ML20069D553

Responses to 820814 Discovery Questions Based on New Contentions Accepted by ASLB Order.Certificate of Svc Encl

ML20069D553
Person / Time
Site:	Midland
Issue date:	09/13/1982
From:	Steptoe P CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Sinclair M SINCLAIR, M.P.
References
NUDOCS 8209210276
Download: ML20069D553 (67)

Text

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DOCKETED U%RC 52 SEP 20 A10i48 I

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bHMin September 13' Igg 2 Mrs Mary P Sinclair 5711 Summerset Street Midland, MI 48640 MIDLAND PROJECT MIDLAND DOCKET NOS 50-329, 50-330 RESPONSE TO INTERROGATORIES Dear Mrs'Sinclair Enclosed are Consumer Power Company's responses to the " Discovery Questions .

For Censumers Power Company On New Contentions Accepted by Board Order, August 14, 1982."

In response to these interrogatories, Consumers Power has made the following interpretations:

1. Interrogatory II 5 has been interpreted as referring to the NRC Staff.

2. Interrogatory II 13 includes a reference to a statement attributed to a Dr Epstein. We do not know whether or not this quote is accurate. In any event, such a statement is purely argumentative and totally inappropriate as a part of a discovery request. Notwithstanding, since the statement was made it should be noted that whatever statement was made by Dr Epstein should be considered in light of the transcript of the construction permit hearing, at pp 8314-8348, 3660-3661.

3. Interrogatory II 1 is a request for production of documents under 10CFRS2.741, rather than an interrogatory. Consumers Power Company will therefore respond to the request within the 30-day period prescribed by 10CFRS2.741(d).

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, u L Philip P teptoe oc0982-0233a100 8 2 0 9 21 o A.'% g. -

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N 4? i UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD In the Matter of Docket No 50-329 OM CCNSUMERS POWER COMPANY 50-330 OM (Midland Plant, Units 1 and 2) Docket No 50-329 OL 50-330 OL September 10, 1982 AFFIDAVIT OF LOUIS S GIBSON My name is Louis S Gibson. I am the Section Head, Safety and Analyses Section, of the Midland Project Licensing Department. In this capacity, my responsibilities are to review or conduct certain plant analyses for the Midland Plant.

I am primarily responsible for providing a response to Interrogatory I, Questions 1, 2 and 3 concerning Mary Sinclair Contention 3. To the best of my knowledge and belief, the above information and the responses to the above interrogatory (ies) are true and correct.

SwornandSubscribedBeforeMeThis/ Day of 1982 Lain , J. A L yNotaryPWhi'ic Jackson County, Michigan My Commission Expires bvAf [ /f[k r i mi0782-0162n100

MP $

Mary P Sinclair Interrogatory I Contention 3 questions the adequacy of the methodology in the DES for deter-mining the possibility of severe accidents at the Midland nuclear plants, and recommends NUREG/CR/2497, as a better basis.

Questions

1. Have any accidents occurred at Palisades and Big Rock that were a part of the data base for NUREG/CR/2497, June, 1982?

2. If so, describe them.

3. If so, explain if they were initiated by:

a. operator error

b. non-safety equipment impacting on safety equipment

c. equipment malfunction

d. not believing readings of non-safety grade instruments

e. instruments giving the wrong reading

f. maintenance during operation that disabled the safety systems

g. minor mishaps

h. failure of safety systems

1. lack of QA control during operation Responses

1. Nine events have occurred at Palisades and Big Rock that were a part of the NUREG/CR 2497 data base.

2. The events mentioned above are described in the attached Licensee Event Reports "LERs" and also the attached relevent pages from NUREG/CR 2497.

miO982-0054a168

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3. The authors of NUREG/CR 2497 have described how the events were l initiated under the heading "cause" and in the " failure sequence  !

description" in the attached pages from NUREG/CR 2497.  !

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'CP Co's own characterizations of the initiating causes of these events are found in the attached LERs. CP Co has not performed any study attempting to recharacterize these events in terms of the categories shown in Discovery Question 3. In addition, CP Co believes it would be potentially misleading to use the terms suggested by Ms Sinclair, miO982-0054a168 s

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I CATECOEIZATION OF ACCIDENT SEQUENCE PREC350RS l I

I NSIC ACCESSION NUHsER: 132958 l DATE OF LER: December 21, 1977 l DATE OF E7ENT: Decumber 11. 1977 SYSTEM INVOLVED: off-site power j I

COMPONENT INVOLVED: switchyard bue "1" CAUSE: spurious stripping relay signal I SEQUENCE OF INTEREST: lose of offsite power

. ACTUAL OCC3 RENCE: reactor trip with lose of offsite power l

REACTOR *.tA 23 Paliaades DCCEET E15ER: ~50-235 ,

REACTOR TTPE: PWE i

DESIGN "F"TRICAL RATINc: 805 gue REACT 01 ACE: 6.8 yr VE3 DOR: Cambuetica Engineering ARCHITECT-ENG UZERS: Bechtel CPERATORS: Consumers Power Co.

I LOCATION: 5 atlee. =* h af Seeth Haven, Mich.

l DURATION: N/A i PUNT QPERATUC CONDITION: 100Z power i

l SAFETT TEATURE TTPE OF FAILmtE: (a) inadequata performance; (b) failed to start; g heedeinoperable; (d)

DISCOVERY METECDs during operation CCHMENT: See also 132943 4

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FRECURSOR DESCRIFTION AND DATA i

f MSIC Accessica Number 132958 Dates December 11, 1977 i

Title:

Complace Loss of Offsite Power occurs at Fa11sedes The failure sequence was

, 1. With the reactor at 100% power, the "1" switchyard bus de-energized because of a spurious signal from the bus stripping relay reemiting in a complete loss of offsite power and consequent loss of main condenser cooling.

2. The reactor was '1y cripped.

3. Both d4==al generators started and provided power to safety-related equipment.

1 1

Corrective actions

1. The "1" bus tripping scheme was modified to . w uo loss of the bus due to spurious action of the "1" bus stripping relay. The specific cause of the  %

stripping relay trip had not been determined. )

Design purpose of failed system or component:

1. Off-site power provides the preferred source of electric power to plant equip-j ment when the unit generator is not in operation. The condenser circulating

, water psamps are normally powered from the off-site power source.

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l j Unavailability of system per WASE 1400:* loss of offsite power: 2 x 10~'Jhr i

! Unavailability of component per WASH 1400 * -

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s Unavailabilitia are in units of per demand D-I. Failure rates are in j

units of per hour BR-

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CATECORIZATION OF ACCIDENT SEQUENCE PRECURSORS 4

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NEIC ACCESSION NUMBER: 132943 DATE OF 121: December 16, 1977

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DATE OF EVENT: Novembac 25, 1977

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SYSTEM INVOLVED: offsite power h CD:! PONE 3T INv0LVED: evitchyard bue "1" CAUSE: bus trip from taaknown causes

I SEQUENCE OF DrfEREST

reactor trip with loss of offsite power 1

l ACTUAL OCCUERENCE: reactor trip with lose of offsite power i

REACTCR 3A

E: Pelisades

. DOCKET stD15ER: 50-235

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l REACTCR TYPE: PWR

, DESIGN ELICTRICAL RATDIC: 805 MWe ,

REACTOR ACE: 6.8 yr f VENDOR: Combustion Engineering t

ARCHITECT-ENGINEERS: Bechtel l

OPERATORS: Consmare Power Co.

i l LOCATION: 5 miles eeeth of SomslkHaven, Mich.

l DURATION: N/A I

j FLANT QPERATDIC CONDITION: SSE power i SAFETT TEATURE TTFE OF TAILUEE (a) inadequate performance; (b) failed to start; g heade inoperables (d)

DISCOVERT METEOD: during operation CCHMENT: .. See also 132958 O

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FREC3501 DESCRIPTION AND DATA I

ESIC Accession Number: 132943 Date: December 16, 1977

Title:

Complete Loss of offsite Power occurs at P=14== des The failure sequence wast ,

1. During notaal operation with the reactor at 85% power, switchyard bus "1" bar- de-energized, causing a complete loss of offsite power and resulting in a loss of main condenser cooling water flow.

l 2. The reactor wee manually tripped.

3. Both diesel ganarators started and provided power to safety-related loads.

5 Corrective actions Nonet the cause of the "1" bus loss was under investigation.

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Design purpose of failed system or components l Cffsite power provides the preferred source of electric power to plant equip-

ment when the unit generator is not in operation. The condenser circulating

water psuspe are normally powered from the off-site source.

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1 Unsvailability of system per WASE 1400:* loss of offsite power: 2 x 10 8/hr Unavailability of component per UASH 1400:* -

Unavailabilitie are in units of per demand D-l. Failure rates are in ,

units of per hour int" -

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CATECOR7% TION OF ACCIDENT SEQUENCE PREC:TRSCRS NSIC ACCESSION NUMB Ms 130119 DATE OF LER: October 18. 1977 DATE OF EVENT: September 24, 1977

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SYSTE:t I::VOLVED: offsite power COMPONE:IT INVOLVED: switchyard "R" bus CAUSE: bus trip from unknown causes SEQUINCE OF INTEREST: loss of offsite power ACTUAL OCCURRENCE: reactor trip due to loss of offsite power REACTOR NA:23 Palisades DCCEET NID1SER: 50-235 REACTOR TYPE: PWR ,

DESIGN ELECTRICAL RATING: 805 NW REACICA AGE: 6.3 yr VE:iDOR: Combustion Engineering i

ARCHITECT-C;CINEERS: Bechtel CPERATORS: Constmars Power Co.

LOCATTON: 3 ut.las south of South Haven. Mich.

DURATION: N/A PLANT CPERATI:!C CONDITION: 100% power SAFETT TEATURE TYPE OF FAILURE: (a) inadequata performanca; (b) failed to start; made inoperable; (d)

DISCOVERT MET 1:0D: during operation COMMENT:

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i PRECUR501 DESCRIPTION AND DATA i

l NSIC Accession Number: 130119 Dates September 24, 1977

Titles A Complete Loos of Offsite Power Occurs at Palisades 1

The failure sequescu

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1. With the reactor at 100% power during an electrical scorn, the switchyard "1" bus de-energized ad caused a complete lose of offsite power and lose of main condenser cooling. ,

! 2. The turbine tripped on high condenser vacuus.' and effected reactor and generator tripe.

