ML20206H277

Forwards Addl Isap Topics for Review,Including Topic 1.64 Re Sys Dependencies on MCC-5 & Topic 1.65 Re Steam Generator Tube Rupture Thermalhydraulic Analysis

ML20206H277
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	04/07/1987
From:	Thomas C Office of Nuclear Reactor Regulation
To:	Mroczka E NORTHEAST NUCLEAR ENERGY CO.
References
NUDOCS 8704150312
Download: ML20206H277 (7)

Text

-. -_

Aprii 7,1987

. Docket No.: 50-213

~

Mr. E. J. Mroczka, Senior Vice President Nuclear Engineering and Operations Northeast Nuclear Energy Company Post Office Box 270 Hartford, Connecticut 06141-0270

SUBJECT:

ADDITIONAL TOPICS TO BE ADDED TO THE HADDAM NECK INTEGRATED SAFETY ASSESSMENT In a letter dated December 12, 1986, Connecticut Yankee Atomic Power Company (CYAPCO) submitted a " final report" for the Haddam Neck Integrated Safety Assessment Program (ISAP). This report provided specific details of Northeast Utilities' (NU) integrated assessment and ranked each of the Haddam Neck ISAP topics.

The staff has reviewed the ISAP topic analyses and the Haddam Neck Probabil-istic Safety Study performed by NU. The staff's review of the Haddam Neck Probabilistic Safety Study (PSS) identified two areas where additional eval-uation under ISAP is warranted. These areas are (1) motor control center five (MCC-5) system dependencies and (2) thermal-hydraulic analyses of ex-tended steam generator tube ruptures.

The staff concludes that CYAPC0 should evaluate the following additional topics in the integrated assessment by PRA methods or thermal-hydraulic analysis and should determine what, if any, cost-effective solutions can be achieved to reduce risk or core melt frequency:

Topic 1.64 System Dependencies on MCC-5 Topic 1.65 Steam Generator Tube Rupture Thennal-Hydraulic Analysis A description of the scope of the integrated assessment for each topic is enclosed.

(original signed by)

Cecil 0. Thomas, Director Integrated Safety Assessment

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5 .p WASHINGTON, D. C. 20555 k.....,/ April 7, 1987 Docket No.: 50-213 Mr. E. J. Mroczka, Senior Vice President Nuclear Engineering and Operations Northeast Nuclear Energy Company Post Office Box 270 Hartford, Connecticut 06141-0270

SUBJECT:

ADDITIONAL TOPICS TO BE ADDED TO THE HADDAM NECK INTEGRATED SAFETY ASSESSMENT In a letter dated December 12, 1986, Connecticut Yankee Atomic Power Company (CYAPC0) submitted a " final report" for the Haddam Neck Integrated Safety Assessment Program (ISAP). This report provided specific details of Northeast Utilities' (NU) integrated assessment and ranked each of the Haddam Neck ISAP topics.

The staff has reviewed the ISAP topic analyses and the Haddam Nect Probabil-istic Safety Study performed by NU. The staff's review of the Haddam Neck Probabilistic Safety Study (PSS) identified two areas where additional eval-uation under ISAP is warranted. These areas are (1) motor control center five (MCC-5) system dependencies and (2) thermal-hydraulic analyses of ex-tended steam generator tube ruptures.

The staff concludes that CYAPC0 should evaluate the following additional topics in the integrated assessment by PRA methods or thermal-hydraulic analysis and should determine what, if any, cost-effective solutions can be achieved to reduce risk or core melt frequency:

Topic 1.64 System Dependencies on MCC-5 Topic 1.65 Steam Generator Tube Rupture Thermal-Hydraulic Analysis A description of the scope of the integrated assessment for each topic is enclosed.

Cada O. O/dpu-w Cecil 0. Thomas, Director Integrated Safety Assessment Project Directorate Division of PWR Licensing-B Office of Nuclear Reactor Regulation

Enclosure:

