory for Decay Heat Removal Condensate is supplied from the Emergency Feedwater Storage Tank (EFWST) during a SB0.
It has been determined from Section 7.2.1 of NUMARC 87-00 that 37,391 gallons of water are required for decay l
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heat' removal for four hours. The minimum permissible emergency
~ feedwater storage tank level per technical specifications provides 55,000- gallons of water, which exceeds the. required quantity for coping with a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SB0. No plant modifications or procedure changes are needed to utilize this water source.
2.
Class 1E Battery Capacity l
1 The ' battery capacity calculation verifies that the Class 1E batteries have sufficient capacity to meet SB0 loads for four hours assuming certain loads not needed to cope with a SB0 are stripped.
These loads i
are identifie ' in the calculation and will be incorporated into plant procedures.
3.
Comoressed Air Air-operated valves relied upon to cope with a SB0 for four hours can either be operated manually or have sufficient backup csurces independent of the preferred and blacked-out unit's Class 1E power supply. Valves requiring manual operation or that need backup scurces for operation are identified in plant procedures.
4.
Effect of loss of Ventilation The calculated steady state ambient air temperature for the steam driven auxiliary feedwater (AFW) pump room during a SB0 induced loss of ventilation is 110.9'F.
The assumption in NUMARC 87-00, Section 2.7.1 that the control room will not exceed 120*F during a SB0 has been assessed.
The control room at the Fort Calhoun Station has been calculated to reach a steady state ambient air temperature of 104.3*F during a SB0.
Therefore, the control room is not a dominate area of concern.
Reasonable assurance of the operability of SB0 response equipment in the above areas of concern has been assessed using Appendix F to NUMARC 87-00.
No modifications or associated procedure 4 sges are required to provide reasonable assurence for equipment ot, ability.
5.
Containment Isolation The plant list of containment isolation valves has been reviewed to verify that valves which must be capable of being closed or that must be operated (cycled) under SB0 conditions can be positioned (with indication) independent of the preferred and blacked-out unit's Class lE power supplies.
Procedure changes are required to ensure that appropriate containment integrity can be provided under SB0 conditions.
Procedures shall be revised to address the following SB0 areas:
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P,.
1 Containment spray header isolation valves outside containment j
must be verified closed, l
HPSI isolation valves outside containment must be verified closed I
if the associated isolation valves inside containment are known to be open at the onset of SBO.
LPSI isolation valves must be verified closed if the associated inboard valves inside containment are known to be open at the onset of SB0.
6.
Reactor Coolant Inventory The ability to maintain adequate reactor coolant system inventory to ensure that the core is cooled has been assessed for four hours. A plant-specific analysis was used for this assessment in accordance with NUMARC 87-00 guidelines and assumptions. The expected rates of reactor coolant inventory loss under SB0 conditions do not result in core uncovery in a SB0 of four hours. Therefore, makeup systems in add'ition to those currently available under SB0 conditions are not required to maintain core cooling under natura1' circulation (including reflux boiling).
During the review process of the results of this analysis, a concern was identified regarding possible leakage through the Control Element Drive Mechanism (CEDM) seals. This condition was not addressed as part of the reactor coolant inventory analysis. OPPD l
is currently addressing this problem and a supplementary analysis will be performed to resolve this issue.
The procedure changes identified in Parts B and C above will be completed within two years after the notification provided by the Director, Office of Nuclear Reactor Regulation in accordance with 10CFR50.63(c)(3).
The NRC Staff will be kept appraised of OPPD's progress in all planned actions discussed above.
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