ML20062M980

Revised Contentions Based on Discovery Per ASLB 820525 Order

ML20062M980
Person / Time
Site:	Midland
Issue date:	08/12/1982
From:	Bishop L HARMON & WEISS, SINCLAIR, M.P.
To:	Atomic Safety and Licensing Board Panel
References
ISSUANCES-OL, NUDOCS 8208200265
Download: ML20062M980 (6)

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USHRC UNITED STATES OF AMEHICA NUCLEAR HEGULATORY COMMISgNO 19 P2:03"-

Before the Atomic Safety and LicensingrBoer L

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aRANcy In the Matter of: )

) Docket Nos. ~50-329 CONSUMEllS POWEH COMPANY ) 50-330

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(Midland Plant, Units I and 2) ) Operating License HEVISED CONTENTIONS OF MARY SINCLAIH BASED ON DISCOVEHY PURSUANT TO 150AHD OllDER OF MAY 25,193 August 12, 1982 Contention 28 deals with the water hammer problem of pressurized water.

reactors of the Midland type. This problem is identified as one of the unresolved safety lasues applicable to Midland I & 2 in the SER, C-1 Babcock and Wilcox(B&W) plants with an internal auxiliary feedwater (AFW) feed ring of the same design as Midland in recent events, have shown a marked susceptibility to internal damage of the feed ring as a result of water hammer. From this, reduced cooling in the steam generators could occur as a result of inadequate AFW Gow following.

loss of normal feedwater flow. (NitC Response to Interrogatory 7) Since this effect involves critical safety systems, the Task A-1 report (Jan. ,1980) states that systematic review procedures in the OL review process will require the Applicant to : 1) address potential water hammer problems in various systems;

2) demonstrate that there are adequate design features and operating procedures to prevent damaging water hammer events; and 3) expand the preoperational testing program to insure that these design features and operating procedures do prevent damaging water hammer events.

However, the SER does not indicate that these criteria have been met by the Applicant. As a result of this omission, the findings required by'10 CFR 5550.57 (a)(3)(1) and 50.57(a)(6) cannot be made.

Contention 30 The degradation of steam tube integrity due to corrosion in-duced wastage, cracking, cduction in tule diameter, and vibration induced cracks 2

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• is a serious unresolved safety problem at the Midland nuclear plant. It is admitted that the chemistry of the cooling water is critical to prevention of steam tube failure, (NUREG-0886). Ilowever, the fact thnt these plants depend on cooling water from the cooling pond increases the likelihood ut corrosion and poor water chemistry because the DEIS states that the plant dewatering system will first be discharged to the cooling pond. (DEIS at 5-2). Thnt means that many wastes, including radioactive materials from leaks and spills on the reactor site, enn enter the cooling pond and disrupt the chemistry of the pond. Theretore, due to this contribution of an undertermined amount and quality of ground dewatering inflows to the cooling pond, the NHC's bland assurance that corrosion is unlikely I

due to the lack of sodium thiosulfate, is unsatisfactory. (NRC Response to Interrogatory D.J.) In fact, due to the contribution of groundwater, the NRC is not fully aware of the likely constituents of the cooling pond, and the findinge required by 10 CFR ll 50.57(a)(3)(1) and 50.57(a)(6) cannot be made.

Contention 31. Numerous non-safety related systema, the feedwater system, main stream system, makeup and pernification system, non-vital electrical power systems,' and the integrated control systems, can adversely affect safety related systems, such as Anticipated Transients Without Scram (ATWS). (NRC Response to Interrogatory 10.c) Since there has been no routine inspection and quality control standards applied to these non-safety systems, and the general quality <

control during construction of even safety related systems has been so poorly done (amply documented in the record of these hearings), there is an even greater probability of ATWS at Midland, llowever, this scenario has not been analyzed in the SER. Furthermore, B&W reactors, such as the Midland reactors, experience the largest pressure rise and thus are the most difficult to modify to achieve ade-quate safety margins to prevent ATWS events. (NUREG-04GO, April,1978, p 46)

Therefore, the findings required by 10 CFR 55 50.57(a)(3)(1) and 50,57(a)(6) cannot be made.