. 3. Both diesel generators started and supplied electric. power to safety-related 1 1 loads.  !

j Corrective actions none I

i Design purpoos of failed system or component:

1 Off-site power provides the preferred source of electric power to plant equipment when the unit generator is not in operation. Tha condenser circulating wecer psesps are moraally powered from the offaite power source.

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Unavailability of system per WASH 1400:* loss of offaite power: 2 x 10 s/hr j Unav=41 aility of component per UASE 1400 * -

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-l Unavailabilitie Failure rates are in .;

units of per hour HR y are in units of per demand D*l. -

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CATECORIEATION OF ACCIDENT SMUENCE PRECURSORS NSIC ACCESSI0lf NUMBER: 97578 DATE OF LIR: November 15, 1974 DATE OF E7ENT: October 17, 1974 SYSTEN INTOL7ED: Offsite Power i

COMPOWENT INVOLVED: Differential Ralays CAUSE: Ieproperly desiseed Differential Relay System SEQUENCE OF INTEREST: Loss of offsite power ACTUAL OCCURRENCE: Loss of offeite power during SIS testing REACTOR NAME: P=Hamdaa DOCIET NUMBER: 50-255

REACTOR ITFE

PWR 1

I i DESIGN ELECTRICAL RATINC: 805 MWe BEACTOR AGE: 3.4 ye i

VENDOR: Combustion Engfaaering l

ARCNTTECT-ENGINEERS: Bechtal r OPERATORS: Consumers Power Coupesy i

l f4 CATION: 5 miles soutir of South Haven, Mich.

I DURATIGi: N/A PLANT OPERATING CONDITION: hot standby f SAFETT FEATURE TTPE OF FAILURE: inadequate performance; (b) failed to start; i , (c) de inoperable; (d) l DISCOVEIT METHOD: testing

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CottG2ET: This event is the second of its type at Paliandes. After the first event (NSIC 71694, 5/17/72), the differential relays were removed panding a design review. They were reinstalled in January 1974 after modification.

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PRECURSOR DESCRIPTION AND DATA -

i NSIC Accesalon Number: 97578 i Dates November 15, 1974

Title:

Loss of Offsite Power During SIS Testing at Feliendes 1

The failure sequence was:

1. With the plant in a hot standby condition, the right channel safety injec-tion test button wee pushed to initiate a quarterly test.

2. Offsite power une lost due to the inadvertent operation of the differential relay system.

3. The diesel generecers started and powered safety-related loads.

i Correceive actions s

The three-phase differential relays were removed from service pending a review and potential redesign of the system. Over-current protection devices r - h d installed to provide transformer protection. . -

Design purpose of failed system or component:

i The differential relays provided overcurrent protection for the startup j transformers.

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Unavailability of system per WASH 1400:

• ffsite O power 2 = 103/hr i Unavailability of component per WASH 1400 -

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Cnave11 abilities are in units of per demand D'I. Failure rates are in ~

units of per hour .HR-1 -

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CATEGORIZATION OF ACCIDElrf SEQUENCE PRECURSCES f

N5IC ACCESSION NUMBER: 71694 DATE OF LER: May 26, 1972 DATE 07 ETENT: May 17, 1972 SYSTEN INTOLVED: Offsite power 1

COMPONENT IETOLVED: Startup transformer CAUSE: Improperly chosen differential relay SEQUENCE OF INTEREST: 14ss of offsite power ACTUAL CCCURRENCE: Iose of offsite power during testing REACTOR NAME: Palisades DOCERT NUMBER: 255 REAC'3R TTFE M j DESIGT ELECIRICAL RATING: 805 sue REACTCE AGE: 1.0 ye 7ENDOR: CE s

ARCHITECI-ENGINEERS: 3echtel OPERATotS: Consteners Power Corp.

LOCATION: 5 miles south ed Semek Haven. Michigan i DURATION: N/A i

Pu NT OPERATING CONDITION: Hot standby a

SAFETT FEATURE TTFE OF FAILURE: ( inadequate perforrance; (b) failed to start; made inoperable; (d)

! DISCOVERY MEISOD: During testing COMMENT: The incompatible current transformer had been installed prior to criticality i

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PRECUR$0E DESCRIPTION AND DATA 4

N8IC Aceseetos Ihamber: 71694 Date: Jesy 26,1972

!' Titles less of offsite Power at P=14=ad==

N failure seepsence was

1. With the piare in a hot standby coidition, the left eh===al safety injectica system test buttaa was pushed to initiate a quarterly test.

2. This resulted in a loss of offsite power due to the spurious operation of a differentiel ralg.ov the 1-2 startup transformer. h actuation of the relay was dua to taabelanced sensing currents from a current transforms.? a casult of the incosoecibility of the installed current transformar with e.he 345 IT - 2.4 17 step-dows situation.

Corrective action:

h differsatial rela 3s were removed from sesrtup trsusformers 1-1 and 1-2. High-eide overcurrent and instantaneens overcurrent relays were installed.

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, Design purpose of failed system or component; I

h differsettal relays provided overcurrent protection for the startup transformer.

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Unave11ab111;y of system per WASH 1400: Offsite power 2 = 10-5/hr Unavailability' of component per WASH 1400

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1 CATEC0112ATION OF ACCIDENT SEQtTENCE FRECU13033 l

• MEIC ACCESSION NUMBER: 65969 l 4

DATE OF LIR: September 16. 1971

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, DATE OF EVENT: September 8. 1971 l i

SYSTE:t I3YCLVED: Raector Coelast System. Reestor Protective System i t 1

CIM!POKENT INVOLVED . Electromatic Relfaf Yalvee CAUSE:

l Opeatag of reactor protective system supply breakers resulted La opening of electromatic relief valves i SEQUENCE OF INTIREST: ses1114CA (open electreestic reliaf valve) i 1

ACTUAL OCCUERENCI: oyes electroestic relief valve iEACTex we, F. n m.. '

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DOCIET NUMBER

50-231 l i

REACTOR TTFE FWE DESICN ELECTRICAL RATING: 805 MWe j IEACTOR ACE: .3 yr j VI3D01: Combustica Engineering I 1

ARCRITECT-ENCINEERS

Bechtel OPERATORS: '

q Comeumere Power Co.

LOCATION

3 milesbeh 4.3ameL EmTea Mich. l l

1 DURATION: N/A '

i FIJlf OPERATIEC CONDITIDE: hoc ehutdove I 1

l SAFETT FEATURE TEFE OF FAZIREE (a) landestusta performance; (b) is11ed te start; l

(c) made inoperable; h failed open l i DISCDVERT METP.00 purias operacios 1

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_ N E DESCEIFTION AirD DATA ESIC ArWAa Numbers 65M9

! Date September 16. 1971 1

Titla

Unclear Wiring Diagreme Iseult in Depreneurisation at Palisades i

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h failure sequence wees

1. With the plaat in hot shutdown, a tacha1cima de-energized breakers to the reactor protective syntes. This resulted in loss of power to the electroustic r=Haf velve pilot valve solenoids opeatas the valves.

i 2.

One of the electromatic relief valves wee imelates because its isolation valve was c3asedt bewever, the open, unisolated relief valve po m itted EC3 blowdown to the quench task.

3.

Safety injectice wee initiated on both safety injectica channels, however channel A ves blocked by the operecer.

4 The operator closed the electromatic relief valve isolacios velve and started j the third charging pteep.

(Continued on attached sheet)

1. The teactor protective system drawings were to be cortected to indicate the "aa-built" plaat conditica ustag standard notation.

2.

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The electromatic reiiaf valve control schema vae to be reviewed to determine if any changes were desirable to lasses the probability of a second incident.

i j Design purpose of failed system or a v - t 1.

The electromatic relief valves provide ICS overpressure protecnica at a pressure below the BCS safety valve set pointa.

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Unavailability of system per WASE 1400 * -

i Unavs11 ability of component per VASE 1400:* relief valve, failure to reseat: 10'*/D i

i e l Uneve11abilitis per hour RE"y are in units of per demand D*I. Failure rates are in unita of . ,

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CATEGORIZATION OF ACCIDENT SEQUINCE PRECURSCRS NSIC ACCESSION NUMBER: 61545 DATE OF LER: September 9, 1971 DATE OF EVENT: September 2. 1971 STinTEM INVOL7ED: offaite power; emergency on-site power l

COMPONENT INTOLVED: switchyard breaker relay'. diesel generator output relay CAUSI: breaker failure of switchyard brasher relay, failure to close for diesel generator

, SEQUENCE OF INTEREST: loss of offeita power ACTUAL OCCUIUtENCE: Loss of offsite power and failure of a diesel generator to load REACTOR NAME: Fati -taa DOCIET NUMBER: 50-235 REACTOR TTFE FWR DESIGN ELECTRICAL RATING: 805 MWe 2

j REACTOR AGE: .5 yr I

7ENDOR: Combustian Engineering ARCHITECT-ENGINEERS: Bechtal OPERAIORS: Consumers Power Campsey .

LOCATION: 5 milee south of South Haven. Mich.

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DURATION: N/A FLANT OPERATING CONDITION: not known

) SAFETT FEATURE TYPE OF FAILURE: (a) inadequate performance; (b) failed to start; ande inoperable; h . failed to loed.

DISCOVERT METHOD: during operation COMENT: -

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PRECURSOE DESCRIPTION AND DATA i

NSIC Accession Ntamber: 61545 Date: September 2, 1971 l

Title:

Loss of offsite Power and Failure of a Diesel Generator to Load at Palisades The failure sequence was:

i 1. The Argenta No. 2 345KY line (one of three 345KY lines at the switchyard) tripped.

! 2. Failure of a " breaker failure relay" associated with the tripped line breaker l

resultad in the tripping of two other breakers on the ring bue, which caused a

'l lose of offsite power.

3. Diesel generator No. 1 started and loaded its safety-related bus.

4. Diesel generator No. 2 started but its breaker did not close util the operator j adjusted the eyechroecope in properacios to close the breaker assually.

Corrective action:

Not specified (only LIE abstract available).

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Design purpose of failed system or component:

Offsite power provides electrical power to safety-related components whom the satic generator is inoperable. The diesel generatore provide a standby source of electric power for safety-related components when both the mit generator and the offsite power sources are not available.

~8 Unavailability of systes per WASH 1400

• Offsite Powert 2 = 10 /hr

~2 Unavailability of component per WASH 1400

• Diesel-generator: 3 x 10 /D Unave11 abilities are in units of per demand D~1 Failure races are in units of per hour H14, .

