ML20206J887

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Isap Public Safety Impact Model Project Analyses Summaries
ML20206J887
Person / Time
Site: Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date: 06/30/1986
From:
CONNECTICUT YANKEE ATOMIC POWER CO.
To:
Shared Package
ML19292F489 List:
References
NUDOCS 8606270304
Download: ML20206J887 (7)
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Docket No. 50-213 i

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Attachment 1

Haddam Neck Plant 1

Integrated Safety Assessment Program

] Public Safety Impact Model Project Analyses Summaries i

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, 8606270304 860613 PDR ADOCK 05000213 P PDR .

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, s ISAP #1.27 Compliance with 10CFR50.46 Safety Isste

. Small break loss-of-coolant accidents (LOCAs) are one of several categories of events which, if not successfully mitigated, will lead to core melt. To ensure that each plant has the capability to provide coolant after a small break LOCA, redundant high pressure injection pumps nust be provided as part of the plant's emergency core cooling system (ECCS). Connecticut Yankee has two diverse means of providing make-up after a small break LOCA. Two 360-gpm centrifugal charging pumps are part of the Chemical and Volume Control System (CVCS). One of these pumps is continually running during normal operation and the standby pump will start on a safety injection (SI) signal if offsite power is available. Two High Pressure Safety Injection (HPSI) pumps which are part of the CY ECCS are also available for high pressure coolant injection. These pumps automatically start on a SI signal, regardless of the availability of offsite power.

To ensure that these redundant and diverse means of high pressure injection are adequate to provide sufficient make-up for all small break LOCAs in all locations, plant-specific calculations are necessary to show that CY has adequate capability to prevent core melt in the event of a small break LOCA.

Proposed Project 10CFR50.46 requires each light-water reactor to be equipped with an ECCS which can provide cooling for all postulated small break LOCAs in the reactor coolant syster (RCS). In addition, plant-specific calculations using NRC-approved models for small break LOCAs are also required to show that the ECCS is capable of mitigating small break LOCAs. The proposed project is an evaluation of the CY ECCS including calculations to show that the ECCS is in compliance with 10CFR50.46. In addition, the proposed project would include any necessary modifications to the ECCS which may be needed to ensure that all small break LOCAs can be successfully mitigated.

CONNECTICUT YANKEE INTEGRATED SAFETY ASSESSMENT PROGRAM J

Analysis of Public Safety Impact Northeast Utilities has performed plant-specific calculations which show that the CY ECCS is in compliance with 10CFR50.46 and is capable of mitigating all small break LOCAs in the injection phase (Reference 1). These calculations only credited the HPSI pumps and did not include the charging pumps. The Best Estimate LOCA Analysis for CY (Reference 2) showed that one HPSI or one charging pump is adequate for all small break LOCAs for both the injection and recirculation phases except for a certain range of break sizes (0.02 to 0.045 2

ft ) in the RCS loop 2 cold leg and for breaks of less than 0.038 ft in the charging line. For these breaks the ECCS was not capable of providing core cooling in the high pressure recirculation mode. At the time the CY Probabilistic Safety Study (PSS) was performed, only the charging pumps were able to provide high pressure recirculation and the only injection path was via the loop 2 cold leg. Since that time modifications have been made to enable l the HPSI pumps to provide high-pressure recirculation in the event that the l charging pumps are unavailable or inadequate (i.e., the small break LOCA cccurs in the loop 2 cold leg). These modifications allow the operator to align the  ;

HPSI pumps to take suction from the residual heat removal (RHR) pumps during l the recirculation mode and inject coolant into the cold legs. This method of recirculation is not single-failure proof, however.

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A sensitivity analysis was conducted to determine the effect of using the HPSI ,

l pumps for high pressure recirculation on the core melt frequency. This 1 analysis incorporated a postulated single failure-proof high pressure l recirculation system with the HPSI pumps taking suction from the RHR punns as a j back-up method to using the charging pumps taking suction from the RHR pumps I for recirculation. New fault trees where constructed and quantified for the l new high pressure recirculation method and event trees for small- and medium-break LOCAs were revised and requantified to determine a new core melt ,

1 frequency.  !

l With respect to meeting the plant-specific small break LOCA calculations j required by 10CFR50.46, Northeast Utilities has performed these calculations for the injection phase of the small break LOCA using procedural and analytical methods which conform to the criteria set forth in 10CFR50.46 and Appendix K (to the extent it is applicable recognizing that stainless steel cladding is i

CONNECTIClTr YANKEE I E EGRATED SAFETY ASSESSME E PROGRAM

, 6 utilized for all but four fuel assemblies) of 10CFR50 (Reference 1). The CY Best Estimate Analysis shows that the ECCS is adequate for all small break j LOCAs in the recirculation phase except for the limited range of breaks previously described. This analysis used the same methods and computer code

! which were employed in the analysis documented in Reference 1 with some minor changes. These changes only affect the analysis of the injection phase. Since there would be no significant difference between the best estimate analysis and an analysis which meets the criteria specified in 10CFR50.46 for the recirculation mode, further analyses of the recirculation mode for small break l LOCAs to show compliance with 10CFR50.46 are unwarranted and would have no 1

impact on public safety.

