ML19209B955

Evaluation of Interactions Between Nonsafety-Grade Sys & Safety-Grade Sys for Zion Station

ML19209B955
Person / Time
Site:	Dresden, Zion
Issue date:	10/05/1979
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML17174A055	List: ML17174A054 ML19209B955 ML19209B961
References
NUDOCS 7910110283
Download: ML19209B955 (24)

Text

'

ATTACEliENT 1 Commonwealth Edison Company Evaluation of Interactions Between Non-Safety Grade Systems and Safety Grade Systems for Zion Station NRC Docket Nos. 50-295 and 50-304 11bh b0

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,79101102 3 1_1

For over a year Westinghouse Electric Corporation has been conducting an investigation into potential interaction scenarios in which the affect of adverse environments, resulting from high energy line breaks, on control systems could lead to consequences more limiting than the results presented in Safety Analysis Reports.

Table 1 summarizes the scope of the Westinghouse investigation which basically consists of a 7X7 matrix of control systems and accidents.

The seven (7) control systems selected for the investigation by Westinghouse included all control systems addressed in current Westinghouse functional requirements.

The sevet (7) accidents ex-amined encompass all postulated High Energy Line break (EELB) en-vironments, including all break locations and a range of break sizes.

Of the forty-nine (49) combinations of control system and accident environments investigated, fifteen (15) interaction scenarios, each denoted by an X in Table 1, were identified which could potentially result in consequences more severe than reported in Safety Analysis Reports.

The fifteen interactions identified are bounded by con-sideration of the following four (4) scenarios postulated by Westinghouse.

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1.

Summary of Westinghouse Postulated Scenario for Steam Generator PORY Control System Following a feedline rupture outside containment in the auxiliary building, the steam generator PORV control circuits are assumed to exhibit a consequential failure due to an adverse environment.

Failure of the PORV's in the open position re-sults in the depressurization of multiple steam generators which are the source of steam supply for the turbine driven auxiliary feedwater pump.

Eventually, the turbine driven auxiliary feed-water pump will not be capable of delivering auxiliary feedwater to the intact steam generators.

Depending upon auxiliary system design, a potential exists that no auxiliary feedwater will be injected into the intact steam generators until the operator takes corrective action to isolate the auxiliary flow spilling cut the rupture.

Commonwealth Edison Ccmoany Evaluation

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This postulated scenario which requires the break to occur outside containment between the penetration and feedline check valve and which involves the steam generator PORV control system as des-cribed above has been evaluated and is considered inconceivable for Zion Station for the fellowing reasons:

a.

Pipe Configuration At Zicn Station, two of the four pairs of feedwater and steam pipes exit the containment on the west side.

The other two pairs exit on the east side.

The power operated relief valves (PORV's) en each steam line are located im-mediately outside the containment and are also separated in the same manner.

PORV failures on all four steam generators in the Westinghouse postulated scenario would require the creation of environ-mental conditions in excess of the equipment qualification in the valve house associated with the line break and the com-munication of this adverse environment to the other valve hcuse on the opposite side of the containment some 220 feet away.

In addition, the communicated environment must also be in ex-cess of the equipment qualifications within the second valve house.

There are several features which act to prevent communication

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between the valvt Nouses.

Due to the Zion Plant layout, the turbine building acts as an energy sink to such a ecmmun-ication.

Also, the energy dissipation from the valve hcuse itself through the cabled down roof sla'o and wall louvers tc the outside will considerably lessen the environment available to be communicated to the second valve hcuse.

Therefere.

Commonwealth Edison has ccncluded that multiple stear gen-erator blowdcwn will not occur via this postule.ted scenario.

113i 131) 14

The following figures illustrate the Zion Station con-figuration in greater detail.

Figure 1.1 shcws the system location of the pressure transmitters and the electric-to-pneumatic converters.

Figures 1.2 and 1 3 depict the piping tunnel and valve house layout as they relate to the component locations.

Each PORV with its integrally mounted electric-to-pneumatic converter is located in its respective upper cbamber of the valve house.

The pressure transmitters are lo,_ted in each respective lower chamber of the valve house.

The figures also include the valve house upper and lower chamber volumes and vent areas.