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r CATEGORIZATION OF ACCIDENT SEQUSCE PRECURSORS NSIC ACCESSION NUMBER: 39024 l

DATE OF LER: March 3, 1972 DATE OF EVETT: Jaenery 15, 1972 SYSTDI INVOLVED: electric power COMPONDIT INVOLVED: relays and circuit breakers CAUSE: line faults induced by a violent storm SEQUENCE OF INTEREST: loss of offsite power ACTUAL OCCURRENCE:

loss of offsite power REACTOR NAME: Big Rock Point DOCKET NUMBER: 155 .

REACTOR TYPE: But DESIGN ELECTRICAL RATING: 7T !We REACTOR AGE: 9,3 yr i

VUfDOR: General Electric ARCHITECT-ENGINEERS: Bechtal i

OPERATORS:Consumere Power Company WCATION: Four miles NE of Charlevoix, Mich.

3 DURATICH: N/A PLANT OPERATING CONDITION: just scrammed SAFITY FEATURE TYPE OF FAILURE: (a) inadequate performance; (b) f ailed to start; nade inoperable; (d)

DISCOVERY METHOD: operational event CCNMENT: -

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D-20 FRECURSOE DESCRIFTION AND DATA NSIC Accession Nuebert 39024 Date: Josuary 23 , 1972

Title:

Loss of Off Site Feuer and Other Failures At Big Anck Point.

The failure sequence was:

1. " Galloping conductors" caused line faults which resulted in the failure of the Gaylord 388 oil circuit breaker (OCB).

2. Other breakers at the substation acted to clear the fault. This, however, left the plant in a no load condition, resulting in a turbine trip on overspeed and subsequently in a reactor trip due to high neutros fluz.

3. The 199 OCS vm manually opened since the 138 kw line ves de-energized inter-mittently Osaspecified reasons) for 20 minutes.

4. The station transferred to the M kv bs.ck-up line, however, the breaker tripped.

A stuck contact of an instantaneous overcurrent relay (1288 OCB) in conjunction with operation of the indervoltass bus fault relay caused the 4 kv line to de-

• • • ** (see attached pass)

Corrective action:

1. An inspection of the transmission lines indicated they had received no damage during event.

2. The faulty trip coil (388 OCB) and the faulty overcurrent relay (1288 OCB) were repaired.

Design purpose of failed system or component:

1. Offsite electric power (both normal and backup) provide power to the station when the unit operator is not inservice.

4 I 2. Belays and circuit breakers are provided to protect electrical components fron excessive and insufficient current and voltage conditions.

i LOOF 2 x 10sygg, Unavailability of system per WASH 1400 *

\ Unavailability of component per WASE 1400

• circuit breakers 11m. # 3/D x 10' /D l relsys l

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units of per hout ER*1, er e--e e-N

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The failure sequence west (continued)

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5. UPea luse of both offeite lines the diesel generator started ar.-

the 2-3 bus.

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6. The 138 kv line was re-energized approximately 20 minutes af ter the turbine j trip. Attempts to close the 199 OCB failed due to falso trip signals from the audio relay equipment.

7. The audio relay equipment wee overridden and offsite power wee restored.

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c.n.,.i ome.ss 212 West WChig.n Avenue. Jackson. Michigan 40201. Area Code S17 788-OSSO December 9, 1977 .

Mr James G Keppler Office of Inspection and Enforcement Region III US Huclear Regulatory Ccemission 799 Roosevelt Road Glen Ellyn, IL 60137

o DOCKET 50-255 - LICENSE DPR PALISADES PLAITT 77 4 LOSS OF OFF-SITE POWER

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On December 7, 1977 G Petit,iean of our Palisades Plant staff contacted D Hunter and requested an extension of one week for the submittal of

""-77-5'- "*** " " ' ""* " **** ' ' " * " ' ' * " * " ' ' " ' * " " '*="*= 8' O 1977. The written Licensee Event Report will now be submitted on December 16, 1977. .

11 1 DavidP,Hoffman(Signed)

ID David P Hoffman Assistant Nuclear Licensing Administrator CC: ASchwencer, USIGC 1

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oq,g Company jPower

%u a.n.r.: ome..: 2 2 ws. usenigen 4..no., a.ca.on, uienigen .osoi . Ar.. coa. sir 7es-oeso December 16, 1977

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Mr James G Keppler Office of Inspection and Enforcer.ent Region III US Nuclear Regulatory Ccz:: mission T99 Roosevelt Road q Glen Ellyn, IL 60137 T DOCKET 50-255 - LICENSE DPR PALISADES PLAUT - ER-TT LOSS OF C OFF-SITE POWER AND PLANT TRIP Attached is a reportable occurrence related to the loss of off-site power and subsequent plant trip at the Palisades Plant.

i t David A Bixel (Signed)

Lf7 David A Bixel

-. Nuclear Licensing Administrator LU CC: ASchwence , USNRC N

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e - c UCENSEE EVENT REPORT Palisades CONTROL BLOCK:l l l l l l l (PLEAss palNT ALL REGutRED INFORMAT10N) 1 6 AM UCENSE NUM8ER P TYPE

-[ l M l A lP l A l L l 114l l 0 l 0 l-l0 l 0 l0 l0 l0 l-l 0 l 0 l l hl 1 l1l1l1l 30 l0l1l 32 7 89 15 25 26 31 T Ea5 ccCxst NuMeER EVENT DATE l1 l1 l2 l 5 l7 l 7 l REPORT CATE l 1l ?l1 l6 l 7 l 7l hCCN'T 57 l

• l *58l [T_j lL l l 0 l 5 l 0 l-l 0 l 2l 5l 5l 00 59 60 61 68 69 74 75 7 6 EVENT DESCamTION g l During non::al steady state operation, the "R" bus became de-energized ausing a com- 80l 7 89 0 l plete loss of off-sitie power and resulting in a loss of main condenser cooling vater. l 80 7 89 lifM l The reactor was manually tricted. The crimarv olant was stabilized in the hot con- l 80 7 89 SE l dition and was borated. LCOs of Tech Snee 3.1.1, 3.7.1 vere exceeded. Event similar l 7 89 60 33 l to ER-77-Oh7. (Contd on attached sheet. ) l paus 60 7 89

'N No'E COMPONENT CCCE VOLATION O lEIAl lFl lZ l Z l Z l Zl Zl Zl lZl l Zl 9l 9l 9l lY l 17 43 44 47 48 27 89 10 11 12 CAUSE DESCRIPTION "R bus loss not known and is still under investigation. l

] l Cause of 00 3'89 c.7. 89 I

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/11 I ST  % POWER OTHER STATUS O oVERY CISCOVERY CESCRIPTICN 1

9 W l0l8 l5 l l N/A 10 12 13 l

44 W

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7 8 RELEA O CF AMOUNT OF ACTMTY LOCATCN OF PELEASE l Secondary water to atmosrhere.

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[jijLEASE l 3h h (Contd belov ) l 10 11 44 45 80 l

PERSONNEL EXPOSURES NUMBER TYPE CESCRIPTCN LQ l0l0l0l Ip l N/A l 7 89 11 12 13 80 N PERSONNEL INJURIES NUMBER DESCRIPTON g[ l0l0l0l l N/A l 7 89 11 12 80 PROBABLE CONSEC.UETCES .

N/A l EE l 60 7 89 LOSS OR DAMAGE TO FAC:UTY TYPE DESCRIPTON 1

7 89 dl 10 N/A 60 l

PUBLICITY DE l N/A l 7 89 8 60 ACCITIONAL FACTCRS

\,J l ( Amount of Activity - Contd) nicrocuries of I-1T1: 29.2 microcuries of T-13't. l 7 89 t0 119 l l 7 89 60

Attachment to Licensee Event Report 77-055 Event Description (Contd)

During the incident, a primary coolant leak occurred in the letdown line. The leak was isolated. Both diesel generators operated normally to supply electrical power during the incident. Electrical power was restored in.2.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

The delay in reporting this event was discussed with Region III. See our letter dated December 9, 1977 (ER-77-55) 1

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'- Region III

,g US Nuclear Regulatory Co=sission 799 Roosevelt Road Glen Ellyn, IL 60137

- DOCKE2 50-255, LIC'5SE DPR PALISADES PLANT - ER-77-05h, ER-77-057 and ER-77-058 .

Attached are three reportable occurrences for the Palisades Plant. Event LO Report 77-058 was a pro =pt reportable event that was identified by T.7X dated December 12, 1977 <

in David P Hoff=an (Signed)

N David P Hoff=an Assistant Nuclear Licensing Administrator CC: ASchwencer, USNRC I

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UCENSEE EVENT REPORT. Palisados CONTROL BLOCX:l l l l l l l (pLEASE PRINT ALL REGutRED INFORMATICN) 1 8 .

EE ucENst Nuusta iYS' Yp"t Q l Ml Il Pl Al Ll1l l0l0l-l0l0l0l0l0l_l0l0l l hl 1l1l1l1l l0l1l 7 89 14 15 25 26 30 31 32 Y ErE ooCxty Nuweta .

