ML19344A500

Safety Evaluation Supporting Amend 29 to License NPF-3.LLL Jul 1979 Technical Evaluation Rept Re Alternate to Keylock Control to Bypass Valves Encl

ML19344A500
Person / Time
Site:	Davis Besse
Issue date:	08/11/1980
From:	Office of Nuclear Reactor Regulation
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ML19344A499	List: ML19344A498 ML19344A500 ML19344A502
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NUDOCS 8008200504
Download: ML19344A500 (7)

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UNITED STATES

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,o NUCLEAR REGULATORY COMMISSION h/

WASHINGTON, D. C. 20555 o

S SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0. 29 TO FACILITY OPERATING LICENSE N0. NPF-3 TOLED0 EDIS0N COMPANY AND CLEVELAND ELECTRIC ILLUMINATING COMPANY DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 DOCKET NO. 50-346 Introduction By letter dated January 11, 1979, the Toledo Edison Company (the licensee) transmitted the technical report " Reliability Study of Davis-Besse Unit 1 Decay Heat Removal System Suction Bypass" in response to License Condition 2.C.(3)(p) and the NRC Safety Evaluation Report (SER) NUREG-0136, Supplement No. I dated April 1977. The license condition required the licensee to submit an analysis of design modification alternatives for the present keylock con-trol in the manual bypass valves DH21 and DH23 around the decay heat removal (DHR) isolation valves and to install the approved modifications before start-up following the first refueling outage. The purpose of this modification is to reduce the likelihood of the bypass path being opened inadvertently when isolation of the DHR system is required.

Evaluation The enclosed Technical Evaluation Report was prepared for us by the Lawrence Livermore Laboratory.

Based on our review of this Technical Evaluation Report, we agree with their conclusions that:

(1) the licensee's analysis of design modification alternatives for the present keylock control of the manual bypass valves DH21 and DH23 satisfies License Condition 2.C.(3)(p) and SER Supplement No.1; and (2) the proposed procedural change entailing the use of one unique key and lock to secure bypass valves decreases the likelihood of the bypass beine opened inadvertently when isolation of the DHR loop is required. The proposed analysis of design modification alternatives is, therefore, acceptable to us.

l Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insigni-ficant from the standpoint of environmental impact and, pursuant to 10 CFR 651.5(d)(4), than an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the l

issuance of this amendment.

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. 4 Conclusion We have concluded, based on the considerations discussed above, that:

(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amend-ment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

Dated: August 11, 1980 l

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TECHNICAL EVALUATION REPORT DAVIS-BESSE NUCLEAR POWER STATION UNIT 1 Alternate to Keylock Control to Bypass Valves Docket Number 50-346 July 1979 Prepared by:

Lawrence Livermore Laboratory D

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July 1979.

1.0 INTRODUCTION

In the Safety Evaluation Report, related to operation of Davis-Besse Nuclear Power Station Unit 1, Supp. No.1 NUREG-0136, dated April 1977, NRO requested that a reliability study be made for a spectrum of hypothesized design modifica-tions to be compared with the present design of the low pressure residual heat recoval system. NRC would evaluate the design modifications to determine if the modifications enhance the safety of the system, and determine that the final system is acceptable to minimize the potential for inadvertent opening of the bypass valves during high pressure operations.

2.0 PROPOSED MODIFICATION:

On January 11, 1979, Toledo Edison transmitted the technical report, " Reliability Study of Davis Besse Unit No.1 Decay Heat Removal System Suction Bypass,"

dated January 5, 1979. The study evaluates the rate of occurrence at Davis Besse Unit 1 of incidents in which the Decay Heat Removal (DER) system is exposed to overpressure due to the improper opening of the DER suction bypass.

This evaluation covers a spectrum of design and procedural options f or the bypass.

The designs include (1) present design, (2) present design plus warning sign, and (3) present design plus flange. Procedural methods include (1) no lock, (2) lock, (3) lock with unique key, and (4) lock with two unique locks.

Four categories of events leading to inadvertent opening of bypass while the unit is above cold shutdown were considered:

1.

Maintenance activities in the vicinity of the bypass may result in inadvertent opening of bypass.

2.

Startup from cold shutdown might be attempted with the bypass left open.

3.

Valve confusion: personnel dispatched to enter containment to che'ck or realign valves might select the wrong valves.

4 Panic:

personnel near the bypass at the tice of what they perceive to be a LOCA or, severe transient might panic and irrationally realign valves.

Results and conclusions were presented for frequency of DER overpressure incidents. The do:inant proble= is maintenance on the pressure relief valve, PSV 4849, which is located on the DHR suction line downstream of the tee at which the bypass line rejoins the principal DER suction line.