Contention 32. The e is no assurance that suitable safety margins can be maintained throughout the design life of the Midland plant with the materials used for reactor vessel fabrication. This makes the Midland reactors unusually sus-ceptable to reactor embrittlement and to pressurized thermal shock (PTS). For 1

example, an investigation following the severe lYrS at the Hancho Seco reactor indicated that the limiting material in the Hancho Seco reactor vessel was fabri-cated using the same weld wire and flux as the limiting material in the Midland reactor vessel beltline and has equivalent chemical composition and fracture toughness properties. This indicates that the staff's conclusions concerning the Hancho Seco reactor vessel beltline materials are applicable to the Midland Unit I reactor vessel beltline materials. (NHC Hesponse to Interrogatory ll.e)

Furthermore, in a memorandum to the Midland file, dated June 14,1977, by G.S. Keeley of Consumers Power Co. and sent to S. II. Howell, et al., described -

a memorandum which A. J. Birkic had written to H. C. Bauman on March 22, 1977, on the status of Midland NSSS-12 reactor vessel girth weld fracture toughness.

(Discovery Hesponse, Consumers Power Co.) This memorandum pointed out that there was 'a chance that the NSSS-12 reactor vessel could have a low level of fracture toughness at the operating temperature after 10 years of operation'.

The low level was with reference to the 50 ft-lb upper shelf criteria of 10 CFR 50, Appendix G & 11 It also indicated that this could possibly be corrected by annealing the vessel which is not now a viable approach although an EPHI R&D effort is under-way." Moreover, Demetrias Basedekas.. NHC reactor safety engineer, in a ' -

memorandum addressed to Chairman Palladino (NHC Response to. Interrogatory ll.a) made the following major points which emphasize the importance of this deficiency.

concerning PTS:

" Substantial uncertainties and non-conservative assumptions

in estimates of consequences and of probabilities cast serious doubts on the validity of conclusions stated by industry and the NHC staff.

The lack of badly needed design information on control and electrical power systems, and related neutronic and thermal-hydraulic parameters for rel,resentative plants (at least one

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for' each NSSS vendor) makes an independent and thorough -

assessment of this issue by-NHC virtually' impossible.

Substantial operation experience w'ith PTS precursor events involving control system and steam generator.

tube fallares, coupled with an understanding of .

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functional and some ' design aspects of control systems .

and components in operating plants, suggest an unaccept-able level of risk associated with a number of older -

pressurized water reactors."

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-4 These points, as well as the fact that the Midland nuclear plants were designed over a decade ago, and contain the same defective material as the Rancho Seco nuclear plant means that findings required by 10 CFR 45 50.57(a)(3)(i) and 50.57 (a)(G) cannot be made.

Contention 35. Assurance of pressure vessel integrity and the ability to detect and adequately size flaws depends, for one thing, on carefully controlling the fabri-cation, welding and examination of welde to minimize the probability of significant weld defects. The affidavits secured by the Government Accountability Project and recently turned over to NRC, especially that of Der.n Darley and one of the anonymous workers, describes extensive failures li. welding. (Midland Daig News, July 20, '82) Therefore, the findings required by 10 CFR 95 50.57(o)(3)(1) and 50.57(a)(G) cannot be made.

Contention 36 Systems interactions, identified as an unresolved safety problem applicable to Midland in the SER (C--4), has special sigr.ifiennee at Midland because the most serious accident resulting from systems interaction failures have occurred in B&W reactors. The serious events nnd their special -

problems with system interaction include the following:

1) The persistant operator disbelief of high temperature data from incore thermocouples and system HTD's was one major, out of many, causes for the TMI-2 accident. This disbelief was based on the rationale that the former were not safety-grade equipment while the latter were outside the calibrated range of the detectors. (NUREG-0000, p 10, and " Daniel Ford, Three Mile Island, Thirty Minutes to Meltdown") In the case of the high temperatures, acceptance of the temperature data as valid might have prompted a higher high-pressure-injection flow rate and a reluctance to subsequently depressurize the plant to use the core flood tanks. (NUREG-0600, p 11) This is one example of non-safety related equipment impacting on safety systems.

