ML19249F320
| ML19249F320 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 10/09/1979 |
| From: | Bixel D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| References | |
| IEIN-79-22, NUDOCS 7910110310 | |
| Download: ML19249F320 (10) | |
Text
,
J Consumers
- Power
.' Company
@(ig General Oroces: 212 We,t Mic..ir,an Avenue. Jac kson. Michigan 49201. Area Code 517788-0550
[
October 9,1979 Director, Nuclear Reactor Regulaticn Att: Mr Dennis L Zie ann, Chief Opere ting Reactors Branch No 2 US N2 clear Regulatory Co==ission Washington, DC 20555 DOCKET 50-155 - LICE'iSE DPR BIG ROCK POINT PLANT - RESPONSE TO QUESTIONS REGARDING INTERACTION BETWEEN NON-SAFETY GRADE A'iD SAFETY GPaDE SYSTEMS NRC Office of Inspecticn and Enforce =ent Infor=ation Notice 79-22, issued Sep-tember l!*,1979, discussed a concern involving potential effects of non-safety grade equipment on safety analyses and perfomance of safety grade equip =ent.
This concern related to the effect on non-safety grade equipment of an adverse environ =ent which =ight be produced by failure of a high energy line.
Consumers Power Co=pany was requested by NRC letter dated September 17, 1979, to evaluate the concerns discussed in ILE Infomation Notice 79-22 as they apply to the Big Rock Point Plant.
Const=ers Power Co=pany was specifically requested to consider whether an unreviewed safety concern could exist and to provide info =ation to enable the NRC Staff to determine whether =odification of License DPR-6 was required.
Consurers Power Co=pany has evcluated the effect that adverse enviren=ents which miF t be created by a high energ line break =iEht have on ncn-safety grade control h
equip =ent at Big Rock Point. The results of this evaluation are reported in the attach =ent to this letter as a = atrix of possible effects of non-safety grade eoui;-
=ert failures and explanatory notes. For purposes of preparing the attached matrix of possible adverse effects, each non-safety grade syste: vas arbitrarily assu=ed to fail with the identified failure = ode regardless of design features which =ight prevent such a failure. The explanatory notes describe how each evaluated failure is bounded within the existing licensing bases for Big Rock Point, regardless of its probability of occurrence, or describe why the failure = ode is considered unlikely.
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2-The attached evaluation concludes that no modifice.? ion of License DPR-6 is need-ed as a result of the concerns discussed in I&E L fbmation Notice 79-22.
David A Bixel (Signed)
David A Bixel Nuclear Licensing Administrator CC: JGKeppler, USNRC 10~i; 089 J
EVALUATION OF CDNCERNS DISCUSSED IN I&E INFORMATION NOTICE 79-22 BIG ROCK POINT The e ffects of the failure of non-safety grade syste=s on safety grade systens,
and the resulting impact on the accident analysis are su=marized in Table 1.
An explanation of each item listed on the Table is given as a series of foot-notes, each corresponding to the numbers found in Table 1, 1
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NAE = No Adverse Effect PAE = Possible Adverse Effect TABLE 1 Control System Failure Mode Recire. Line Break Steam Line Break FW Line Break Initial Presture a) Signals for Increased NAE (1)
NAE (1)
NAE (1)
Regulator Flow (a decreased pres-sure) 1.e. admission valves open b) Signals for Decreased NAE (1)
NAE (1)
NAE (1)
Flow (or increased pressure) 1.e. admis-sion valves close Bypass Valve a) Valve Fails Open NAE (1)
NAE (1)
NAE (1)
Control b) Valve Fails Closed NAE (1)
NAE (1)
NAE (1)
Recire Pump a) Pumps Continue to Run NAE (2)
NAE (2)
NAE (2)
Control b) Pumps Trip NAE (2)
NAE (2)
NAE (2)
Recire Pump a) Signaled to Open NAE (3)
NAE (3)
NAE (3)
Valves b) Signaled to Close PAE (3)
PAE (3)
PAE (3)
FW Control a) Signaled to Decrease NAE (b)
NAE (4)
NAE (4)
System b) Signaled to Increase NAE (4)
NAE (4)
NAE (4) c) Loss of FW Heaters NAE (4)
NAE (4)
NAE (4)
,f Control Rod a) Sig. to Withdraw PAE (5)
PAE (5)
PAE (5) b) Sig. to Insert NAE (5)
NAE (5)
NAE (5)
C3 sg)
Rx Vessel llead a) Sig, to Open NAE (6)
HAE (6)
NAE (6)
Vent b) Sig. to Close NAE (6)
NAE (6)
NAE (6)
NAE = No Adverse Effect PAE = Possible Adverse Effect TABLE 1 (Continued)
Control System Failure Mode Recire. Line Break Steam Line Breat FW Line Break Emergency Condenser a) Sig. to Activate NAE (7)
NAE (7)
NAE (7) b) Sig. to Isolate NAE (7)
NAE (7)
NAE (7) m CD w:
C, CD
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1.