EvEur caTE aEpon? OATE O 1 CCN'T l' " l

• l l ml W l n l c: l o l-l n I ? l c; l c: l l1l ? l1l1l 7l 7l l1 l2 l2 l1lT l T l 7 8 57 58 59 60 61 68 69 74 75 60 EVENT DESCRIPTION HE l Durine nor s1 steady-stat'e operation, the 'R' bus became de-energised, causing a l 7 89 CO h l -~- 1 ac a loss of nffs'ite never and resultine in a loss of cain condenser cooling l 7 89 80 0 l.,,,+._ mw . . . . + m. v o n -am. 11v +,.M,ad, mk. eri :ary niant 'vas stab 414.. A 4- +w kat I 7 89 80

".1D l condition and was borated. Technical Specifications 3.1.1.a and 3.7.1 were violated l 7 89 EO 6l h*k Adasal canarators onarsted nor'slly to surely electrical power during the 1.5 l 7 89 p, vE , SO

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NI $co'E CcupoNENT COCE F A VCLATCH (ER-77-58) gg l El A l Q l Zl Zl Zl Zl Z l Z l lZ l lZ l9 l9 l9 l lY l ,

7 89 to 11 12 17 43 44 47 48 Q CAUSE DESCRIPT1CN O6 l The cause of the 'R' bus loss is unknown at this. time. To mitigate the consequences l 0

f of future 'R' bus losses, the tripping scheme for the 'R' bus has been codified so 7 89 60 1o l that a signal frc: the stripping relay (h86 s-x) vill trip neither the 3h5 KV supply , p 89 racvTY utTwoo op (cont'd) 60 STATUS  % POWER OTHER STATUS CISCOVERY OtsCOVEAY DESCAiPTCN NA 1 lEl 1110101 l NA I l al i l 7 8 9 10 12 13 44 45 46 CO CT Cc4 TENT RELEASED aucuNT QF 4CTMTY LOCATON OF AELEASE 12 l Gl OF WRELIASE l 15 Tcont di l l secondary water to at=csphere l If> 8 9 10 11 44 45 80 PERSONNEL EXPOSURES

.= NUMBEA TYpf CESCA'PTON ,

J,A 13 l0l0l0l gl l If 89 11 12 13 80 PERSONNEL INJURIES NuMEER CESCAipTCN 0@1 7 89 inl0101 11 12 1 NA 80 l

PROBABLE CONSEO,UENCES ,

gl NA l 7 89 60 LOSS OR OAMAGE TO FAC:UTY TYPE CESCAPTON NA i 7 89 10 -

60 PUBLICITY

$l 7 89 NA 80 l

ACO!TICNAL FACTCAS l (1 onnt of Activity- cont'd) nicrocuries of I-L31; ".5 microcuries of I-133 l

, 89 60 MI l 7 89 tiu

Event Description, (cont'd):

hours that off-site power was not available. This event similar to ER 77-055 (ER 77-058)

Cause Description (cont'd):

breaker 27-R8 nor the lov side breakers feeding the 2h00 and h160 volt buses.

The 27-R8 breaker vill trip from the transformer bank differential relays, and thermal trips for the feeder breakers have been retained. This scheme retains electrical fault protection, yet vill prevent plant trips from spurious action of the 'R' bus stripping relay.

To determine the source o'f the 'R' bus stripping signals, the h86 S-X relay has been instrumented so that recorder traces can be obtained in the event future trips occur.

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w DfU o.ner.e onice.: 2i2 w..: uienig.a av.no..s.ca.on, uien.g.n .osoi . 4,.. coa. siv res.osso March 3, 1972 Dr. Peter A. Morris, Director Rei Docket 50-155 Division of Reactor Licensing DPR-6 (ZEK)

United States Atomic Energy

' Commission Washington, DC 20545

Dear Dr. Morris:

V This is intended to apprise you of the loss of an off-site Os power incident which occurred at our Big Rock Point Nuclear Plant on January 25, 1972. During the outage caused by the loss of off-site 4 power, a two-to-three-foot drop in the spent fuel pool water level O was experienced and one of two d-c operated emergency condenser valve (Mo-7063) failed after opening automatically as required by reactor i pressure conditions.

The loss of off-site power was attributed to a combination of unusually severe weather conditions and several equipment failures.

p On the evening of January 24, 1972, an intense storm system passed through the area which consisted of rain that later changed to v heavy snow as the te=peratures fell. High winds on the following day caused the ice laden power lines to dance and sway. This resulted in N a phenomenon called " galloping conductors" in which line faults occur-red as the lines move relative to one another. The location of these line faults was calculated to be approximately 10 miles from the Gaylord Substation and on the 138 kV power line (see attached sketch).

The Gaylord 388 oil circuit breaker (OCB) operated 12 times to clear the momentary line faults. On the thirteenth fault, the trip coil burned out and the 388 OCB failed to operate. Relays in substations feeding into and out of the Gaylord Substation caused breaker operations at their respective locations to clear the fault. As a result, the Big Rock Point Plant became momentarily isolated from the rest of the system and with essentially no load on the generator the unit tripped off on overspeed (116 OCB opened). This occurred at approximately 1304. The reactor scrammed due to high flux.

Since the fault occurred on the Gaylord side of the Emmet 488 OCB, a trip signal was not sent to the Big Rock Point 199 OCB O end e 1oea reaection dia not occur. acwever, the 199 OCB wes opened manually since the 138 kV line was de-energi::ed intermittently over M2G34

Dr. Peter A. Morris 2 March 3, 1972 O

a span of approxi=ately 20 minutes. Nomally, the station power would have automatically transferred on undervoltage to the 46 kV source to supply the station power equipment. However, a stuck contact of an instantaneous overcurrent relay in the Emmet 46 kV bus protection relay scheme coupled with the operation of the undervoltage (UV) bus fault detector relay (which would have reopened had the fault cleared within a few cycles - fault lasted for 69 cycles) caused the 1288 OCB to trip and thus de-energize the 46 kV line to Big Rock. The two relays are connected in series and both must be closed simultaneously for a few cycles to cause the breaker to trip. The line was de-energized for approximately two hours until repairmen, who were hampered by consid-erable blowing and drifting of snow, could make the essential repairs and return the system to normal. The overcurrent relay had failed closed prior to the undervoltage condition occurring; it should not have operated under the conditions of the incident.

Upon the loss of both the normal and backup supplies of

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an station power,.the diesel generator started and closed onto bus 2-B to provide power for operation of essential emergency equipment.

Approximately 20 minutes after the turbine trip (1324), full 4 potential was restored to the 138 kV line. However, when attempts were o made to reclose the 199 OCB, a false tripping signal (from the audio tone

. relay equipment) caused the breaker to i= mediately retrip. The audio s' tone control was then put into the "off" position defeating the tripping signal and the 199 OCB was then reclosed successfully. The tone controls were then reconnected and station power loads were returned to their 4 normal station power supply at 1353 The diesel generator was not shut down until approximately 1507 Extensive inspection of the line section between the calculated T fault location and the Gaylord Substation did not produce any evidence of damage. The faulty trip coil (388 OCB) and the faulty overcurrent N relay (1288 OCB) have been repaired. The false tripping signal observed o in the tone relaying equipment is being investigated as is the trip scheme for ths 199 OCB.

In summary, the simultaneous loss of the normal and backup off-site station power supplies was caused by extremely severe and unusual weather conditions and two equipment failures. The length of time that off-site power was lost was extended by difficulties in get-ting substation operators to the substations and further compounded by a false tripping signal in tone relaying equipment. It is not considered possible that a plant-initiated event would cause a loss of the normal station power supply because of the size of the plant with respect to the system size. The plant provides up to 71 We (net) to Michigan Power Pool system of approximately 10,000 We, The 46 kV station power supply was installed in March 1968 to h provide a redundant station power supply to the Big Rock Point Plant.

Since that installation, the loss of off-site power experienced January 25, 1972 is the only instance where both off-site power supplies werc not 542635  ;

Dr. Peter A. Morris 3 March 3, 1972 O available at the same time. From the review of this occurrence, it is considered highly unlikely that a similar incident will ever occur again.

During the reactor pressure transient resulting from the turbine overspeed trip on January 25, the two emergency condenser out-let valves (MO-7063 and MO-7053) opened automatically as designed.

Approximately two and one half hours later, when an attempt was made to shut MO-7063, the valve failed to operate. The emergency condenser inlet valve MO-7062 in the affected loop was shut to maintain reactor pressure. MO-7053 operated normally.

The motor unit of MO-7063 was disassembled and taken to a local motor rewinding shop for inspection and possible repair. The insulation on shunt and series fields was found burned to the extent that it required rewinding. However, there were no grounds or shorts found in either winding which suggested the burnup may have been caused 4 . by excessive running of the motor beyond its shutoff limits. Upon re-installation of the motor, the torque switches were inspected for pro-

& per settings. The open torque switch was adjusted to allow for a wider margin between the valve automatic stop and manual stop.

It was concluded that the failure of the valve operator motor o was probably due to an improperly set torque switch. A replacement

,> motor for the valve operator has been ordered. When it arrives, it will be installed and the present motor returned to the vendor for a detailed inspection. This motor failed and was revound once previously.

The previous failure occurred while maintenance was being performed 4 on the motor and was due to a maintenance error.

Os The'first indication of an abnormal spent fuel pool wate?

T level was at 0100 on January 26, 1972 when the Control Room Operator observed a gradual increase in the fuel pool area monitor readings N (approximately 12 mR). A visual inspection followedi The fuel pool O water level was discovered to be two to three feet below the normal overflow to surge tank level. Detailed investigations of this de-crease in fuel pool water level revealed the cause to be a siphoning action created by the piping configuration and valve alignment.

At the time the outage occurred, the fuel pool was valved for recycle through the radwaste system. However, upon loss of the normal station power, the fuel pit, radwaste and treated was% pumps ceased to operate and the fuel pit to radwaste isolation valves CV-14027 and CV 14117 closed automatically.

The critical elevations in the piping sequence are as follows:

1. Discharge Pipe at Bottom of Pool - 601'-6" h 2. Highest Elevation of Pipe at pool Surface - 630'-6" 542636

Dr. Peter A. Morris 4 March 3, 1972 0 Elevation of Sphere Isolation Valves - 593'-6" 3

4. Elevation of Pipe Where It Opens Into Clean Waste Receiver Tanks - 590' the hydraulic head was available to create a siphoning action Thus,6" (11'- head) from the fuel pool to the clean waste receiver tanks when the isolation valves were reopened. (See Attachments 2 and 3.)

With the loss of two to three feet of water, equipment hanging on the side of the pool such as flange protectors, sample specimen racks and vacuum hoses were exposed to some extent. A fuel bundle was located in the elevator which was at the top of its vertical travel. The elevator has a pneumatic stop at seven feet below normal water surface and a mechanical stop at six feet. The decrease in water shielding allowed an increase in radiation dose rate in the vicinity of the fuel pool.

N . .

After the discovery, the valve to radwaste ("T" handle valve

& in fuel pit pump room) was closed and the fuel pool level restored via the waste hold tanks.

o Two to three feet of water represents approximately 7,800 to

,. 11,700 gallons. Since water was being processed to the condensate storage tank and the reactor valved for blowdown, the increasing amount of water to the radwaste system from the fuel pool was not immediately evident. Each of two clean waste and waste hold tanks has a high level 4

alarm set at approximately 90%. The condensate storage tank high. level alarm annunciates at about 95%. The storage tank level is also re-p corded on a chart on the front panel of the control room. Since these tanks are monitored frequently, it is inconceivable that the tank levels v would have continued to rise unnoticed. Even assuming that they did, the spent fuel pool area monitor is set to alarm at 15 mR. It was N reading approximately 12 mR and rising slowly when low pool level was first detected.

