IR 05000528/1985027

Enforcement Conference Rept 50-528/85-27 on 850808.Major Areas Discussed:Problems Associated W/Implementation of post-accident Sampling Sys Required by NUREG-0737,Item II.B.3

ML20137M442
Person / Time
Site:	Palo Verde
Issue date:	08/28/1985
From:	Sherman C, Wenslawski F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20137M421	List: IR 05000528/1985027 ML20137M417
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM 50-528-85-27-EC, NUDOCS 8509130235
Download: ML20137M442 (4)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION V

Report N /85-27 Docket N License N NPF-41 Licensee: Arizona Nuclear Power Project P. O. Box 52034 Phoenix, Arizona 85072-2034 Facility Name: Palo Verde Nuclear Generating Station Subject: Enforcement Conference, NRC Region V Office Prepared by: $ @ Q t ,u kn Ts/9yJ Rd_

Ddte Signed C.1.{Rrman,~ Rad {pionSpecialist Approved by: 6e Omm einin-Ddte Signed F.A.Wegslawski,Chipf7 Radiation an mergency IVparedness Branch Summary:

An NRC Enforcement Conference was held at the Region V office on August 8, 1985 to discuss problems associated with the licensee's implementation of the post accident sampling system required by NUREG-0737, item II.B.3 and the facility technical specifications as described in NRC Inspection Report No. 50-528/85-22. This report, transmitted to the licensee on July 31, 1985 identified an apparent violation of facility technical specification 6.8. In addition, the licensee's' technical specification compliance record was reviewe Results:

NRC reviewed the inspection findings involving relocation of the post accident containment atmosphere sample point and expressed concerns regarding the control and quality of technical work. The accuracy of licensee submittals to NRC and the importance of observing all technical specification requirements was also discussed. The licensee agreed that the inspection report was factual. end described corrective actions which would address NRC concern h W O G

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Details 1. Attendees: Arizona Nuclear Power Project E. Van Brunt, Jr., Executive Vice President J. Haynes, Vice President W. E. Ide, Manager, Corporate QA/QC R. Eutler, Manager, Technical Services A. Gehr, Legal Counsel J. Bynum, Director Nuclear Operations Zeringue, Technical Bingham, Bechtel Power Corporation T. Quon, Licensing B Nuclear Regulatory Commission J. Martin, Regional Administrator R. Scarano, Director of Radiological Safety and Safeguards Program L. Miller, Chief, Reactor Projects,Section II J. Crews, Senior Reactor Engineer F. Wenslawski, Chief, Radiation Protection and Emergency Preparedness Branch A. Johnson, Enforcement Officer B. H. Faulkenberry, Deputy Regional Administrator R. Zimmerman, Senior Resident Inspector C. Sherman, Radiation Specialist A. Hon, Project Inspector Background During the fall of 1984 the licensee confirmed that the Post Accident Sampling System (PASS) described to NRC could not perform its intended function due to design deficiencies and equipment malfunctions. A task force created by the licensee implemented a design change to establish the liquid sample capability and elected to use RU-1 (the normal containment air sample system) to obtain the post accident containment air sampl This shift to RU-1 was not described in the licensee's December 5, 1984 letter requesting scheduler exemption for PAS In a letter dated May 22, 1985 the licensee informed NRC that the PASS would be operable prior to exceeding 5% power, and in a June 13 letter advised NRC that they had made operable a PASS which meets the provisione of NUREG-0737 Item II.B.3, "Postaccident Sampling Capability".

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NUREG 0737 Item 11.B.3 states in part that a design and operational review of the containment atmosphere sample line system must be performed to determine the capability of personnel to promptly obtain a sample under accident conditions without incurring a radiation exposure to any individual in excess of the dose criteria expressed in 10 CFR 50 Appendix A, General Design Criteria (GDC)-1 In an onsite inspection conducted June 24 to July 12; July 23-24 and subsequent telephone conversations July 25-31, 1985 the regional staff, identified that the decision to use the normal containment air sample system was not specifically reviewed against the NUREG-0737 criteria; GDC-19. The inspector had observed the licensee's calculations indicated radiation exposure rates could be as high as 1000 rem / hour and that procedures and training may have been inadequate even for entry into 100 rem /hr field . Discussion During the conference , the NRC staff reviewed the inspection findings and apparent violations reported in Inspection Report No. 50-528/85-2 Based on these findings the following areas of concern were brought to the licensees attention: failure to communicate management expectations to the staff regarding quality of technical work; the lack of fornal method to review technical work of this nature; the need to have a substantive independent verification for technical work; role of cuality assurance licensing in this area; and the basis for management signing correspondence to the NR The licensee discussed their initial corrective actions to reestablish a containment atmosphere sample point and reaffirmed their commitment (letter dated July 26 1985) not to return Unit 1 to power until the sampling capability is insured. With respect to the condition that allowed the violation to occur, the licensee concluded that a lack of formality in operating the PASS. task force was the root caus In terms of the accuracy of the licensee's May 22 and June 13, 1985 submittals they explained the misinformation was the result of an ineffective review process and was in no way a deliberate attempt to mislead the NRC or the result of careless disregard of NRC requirement In response to their evaluation of the cause and NRC's expressions of concern the licensee described three measures to prevent recurrence: Future task forces will be formalize . A-program including procedures and training will be established to assure that independent review of technical work not covered by the existing design change system will be performe . Submittals to NRC will be reviewed to verify that they are accurate, complete and address all aspects of and correctly indicate the conformance with the basic regulatory criteria being discusse . . -

. - 3 The licensee provided their position on safety significance. Based on their retrospective evaluation of the RU-1 sample point the licensee concluded that at the time of discovery the inventory of core radioactivity could have resulted in dose rates of 129 rem /hr. In addition, with current burnup, the peak clad temperature following a LCCA would be 1325'F and core damage would not result. The maximum dose rate with equilibrium core inventory would be 262 rem / hour.The licensee believed that using 5 people, a sample could be obtained with the maximum

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exposure to any individual being 4.77 re . Other NRC Concerns In addition, the NRC staff directed the licensee attention to the numerous recent LER's which have reported failures to observe the technical specification requirements. Several of these LERs indicated misunderstanding of the specifications, or inadequate review prior to changing operating modes. The staff stressed that significantly improved performance is expecte . Conclusions The Regional Administrator characterized the technical work related to this matter as informal, undocumented , unreviewed, and stated that it was not surprising that work conducted in this fashion would come out wrong. The Senior Reactor Engineer stressed the importance of substantive independent evaluations and the importance of individual sign-off. The Regional Administrator additionally characterized the problem as a management failure to communicate expectations to their staff; failure of QA and licensing to take a more active role; and failure to assure that communications to NRC are complete and accurate.

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