High Enerry Line Break with Failure of IPR or Bvtass Valve Centrol Caly high energy line breaks outsi6(1)f containment can affect the IPR or o
bypass valve control system. These breaks range from the main steam line break, (MSLB) in which complete severence of the main steam line is considered, to a feedvater line break, which is in effect a loss of feedvater transient.
Failure of the IPR (which controls the position of the turbine admission valves) or the bypass valve in the open position would result in increased steam flov from the primary system. Steam flow from the system is limited on the high end by the flow area of the main steam line. Since the bypass valve vid admiss-ion valves are fed by the main steam line, opening of one or both cannot result in a greater steam flov fro = the primary system than that found in the main steam line break analysis. This event is therefore bounded by the !GLB enalysis.
Failure of the IPR in the closed position in the event of a break outside contain-cent with coincident failure of the bypass valve results in the same sequence of events as the turbine trip without bypass. Reactor would scram on a high flux s1 gnal.
A feedwater line break outside of containnent has the same effect on the primary system as a loss of feedvater event. Primary coolant is not discharged from the break because the feedvater check valves are located inside containment.
The steam line break analyses all assure that loss of feedvater occurs coincident with the break. The failure of the IPR and/or bypass valve in the open position as a result of a feedvater line break outside of containment is bounded by the steam line break analyses.
The size of the " break" in this case vould be the flov area of the ad=ission valve and the bypass valve. Failure of the IFR and bypass valves in the closed position is effectively a loss of feedvater event without bypass.
In this case : the e=ergency condenser vould li=it pritary system pressure, end begin syste-t cooldown.
2.
High Enerry Line Break with Failure of Pecirculation Punt Control Only high energy line breaks inside of containment can affect the recircula-tion p ep centrol system. Since the Big Rock Point recirculation pu ps are cen-stant speed pumps (variable speed centrol system does not exist), the purps can only fail in the running condition, or in the tripped condi icn. All safety analyses assu=e recirculation pump trip coincident with a break, as this is considered the volst case. The continuatien of cne or both recirculation pu=ps G G: 7 1 ' f I
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2-to run during an accident would result in higher mass flow through the core and therefore a greater thermal margin than is esiculated in the accident analysis.
No adverse effect on a high enelgy line break is produced by failure of the recirculation pump control system.
3.
High Energy Line Break with Failure of Recirculation Pu=n Valves Only high energy line breaks inside of containment can affect the recircula-tion pu=p valves. The pump discharge and auction valves are assumed to mmain open in all of the safety analyses. No adverse effect is pmduced if the valves fail in this position.
Failure of the valus in the closed position such that the recirculation loops were isolated could have an adverse effect on at.cident mitigation. How-ever, it is considend highly improbable that this situation can occur due to the configuration of the Philadelphia Licitorque Motor Operator. This operator provides a heavy duty industrial grade drip proof boundary. To close the valves, a failure vould have to involve a simultaneous, low resistance short on all three phases of the h80 V rotor starter contacts or a hot vire-to-vire 120 V short suffi-cient to energize the coil of the "close" contacter. The possibility of control circuit contact shorts causing failun vill be precluded by placing the control room hand switch for each of these valves in the pull-to-stop position.
