Information Notice No. 83-64: Lead Shielding Attached to Safety-Related Systems Without 10 CFR 50.59 Evaluations

                                                           SSINS No.:  6835 
                                                           IN 83-64        

                               UNITED STATES 
                        NUCLEAR REGULATORY COMMISSION
                    OFFICE OF INSPECTION AND ENFORCEMENT 
                          WASHINGTON, D.C.  20555  
                                     
                             September 29, 1983

Information Notice No. 83-64:   LEAD SHIELDING ATTACHED TO SAFETY-RELATED 
                                   SYSTEMS WITHOUT 10 CFR 50.59 EVALUATIONS 

Addressees: 

All nuclear power reactor facilities holding an operating license (OL) or 
construction permit (CP). 

Purpose: 

This information notice is provided to inform licensees of an event at a 
pressurized water reactor (PWR) where significant quantities of lead 
shielding were installed on safety-related systems without a proper 
engineering evaluation as required by 10 CFR 50.59. Licensees are devoting 
increased attention and resources to reduce radiation fields in an effort to
minimize workers' exposure. The NRC encourages these ALARA efforts; however,
this event and other similar occurrences illustrate a need to reemphasize 
the requirements of 10 CFR 50.59. No specific action or response is 
required. 

Description of Circumstances: 

During a routine inspection at the Maine Yankee Atomic Power Station on June
8, 1983, an NRC inspector noted that portions of safety-related piping 
(three-inch hydrogenated waste header pipe) in the primary auxiliary 
building was covered with lead shielding. Discussions with the plant 
engineering staff revealed that licensee safety evaluations to support this 
plant modification had not been performed. Since the licensee had no formal 
control mechanism to govern the installation, use, and accounting of 
temporary shielding (in the 1974-75 period), no records existed to document 
the dates and locations of shielding installations. The shielding was placed 
on plant systems during the 1974-1975 period when high fuel element failure 
rates led to increased radiation fields throughout the plant. 

After the 1982 refueling outage, the licensee had initiated a program to 
identify and remove temporary shielding installed on systems inside the 
containment building, but failed to broaden this effort to other plant 
areas. Recently implemented improvements in the maintenance and design 
program would currently prevent shielding installation without required 10 
CFR 50.59 evaluations. The controls and procedures currently in place as 
part of the facility Quality Improvement Program should prevent any 
reoccurrences. 




8308300349 
.

                                                       IN 83-64            
                                                       September 29, 1983  
                                                       Page 2 of 3         

In response to a Regional Confirmatory Action Letter, the licensee initiated
the following corrective actions to identify, the extent of the problem and 
to resolve the safety concerns: 

1.   Inspection of all safety-related piping in radiological controlled 
     areas outside containment to identify any shielding affixed to or which 
     could affect these systems (e.g., lead insecurely attached to non-
     safety-related system such that it might fall onto a safety-related 
     system). 

2.   Removal of all identified shielding and documentation of location and 
     quantity. 

3.   Identification of any system degradation problems evident after 
     shielding removal. 

4.   Description and verification of the effectiveness of actions taken 
     during and after the 1982 refueling outage to identify and remove lead 
     shielding from piping inside containment building. 

5.   Performance of inspections and engineering analyses of the affected 
     systems to ensure their operability under design-basis event 
     conditions. 

Detailed inspections were conducted in all accessible areas outside 
containment and the licensee identified 18 to 20 locations where quantities 
of lead shielding weighing between 10 and 380 pounds had been installed. By 
June 21, 1983 all the identified lead shielding had been removed from the 
safety-related systems.  Because of improved fuel integrity performance, 
radiation surveys conducted after the shielding was removed indicated only 
two or three of the affected locations would still need any additional 
shielding. Since only cursory inspections were conducted inside containment 
because of high radiation dose rates, system walkdowns inside containment 
will be performed during the next unscheduled plant outage. The licensee 
found no visual evidence of permanent degradation to piping or its supports.

Any future permanent shielding modifications will be handled as design 
changes. The licensee also plans to develop a program for control of 
temporary shielding. Since the temporary shielding had not been readily 
discernable from other pipe coverings/surroundings, brightly colored 
temporary shielding materials will be used to enhance identification. At the
request of NRC's Region I, the licensee agreed to check the concrete anchor 
bolt pre-load torque on the piping supports for the affected systems. 

Discussion: 

Failure to analyze for possible seismic/structural effects (both 
dynamic/static) of lead shielding on safety-related systems constitutes an 
unreviewed safety question. Maine Yankee safety-related systems (e.g., 
safety injection trains) were modified with additional shielding without 
supporting engineering evaluations to ensure system operability under 
design-basis event conditions. 
.

                                                         IN 83-64          
                                                         September 29, 1983 
                                                         Page 3 of 3       

Although focused on radioactive waste treatment systems, IE Circular No. 
80-18, "10 CFR 50.59 Safety Evaluation for Changes to Radioactive Waste 
Treatment Systems," provides general guidance/clarification regarding the 
requirements of 10 CFR 50.59. If you have any questions regarding this 
matter, please contact the Regional Administrator of the appropriate NRC 
Regional Office, or this office. 


                              Edward L. Jordan, Director 
                              Division of Emergency Preparedness 
                                and Engineering Response 
                              Office of Inspection and Enforcement 

Technical Contact:  J. E. Wigginton, IE 
                    (301) 492-4967 

Attachment: 
List of Recently Issued IE Information Notices 
.