Also given are pipe tunnel dimensions and valve house separational distances.

b.

Pump Diversity and Redundancy Figure 1.4 describes the piping configuration associated with each unit's two motor-driven and one steam-driven auxiliary feedwater pump.

There are two separate supply headers and throttle valves regulate the flow to each steam generator and restrict flow lost to the 'oroken loop.

If the postulated scenario resulted in the depressurization of all four steam generators, auxiliary feedwater would still be supplied via the two remaining motor-driven feedwater pumps.

Assuming the loss of the turbine driven pump and single failure taken on one motor driven pump and the postulated depressurization to atmospheric pressure of all steam gen-erators, auxiliary feedwater from the second motor-driven pump will feed equally into all four lines.

The Zion Station FSAR loss of normal feedwater accident (Section 14.1 9) assumes only one motor driven auxiliary feedwater pump is available and supply is to two steam gen-erators.

Therefore, the loss of the turbine driven auxiliary feedwater pump does not affect the results cf the analysis and the core, reactor coolant system and steam system will not be adversely affected.

c.

Material Characteristics of Feedwater Pipe At Zion Station, the feedwater pipe consists of low alloy carbon steel SA333, Grade 6, Schedule 120, 16 inch OD pipe.

1 This material was supplied with impact testing at 0 F, trans-i 0

verse tension test, flattening test and chemical check analysis.

In addition, the Zion pipe is superior to that which is tynically specified for this application, i.e.,

Schedule 60 'r 30 A106, Grade 3 carbon steel.

Therefore, Commonwealth Edison does not consider an abrupt failure of the Zion feedwater piping to be credible.

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1 d.

PORV Design The Zion Station FORVs are Crosby 6 X 8 inch angle valves with a Limitorque SMB-0 motor operator and WKM air operator (Black-Sivalls & Bryson, Inc.).

These PORVs are operated either by an electric motor or by an air piston.

The air piston is designed to close the valve on loss of air supply.

Under adverse environmental conditions neither of these valve drives are expected to fail open.

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2. Summary of Westinghouse' Postulated Scenairo for Main Feedwater Control System Following a small feedline rupture, the main feedwater control system malfunctions in such a manner that the liquid mass in the intact steam generators is less than for the worst case presented in Safety Analysis Reports. The reduced secondary liquid mass at time of automatic reactor trip results in a more severe reactor coolant system heatup following reactor trip. Commonwealth Edison Company Evaluation The postulated scenario involving the main feedwater control system as described above has been evaluated and will not occur at Zion Station for the following reasons: a. Zion FSAR Analysis Section 14.1.9 of the Zion FSAR addresses the loss of normal feedwater accident for the purpose of demonstrat-ing that the auxiliary feedwater system is capable of removing the stored and residual heat and thus preventing either overpressurization of the reactor coolant system or the uncovering of the core. The results of the analysis presented in the FSAR conclude that the loss of normal feedwater does not adversely affect the core, reactor coolant system cr the steam systems. The results of che Westinghouse postulated scenario are less severe than that already analyzed in the Zion FSAR. This is apparant when the aosumptions made for the FSAR analysis are examined, i.e.

1) The initial water level in all stecm generators at the time reactor trip occurs is at the lowest level which will result in reactor trip and automatic initiation of auxiliary feedwater flos.

2) Only one motor driven feedwater pump is available at one minute after the accident with the flow rate conservatively assumed to be 410 gpm, and

3) That auxiliary feedwater is delivered to two steam generators.

t i b. Piping Configuration For this scenario to be valid the break must occur between the steam generator nozzle and feedline check valve. However, at Zion Station the feedwater piping configuration prevents all four steam generators from being affected simultaneously because the feedwater regulating valves are located in the turbine building )1b 1-11