O To prevent any further reoccurrence of the problem, a siphon breaker has been added to the inlet piping of the spent fuel pool.

Three 7/16" holes were drilled in a horizontal pattern on the vertical section of inlet piping about 2" below the normal fuel pool level. On February 2, a test of the siphon breaker was conducted under conditions similar to those of January 25 and the siphoning action ceased when the pool level reached the three holes.

The addition of the siphon breaker eliminates any potential for losing the spent fuel pool water level. However, assuming that a loss of fuel pool water level were to again occur, the spent fuel pool l

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Dr. Peter A. Morris 5 March 3,1972 O . area monitor remote indicators and alarms and radwaste system tank level indicators and alarms would provide early notice of the ab-normal condition.

Yours very truly, Ralph B. Sewell (Signed)

RBS/dmb Ralph B. Sewell Nuclear Licensing CC: BHGrier, Administrator USAEC so ~~

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September 9, 1971 Dr. Peter A. Morris, Director , Re: Docket No 50-255 Division of Reactor Licensing License No DPR-20 US Atomic Energy Comission Washington, DC 20545

Dear Dr. Morris:

• This letter is written to apprise you of a recent incident g involving the emergency power supply at this facility.

'; At the time the incident occurred, the plant was in a cold shutdown condition with primary coolant system at refueling boron concentration and atmospheric pressure. The shutdown cooling system O.. O .

was in operation.

The 345 kV Argenta No 2 line tripped at 0706 on Septe=ber 2, 1971. A breaker failure relay operation on the 27R8 air blast breaker tripped' breakers 29H9 and 25R8, clearing the 'R' 345 kV bus and resulting in the loss of power to the start-up transformers.

M (Refer to attached Drawing E-501.) The main bank has not been routinely energized during the current low power operation resulting C in a single source of off site power.

' LO The emergency diesel generators both s :arted and quickly M. achieved rated speed with the 1-1 unit properly closing in on the dead 2400 V bus and picking up load as designed. (Refer to attached Drawing E-1.) However, the 1-2 diesel generator unit failed to close in on the dead bus automatically until the operator turned the synchroscope plug in preparation for closing the breaker manually.

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Dr. Peter A. Morris 2 US Atomic Energy Commission September 9, 1971 O CausE OP INCrDEnT The cause of the loss of power to the station was determined to be a faulty breaker failure relay on the 27R8 breaker which caused the 'R' bus to be isolated. The exact cause of the diesel generator failure to close automatically on the dead bus is still under investi-gation at this time.

CORREt;nV8 ACTION PLANNED A review of the viring scheme and actual installation is in Progress to determine the exact mode of failure. Test operation under simulated loss of offsite power vill be conducted to determine any deficiency and/or demonstrate the operability of the complete system.

Yours very truly, e eA N

IMH/ERC/mho Robert L. Haueter

- ' Electric Production Superintendent - Nuclear Oo cc: anorier USAEC Compliance, COO to C

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September 16, 1971 Dr. Peter A. Morris, Director Re: Docket No 50-255 Division of Reactor Licensing License No DPR-20 US Atomic Energy Commission Washington, DC 20545

Dear Dr. Morris:

O This letter is written to apprise you of a recent incident g involving the primary coolant system at this facility.

5 At the time the incident occurred, the piant was in a hot shutdown condition with the primary coolant system at refueling boron QO concentration, 532 F and 2100 psia. Three of the four primary coolant pumps were in operation and preparations underway to bring the reactor critical for operator training.

At 1335 on september 8, 1971, a technician de-energized the breakers to the reactor protective system to install a minor modification. This act de-energized the feed to the electromatic re-M lief valve pilot valve solenoids allowing the valves to open.

C The primary system pressure decreased to a low point of tf) approximately 1280 psia over a period of 2 - 3 minutes until the blowdown was terminated by closure of the motor operated valves W used to isolate the open relief valve.

The system pressure and temperature were back to normal in approximately one hour.

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Dr. Peter A. Morris 2 US Atomic Energy Commission September 16, 1971 The rate of depressurization of the reactor vessel and the pressurizer approximated the design rate for a pressurizer safety valve operation. The number of occurrences allowed for in the design was 200. Therefore, we conclude that the integrity of the primary coolant system has not been compromised by this incident.

SEQUENCE OF EVENTS 1335 Breaker #13 on panels Y10 and Y30 de-energized to allow for a minor modification to the reactor protective system. Refer to attached drawing E-8.

Relays K1 and K3 which feed the electrically operated pressurizer relief valve bi-stables (RV1042B and RV1043B) de-energized.

1335 RV1043B opened and released steam to the quench tank starting a decrease of primary system pressure. RV1042B had previ-

~ ously been isolated by means of the motor operated isolation valve and did not pass steam during the incident. Refer to y attached drawing M-201.

? 1336 safety injection signal (sis) actuated on both channels, however the A channel was blocked by action of the operator.

]O upon realization of the cause of the pressure drop. Safety injection pumps started.

1336 Operator started MOV1043A closing.

1337 Charging pump P55c manually started.

M 1337 System pressure decay turned around at a low point of 1280 C psia as MOV1043A closed stopping flow of steam to the quench tank via RV1043B.

Lf) 1338 stopped safety injection pumps.

N 1343 Breaker #13 on panels Y10 and Y30 closed and power restored to relief valve control circuits.

1344 Returned charging and letdown to normal.

1400 As the system pressure recovered to a point above 1700 psia, the operator reset the safety injection channel B and returned all equipment to normal.

1430 Frimary coolant system pressure and temperature back to normal.

The coolant te=perature and pressurize level swings of 30 F 0

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Dr. Peter A. Morris 3 US Atemic Energy Co::: mission September 16, 1971 O

and 4 percent respectively are considered insignificant. -

CAUSE OF INCIDENT The basic cause of this incident was the non-standard desig-nation of contacts (as used by the architect engineer on the drawing) in the control circuit to the power operated relief valves. The technician was misled by the 'a' contact designation as shown on the drawing when in fact the circuit is vired using 'b' contact (s).

This led him to believe the relief valves would not open when the power was removed by opening breaker #13 on the panels Y10 and Y30. -

CORRECTIVE ACTION PLANNED

/

The drawing (s) will be corrected to indicate the "as built"_ _,

/'

condition using standard notations.

ps A review of the control scheme design will be conducted by Consumers and the Reactor Supplier personnel to determine if any '

changes are desirable in order to lessen the probability of a second incident. _,

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The plant operators will be furnished with revised procedures ,<' ,

hO indicating steps to be utilised to minimise the blowdoun if it should occur in the future.

Yours very truly, _

s.

I R L Hauct:r -

-f 00 r r IMH/ERC/mho Robert L. Haueter , 7I If1 Electric Production , '.

Superintendent - Nuclear -

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Power company Generse Offices: QM West Mich4 gen Avenue, Jackson. Michigan 49201. Area Code 517 788-0550 September 29, 1971

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s, Dr. Peter A. Morris, birector

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Re: Docket No 50-255 I.

1. Division of Reactor Licensing -

License No DPR-20 US Atomic Energy Co::: mission -

Palisades Plant Washington, DC '

,20545 - J<

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Dear Dr. Morris:

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This lettst vill confirm our recent discussion with the

, 'p ~l USAEC Compliance OHice, Chicago, and provide supplemental informa-O tion for our letter dated September 9, 1971 concerning the incident involving

/

the emerginney power supply at Palisades.

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.. - , - ' - ,,s " An inyestigation of the incident revealed that all four

]'d 345 kV lings into'the substation were energized at the time of the incident c4 L' hat lightning faulted the 3k5 kV Argenta No 2 Line.

The Argenta end of the line and one breaker at Palisades cleared i

s properly. The 27R8 air blast breaker'at Falisades failed to trip by primary' relaying and after a few cycles the breaker failure relays operated to clear the 27R8 breaker and "R" bus. Relaying for these conditions was correct. This is contrary to our earlier to - information as reported in our September 9, 1971 letter.

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The '27R8 air blast breaker circuitry was tested and the

' breakers repeatedly tripped in an effort to again initiate the

[f)alfunction. This malftu:ccion could not be repeated. If further

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testing proves' futile, certain components of the primary relays y ,' , on this circuit will be replaced as a precautionary measure to

, possibly prevent a future occurrence of the above.

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,. A viring error was found to have caused the failure of

/ ,', the number 1 2 diesel generator to automatically close in on the v '; 'T dead 2400' volt: 1D bus after power was lost from the startup trans-i former. Contrary to the electrical drawings, the permissive con-

/ tacts to breaker number 152-213(dieselgenerator)werewiredto O f m c2e23

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Dr. Peter A. Morris 2 US Atomic Energy Commission September 29, 1971

'O " movable" type contacts rather than stationary contacts in breaker cell number 152-203 (station power to 1D bus). Thus, continuity was broken to the diesel start circuitry with breaker number 152-203 in the " racked out" position.

• Breaker number 152-203(stationpowerto1Dbus)was placed in the test position and caution tcgged thus, placing the

" movable" contacts into the circuit that will allow the diesel generator to close into the II) bus normally.

The wiring error will be corrected before power operation.

Yours very truly, R L Hm:: g

'l g ERC/mho Robert L. Haueter Electric Production a Superintendent - Nuclear CC: BHGrier

'? USAEC Compliance, 000 0'

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• At the time of this incident no power was being fed from the main transformer bank via the station power transformers because of plant status and the low power demand for station auxiliaries.

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.j c..n. . omc..: sts w..: unemo.n 4 .nu..s.ca.on, u.cmo.n 4o20: . Ar.a Coc. 517 788-0550 November 15, 1974 Mr. James G. Keppler, Director Re: Docket 50-255 Region III, Directorate License DPR-20 of Regulatory Operations Palisades Plant US Atomic Energy Commission UE-2-Th 799 Roosevelt Road Glen Ellyn, Illinois 6013T

Dear Mr. Keppler:

is Attached is an Unusual Event Report (UE-2-Th) covering a problem

. with the start-up transformer differential current protection relays. This unusual event is similar to the one reported on May 26, 1972. A prelimi-O nary investigation of the present problem has indicated that our recent

• O revision of the differential relay arrangement has not completely resolved the original problem. The investigation is continuing and we are confi-O ae=* **=* = eaea" *e revi=to= * **e airrere=*1 1 re1 71#8 =7=*e= e*= ae made and spurious operation eliminated. ,

Adequate protection of the start-up transformer is being provided by the overcurrent protective device.

