ML19209B685

Forwards Response to NRC 790917 Ltr Re Interaction Between nonsafety-grade Sys & safety-grade Sys.Discusses Steam Generator Power Generated Relief Valve Control Sys & Main Feedwater Control Sys

ML19209B685
Person / Time
Site:	North Anna
Issue date:	10/05/1979
From:	Stallings C VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	Harold Denton Office of Nuclear Reactor Regulation
References
776-091779, 776-91779, NUDOCS 7910100317
Download: ML19209B685 (8)

Text

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Vino n NI A E LEC~rltIC AN D POW E lf C O bt1*A N Y n rcirwon n,vlHOrN I A ullu61 October 5, 1979 Mr. Harold R. Denton, Director Serial No. 776/091779 Of fice of Nuclear Reactor Regulation P0/FHT:baw U. S. Nuclear Regulatory Commission Docket Nos: 50-338 Washington, D. C. 20555 50-339 License Nos: hTF-4 CPPR-78

Subject:

North Anna Power Station Units 1 and 2 Interaction Between Non-Safety Grade Systems and Safety Grade Systems

Dear Mr. Denton:

We have reviewed your letter of September 17, 1979, on the subject issue and in accordance with 10 CFR 50.54 (f) are providing the attached response for North Anna Power Station Units I and 2.

Very truly yours,

f. . We C. M. Stallings Vice President-Power Supply a7d Production Oper.:tions llB 250 7 n 10100 3 !-

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COMMOL3LTH OF VIRGINIA )

) S. S.

CITY OF RICHMOND )

Before me, Notary Public, in and for the City and Commonwealth aforesaid, today personnally appeared C. M. Stallings, who being duly sworn, made oath and said (1) that he is Vice President-Power Supply and Production Operations, of the Virginia Electric and Power Company, (2) that he is duly authorized to execute and file the foregoing statements in behalf of that Company, and (3) that the statements are true to the best of his knowledge and belief.

Given under my hand and notarial seal this 5th day of October, 19 79 .

My Commission expires .)-tj i < w r v 2.o e4</ .

A 9? Notary O$b/ Public P

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Attachment, page 1 of 6 North Anna Power Station Units 1 and 2 Environmental Interaction of Safety and Non-Safety Grade Systems In response to your ORDER pursuant to 10CER50.54(f) issued on September 17, 1979, information is provided which will enable the staf f to make a determination on the su'aject of potential unreviewed safety question concerning interaction between non-safety grade syatems and safety grade systems that were discussed in IE Information Notice 79-22, issue (

September 14, 19 79 .

Westinghouse, as a part of its environmental qualification activit:s s for IEEE 323-19 74, reviewed original assumptions it made for safety analysis reports. Specifically, could a severe environment cause a failure of a non-protection grade component that was previously assumed to remain "as is" and alter the results of the design basis analysis? Wee 3house addressed the f ailure of a control system due to an adverse enviro snside or out-side containment following a high energy line rupture wl..c.i could negate a protective function performed by a safety grade system. They determined that potential interactions existed for the following systems in conjunction with a feedline rupture event:

1) Steam generator power operated relief valve control system

2) Pressurizer power operated relief valve control system

3) thin feedwater control system i 135 2S2

Page 2 of 6 They further dete rmined that a potential interaction existed for tr e automatic rod control system in conjunction with an intermediate steam-11ae rupture event.

These four consequential failures were found to violate internal Westing-house safety analysis criteria. Specifically, hot leg boiling could occur following a feedline rupture with a consequential failure and minimum DNBR could f all below 1.30 prior to a reactor trip following an intermediate steamline rupture with a consequential failure.

A review of the North Anna Final Safety Analysis Report was performed to determine if this new information was outside the bounds of the FSAR.

Hot leg boiling was permitted by the analysis of a feedline rupture event (FSAR section 15.4.2.2). As previously stated, recent Westinghouse criteria did not permit this result. Therefore, while a NSSS vendor criterion has not been met, this new analysis does not exceed the existing FSAR analysis.

For North Anna, an intermediate c' ramline rupture event is directly comparable to an excessive load increase incident (FSAR Section 15.2.11).

This new information does not exceed the existing FSAR analysis. Specifically, three reactor protection circuits are provided (the first is assumed to fail. in this new analysis):

Over temperature Delta - T .