Consumers Power Company intends to further evaluate the environmental qualification of the motor starters associated with these valve operators.
h.
Hich Enercy Line Break with Failure of the Feedvater Control System High energy line breaks inside and outside of containment can affect the feedvater control system. The failure can take the form of decreasing feed-water flov, increasing feedvater flev, or decreasing feedvater heating. In all of the accident analyses feedvater is assumed to be lost coincident with the break and therefe re no adverse effect exists for this case. Emergency cooling water vould normally be supplied fmm an outside water source (the fire system) at Big Rock Point. Therefore, failure of the feedvater syste=
such that it continued to supply water to the reactor for large breaks has the same effect as the ECCS, with the exception that recirculation cora spray systems may have to be actuated sooner. No adverse effect on accident nitiga-tien is produced.
Failure of the FW system in this manner for reall breaks 1
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3-such that the W system would override break flow is an event whi:h could also occur during non-accident conditions. Operator action would be required in both cases to prevent 7,he steamline from flooding, but no adverse effect vould be pro-duced on mitigation of the sean break consequences.
Loss of W heating as a result of a high energy line break would increase the rate of depressurizing and cooling down the system. This vould be a ne~,li-gible effect for the large break, and would revert to a loss of W heaters event for a small break. In the case of the small break, a scram may occur on high flux rather than lov drum level. In either case, no adverse effect on accident
=itigation is produced.
5 High Energy Line Break with Failure of Centrol Pod Dri*.e System Failures of the control rod drive system which would cause rods to insert or withdrav vould requize failure of the control solenoids. Early insertion of control rods as a result of a system failure would have a desirable effect
_or Cl breaks. A failure which caused one or more rods to wit:1 draw could have an adverse effect, but such a failure is considered incredible. The control rod drive system is designed such that withdrawal of a control rod requires a nu=ber of control solenoid operations, eacn of which must occur in the proper sequence. The possibility of failures occurring which exactly dupli-cate this sequence of cperatiens is considered incredible.
6.
High Energy Line Break with Failure cf Peactor Vessel Head Vent Valve Only high energy line breaks inside centainment can affect the reactor vessel head vent valve. The vent valve is a 11/2" motor-operated valve which vents to the steam drum. The valve is normally closed but nas a 3/h" bypass around it which is always open_
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L-Failure of the valve in the open position as a result of a high energy line break would produce no adverse effect because of the valve's small size and because the valve would vent water to the primary system and no loss of primary system inventory would occur.
7 High Enerry Line Break with Failure of the E=ereeney Condenser Only high energy line breaks inside containment can cause failure of the emergency condenser. The emergency condenser is not required to operate for eny breaks considered in the LOCA analyses and therefore failure in the iso-lated mode does not produce a more adverse effect than analyzed. Operation of the emergency condenser at the time of a break inside containment would inemase the rate of depressurization and cooldown of the system. For the case of a large break, this would be an insignificant contribution to depressur-ization. In the case of a small break, depressurization vould occur sooner than curre' c LOCA analysis predicts, pe:=itting earlier core spray initiation.
Therefore, no adverse effect occurs because of emergency condenser failure <
Conclusion Based on the consideratiens discussed above, it is concluded that the concerns identified in Inspection and Fnforcement Infer =ation Ilotice 79-22 are addressed within the existing Big Rock Peint accident analyses. No codi-fications to License DPR-6 am required.
CONSUMERS POWER CCMPANY By R B DeWitt (Signed)
Sworn and subscribed to me on this R B DeWitt, Vice President 9th day of October,1979 Nuclear Operations Dorothy H Bartkus (Sicned)
Dorothy H Bartkus Jackson County My cor=ission expires March 26, 1983.
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(1) Locations in the plant where an adverse envimn=ent could exist were identi-fied through review of a report on pipe breaks outside of containment.
(Effect of Cc=cartment Pressurization Due g Pine Syste= Break Outside Containment -
Bechtel, April,1973). Equipnent not located in the areas specified in the report were not considered to be affected by the break.
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