As stated cc: See next page

6 Mr. Edward J. Mroczka Connecticut Yankee Atomic Power Company Haddam Neck Plant cc:

Gerald Garfield, Esquire Kevin McCarthy, Director Day, Berry & Howard Radiation Control Unit Counselors at Law Department of Environmental City Place Protection Hartford, Connecticut 06103-3499 State Office Building Hartford, Connecticut 06106 Superintendent Haddam Neck Plant Richard M. Kacich, Manager l RFD #1 Generation Facilities Licensing Post Office Box 127E Northeast Utilities Service Company East Hampton, Connecticut 06424 Post Office Box 270 Hartford, Connecticut 06141-0270 Wayne D. Romberg Vice President, Nuclear Operations Northeast Utilities Service Company Post Office Box 270 Hartford, Connecticut 06141-0270 Board of Selectmen Town Hall Haddam, Connecticut 06103 State of Connecticut Office of Policy and Management ATTN: Under Secretary Energy Division 80 Washington Street Hartford, Connecticut 06106 Resident Inspector Haddam Neck Nuclear Power Station l c/o U.S. NRC P. O. Box 116 East Haddam Post Office East Haddam, Connecticut 06423 Regional Administrator, Region !

U.S. Nuclear Regulatory Commission 631 Park Avenue -

King of Prussia, Pennsylvania 19406

i ENCLOSURE 1

Additional Haddam Neck ISAP Topics Topic 1.64 - System Dependencies on MCC-5

Background

One of the insights gained from the Haddam Neck PSS and its review is that failure of MCC-5, both as an initiating event and as a failure sub-sequent to other intitiating events, is a significant contributor to the total core melt frequency of the Haddam Neck plant. The loss of power on 1 MCC-5 is a significant contributor to estimated core melt frequency, pri-marily because of the dependence of reactor coolant pump (RCP) seal cooling on MCC-5. Both the RCP seal cooling (provided by the charging system) and thermal barrier cooling (provided by the component cooling water system) are dependent on MCC-5 either directly or indirectly through other support systems such as the control air system.

Additionally, several accident mitigation systems are dependent on MCC-5.

These systens include the service water system, the high pressure safety injection system, the low pressure safety injection system, the residual heat removal system, the main feedwater system, the emergency boration system, and others which are all at least partially dependent on MCC-5.

, In the Haddam Neck PSS, the failure of MCC-5 contributes approximately 6E-5 per year of the total core melt frequency. This figure is slightly more than 10% of the total core melt frequency. Of the 6E-5 per year, i

3E-5 per year is the result of the loss of MCC-5 as an initiating event.

The staff's reanalysis of MCC-5, based on the data developed during the review of the Haddam Neck PSS, estimates that failure of MCC-5 contributes

) approximately 1E-4 per year to the total core melt frequency.

Proposed Project Under this topic, CYAPC0 should analyze different options designed to re-duce the contribution failure of MCC-5 makes to core melt frequency. The licensee has already proposed to build a new switchgear room. The con-ceptual design for the new switchgear room includes a new 480 V bus and 480 V MCC. Several components necessary for charging system operation (including some component cooling water system components) are part of the proposed set of components to be fed from the new switchgear room. Many of these components are being transferred from MCC-5. However, the proposed MCC does not appear to supply power to any air compressors or semi-vital AC distribution panels. Thus, the new MCC would not create a redundant power supply to the above system. In addition, the conceptual design for this switchgear room does not appear to include an alternative power supply.

a l The objective of this topic is to assess the benefits associated with addition of loads and power supplies to the new MCC such as (1) the air compressors, (2) the semi-vital AC distribution panels, and (3) an alter-native power supply.

Based on information in the Haddam Neck PSS, it appears that elimination of MCC-5 as an initiator would require that semi-vital AC and the instru-ment air system not fail on a loss of MCC-5. This modification could be accomplished either by splitting MCC-5 into two MCCs with no automatic bus transfers or by altering the loads powered by MCC-5 and the new MCC as discussed above. With either of these modifications, it dcas not appear that a loss of MCC-5 alone would result in a plant trip. Additional efforts

to separate other loads currently powered by MCC-5 could lead to further I estimated core melt frequency reductions.

Analysis i Currently, it appears that both semi-vital AC power and the control air systems fail on a loss of power to MCC-5. Both the preferred and alter-

,t native power feeds to the semi-vital AC panels are through MCC-5. All three control air compressors are powered through MCC-5. A review of the loads on MCC-5 suggests that the loss of these two systems is the reason the plant will trip on a loss of MCC-5. Therefore, it would seem possible to eliminate the loss of power on MCC-5 as an initiating event if the semi-vital AC power cystem and the control air system were not solely dependent on MCC-5 for power. This action could be achieved by chang-l ing the alternative source of power for the semi-vital AC distribution

panels and by moving one control air compressor to a different MCC.