This relief valve =ust be recoved occasionally for bench testing, and it is a plausible error for the maintenance personnel to open the bypass after reinstalling the pressure relief valve. Maintenance on PSV 4849 can only be perforced while the reactor coolant syste= is between hot shutdown and ccid shutdown. Therefore, the risk of exposing the OHR to damaging over-pressure or initiating a severe interfacing systers LOCA is cuch less than fcr accident sequences applicable to periods of power generation.

3.0 REASON FOR CHANGES:

SER Supple ent No. 1 (p.5-5 and p.E-3) sea:es tha: :he license condizion 2.:(3)(p) requires that the licensee submit an analysis of design codification l

al:ernatives for the presen key lock control in the :anual bypass valves 05:1 and CH23 around the DER suction line valves to decrease the likelihood

July 1979.

3.0 REASON FOR CHANCES:

(cont'd) of the bypass path being opened inadvertently when isolation of the DHR loop is required. (See diagram.) The submitted analysis and installation of approved design modifications shall be completed prior to startup following the first scheduled refueling outage.

The bypass loop contains two manually operated valves around the DHR suctiou line valves. The normally closed bypass valves would be opened in the event of a spurious closure of one of the DHR syste= suction line isolation valves during system operation. NRC requires that further attention be given to the means employed for isolation of the low pressure residual heat removal system f rom the primary system while the latter is pressurized, and that reliable means be developed to assure such isolation. Present procedures have a chain and padlock. The key opens no other valves, but does open certain restricted Orea doors. The two manual isolation valves are in series on the bypass line.

4.0 REVIEW OF LICENSEE'S SUBMITTAL The Toledo Edison's technical report, " Reliability Study of Davis Bese Unit No. 1 Decay Heat Removal System Suction Bypass," dated January 5, 1979, gives results of occurrence rate of incidents in which inadvertent opening of the bypass exposes the DHR to pressures greater than the design pressure for each of the 12 design and procedural options. For all 12 options the dominant accident sequence is associated with maintenance on PSV 4849. The presence of the pressure relief valve is useful in reducing the risk posed by startup with the bypass left open and to protect against RCS overpressure if high pressure injection occurs while the RCS is in cold shutdown. Therefore, Toledo Edisca does not reco= mend the elimination of the pressure relief valve. Rather, one of several more stringent administrative procedures applied to the present design would reduce the prob bility of DHR overpressure to a very lowlevel;i.e.,lessthan4.0x10~9 per year.

The Toledo Edison technical report states that the NRC has no clear-cut policy on a probabilistic criterion for the acceptability of design provisions to avoid interfacing systems LOCA. However, a criterion can be inferred from the dis-position of the overpressurization event leading to the interfacing systems LOCA problem that arose in the Reactor Safety Study (RSS).

The RSS estimated thefrequencyofaninterfacingsg/ year stems LOCA at the low pressure safety injection check valves at 4 x 10-The NRC responded by suggesting design changes which reduce the probability of this event by a factor of 10, to about 4 x 10-7/ year and by promulgating Standard Revier Plan 6.3, "E ergency Core Cooling System," which endorges the fix at Surry as adequate.

3y i= plication, then, a frequency of 4 x 10~

per year is sufficiently safe.

5.0 00NC'_L'S ION :

The Toledo Edison technical report, dated January 11, 197?, concludes that the present design and procedures offer sufficient prctecti:n for the health inc safety of the public.

However, present design and procedure do not meet the criterion inferred from WASH-1400, the accident sequence which fails to 1

July 1979 '

5.0 CONCLUSION

(cont'd) meet the criterion associated with shutdown when the risk is much reduced.

In order to improve safety and meet the inferred acceptance criterion with-out question, Toledo Edison is prepared to implement procedural option 3 entailing the use of one unique key and lock to secure the bypass valves.

The Toledo Edison technical report fulfills the NRC Safety Evaluation Require-ments (April 1977) for the analysis of design modification alternatives for the present key lock control of the manual bypass valves DH21 and DH23. The proposed procedural change entailing the use of one unique key and lock to secure bypass valves decreases the likelihood of the byrass being opened inadvertently when isolation of the DER loop is required. This unique key and lock procedure will, be implemented prior to startup following the first regularly schedulad refueling outage. NRC Safety Evaluation Report (April 1977) requirements are being met.

Therefore, I see nothing technically wrong with the alternative to the key lock control procedure.

6.0 REFERENCES

1.

Safety Evaluation Report py NRC related to operation of Davis-Besse Nuclear {,awerStaticuUni%1.

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Letter.. Toledo Edison to P.eid, January 11, 1979.

Enclosed submittal:

Reliabiliry Study cf Davis-3 esse Unit No. 1 Decay Heat Removal System Suction Bypass.

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