and thus, hot directly affected by the postulated pipe break. (See Figures 1. 2, 1.3 and 2.1 which illustrate the system components and their locations.) Note the steam generator level and main steam flow transmitters are located inside the containment but outside of the missile barrier. Also, the main steam pressure transmitters are lochted in the lower chamber of each respective valve house. In addition, the feedwater flow transmitters are located at the end of the pipe tunnel as it opens into the turbine building. Therefore, the separation of all these devices makes the simultaneous failure of all feedwater control systems inconceivable. c. Material Characteristics of Feedline Pipe At Zion Station, the feedwater pipe consists of low alloy carbon steel SA333, Grade 6, Schedule 120, 16 inch OD pfpe. This material was s~pplied with impact testing u at0 F, transverse tension test, flattening test and chemical check analysis. In addition, the Zion pipe is superior to that which is typically specified for this application, i.e., Schedule 60 or 80 A106, Grade B carbon steel. Therefore, Commonwealth Edison does not consider an abrupt failure of the Zion feedwater piping to be credible. d. Operator Action Section 4.2 of WCAP-9600, Report on Small Break Accidents for Westinghouse NSSS System, describes transient analyses for a postulated loss of all main and auxiliary feedwater + (no pipe rupture). Following a loss of all main and auxiliary feedwater, the operator is not required to take action for at least 4,000 seconds following the loss of all feedwater to prevent the core from uncovering. With a feedline rupture assumed conincident with the assumptions made in WCAP-9600, the operator continues to have at least 2800 seconds (47 minutes) before corrective action must be taken to inject auxiliary feedwater into the intact steam generators to prevent core uncovering. { 1-12

i l STEAM STEAM STEAM STEAM FEED FEED PRESSURE FLOW FLOW PRESSURE FLOW FLOW LEVEL p[33gg l U'SE 1 3 C ej, l 33 t O N EGO

• /. PRES $URE PROGRAMMED LEVEL

+ LEVEL + e o ERROR a SUMMER AMPLIFIER : SUMMER P LOW TAV 554' F s VENT AND REACTOR TRIP SAFETY INJECTION OR H1-H1 LEVEL IN --+@~ VENT ONE STEAM GENERATOR STEAM GENERATOR g LEVEL CONTROL FEEDWATER tri REOULATING ~' VALVE ~ A Figure 2.1 1-13

3 Summary of Westinghouse Postulated Scenario for Pressurizer PORV Control System Following a feedline rupture inside containment, the pressurizer PORV control system malfunctions in such a manner that the power operated relief valves fail in the open position. Thus, in ad-dition to a feedline rupture between the steam generator nozzle and the containment penetration, a breach of the reactor coolant system boundary has occurred in the pressuruzer vapor space. Commonwealth Edison Comoany Evaluation The postulated scenario involving the pressurizer PORV control system as described above has been evaluated for Zion Station and is very unlikely to occur for the following reasons: a. Piping Configuration At Zion Station, the postulated scenario is highly unlikely because of the distance from the PORVs and their control components to the feedline break. Figures 3 1, 3.9, and 3 3 are general arrangement drawings showing the rel ?.icn of the pressurizer to the four steam generators. The feedwater nozzle to steam generator elevation is approx-imately at the 631' elevation and the pressurizer PORV ele-vation is approximately at the 648' elavation. The PORVs are shielded from the feedwater lines by the concrete barrier (coffin) surrounding the pressurizer. The pressure sensing control instrument (s) are located at the 560' basement elevation outside the missile barrier well below the postu-lated break location. b. PORV Design Figure 3.4 depicts the pressurizer PORV control system. The valves are operated by an air piston. The air piston is de-signed to fail close the valves on a loss of air supply. Under adverse environmental conditions the PORVs are not ex-4 pected to fail open. i L c. Material Characteristics of Feedwater Pipe For the postulated scenario to be valid, the break must occur inside the ccntainment between the steam generator nozzle and the containment penetration. At Zion Station, this postulated break is unlikely to occur in this specific location because the feedwater pipe consists of low alloy carbon steel SA333, Grade 6, Schedule 120, 16 inch OD pipe. This material was li31 14]}

supplied with impact testing at 0 F, transverse tension test, flattening test, and chemical check Lnalysis. In addition, the Zion pipe is superior to that which is typ-ically specified for this application, i.e., Schedule 60 or 80 A106, Grade B carbon steel. In addition, this pos-tulated scenario requires a double ended break to lead to the limitin'; consequences. Based on the above material char-acteristics, Commonwealth Edison does not consider an abrupt failure of Zion feedwater piping to be credible. Smaller breaks are more conceivable and permit longer operator action times to close the isolation valves installed at Zion Station. The t.nsequences of this postulated scenario are not applicable to Zion Station due to its current mode of operation. The PORVs on the pressurizer are considered to be operator con-venience items. Due to leakage problems associated with these valves, Zion Station revised operating procedures to isolate the PORVs during power operation by utilizing the upstream block valves. Since the Zion safety analyses do not take credit for the function of the PORVs in the safety relief of the pressurizer, there is no compromise to the reactor coolant system by operating with block valves closed. I 15i ! 41]L e .I i 1-15

