ML19209B986

Responds to 790917 Ltr Re Potential Unreviewed Question on Interaction Between nonsafety-grade & safety-grade Sys,Per IE Info Notice 79-22.Submits Rept of re-evaluation of Potential Impact of Nonsafety Sys on Safety Functions

ML19209B986
Person / Time
Site:	Brunswick
Issue date:	10/04/1979
From:	Utley E CAROLINA POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
GD-79-2480, NUDOCS 7910110358
Download: ML19209B986 (6)

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October 4, 1979 FILE: NG-3514 (B) SERIAL: GD-79-2480 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulc. tory Commission Washington, DC 20555

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BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-325 AND 324 LICENSE NOS. DPR-72 AND 61 HIGH ENERGY LINE BREAK ENVIRONMENTAL EFFECTS ON CONTROL SYSTEMS

Dear Mr. Denton:

This letter responds to your September 17, 1979, letter on the subject of a " potential unreviewed safety question on interaction between non-safety grade systems and safety grade systems." This potential problem was further addressed in IE Information Notice 79-22, dated September 14, 1979. This report also contains the more specific and comprehensive information and analysis requested by your staff during a September 20, 1979, meeting on this issue.

The assessment has not identified any impact on safety actions or analysis conclusions which would increase the consequences (calculated peak cladding temperature, peak containmenc pressure, peak suppression pool temperature, or radiological release) of any SAR events. In particular, the assessment concludes that:

1. No previously performed safety analyses would be adversely affected by the failure of non-safety equipment due to environmental effects of high energy pipe breaks (HEPB's),

and

2. No previously identified safety limits would be violated by the subject effects.

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The attached tables identify the systems examined at the Brunswick Plant and their potential for failures which could affect [i safety system performance for a variety of postulated high energy pipe breaks, locations, and sizes. Table 1 describes those non-safety systems which have been examined and found to have at least the potential for interaction, while Table 2 lists those systems for which no system failure can affect a safety system response.

It will be noted that there are no entries where a postulated non-safety system failure could adversely affect safety system g\

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Mr. Harold R. Denton performance. This results from the almost complete decoupling of the BkB nuclear steam supply and containment system from non-safety B0P equipment and functions.

A number of observations should be made even in light of the successful evaluation.

1. It should be noted that the criteria and suggested NRC Staff evaluation basis involved in this assessment are new recently evolved requirements from RG 1.70, Rev. 2.

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Previous plant design bases for non-safety equipment established a " fail as is" mode rather than the present

" fail in worst position." This is a rather arbitrary and extremely conservative requirement.

2. Evaluation of plant safety as regards HEPB's have been conducted in recent years. Comprehensive analyses were submitted to the NRC Staff and their approval was documented in individual plant SER's. Reevaluation here for more severe criteria has confirmed the previous safety audit.

3. The BWR includes a number of inherent characteristics which are specifically important to this issue:

a. Thorough evaluation of outside containment line breaks for radiological reasons has resulted in a set of comprehensive, sensitive leak detection and isolation systems on BkR's;

b. The BkB does not depend on non-safety equipment for safety actions;

c. The separation of protection systems from control systems has long been a rule relative to safety function reliability;

d. As previously noted, IEPB analyses have been performed and verified physically at BkB facilities;

e. The Bb3 has treated intersystem relationships in ,

considerable detail in a standard SAR section, the i Nuclear Safety Operational Analysis (NSOA). This i systematic evaluation of the BhR system has proven to I-be very valuable relative to environmental impacts effects analysis;

f. Transient and accident analyses of BkR's are conservatively bounded in most cases with respect to non-safety system performance.

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Mr. Harold R. Denton In summary, this submittal is the result of.an extensive reevaluation of the potential impact of non-safety systems on safety functions. The previously approved safety evaluations remain valid.

Further dialogue or discussions in this area, if necessary, should be conducted after the Lessons Learned T.sk Force recommendations and the Bulletin & Orders analysis tasks are resolved.

Yours very truly, k k' N E. E. Utley ~

Executive Vice President Power Supply & Customer Services JSB/CSB/jnh*

Sworn to and subscr'ibed before me this 4thfay of October, 1979.

My commission expires: October 4, 1981.

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TABLE 1 ENVIRONMENTAL INTERACTION .

LOCA ,

Inside Main Steam Line Feedwater Breaks RWCU RCIC llPCI Inside Inside Reactor Turbine Reactor Turbine Non-Safety Systems Location Small Large Bldg. Bldg. Inside Bldg. Bldg. Sml Lrg Outside Outside Outside Recirc System Pumps DW 2 2 4 4 2 4 4 2 2 4 4 4 Valves & Opers. DW 3 3 4 4 3 4 4 3 3 4 4 4 MG Sets TB 4 4 4 4 4 4 4 4 4 4 4 4 MCC TB 4 4 4 4 4 4 4 4 4 4 4 4 Flow Con. Sys. CR 4 4 4 4 4 4 4 4 4 4 4 4 Con. Inst.

