Information Notice No. 96-39: Estimates of Decay Heat Using ANS 5.1 Decay Heat Standard May Vary Significantly

WASHINGTON, D.C.  20555-0001

July 5, 1996

                               HEAT STANDARD MAY VARY SIGNIFICANTLY


All holders of operating licenses or construction permits for nuclear power


The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to the sensitivity of analytical results to input
parameters used with American Nuclear Society standard 5.1 on decay heat
(Reference).  It is expected that recipients will review the information for
applicability to their facilities and consider actions, as appropriate, to
avoid similar problems.  However, suggestions contained in this information
notice are not NRC requirements; therefore, no specific action or written
response is required.


American Nuclear Society standard 5.1 (ANS 5.1 standard) for decay heat
generation in nuclear power plants provides a simplified means of estimating
nuclear fuel cooling requirements that can be readily programmed into computer
codes used to predict plant performance.  The ANS 5.1 standard models the
energy release from the fission products of U-235, U-238, and Pu-239 using a
summation of exponential terms with empirical constants.  Corrections are
provided to account for energy release from the decay of U-239 and Np-239, and
for the neutron activation of stable fission products.  Although the empirical
constants are built into the standard, certain data inputs are left to the
discretion of the user.  These options permit accounting for differences in
power history, initial fuel enrichment, and neutron flux level.

Description of Circumstances

During a review of decay heat estimates calculated using various codes for the
same plant, the staff found that the predicted decay heat varied considerably. 
This was unexpected because the analyses were made using the 1979 ANS 5.1
standard and were to be used in "best-estimate" thermal-hydraulic analyses.  

The staff compared calculations of decay heat using MELCOR, TRAC, RELAP, and
a vendor code for a pressurized water reactor operating at 1933 megawatts
thermal (MWT).  The staff made 3 calculations using RELAP but varied certain 
inputs.  These calculations appear as RELAP1, RELAP2, and RELAP3 in the
following tables.  The decay heat estimates in MWT calculated for 2 hours 

9606250199.                                                            IN 96-39
                                                            July 5, 1996
                                                            Page 2 of 3

after shutdown of the reactor using the various codes and allowable inputs
were as follows:

                              MELCOR -      20.70 MWT
                              Vendor code - 22.15 MWT
                              RELAP1 -      22.02 MWT
                              RELAP2 -      26.13 MWT 
                              RELAP3 -      20.82 MWT
                              TRAC -        22.02 MWT
                              ORIGEN -      20.90 MWT

It should be noted that the differences in predicted decay heat resulted from
the parameters selected for input and not from the codes themselves. 

The last entry in the table was not calculated by the ANS 5.1 standard but was
calculated by the ORIGEN computer code.  ORIGEN does not use empirical methods
to calculate decay heat but tracks the buildup and decay of the individual
fission products within the reactor core during operation and shutdown. 
ORIGEN also includes the effect of element transmutation from neutron capture,
both in fissile isotopes and fission products.  Because ORIGEN is a rigorous
calculation of all decay heat inputs, it was used in the calculations for
decay heat in attached Figure 1 and is contrasted with attached Figure 2 using
decay heat internally calculated by RELAP to the ANS 5.1 standard. 


The staff found that the different decay heat estimates occurred because the
ANS 5.1 standard was not fully utilized in the selection of inputs.  Attached
Table 1 shows the various inputs to the codes to estimate decay heat. 
Attached Table 2 shows the effect of different assumptions on best estimate
calculations of decay heat 2 hours after shutdown of the reactor.

The variation in peak cladding temperatures with various predictions of decay
heat may be seen by comparing the attached Figures 1 and 2.  These figures are
plots of the peak cladding temperatures for a hypothetical beyond-design-basis
loss-of-feedwater event.  For Figure 1, values of decay heat predicted by the
ORIGEN code were input directly into RELAP.  Figure 2 is the same event with
the decay heat internally calculated by RELAP using the ANS 5.1 standard. 
Input assumptions to the standard were those that produced the highest values
of decay heat denoted in Table 1 as RELAP2.  The difference in predicted peak
core cladding temperatures (approximately 250� K [630� F]) demonstrates the
importance of carefully selecting required input parameters when using ANS 5.1
because analytical results may be significantly affected.  Depending on the
input parameters selected the results may be conservative or nonconservative..  
		                                                       IN 96-39
                                                            July 5, 1996
                                                            Page 3 of 3

This information notice requires no specific action or written response.  If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.