T Yours very truly, f.O Ralph B. Sewell (Signed)

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DAB / map Ralph B. Sevell Nuclear Licensing Administrator CC: Directorate of Licensing

!!cf.I;C Washington, DC N

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O UNUSUAL EVENT REPORT Palisades Plant

1. Unusual Event: UE-2-74 2A. Report Date: November 15, 1974 2B. Event Date: October 17, 197h 3 Facility: Palisades Plant, Covert, Michigan

4. Identification of Occurrence: Start-Up Transformer Trip During SIS Test 5 Condition Prior to Occurrence: The plant was at hot standby.

6. Description of Event: The quarterly safety injection system (SIS) test was in progress with the left channel part of the test having been suc-00 cessfully completed. When the right channel test was initiated, off-site power was lost. The, diesel generators started and automatically closed to provide plant power. Off-site power was restored in about one half g how.

O 7 Designation of Apparent Cause of Occurrence: The installation of the current transformers associated with the differential relays appears to be the cause of the problem. This is complicated because of incom-patability problems associated with 345 kV to 2.h kV step-down situation.

8. Analysis of Occurrence: Operation of the SIS vould have been dependent V on the plant emergency diesel generators. The SIS is designed to oper-

_ ate off of the diesel generators and would have performed properly. l g 9 Corrective Action: The three-phase differential relays have been removed from service pending an investigation by our Relay Protection Department. ,

gq .Overcurrent protection devices remain installed and provide adequate transformer protection. Even so, because of the added transformer pro-tection available utilizing a differential relay system, we plan to rein-stall the differential relay system when a suitable design can be achieved.

The Safety Audit and Review Board vill conduct a review of the differen-tial relay system and any appropriate testing program prior to plant oper-ation (at power) with differential relays in service on the start-up transformer.

The SIS test has been successfully completed with the differential relays out of service.

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10. Failure Data: This same failure occurred in May 1972 under similar cua-ditions. At that time, the differential relay system was removed fro h service' and was not reinstalled until after it was modified in Ja  ;/i 1974. y e

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Mr James G Keppler Office of Inspection and Enforcement Region III US Nuclear Regulatory Co==ission 799 Roosevelt Road

v. Glen Ellyn, IL 60137

..g DOCKET 50-255 - LICENSE DPR ~

PALISADES PLMIT - ER-TT-Oh7 -

A t ch d 1 -day eportable occurrence relating to the loss of off-site The report due date was power at the Palisades Plant on Septe=ber 24, 1977 extended one week as previously discussed with Mr R Warnick.

An attachment to the LER is provided for additional details.

LD

- David P Hoffman (Signed)

Ln David P Hoffhan

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• Assistant Nuclear Licensing Administrator CC: ASchwencer, USNRC e

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• e NITACF2ETT TO LER 77-Oh7 The cause of the 'R' Bus loss is not known. Loss of off-site power causes a loss of main condenser cooling vater. Thus, the main turbine tripped on high back pressure and tripping of the =ain generator and reactor occurred. Both emergency diesel generators started i=nediately, loaded properly and performed satisfactorily throughout the incident. *he secondary system was isolated, primary system stabilised and the plant functioned as designed during the incident. -

Atmospheric dumps were operated as necessary to maintain PCS temperature. During the incident, power was lost to the Security System and extra Security personnel were called into the plant. Technical Specifications 3.1.1 and 3.7.1 vere violated.

3.1.1. At least one primary coolant pu=p or shutdown cooling pu=p shall be in operation whenever a change is being made in the boron concentration of the primary coolant.

It is conservative and prudent to borate the plant to shutdown condition after a reactor trip. Boron sa=ples during the incident and after restoring primary g

coolant flow were as anticipated and verified that no stratification occurred.

.o 3.7 1 The primary coolant system shall not be heated above 3250F or maintained above 3250 F if the following electrical systems are not operable.

a. Station power transformer 1-2.

b. Start-up transformer 1-2.

1. 2h00 'V' Bus 1-E.

Throughout the incident the plant was maintained in hot shutdown. Power was m restored after h hours and h5 minutes. It was prudent to maintain the PCS hot rather than add to the Operators workload as power was restored shortly, and o - plant conditions were stabilised. This is permitted by C.E. Standard Technical Spccifications. Changes to 3.1.1 and 3.7.1 vill be submitted for approval.

The plant operated as analysed and no failures of safety-related systems occurred as a result of this incident.

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UCENSEE EVENT REPORT Palisades CONTROL BLOCK:l l l l ! l l (PLEASE PRINT ALL REQUIRED INFORMATION) 1 6 E UCENSE NUMBER PE g3 lu II IP l A lL l4 l 7 89 14 l0l0l-l0l0l0l0l0l-l0l0l 15 25 l h l1 - l1 l1 l1 l 26 l0l1l 30 31 32 TYy enc $ CoexET NUMBER EVENT DATE REPORT OAT 2 0 1 CON'T l'

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• l l Tl lL l l 0 l 5 l 0 l-l 01215 l 5 l l0l9l2lhlTl7l l1 l 0 ll l a l 717 l 7 8 57 58 59 60 61 68 69 74 75 60 EVENT DESCRIPTION gg l During an electrical stor=, the 'R' Bus was de-energized causing a co=plete loss '

l 7 89 . SO HE I of off-site power, resulting in a loss of =ain condenser cooling vater and ulti=ately l 7 89 60 ME l a plant trip. Primary plant was stabilized in the hot condition and was borated. l 7 89 GO o5 l Event nonrepetitive. Tech Specs 3.1.1 and 3.7.1 vere violated. Electrical power l 7 89 c0 Em [ vas restored after k.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> and the plant returned to Tech Spec li=its. (ER-77 h7) -l 7 89 pauf 60 C E CCMooNENT CODE M F VCtATCN Q lElAl ] l Zl Zl Zl ~2l Z l Z l lZl lZ l9 l9 l9 l lYl 7 89 10 11 12 17 43 44 47 48 9 CAUSE CESCRIPTION .

Oe l Exact cause of loss of electrical power is not known. A change to the Tech Specs l 7 89 60

@ l vill be =ade to make this type of event a nonviolation. l 60

, C]8 9I l s US  % POWER OTHER STATUS C VERY CISCOVERY DESCA@TCN 1

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80 PERSONNEL EXPOSURES IO NUMBER TYPE CESCRGTICN Q l 0l 0l 0l gl 7;89 11 12 N/A 13 l

80 PERSONNEL INJURIES uuueER CEsCamfoN 3[ l 0l 0l 0l l N/A l 7 89 11 12 80 PROBABLE COUSEQUENCES ,

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i' ;;h o . CADS!!mSiS Power O (%g, company b General Offices: 212 West u6cNgan Avenue. Jackson, uscrugan 49201 Area Coce 5t7 780 0550 May 26, 1972 Mr. E. J. Bloch, Director Re: Docket 50-255 Directorate of Licensing License No DPR-20 United States Atomic Energy P,. , f 9

.2 () -O Commission Washington, DC 20545

Dear Mr. Bloch:

This is to apprise you of two incidents which occurred O recently at the Palisades Plant. The first occurred at 1808 on May 17, 1972 and involved the 1-2 start-up transformer protective O relays and the second occurred at 0120 on May 18,1972 and in-g volved emergency diesel generator No 1-1.

At the time both of these incidents occurred, the plant i?

Q- was in a hot standby condition with the primary system at 530 F, 2100 psia and the coolant boron content at refueling concentration.

At 1808 on May 17, 1972, the safety injection system (sis)

~

test button was pushed to initiate a quarterly test on the left chan-nel of the safety injection system. This resulted in a loss of out-M- side power as a differential relay on the 2400-volt start-up trans-former operated and cleared the 345 kV "R" bus in the switchyard, The diesel generator started automatically but required g., manual synchronization to the 2400-volt safeguard buses 1C and 1D.

This was as designed and due to the test button depression during N the auto cycle.

Preparations were made to back feed off-site power through the normal station power transformers while the tripping of the dif-ferential relay was being investigated.

The actuation of the differential relay was spurious and was due to unbalanced sensing currents frem the current transformer with load on the start-up transformer. This unbalance was due to the incompatibility of the installed current transformer to the 345 kV to 2.4 kV step-down situation. This unbalance plus the starting currents of the pump motors initiated by the test signal caused the relay to operate.

10200S94

_.l Mr. E. J. Bloch 2 Docket 50-255, License No DPR-20 May 26, lgr72 O At the time the relay operated, the load on start-up trans-former No 1-2 was 4.4 MVA plus the starting current of 1750 horsepower of the motors. From test data and the above load data, the 345 kV current transformers require approximately .8 ampere of excitation current to support the voltage requirement of the current transformer.

With .8 ampere excitation current, the current transformer had 32%

error. The 32% error was enough to cause the differential relay to operate.

To provide protection for start-up transformer No 1-2, the differential relays have been removed and high side overcurrent and instantaneous overcurrent relays have been installed. The high side overcurrent relays will not trip incorrectly for any current trans-former saturation that occurs.

The protection for start-up transformer No 1-1 was reviewed and the same saturation problem was found to exist for its differen-tial relay. The same changes were made for the No 1-1 transformer as for the No 1-2 transformer.

O s The protection for station power transformer No 1-1 and Ch No 1-2 has been reviewed. Current transformer saturation is not U considered to be a problem in the application of differential pro-tection to these banks, because of the different voltage ratio of

_ these transformer banks.

The load that existed on the No 1-2 start-up transformer at the time of trip was the maximum expected during plant operation.

The SIS test has been conducted successfully several times before but at the time they were conducted, loads on the transformer were less.

to t Prior to returning the plant to service, a safety injection C system test will be performed which involves simultaneous actuation of both halves of the safety injection system with the plant in a 10 hot standby condition, thereby subjecting the starting power source g to the maximum loads they will see.

At 0120 on May 18, 1972, the diecel generator No 1-1 tripped off due to loss of fuel supply. The engine had operated for over seven hours but ran out of fuel as the level switch on the engine fuel res-ervoir failed and prevented the flow of fuel from the day tank.