Tmitters RB 4 4 2 4 4 2 4 4 4 4 2 4 Feedwater Delivery Flow Elements TB 4 4 4 2 4 4 2 4 4 4 4 4

_ Level DN/RB 2 2 4 4 2 4 4 2 2 4 4 4

- Pumps TB 4 4 4 2 4 4 2 4 4 4 4 4 N Valves & Opers. TB 4 4 4 2 4 4 2 4 4 4 4 4 e MCC TB 4 4 4 4 4 4 4 4 4 4 4 4 Flow Con. Sys. CR 4 4 4 4 4 4 4 4 4 4 4 4 u W lteating TB 4 4 4 2 4 4 2 4 4 4 4 4

"

Instrument Air TB/RB 4 4 2 2 4 2 2 4 4 4 4 4 O Con. Inst.

Tmitters TB/RB 4 4 2 2 4 2 2 4 4 2 2 2 Turbine Pressure Controls liypass Valves TB 4 4 4 2 4 4 2 4 4 4 4 4 Pressure Sensors TB 4 4 4 2 4 4 2 4 4 4 4 4 Control System CR 4 4 4 4 4 4 4 4 4 4 4 4 EllC TB 4 4 4 2 4 4 2 4 4 4 4 4 Neutron Monitoring LPRMS & Cables DW/RB 2 2 2 4 2 2 4 2 2 2 2 2 IRMS & Cables DW/RB 2 2 2 4 2 2 4 2 2 2 2 2 RPIS/RW Blk. Hon.DW/RB 2 2 2 4 2 2 4 2 2 2 2 2 Reactor Protection Turbine Scram TB 4 4 4 2 4 4 2 4 4 4 4 4 MI. Set CB 4 4 4 4 4 4 4 4 4 4 4 4 Reactor Man. Con. hB/CR 4 4 2 4 4 2 4 4 4 2 2 2 SRV Sys.(Non-ADS) DW/RB 3 3 4 4 3 4 4 3 3 4 4 4 RBCCW RB , ,,,

4 4 2 4 4 2 4 ,4 4 4 4 4

• LOCA Inside Main Steam Line Feedwater Breaks RWCU RCIC IIPCI Inside Inside Reactor Turbine Reactor Turbine Non-Safety Systems Location Small Large Bldg. Bldg. Inside Bldg. Bldg. Sm1 Lrg Outside Outside Outside RWCU DW/RB 3 3 2 4 3 2 4 3 3 2 4 4 Circ Water TB 4 4 4 2 4 4 2 4 4 4 4 4 liVAC All 2 2 2 2 2 2 2 2 2 4 2 2 SLC DW/RB 3 3 2 4 3 2 4 3 3 4 4 4 AC Aux. Electric RB/TB 4 4 4 4 4 4 4 4 4 4 4 4 Cond. Tfer. &

Storage-Demin.

Water TB/RB 4 4 4 4 4 4 4 4 4 2 4 4 Main Turbine &

Contr. TB 4 4 4 ~! 4 4 2 4 4 4 4 4 Main Cond. &

Control TB 4 4 4 2 4 4 2 4 4 4 4 4 Instrument Air Compressors TB 4 4 4 4 4 4 4 4 4 4 4 4 Controls TB/RB/DW 4 4 2 2 4 2 2 4 4 4 4 4 Fire Protection TB/RB/DB 4 4 2 2 4 2 2 4 4 2 2 2 CRD liydraulic (Non-Scram) RB 4 4 2 4 4 2 4 4 4 4 2 4 RV llead Vent DW 2 2 4 4 2 4 4 3 3 4 4 4 Suppresston Pool Temp. Mont. RB/Torrus 3 3 4 4 3 4 4 3 3 4 4 4

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Level Mont. RB/Torrus 3 3 4 4 4 4 3 3 4 2 2 1 - Environmental induced malfunction may provide an adverse response 4 - System will nec experience adverse environment 2 - Environmental induced malfunction will not provide an adverse response 5 - No systes. isi ere can affect safety system 3 - System is qualified for adverse environment response i1nna siI

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TABLE 2 System Any High Energy Break Lighting 5 Communications 5 Service Air 5 ,

Equipment Drain Piping 5 Drywell Temp. Monitoring 5 Under Vessel Maint.enance Equipment 5 Process Computer 5 Arec, Radiation Monitoring 5 Process Radiation Monitoring (Non-safety Part) 5 Sampling Systems 5 Maintenance Monorails 5 Environs Monitoring 5 Potable Water 5 Screen Wash 5 Hydrogen Cooling 5 Condenser Priming 5 TBCCW 5 Stator Cooling 5 Offgas 5 Radwaste 5 if i

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