2) At Crystal River, ad accident on February 26, '80, is of interest because of systems interaction where the integrated control system input, the PORV positioning, the instruments usul for manual control of ECCS and the entire non-nuclear instrumentation (NNI) power supply depended on onc and 24 VDC line within the NNI power supply system. (NUR EG-0667)

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3) At Davis-Hesse I on April 19, 1980, maintenance activities allowed an elimination of redundant po'ver supplies that were supporting the decay heat removal function. Concurrent con-8truction activities caused the loss of working pov.c r supply and subsequently decay heat removal was lost for over two hours. (USNHC IE Information Notice 80-20, May 8,1980)

(N!!C Ilesponse to Interrogatory 15.e)

In spite of this repeated history of system interaction problems at Il&W reactors, the staff Sell specifically fails to require a comprehensive program to reportedly evaluate :11 ayrtems which could interact. (SEH at C-12.) Moreover, the apparent use of non-safety grade materials for safety grade functione at Midland significantly increases the risk of adverse system interactions. (floward affidavit).

Contention 40 ricals with lack of adequate qualification methods to satisfy the rquirements for safety related equipment.

Contrary to NRC Response to interrogatory 10 (a), a Commission decision in the UCS Petition for Emergency and H(medial Action (CLI 80-21, May 27,1980),

11 NRC 701, requires that all plants under licensing review must meet the equiva-lent of the IEEE 1971 Standard in order to satiary GDC 4 (10 CFH 50, Appendix 4).

In fact, the SEh admits that this standard has not been met, (SEH p 3-3G) Thus, absent further action, the findings required by 10 CFH ll 50.57(a)(3)(i) and 50.57 (n)(G) cannet be made.

Contention 45 There is no assurance that offsite power is suffleiently reliable to ensure the maintenance of safety functions during accident condittor.s. In one of the anonymous GAP affidavite, an electrician described the pcor quality control that has gone into the electrical work at the Midland nucicar plant. Ile stated that the cables shor substituted control cables whcn the correct type was unavailable, i

Ile explained that a cable design may have called for three shleided pairs of 16-guage wire but the enble shop in which he worked wouid use six stranded 16-guage wire with the shielding around the entire bundle. (Midland Daily News, June 28,'1982)

These types of electrical cable deficiencies built into many party of the plant do not comply with the General Design Criteria, therefore the findings required by 10 CFR II 50.57(a)(3)(1) and 50.57(u)(6) cannot be made.

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Contention 50. The occupational eximure of regular workers or transient workers at the Midland nuclear pinnt ennnot be controlled as the NitC Ilesponse r to Interrogatory 29(a) states, because of th( extensive quality control failures that the disclosures of Zack Co. employees and Dean Darley indicate have been built into the heating, ventilating and air conditioning system at the Midland nuclear plant. Therefore, the findings required by 10 C FR ll 50.57(a)(3)(i) and 50.57(a)(G) ennnot be made.

Contention 52. The reliability of the emergency onsite diesel generator at Midland is seriously in question. The N11C staff has stated that: "The excessive  ;

settlement and cracking of the diesel generator bulhling due to improperly compacted soll enn seriously and adversely affect diesel generator performance since this can enuse excessive differential movement between diesel generator and building

found a tion s . " (NilC Ilesponse to Interrogatory 31.d) Alsq there is concern at ,

Midland for damaging fuel oil and service water lines entering and exiting the building. Therefore, the findings required by 10 CF11 il 50.57(a)(3)(1) and 50.57 -

(a)(G) cannot be made.

Respectfully submitted,

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l-Le e L.. Bishop J. d l' Ilarmon & Weiss 1725 I Street,1 NW #506 Washington, D.C. 25006' (202) 833-9070 Attorney for Mary Sinclair B

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