1135 253 NIS Power Range (High Flux)

Overpower Delta-T

Page 3 of 6 Further clarification of the envelope of protection provided by over-temperature Delta-T and overpower Delta-T is shown in FSAR Figure 15.1.1. (Note that this new analysis by the NSSS vendor did not consider other protection circuits which would serve to mitigate this event, i.e.,

Safety Injection Trip from Steamline Differential Pressure or Containment High Pressure.)

Information is provided below for North Anna control systems with respect to the specific concerns of IE Information Notice 79-22, issued September 14, 19 79 .

Steam generator power operated relief valve control system: These valves are located in the Main Steam Valve House. They could be subject to an adverse environment in the event of a main feedwater line rupture in this building. (The new Westinghouse analysis assumes a break between the containment penetration and the first upstream check valve. This run of pipe is very short and is ANSI B31.7 piping.)

The steam generator PORV's fail to a closed position upon a loss of air or electricci signal. These valves are air operated with an electrical to pneumatic interface device which converts the electrical control signal to a pneumatic control signal. FSAR Supplement S10.19 states that the auxiliary feedwater system meets the guidelines of Branch Technical Posti APCSB No. 10-1. Additionally, FSAR Supplement response 10.19b.2, in ef fect, already addressed the new Westinghouse analysis in that it discusses the loss of the turbine-driven pump and one motor-driven pump.

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Page 4 of 6 Additionally, FSAR Supplement S15.18 indicates that one stean generator is sufficient to provide cooling and that FSAR 15.4.2.2 is the bounding analysis for the one-on-one auxiliary feedwater system.

Pressurizer power operated relief valve control system: These valves are located inside the containment. They could be subject to an adverse environment in the event of a main feedwater line rupture in this building.

The pressurizer PORV's fall to a closed position upon a loss of air or electrical signal . These valves are air operated with solenoid valves in the air line.

An additional design feature in provided that could mitigate the ef fects of a RCS depressurization if a PORV were open. Air is removed from the control air system for these valves by a signal derived from protection grade equipment upon RCS pressure (sensed in the pressurizer) falling below a preset value.

Main feedwater control system: These valves are located in the Service Building and are upstream of the containment penetrations and check valves.

They could be subject to an adverse environmeat in the event of a main feedwater line rupture in this building. (Note that a break in this area would be upstream of the check valve for each line. FSAR 15.4.2.2 states that this break location would be treated as a loss of normal feedwater which is covered by FSAR 15.2.8. This section assumes that the steam generators are at a nominal 0% indicated level. This is, then, already consistent with the new Westinghouse analysis for this consequential failure.)

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f Page 5 of 6 The feedwater regulating valves fail to a closed position upon a loss of air or electrical signal. These valves are air operated with solenoid valves in the air line and an electrical to pneumatic interface device which converts the electrical control signal to a pneumatic control signal.

Automatic rod control system: The excore power range detectors are assumed to fail in the new NSSS vendor analysis. These detectors are located with-in the neutron shield tank. The detectors are lif ted into dry spaces within the tank. These spaces are " dead-ended;' and, thus, the 'etectors would not be exposed to a severe environment. Additionally, the largest opening into the area under the vessel is at the RHR mezzanine, which is above the bottom of the shield tank. This further reduces the potential for an adverse environment to affect the detectors and cable connectors. Incore detector guide tubes (and thimbles) pass through much of this opening. NIS cable is run in conduit from the detector to the containment electrical penetration. The detectors themselves do not have an environmental qualification. However, they can operate at 175* for eight (8) hours and probably could operate at an elevated temperature for shorter durations. The cable has been qualified for 300*F for 15 minutes followed by 252*F for the period of 15 minutes to 13 days following the event. (The cable is radiation qualified to 2 x 10 rads and is qualified for a spray solution of boric acid and sodium hydroxide.)

Additionally, FSAR 6.2.2 states that the containment design temperature is exceeded for a major steam line break for only one (1) minute. The break size assumed by Westinghouse in this new analysis is 0.1 to 0.25 square feet at a reactor power level of 70 to 100Z. This break is small compared ll$LJ *) C 4 CJU

Page 6 of 6 to the large Fhin Steam line break. Thus, the peak containnent temperature would be reached later in the event since the blo.idown rate is lower.

Consequently, the NIS power range detectors should remain operable to provide the protection for excessive core power.

Long Tern Action Licensed reactor operators .ad licensed senior reactor operators will review this response so that they will be informed of new infornation generated by West inghou.,e.

This review is intended to provide an understanding of recent NSSS vendor work and to develop an awareness of systen interactions.

Based on this review we have concluded that this does not constitute a significant safe ty concern. We will continue the investigation and will previde additional information as it becomes available,

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