Control of Heavy Loads (Generic Letter 81-07)

                                UNITED STATES
                        NUCLEAR REGULATORY COMMISSION
                           WASHINGTON, D.C. 20555

                              February 3, 1981

TO ALL LICENSEES OF OPERATING PLANTS AND APPLICANTS FOR OPERATING LICENSES 
     AND HOLDERS OF CONSTRUCTION PERMITS* 

SUBJECT:  CONTROL OF HEAVY LOADS (Generic Letter 81-07) 

Gentleman: 

By our leter dated December 22, 1980, you were requested to review your 
controls of the handling of heavy loads to determine the extent to which the
guidelines of NUREG-0612 are presently satisfied at your facility and to 
identify the changes and modifications that would be required in order to 
fully satify these guidelines. 

To expidite your review, three enclosures were included with the letter. One
of the enclosures was Request for Additional Information on Control of Heavy
Loads (Enclosure 3). We have found that five pages from Enclosure 3 were 
missing due to a reporduction error. The missing pages are enclosed with 
this letter. In addition the December 22, 1980, letter on Page 2 in Item 1 
required that information identified in Section 2.1 through 2.4 of Enclosure
3 be included in a report documenting the results of your review. This 
requirement should be modified to read:  "Sections 2.1 through 2.4 for PWR 
plants and Sections 2.1 through 2.3 for BWP plants." 

Because of these errors we are extending the Enclosure 2 90-day 
implementation requirement to May 15, 1981. 

                                        Sincerely, 


                                        Darrell G. Eisenhut, Director 
                                        Division of Licensing 

Enclosure:
"Enclosure 3" missing pages

*With the exception of licensees for Indian Point 2 and 3, Zion 1 and 2 and 
Three Mile Island 1 
.

                                                            Attachment (4) 

                       ANALYSIS OF PLANT STRUCTURES 

The following information should be provided for analysses conducted to 
demonstrate compliance with Criteria III and IV of NUREG 0612, Section 5.1. 

1.   INITIAL CONDITIONS/ASSUMPTIONS 

     Discuss the assumptions used in the analysis, including: 

          a.   Weight of heavy load

          b.   Impact area of load

          c.   Drop height

          d.   Drop location

          e.   Assumptions regarding credit taken in the analysis for the
               action ofimpact limiters 

          f.   Thickness of walls or floor slabs impacted

          g.   Assumptions regarding drag forces caused by the environment

          h.   Load combinations considered

          i.   Material proporties of steel and concrete

2.   METHOD OF ANALYSIS

     Provide the mthod of analysis used to demonstrate that usfficient load-
     carrying capability exists within the wall(s) or floor slab(s). 
     Identify any computer codes employed, and provide a description of 
     their capabilities. If test data was employed, provide it and describe 
     its applicability. 

3.   CONCLUSION 

     Provide an evaluation comparing the results of this analysis with 
     Criteris III and IV of NUREG 0612, Section 5.1. Where safe-shutdown 
     equipment has a ceiling or wall separating it from an overhead handling
     system, provide an evaluation to demonstrate that postualted load drops
     do not penetrate the ceiling or cause secondary missiles that could 
     prevent a safe-shutdown system from perfoming its safety function. 
.

                   (3)  A description of any Engineered Safety Feature 
                         filter system which includes information sufficient
                         to demonstrate compliance with the guidelines of 
                         USNRC Regulatory Guide 1.52, "Design, Testing, and 
                         Maintenance Criteria for Engineered Safety Feature 
                         Atmosphere Cleanup System Air Filtration and 
                         Absorption Units of Light-Water-Cooled Nuclear 
                         Power Plants." 

                    (4)  A discussion of any intial conditions (e.g., manual
                         values locked shut, containment airlocks or 
                         equipment hatches shut) necessary to ensure that 
                         releases will be terminated or mitigated upon 
                         Engineered Safety Feature actuation and the measure
                         employed (i.e., Technical Specification and 
                         administractive controls) to ensure that these 
                         intial conditions are satisfied and that Engineered
                         Safety Feature systems are operable prior to the 
                         load lift. 

2.   METHOD OF ANALYSIS 

     Discuss the method of analysis used to demonstrat that post-accident 
     dose will be well within 10CFR100 limits. In presenting methodology 
     used in determining the radiological consequences, the following 
     imformation should be provided. 

          a.   A description of the mathematical or physical model employed.

          b.   An identification and summary of any computer program used in
               this analysis. 

          c.   The consideration of uncertainties in calculational methods, 
               equpment performance, instrumentation response 
               characteristics, or other indeterminate effects taken into 
               account in the evaluation of the results. 

3.   CONCULSION 

     Provide an evaluation comparing the results of the analysis to 
     Criterion I of NUREG 0162, Section 5.1. If the postulated 
     heavy-load-drop accident analyzed bounds other postulated heavy-load 
     drops, a list of these bounded heavy loads should be provided. 



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.

bound other postulated heavy-load drops, alist of these bunded heavy loads 
should be provided. 



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