1 To maintain the same reliability for the power supplies to these relocated electrical loads, it would be necessary to place these loads on an MCC with characteristics similar to MCC-5; that is, an MCC with pre-2 ferred and alternative power supplies and an automatic bus transfer.

l Eliminating the loss of power on MCC-5 as an initiator results in elimina-4 tion of the 7E-5 per year contribution to core melt frequency, based on the

! results of the Haddam Neck PSS review. However, a small increase in the loss of semi-vital AC power initiating event frequency would be expected.

Since the loss of power on MCC-5 was an initiating event that resulted in the loss of semi-vital AC, the loss of power to the semi-vital AC power system (MCC-5) was not evaluated as part of the loss of semi-vital AC power as an initiating event. With the new design, loss of MCC-5 and failure of the semi-vital AC power automatic bus transfer would result in loss of semi-vital AC power. This additional contribution to the semi-vital AC power initiating event frequency would be 1

F(sv) = F(MCC-5) x P f

.' where F(sv) is the change in the loss of semi-vital AC power frequency F(MCC-5) is the loss of MCC-5 frequency = 2E-3 per year P is the probability of the failure to transfer on loss of power to semi-vital AC distribution panels = 2E-2 (from the Haddam Neck PSS).

The increase in the frequency of loss of semi-vital AC power would be 4E-5 per year. This corresponds to an increase of 3E-7 per year in the core melt frequency. In the Haddam Neck PSS the loss of semi-vital AC power contributed 2E-5 per year to the core melt frequency, with an in-itiating event frequency of 2E-3 per year. This is negligible when compared to the potential reduction gained through the elimination of the loss of MCC-5 as an initiating event.

Such a design change would also help reduce the contribution of the failure of MCC-5 as a support system to the core melt frequency. With the cur-rent plant configuration, the charging system is shut down on a loss of MCC-5. The letdown line isolates and cooling to the RCP seals via the charging system is lost. With design modifications, seal cooling need not be lost because of a failure of MCC-5. Since control air and semi-vital AC are not lost on a failure of MCC-5, it appears that the charging system would not be tripped on a loss of either MCC-5 or the MCC located in the new switchgear room. Based on the results of the Haddam Neck PSS review, this change has the potential to futher reduce the core melt frequency by approximately 3E-5 per year. (It should be noted that the total possible reduction based on the Haddam Neck PSS is nearly 6E-5 per year, with a reduction of a little more than 3E-5 per year resultin event.) g from the elimination of the loss of MCC-5 as an initiating Topic 1.65 - Steam Generator Tube Rupture Thermal-Hydraulic Analysis

Background

l In the Haddam Neck PSS, t N event tree for the steam generator tube rupture (SGTR) appears to have been developed based on the assumption that if the secondary steam relief valves are stuck open and loop isolation valves are not closed, depressurization of the primary system along with long-term decay heat removal results in a safe condition.

The thermal-hydraulic analysis of this event provided in the Connecticut Yankee Best Estimate LOCA Analysis, NUSCO-150, only carries the evaluation of this event to the point at which the Refueling Water Storage Tank is depleted. This is approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the initiation of the ac-cident. Thus, the evaluation is not carried out long enough to determine if the primary pressure could be reduced sufficiently to preclude con-tinuous loss of coolant to a steam generator that is essentially at

' atmospheric pressure. In the absence of this thermal-hydraulic analysis, the staff assumes that if the secondary steam relief valves are stuck open and the primary loop is not isolated, the coolant will be lost to the atmosphere, and core melt will result in about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the initiation of the accident. As a result of this change, the SGTR core melt frequency in the staff's model increased by a factor of 5 compared to the PSS.

Proposed Project In the staff's revised model generated as a result of the review of the Haddam Neck PSS, the assumption that an SGTR coincident with a stuck-open secondary steam relief valve and failure of primary loop isolation could lead to core meltdown resulted in a new dominant core melt sequence. This sequence in the Haddam Neck PSS review is described as follows: SG cooling available, high pressure injection available, SG isolation successful, secondary safety relief valve fails to reseat, reactor coolant loop isolation fails. The frequency of this sequence was estimated to be 4E-5 per year. Thus, if the thermal-hydraulic analysis indicates that this sequence does not lead to core meltdown, the total estimated core melt frequency in the staff's revised model would be reduced by 4E-5 ;r year.

The staff proposes that CYAPC0 perform the necessary thermal-hydraulic analysis for the above sequence. The evaluation should be carried out until either a stable core condition or a core melt situation is reached.

The results of this thermal-hydraulic analysis would then be used to make decisions about any other actions related to this topic.

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