The level switch which failed and prevented continued op-eration of the diesel generator No 1-1 has been replaced. In addition, the monthly operational tests on the diesel generator unit have been revised to include a functional check of the lev 11 switches which O .

10200S95

1

- *. *4 l Mr. E. J. Bloch 3 Docket 50-255, License No DPR-20 May 26, 1972

0- actuates the fill valve and the low-level alarm. This is accomplished by shutting off the manual valve and allowing the reservoir level to drop until the switches are actuated.

Yours very truly, Ralph B. Sewell (Signed)

RBS/dmb Ralph . B. Sewell Nuclear Licensing CC: BHGrier, Administrator USAEC i

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD In the Matter of Docket No 50-329 OM CONSUMERS POWER COMPANY 50-330 OM (Midland Plant, Units 1 and 2) Docket No 50-329 OL-50-330 OL September 10, 1982 AFFIDAVIT OF DAVID A S0MMERS My name is David A Sommers. I am a Section Head in the Midland Safety and Licensing Department. In this capacity, my responsibilities are supervising and coordinating the review of environmental licensing and radiological safety issues for the Midland Project.

I am primarily responsible for providing a response (s) to Interrogatory II, Questions 5, 6, 7, 8, 9, 10, 11 and 13 concerning Mary Sinclair Contention 5.

To the best of my knowledge and belief, the above information and the responses to the above interrogatory (ies) are true and correct.

05 Sworn and Subscribed Before Me This y of 982 l

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potary P@c Jackson County, Michigan My Commission Expires b ,

miO982-2670a168

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD In the Matter of Docket No 50-329 OM CONSUMERS POWER COMPANY 50-330 OM (Midland Plant, Units 1 and 2) Docket No 50-329 OL 50-330 OL July 12, 1982 AFFIDAVIT OF DONALD H EVANS My name is Donald H Evans. I am an Engineering Supervisor for Bechtel Associates Professional Corporation. In this capacity, my responsiblities are for Geotechnical support to the Ann Arbor Power Division for Hydraulic, Hydrologic and Hydrothermal Analyses.

I am primarily responsible for providing a response to Interrogatory Questions 2, 3, 4 and 12 concerning Mary Sinclair Contention 5. To the best of my knowledge and belief, the above information and the responses to the above interrogatory are true and correct.

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/- M SwornandSubscribedBeforeMeThis(( y ofh 982 lin110 Y ,a g Notary P $ ic JacRson County, Michigan My Commission Expires ,8 k,,

mi0782-2268a100

Mary P Sinclair Interrogatory II Contention 5 deals with questions about the adequacy of the basis of the data in the Monthly Cooling Pond Performance Tables on the cooling pond provided in the DES (4-7, 4-8).

Questions

1. Provide all internal correspondence on the cooling pond between Consumers Power Co., Bechtel and the NRC.

2. What data base did you rely on to arrive at 380 acres as an adequate size for a cooling pond for the Midland plants?

3. Who supplied this data base?

4. What experience in cooling pond operation are you relying on for your assurance that this pond is workable and adequate to its tasks?

5. What qualifications did the staff have that you relied on for determining the adequacy of this pond? Provide information on their backgrounds in terms of not only their education, but also their experience for this evaluation.

6. Have you sent any observers to the Dresden pond which was identified by the NRC staff as most closely identical in performance to the Midland plants because of climate and meteorological conditions?

7. If so, when did they make their observations?

8. If so, please provide their report.

9. What are the differences in size between the Dresden pond and the Midland plant?

10. Are their one or two reactors at Dresden?

11. What will be the differences in heat load during full operation between the Midland nuclear plants and Dresden?

12. Explain the source of data for each factor in the Monthly Cooling Pond Performance Tables.

13. What studies have been made to determine the effect of the fogging on the people in the area and the Bullock Creek elementary i school? (Dr. ' Edward Epstein, the meteorologist from the University of Michigan who was our expert witness on fogging, discussed this at a seminar of the nuclear engineering department in October, 1972, and said, "I don't know how those people are going to live.")

miO982-0054bl68

2 Responses

2. Cooling pond sizes are set by the allowable condenser inlet temperature, heat rejected by the plant, and natural heat energy entering and leaving the pond considering the varying meteorological conditions at the pond location. Analytical and physical model studies were performed for the Midiand Plant Units 1 and 2 to support the adequacy of the cooling pond size.

Listed below are the more significant studies that give the data base to arrive at 880 acres as an adequate size for the Midland cooling pond. These reports contain various data bases and assumptions used for evaluating cooling pond performance,

a. Cooling System Study (May 1968)

The cooling pond alternative was recommended by Bechtel against other types of plant cooling systems.

b. Cooling Pond Calculations for Midland Plant (February 1969)

A Bechtel thermal performance comparison between 625 acre and 880 acre cooling pond sizes provided factors considered in pond sizing.

c. Midland Condenser Optimization Study (March 1969) l Within this Bechtel study the 880 acre pond size was recommended based o'n optimizing of the condenser performance.

l i-l miO982-0054bl68

3

d. Midland Cooling Pond Model Studies (May 1971)

Recommendations based on analysis of preliminary cooling pond model studies performed at Alden Research Laboratories to support the design of the Midland cooling pond.

e. Cooling Pond Thermal Performance Summary Report (August 1973)

This is a combined progress and thermal performance report summarizing what Bechtel has done on cooling pond design since 1968. It also summarizes the thermal performance of the 880 acre cooling pond.

Additional studies have been conducted since these were published which evaluate pond thermal performance but were not used in selecting the 880 acre pond size.

3. Refer to the list of studies provided in response to question 2. The thermal performance studies performed by Bechtel to support the cooling pond size were based on meteorological data from the following:

U.S. Weather Bureau Michigan Data for Midland, Flint, East Lansing and/or Tri-City Airport 1

Alden Research Laboratories provided data from physical model studies under subcontract to Bechtel.

4. The principles and methodology used in the Midland cooling pond thermal performance analysis are well accepted in practice and have been verified through operation or tested against field data at a minimum of eight miO982-0054bl68

• e O 4

plants. These plants, some of which are located in Nebraska, Virginia, Illinois, South Carolina and South Dakota, cover a wide range of climatic conditions and geographic locations, including the Dresden Plant in Illinois.

In addition to many detailed analytical evaluations of the Midland cooling pond, summarized in part by the Reference, a physical model study was perfo rmed. Physical models provide a better understanding of the physical phenomena in a complex environment. When physical model studies are used in conjunction with analytical evaluations they provide added assurance that the design will accomplish the desired results.

Simple analytical techniques, similar to those used in the early stages of evaluating the Midland pond thermal performance, were applied to field data from the Four Corners plant. One approach included the use of temperature distribution information developed by the physical model studies of the i

Midland cooling pond. The results are reported in the Reference and are reasonable for this limited comparison. The methodology used to determine the Midland- cooling pond transient thermal performance was no way affected, adjusted or altered based on the above comparisons with the Four Corners field data. Experiences in cooling pond operation at the Four Corners plant has no affect on the Midland cooling pond thermal performance ' analyses.

Reference:

" Cooling Pond Thermal Performance, Summary Report" Midland Plant Units 1 & 2, Bechtel Incorporated, for Consumers Power Company, August 1973.

miO982-0054bl68

i e

5. Consistent with the definitions provided by the intervenor in the preface to her interrogatories, CP Co interprets this question as referring to the NRC Staff, not the CP Co Staff; hence we have no direct response.

6. Neither Consumers Power Company nor Bechtel has sent personnel to the Dresden cooling pond for the purpose of observing climatic or meteorological conditions associated with pond operation. However, Consumers Power has retained Murray and Trettel.Inc, for the Midland fog and ice monitoring studies. Mr J P Bradley of Murray and Trettel was a principal investigator / project manager for Commonwealth Edison's steam fog impact studies for Dresden. In this capacity, Mr Bradley participated in the field observations at Dresden.

7. As noted above, no field observations were made by Consumers Power or Bechtel personnel. The Murray and Trettel observations, in which J P Bradley participated, occurred during the periods of December 1971 to March 1973 and November 1977 to March 1978.

, 8. Murray and Trettel has issued the following two reports documenting their field observations at Dresden:

4 l

1. Report on Meteorological Aspects of Operating the Cooling Pond and Sprays at the Dresden Nuclear Power Station, Murray and Trettel Inc, Chicago, IL, August 1973 1001-1005.

i

2. Report on Steam Fog Impact Engineering at Dresden Nuclear Power Station Report #1183 Murray and Trettel Inc, Chicago, IL, May 1978.

miO982-0054bl68

6

9. The Dresden Station cooling lake is 1275 acres (Dresden FES-OL, Section 3.4.3.a, November 1973).

The Midland Plant cooling pond is 880 acres (Midland FES-OL, Section 4.2.4.2, July 1982).

10. There are three nuclear reactors at Dresden. Unit 1, which does not have its heat load rejected to the cooling pond, was shut down in late 1978 or 1979 and has not operated since then. Units 2 and 3 are the other two reactors at Dresden, which do have their heat load rejected to the cooling pond.

11. During full operation of the Midland Plant Units 1 and 2, the heat 9

load rejected to the cooling pond can vary from 7.69 to 9.05 x 10 Btu /hr dependent on the amount of process steam being sent to Dow Chemical Company (Midland FES-OL, Table 4.2, July 1982). This variance in heat load rejection represents the range of plant operations from maximum guaranteed process steam load to back-end limited on Unit I with Unit 2 valves wide open.

The heat load rejected to the Dresden cooling lake by Dresden Units 2 and 9

3 at design conditions is 11.2 x 10 Btu /hr (Dresden FES-OL, Section 3.4, November 1973).

Comparing the respective design heat loads to cooling pond acreage, 6

Midland can vary from 8.74 to 10.28 x 10 Btu /hr/ac as contrasted with Dresden 0

at 8.78 x 10 Btu /hr/ac.

12. The factors referred to in the question are condenser inlet temperature; average pond surface temperature; total evaporation; percent l

miO982-0054bl68

7 i

imposed heat load lost by evaporation. All of these are determined from the principles of cooling pond thermal performance. These well accepted principles are based on a balance of energy added to, subtracted from, and stored in the water body.

Energy added to a cooling pond consists of net solar radiation, net atmospheric radiation and plant heat load rejected to the circulating water at the condenser. Energy removed from the cooling pond consist of back radiation

')

and evaporation and conduction processes. This outgoing energy is a function of the water surface temperature.