                                          signed by

                                    Brian K. Grimes, Acting Director
                                    Division of Reactor Program Management
                                    Office of Nuclear Reactor Regulation

Technical contacts:  W. L. Jensen, NRR 
                     (301) 415-2856

                     J. L. Birmingham, NRR
                     (301) 415-2829

1.  Table 1, "Decay Heat Options Using ANS 5.1, 1979"
2.  Table 2, "Relative Importance of Options 7200 Seconds 
              (2 hours) After Reactor Shutdown"
3.  Figure 1, "Peak Core Cladding Temperature Calculated by ORIGEN" and
    Figure 2, "Peak Core Cladding Temperature Calculated by ANS 5.1"
4.  List of Recently Issued NRC Information Notices

"American National Standard for Decay Heat Power in Light Water Reactors," 
American Nuclear Society Standards Committee Working Group, ANS 5.1, Approved
August 29, 1979.
.                                                                  Attachment 1 
      IN 96-39
                                                                  July 5, 1996
                                                                  Page 1 of 1

               Table 1.  Decay Heat Options Using ANS 5.1, 1979

  Si. Power     
.   Yes.  0.6. Not ANS. 1.347. 3 Yrs..  3 .RELAP1.   No.  NA. MAX. NA. Infinite.  All
  U235.RELAP2 .   Yes.  1.0. MAX. NA. Infinite.  All
 .   No.  NA. No. NA. Infinite.  All
.   Yes.  0.526. Yes. 0.713. 1.6 Yrs..  3.TRAC.   No.  NA. Yes. 1.0. Infinite.  All

Actinides:  Neutron capture by U238 produces U239 which decays into Np239
which also decays, adding to the total decay heat.  Not all the above code
inputs included actinide decay.

R-factor:  The actinide production multiplier.  The standard states that the
value of R shall be supplied and justified by the user.

G-factor:  A decay heat multiplier to account for the effect of neutron
capture in fission products.  The standard provides the option of using a
maximum value table for the G-factor or a best estimate equation for the first
10,000 seconds.

Si:  Fissions per initial fissile atom.  Si is a multiplier applied to the
G-factor equation.

Power History:  Length of full-power operation before shutdown. 

Fissile Elements:  The standard permits decay power to be fractionally
attributed to the fission products of 3 fissile isotopes U235, U238, and
Pu239.  The vendor input attributed the fractional fission product power as
0.487 from U235, 0.069 from U238, and 0.443 from U239.  The fractional fission
product power input to MELCOR was 0.647 from U235, 0.0425 from U238, and 0.31
from Pu239.  The other code inputs assumed all fission product power came from
U235..                                                                  Attachment 2 
      IN 96-39
                                                                  July 5, 1996
                                                                  Page 1 of 1

                   Table 2.  Relative Importance of Options 
                 7200 Seconds (2 hours) After Reactor Shutdown

Complete Set For:      Options as Input and Varied           Decay Heat in MWT 
MELCOR              MELCOR assumptions (see Table 1)                     20.70
                    Increase R from .526 to .6                           21.03
                    Increase Si from .713 to 1.347                       21.13
                    Operation time from 1.6 years to 3 years             21.40
                    Vendor fissile isotopes                              21.12
                    Vendor heavy element equation                        21.86
Vendor code         Vendor G-factor equation                             22.15
                    ANS 5.1 G-factor and heavy element eqs.              21.12
                    No actinides                                         18.65
                    Gmax rather than equation                            19.44
                    Infinite operation                                   20.53
RELAP1              All fission from U235                                22.02
RELAP2              Add actinides with R=1                               26.13
RELAP3              Remove actinides and G-factor                        20.82
TRAC                Include G-factor equation with Si=1                  22.02

Note:  This table illustrates the relative importance of the various options
contained in the 1979 ANS 5.1 standard.  Starting with the decay heat rate
calculated by MELCOR using the inputs listed in Table 1, the options were
changed in sequence to those used in the vendor computer code.  The options
were then changed to those used in the RELAP1 analyses, the RELAP2 analyses,
the RELAP3 analyses, and the TRAC analyses. 


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