Information Notice No. 97-06: Weaknesses in Plant-Specific Emergency Operating Procedures for Refilling the Secondary Side of Dry Once-Through Steam Generators

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                     OFFICE OF NUCLEAR REACTOR REGULATION
                         WASHINGTON, D.C.  20555-0001

                                 March 4, 1997


NRC INFORMATION NOTICE 97-06:  WEAKNESSES IN PLANT-SPECIFIC EMERGENCY          
                               OPERATING PROCEDURES FOR REFILLING THE          
                               SECONDARY SIDE OF DRY ONCE-THROUGH STEAM        
                               GENERATORS

Addressees

All holders of operating licenses or construction permits for nuclear power
reactors with once-through steam generators (OTSGs).

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to a potential need for guidance in plant-specific
emergency operating procedures (EOPs) regarding the refilling of dried-out
OTSGs.  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.

Description of Circumstances

On May 19, 1996, the reactor at Arkansas Nuclear One, Unit 1 (ANO-1) tripped
as a result of circuitry problems in the main feedwater control system.  The
reactor trip was complicated by a main steam safety valve (MSSV) which opened,
as designed, in response to an increase in pressure in the associated OTSG,
but failed to reseat after the steam generator depressurized.  Consequently,
the steam generator boiled dry and the reactor coolant system (RCS) cooled
down rapidly.  The MSSV was later gagged shut.  

Operators had been trained to respond to this type of event and, therefore,
were familiar with the generic guidance to refill the dried-out steam
generator.  However, the plant-specific "overcooling" EOP (the EOP that
addresses situations in which the RCS experiences such a cooldown) contained
no guidance for refilling the steam generator.  Operators consulted with their
support staff who then drafted a modification to the "overcooling" procedure
to provide guidance on refilling the steam generator.  After obtaining
concurrence on the modification from the OTSG vendor and approval of the
revised procedure from the ANO-1 plant safety committee, the licensee
implemented the procedure.  Operators used the revised procedure to refill the
dried-out steam generator and proceeded to stabilize the plant in the
hot-shutdown mode of operation.


9703040265.                                                            IN 97-06
                                                            March 4, 1997
                                                            Page 2 of 3


Generic guidelines for mitigating overcooling events in nuclear power reactors
designed by the Babcock & Wilcox Company (B&W) have been provided by the B&W
Owners Group.   Section III.D, Step 7, of the Generic Emergency Operating
Guidelines (Volume 1) provides guidance for refilling a dried-out steam
generator during events that result in overcooling of the RCS.  In addition,
Section III.G of the Bases (Volume 3) discusses steam generator tube axial
load concerns associated with cooldown scenarios.  Items 3.6 and 3.8 of
Section III.G provide tensile and compressive tube-to-shell temperature
difference limits and strategies for minimizing tube compressive and tensile
stresses during transients.  At the time of the May 19, 1996 event, the ANO-1
licensee was in the process of revising the plant's "overcooling" EOP to bring
it into agreement with the guidance document.  In addition, lessons learned
from a similar event that occurred at Oconee Unit 3 on August 10, 1994, were
also being incorporated into this revision.  Had these changes been completed,
operators would have had the guidance necessary to refill the steam generator
and would not have had to wait for the procedure to be revised.

Discussion

During normal operation, the water and steam in a steam generator provide
thermal communication between the shell and the tubes.  When a steam generator
becomes dry, the thermal communication between the shell and the tubes is
lost.  Therefore, temperature changes in the RCS (including the steam
generator tubes) are not transferred to the shell of the steam generator,
allowing the temperatures to diverge rapidly.  The resultant tube-to-shell
temperature difference, and consequent differential thermal expansion between
the tubes and the shell, can place unacceptable stresses on the tubes.  These
stresses can be either tensile or compressive during the dry-out /refill
event.  The loads are of a tensile nature when the tubes are cooled faster
than the shell and of a compressive nature when the shell is cooled faster
than the tubes.  The differential pressure across the tube wall and the
preload at fabrication also contribute to the tube axial loads.

The results of an analysis to determine the response of a dry, pressurized
OTSG to refill under various conditions were used to develop plant operating
procedures and guidance for refilling a dry OTSG.  Various combinations of
initial OTSG tube temperatures and auxiliary feedwater (AFW) flow rates were
analyzed and the thermal-hydraulic results were used to determine acceptable
AFW refill flow rates for various tube-to-shell temperature differences and
initial tube temperatures.  The allowable normal operating tube loads were
established using NRC-developed minimum wall thicknesses for degraded tubes
and the acceptance criteria provided in NRC Regulatory Guide 1.121, "Bases for
Plugging Degraded PWR Steam Generator Tubes" and the ASME Boiler and Pressure
Vessel Code.
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                                                            March 4, 1997
                                                            Page 3 of 3


B&W analyses have demonstrated that the steam generator tube loads are bounded
by loads occurring during routine plant heatup and cooldown.  Therefore, if
up-to-date procedures and guidelines are adhered to during the dry-out/refill
event, there should be no increase in the likelihood of a tube failure. 
During the ANO event, the tube-to-shell temperature difference exceeded the
allowable limits and, therefore, the loads imposed by the event were analyzed
to ensure continued operability of the steam generator. 

As a result of recent operating experiences, the B&W Owners Group is currently
reviewing the present guidance on refilling dried-out OTSGs.  Generic
guidelines and plant-specific EOPs are intended to reflect operating
experience and the lessons learned from such experience.  Therefore, the
generic guidelines may change after the B&W Owners Group completes its review. 


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 D.B. Matthews

                                       Thomas T. Martin, Director
                                       Division of Reactor Program Management
                                       Office of Nuclear Reactor Regulation

Technical contacts:  Mohammed A. Shuaibi, NRR   
                     (301) 415-2859                   
                     E-mail:  mas4@nrc.gov                  

                     Jai Rajan, NRR
                     (301) 415-2788
                     E-mail:  jrr@nrc.gov

                     Chu-Yu Liang, NRR                
                     (301) 415-2878                   
                     E-mail:  cyl@nrc.gov