Data used as input to the computations that evaluated the energy balance for the Midland cooling pond include:

- dry bulb temperature

- dew point temperature (or relative humidity) 1

- wind speed

{ - cloud cover

- barometric pressure

- solar radiation 4

These data were available from Michigan weather bureau data at Midland (Dow Chemical), Flint, East Lansing or the Tri-City Airport.

l A more' complete discussion of the methods of analysis, data, and studies are presented in the referenced report.

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Reference:

" Cooling Pond Thermal Performance, Summary Report", Midland Plant Units 1 & 2, Bechtel Incorporated, for Consumers Power Company, August 1973.

13. Consumers Power Company has had two separate analytical studies conducted on cooling pond fog and its environmental impact in the vicinity of the Midland Plant cooling pond. These studies, their mode of publication or discussion and their availability are listed below.

a. Bechtel Company, The Environmental Effects of the Midland Plant Cooling Pond: Interim Report, June 1971. This study was included as an attachment to Section 4.3 of the Applicant's Supplemental Environmental Report (ASER) and was available since published in the ASER in October 1971.

The final version (Summary Report) was issued on April 28, 1972. The Bechtel study (both interim and summary reports) were available at and thoroughly discussed (even by the intervenor's witness) at the ASLB Construction Permit Hearings in 1972.

b. D J Portman and M R Weber, Fog and Plumes From Power Plant Cooling Systems in the Tri-Cities-Saginaw Bay Area, pp 1-5, 44-65 and 68-71, June 1975. This study was included as Appendix 5.1C of the Environmental Report ER-OLS and has been available since published in the ER-OLS in April 1978.

The issue of cooling pond and its environmental impact has also received coverage in the following other documents listed below.

miO982-0054bl68

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a. Consumers Power Company, Applicant's Supplemental Environmental Report, Section 4.3 " Fogging Effect of Cooling Pond Operation,"

Section 5.1.3.1.K " Environmental Costs (Items 11.1 and 11.2),

October 19, 1971 (as amended through January 7, 1972). This information has been available since published in October 1971 and was thoroughly discussed at the ASLB Construction Permit Hearings in 1972.

b. US Atomic Energy Commission, Final Environmental Statement Related to the Construction of Midland Plant Units 1 and 2,Section V.A.2

" Fogging and Icing,"Section VII " Adverse Impacts That Cannot Be Avoided,"Section XI.B.3 " Summary of Cost-Benefit Analysis: Impact on Air and Land,"Section XII.B " Discussion of Comments Received...: Fog Occurrence and Extent," March 1972. This information has been available since published in March 1972 and was thoroughly discussed at the ASLB Construction Permit Hearings in 1972.

c. Consumers Power Company, ER-OLS, Section 5.1.4.1 " Frequency of Fog Occurrence, Section 6.1.3.1.8 " Fog and Ice Monitoring," Section 6.2.3.1.2 " Operational Fog and Ice Monitoring Program," Section 6.2A-3.1.1.2.3 " Fog and Ice Formation," Section "NRC Questions and

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Responses" Met 2, Met 3, Met 4, Met 5, Met 6, Met 7, Met 8, Met 9, Met 10, Met 11, Met 12, submitted for docketing April 12, 1978 (as amended through Revision 13 - December 1981, January 4,1982). This information has been available since published in April 1978.

miO982-0054bl68

. n UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD 1

l In the Matter of Docket No 50-329 OM  !

CONSUMERS POWER COMPANY 50-330 OM (Midland Plant, Units 1 and 2) Docket No 50-329 OL 50-330 OL September 8, 1982 AFFIDAVIT OF PIIILIP A DI BENEDETTO My name is Philip A Di Benedetto. I am the Engineering Manager at Hutech ,

l Engineers-Bethesda. In this capacity, I am presently involved in providing technical consulting services to several utilities on equipment qualification ,

program development.

i I am primarily responsible for providing a response to Interrogatory III, d

Question 3, concerning Contention 7. To the best of my knowledge and belief, the above information and the responses to the above interrogatory are true and correct.

y / M Sworn and Subscribed Before Me This f Day of/ w) ,1982-e leu 1 l Motary Public Washtenaw County, Michigan i

My Commission Expires V lptt(m l u/3 6 /9[ 1 PE7ZRtY A. 2E033

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l UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD In the Matter of Docket No 50-329 OM CONSUMERS POWER COMPANY 50-330 OM (Midland Plant, Units 1 and 2) Docket No 50-329 OL 50-330 OL September 10, 1982 AFFIDAVIT OF PETER W JACOBSEN My name is Peter W Jacobsen. I am a Senior Engineer in Technical Services Section of Midland Design Production Department. In this capacity, my responsibilities include coordination of the environmental qualification program.

I am primarily responsible for providing a response to Interrogatory III, Questions 1 and 2 concerning Mary Sinclair Contention 7. To the best of my knowledge and belief, the above information and the responses to the above interrogatory are true and correct.

1 T

SwornandSubscribedBeforeMeThis/[ y of 982 i

L g Notary p lic fackson County, Michigan My Commission Expires

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l mi0782-2265c100 l

. e Mary P Sinclair Interrogatory III Contention 7 deals with the effects of low doses of radiation on polymer cable insulation and jacketing and the synergistic effects of radiation and tempera-ture in degrading those materials. -

Questions

1. Send all correspondence with the NRC on this study and its implications for the Midland nuclear plants.

2. What percentage of the plant's electrical wiring system is not accessible to inspection once the plant is started?

3. For what percentage of the estimated lifetime of these plants can the cable insulation be expected to re;ain its integrity before its degradation by synergistic effects makes it unsafe?

Responses

1. There has been no formal technical correspondence with the NRC on this study and its implications for the cable used at Midland to date.

However a letter from J E Brunner to M Wilcove dated August 30, 1982 indicating we would be requesting information from the NRC Staff has already been copied to Ms Sinclair. The letter requesting technical information on the Sandia Reports NUREG/CR-2156 and NUREG/CR-2157 is being prepared by Consumers Power Company. As part of normal distribution, Ms Sinclair and all other persons on the service list will be sett copies of this letter.

2. A detailed assessment of all electrical cables in the plant to determine accessibility for inspection has not been conducted. However, a review of all cable types most likely to be affected by the phenomenon reported in the Sandia Study, in their worst case application (ie, highest potential for accelerated damage due to synergistic effect as discussed in the Sandia Reports) has shown that these cable types are miO982-0054c168

. e 2

accessible for inspection with varying degrees of difficulty. Such difficulties include work in high radiation areas, possible need for plant shutdown, whether or not the cable can be inspected in place, etc, and depend on the method of inspection. '

3. It is anti-:ipated that all Class 1E cable within the Midland plant will maintain its electrical integrity throughout the 40 year life of the plant. This is based on the testing and analyses performed, as well as operating experiente throughout the nuclear industry.

Testing performed for Consumers Power Company on actual cable used in safety circuits his demonstrated a qualified life of 40 years and has further demonstrated the ability of the cable to perform its safety function following simulation (by test) of a design basis event (e.g.,

loss of coolant accident). This testing did not take into account the synergistic effect of low dose rate and mildly elevated temperature suggested by the Sandia Reports. However, a preliminary analysis of_the Sandia Reports indicates that although some accelerated degradation is caused by the low dose rate and mildly elevated temperature synergistic effect, the total amount of degradation does not impact the Midland cables' integrity or their ability to perform their safety function during or after the design basis events. On this basis, we expect to show that although some additional. degradation is attributable to the synergistic effect of low dose rate and mildly elevated temperature its impact on the overall life of the cable is not significant.

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i i Praels J Kalley, Esq Atomie Safirty k Licensing- -

Attorney General of the- AppeeL Panel l Stacecof Miah4=== ~ T S' Etaclear Regulatory Casr Carole Steinberg,.Est Weakingtos,.D.c 20555

! Assistant. Attorner General -

Environmental Protectica'Div . Mr C R Stephens (3) l 720-Lar *=4 m a'r Chief , Docketing k Servicess 1'===4 ne.,. MT. h8913 7 5 Etacleae Begulatory Comt 01Tice of the Secretary ltrren.W Cherry, Esq Washingtcar,. D C' 20557 i One IBK Plase

! Suite 450I Ms Mary Staal=4" l Q11 case, Iri 60611 -

5711 Summerset. Street l Midland, MI kS6ko

Me mad =T1 E Marshall l

RFD.10 VfT14== D: Patree, Est .

l Nf A7 and , Mr.486ko . Counsel' for the EBC Staff I  : 7 S Nualear Regulatory Casm i Charlen,Bechhoefer,.Esq -

Weahington ,D C 20555 3 Atmic Safety er tia===4=r Board Panel Ateic Safety Ir Licensing.

U S Nnal- Regulaterr Ccum. Board. Panel u..h4*=tany D C: 20555- 'E S Nuclear Regulatory Cama

- Washington, D C' 20551 Dr Frederick.P Cowan. -

i 6152 Y 7erda Trail

• Barbarn Stamiris

.Atp B-127 .

5795: North- River Road.

Boca;Raton,.FL 33k33 Rtr 3 3

Freeland, MI kS623 .

Jerry Harbour Atomic Safety & Licensing Boisrd Panel

-l Carroll I Mahaney U S Nuclear Regulatory Camm i Babcock & Wilcox _

W--h4== ton, D C 20555

! PO Box 1260 -

i Lynchburg, Virginia 2k505 Lee L Bishop -

Raraca & Weiss James E Brunner,.Esq 1725 "I" Street,.NW #506 Consumers Power Company W==h4*= ton, DC 20006 212 West Michigan Avenue Jackson, MI k9201 M I Miller, Esq Isham, Lincoln & Beale Mr D y Judd One First National Plaza Babcock & Wilco:t. Suite 4200 PO Bc:c 1260 Chicago, IL 60603 Lynchburg, VA 24505 John Demeester, Esq Steve Gadler, Esq Dow Chemical Bldg 2120 Carter Avenue Michigan Division St Paul, MN 55108 Midland, MI k86ho

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CERTIFICATE OF SERVICE I hereby certify that copies of the attached responses of Consumere Power l Capany to Discovery Questions of Interviewer Mary P Sinclair vere sent by U S Mail, first class, postage prepaid, to the attached service list this 15th day of September, except for Mary sinelair, who was served by Federal Express.

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