Results The sensitivity analysis which assumes a single-failure proof high pressure recirculation method using the HPSI pumps as a back-up method to using the charging pumps for recirculation shows a decrease in the core melt frequency of 1.57 x 10-4, a 28.6% decrease from the present core melt frequency of 5.48 x 10 . This decrease in core melt frequency is larger than the 14% of the core melt frequency which is contribreed by the small break LOCAs in the RCS loop 2 cold leg because using the FPSI pumps as a back-up to the charging pumps increases the reliability of providing recirculation for all small- and medium-break LOCAs.

All of the core melt sequences affected by this change are in consequence category 5 (mean consequence: 2.8 x 103 man-rem). The resulting public risk is calculated as follows:

R =Tx AP1 xC 1 where h = total change in public risk, man-rems T = remaining plant life, 20 years i AP 1

= change in core melt frequency, 1.57 x 10-N/ year i l

l CONNECTICUT YANKEE INTEGRATED SAFETY ASSESSMENT PROGRAM

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= offsite public consequences associated with category l 3

5 - 2.8'x 10 man-rems l R = (20 years) x (1.57 x 10# year-l)x (2.8 x 103 ) j

~ 9 man-rems l l

Based on the Public Safety. Impact Model, this reduction in risk is equivalent  !

to a rank of 0.2 on a scale of -10 to +10.

References

1. W.G. Counsil letter to John A. Zwolinski, "Haddam Neck Plant Small Break I

LOCA Topical Report, TMI Action Plan Items II.K.3.5, II.K.3.30, and II.K.3.31," Docket No. 50-213, dated December 19, 1984.

2. J.F. Opeka letter to Christopher I. Grimes, "Haddam Neck Plant Probabilistic Safety Study - Summary Report and Results," Docket No.

50-213, dated March 31, 1986.

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CONNECTICUT YANKEE INTEGRATED SAFETY ASSESSMENT PROGRAM l

ISAP #2.06 Evaluation of RCS Loop Isolation Valves to Mitigate Steam Generator Tube Rupture Safety Issue The RCS Loop Isolation Valves can be used to isolate a primary loop in cases of RCS pump seal failure or other problems associated with one loop. Each primary loop is equipped with two motor-operated isolation valves at either side of the reactor coolant pump (RCP) and the steam generator. This project was initiated to evaluate the acceptability of utilizing the RCS loop isolation valves (LIV's) to mitigate the consequences of a steam generator tube rupture (SGTR) event, particularly where a steam generator safety relief valve is stucl: open.

The LIV's at CY are capable of closing against a maximum differential pressure of 500 psi. CY transient analyses have shown that only in situations where the high pressure steam dump system is unavailable might this limit of 500 psi be exceeded following reactor trip. (The possibility exists of opening the pressurizer PORVs but this was not considered here). High pressure steam dump to condenser will eventually be unavailable following such events as loss of offsite power (LOSP), loss of condenser vacuum or loss of control air. The high pressure steam dump system consists of ten air-operated valves (A0Vs).

Proposed Project l

The proposed project concerns the qualification of the valve and its operator to function in the presence of a possible pressure drop across the valve disk.

Since LIV's take about three minutes to close, a pressure drop may develop when the valves are nearly closed and the steam generator side of the valves begins i

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. to depressurize due to the leakage into the generator. (It should be noted ,

that credit for closure of the LIV for non-LOSP events was taken in the CY PSS for cases where a safety relief valve on the faulted steam generator is stuck open). This analysis evaluates the effect of a failure of the LIV's to close.  ;

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CONNECTICUT YANKEE INTEGRATED SAFETY ASSESSMENT PROGRAM

l Analysis of Public Safety Impact The dominant contributor to the unavailability of steam dump to condenser is a LOSP. Loss of air or other causes of failure of the A0Vs are relatively small. There is no reason to believe that there is any correlation between a SGIR event and a LOSP, or any of the other failures. They can therefore be treated independently. A SGTR event followed by LOSP for about one hour has a frequency of [ Reference 1]:

1.7E-2 yr-I x (.17 yr /(365 days /yr x 24 hr/dy) x 1 hr) x 0.5

= 1.7E-7 yr .

where a factor of 0.5 is conservatively applied for probability of not recovering offsite power in one hour. The period of one hour is chosen conservatively, since the amount of time required for depressurization of the primary system is only a few minutes. This frequency is then multiplied by the probabilities of various sequences leading to core melt (plant damage state V2) {

that involve the availability of the loop isolation valves. The probability of failing to utilize the loop isolation valves is set equal to one in all of these sequences.

Results The frequency of the V2 plar.t damage state is decreased by 1.0E-9 yr-l if the LIV could be used under all conditions of SGTR coincident with loss of high pressure steam dump capability. The decrease in public risk is given by:

R = 1.0E-9 yr-l x 1.6E+6 man-rem x 20 yr = 3.2E-2 man-rem This project has an ISAP score of less than 0.001.

References

1. J.F. Opeka letter to Christopher I. Grimes, "Haddam Neck Plant Probabilistic Safety Study -

Summary Report and Results," Docket No.

50-213, dated March 31, 1986.

CONNECTICUT YANKEE INTEGRATED SAFETY ASSESSMENT PROGRAM

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