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4. Summary of Westinghouse Postulated Scenario for Rod Control System Following an intermediate steamline rupture inside containment, the automatic rod control system exhibits a consequential fail-ure due to an adverse environment which causes the control rods to begin st epping out prior to receipt of a reactor trip sigr.al on overpower delta-T. This scenario results in a lower DNB ratio than presently presented in Safety Analysis Reports. Commonwealth Edison Company Evaluation The postulated scenario involving the rod control system as ^ described above has been evaluated for Zion Station and is con-sidered inconceivable for the following reasons: a. Rod Control System Design At Zion Station, the rod control system design precludes the possibility of the postulated failure scenario. Figure 4.1 contains the logic diagram associated with the power range excore detectors as they relate to the rod con-trol system. The Nuclear Instrumentation System input to the rod control system passes through an auctioneering cir-cuit which constantly selects the highest of the four ch'annels. Failure of a single channel will not produce any adverse con-sequences since the rod control system will be operating on one of the other three channels. b. NIS Instrumentation At Zion Station, Commonwealth Edison has determined that all potential NIS failures will initiate reactor trip, thus pre-cluding rod withdrawal. Figure 4.1 contains the logic dia-gram associated with the power range excore detectors as they relate to the negative flux rate reactor trip. The power range detectors are qualified to 300 F at 100% humidity. These qualifications exceed the anticipated ad-verse environment from the postulated scenario. All conceivable failure modes taken on a given power range { excore detector and its associated cabling result in total channel failure. The following failures were also included: 1) High voltage detector input to the shield or high voltage detector input to ground will produce zero output from the dctector; and 2) Detector output to the shield or detector output to ground will produce zero input to the summing amplifier. Thus, the failure of the cabling or connectors will only produce a negative flux rate trip for the failed channel. 1-20

As described above, the failure of one channel will place the negative rate trip bistable in an armed position, i.e., one channel out of four in a trip mode. The reactor will trip on a two out of four logic, i.e., if two or more channels are failed. Therefore, the Zion units are protected against multiple channel failures within the power range excore de-tectors. c. Detector Configuration At Zion Station, physical separation and installation of the detectors precludes all four channels being simultaneously affected by an adverse environment. Figure 4.2 details the placement of the power range excore detectors around the upper and lower reactor core centerl ne. Figure 3 1 shows the general arrangenent relating col' centerline to steam pipe location. It is quite apparent that a torturous flow path must be followed for the adverse environment created by the postulated steam break to reach the detector locations. 113i(!hlC) .I i 1-21

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ZION UNIT I NUCLEAR INSTRUMENT SYSTEM DETECTOR POSITIONS N32 SOURCE N36 INTERMEDIATE WHITE N SET II LOOP D-3 Q LOOP B-4 O' N42 POWER N44 POWER h YELLOW WHITE SET II SET IE A GH R 15 i4--_-- -l _ _ _ _ 13 - - - - '- O - - - - - - O se^ar O sec SPARE s 270* 8 90* 3---- ,O----- r 2 _ - - - .t - - - - N43 POWER N41 POWER h h RED BLUE SET III 18 0

• O LOOP C-2 N31 SOURCE LOOP A-t -

N35 INTERMEDIATE RED ~ SET I i TYPICAL CROSS-SECTION. I A [ POWER ~ REACTOR CORE F. OE CIC / ~ ~ ~ ~ ~ IN T E R t.l E DI A T E N 41,4 2,4 5 4 4 RANGE N35 $ 36 %n UPPER HALF U (UlCle BF /"7-P O W E R 3 RANGE SOURCE RANGE 5' (

U N 41,4 2,4 3,44 N3I( 32 I

LOWER HALF Figure 4.2 f}}f l l-23}}