United States Nuclear Regulatory Commission - Protecting People and the Environment

ACCESSION #: 9508080129

BW
B&W NUCLEAR TECHNOLOGIES      JHT/95-79          3315 Old Forest Road
                              July 31, 1995            P.O. Box 10935
                                             Lynchburg, VA 24506-0935
                                              Telephone: 804-832-3000
                                              Telecopy:  804-832-3663

Document Control Desk
U.S.  Nuclear Regulatory Commission
Washington, DC 20555-0001

Subject:  Potential Safety Concern on Post SBLOCA Core Recriticality
          following Moderator Dilution by RC Pump Bump

Gentlemen:

The purpose of this letter is provide you with preliminary information on
a Potential Safety Concern.  The concern relates to the potential for
post SBLOCA core recriticality caused by a reactivity insertion due to
moderator dilution following an RC pump bump or restart of natural
circulation for the B&W-designed operating plants.

Discussions for resolution of this concern are being held with the B&W
operating plant owners.  However, BWNT and the B&W Owners Group believe
it is prudent, at this time, to provide the NRC with information on the
nature of the concern.  This letter is not a notification of a
substantial safety hazard under 10CFR21.

The concern is that during recovery from SBLOCA the plant may spend some
time in boiler/condenser mode of cooling and may accumulate substantial
deborated water in the RC pump suction piping.  In addition, the boron
concentration of the coolant in the reactor vessel downcomer and lower
head region may decrease as steam flow through the reactor vessel vent
valve is condensed on the ECCS injection flow.  If, after this
accumulation, an RC pump is started or bumped or natural circulation is
reinitiated, the boron concentration of water entering the core may not
be sufficient to assure a continued subcritical condition.  Preliminary
calculations of core boron concentration indicated that, under 10CFR50.46
Appendix K conditions, a return to critical or even prompt criticality is
possible.

BWNT is currently performing preliminary calculations that will predict
the reactivity change due to the moderator dilution under realistic plant
initial conditions; 1979 ANS decay heat, non-infinite operation or Xenon
buildup, full ECCS availability, natural circulation flow rates, and
appropriate feedback.  It is expected that reasonable results will be
obtained for reactivity insertion rates under natural circulations
conditions.  Further studies will be performed considering pump bump and
pump restart conditions.  Attached is a more detailed description of the
concern, including an evaluation which explains why no real safety
concern is believed to exist.  As described in that evaluation, this
conclusion is preliminary and requires further analytical work to
document this conclusion.

                                   -2-

BWNT will provide information to the NRC concerning the evaluation and
resolution schedule for this concern at a later time.

If you have any questions concerning this matter, please contact the
undersigned at 804-832-2817 or Robert Schomaker at 804-832-2917.

                                        Very truly yours,

                                        J. H. Taylor, Manager
                                        Licensing Services
RJS/bcc

Attachment

c:   R.  C.  Jones/NRC
     A.  C.  Attard/NRC
     R.  J.  Schomaker/OF57

                    Evaluation of Safely Significance

               Moderator Boron Dilution Core Recriticality
                       Preliminary Safety Concern

1.   Problem Statement

     During SBLOCA and other events that evolve to boiler/condenser
     operation, deborated water can accumulate at the RC pump suction.
     If, after sufficient accumulation, an RC pump is started or bumped
     or natural circulation is reinitiated, deborated water may enter the
     core, possibly leading to a criticality incident.  The concern was
     initially identified by BWNT as PSC 8-81, which was resolved in
     1985.  The calculations used to resolve PSC 8-81, however,
     incorporated assumptions that are no longer valid for the B&W-
     designed NSS.  Therefore, it is necessary to revalidate that a
     severe core criticality incident will not follow the return of
     circulation following an extended period of boiler/condenser
     operation.

2.  Detailed Description

     During the course of a small break LOCA and any other transients
     that involve the partial loss of reactor coolant system (RCS)
     inventory, the steam generators may provide an energy sink for part
     or all of the core decay heat via boiler/condenser operation.  In
     this mode, steam generated in the core through boiling is passed
     through the hot legs to the steam generators and condensed.  The
     condensate is returned to the core through the cold legs.  Because
     boron volatizes with the steam production at a concentration
     substantially below that of the source, the concentration of boron
     carried with the steam is greatly reduced with the result that the
     boron concentration downstream from the steam generator is gradually
     reduced.  Evaluation of the boron carryover fraction with the MULTI-
     Q code shows that the fraction is temperature dependent and limited
     to ten percent for the conditions of interest.

                                   -2-

     Additionally, if the ECCS HPI system is in operation, steam
     accumulating in the reactor vessel upper head and upper downcomer
     (vent valve flow) will be condensed in the downcomer or cold leg.
     At high pressures, the amount of condensate approximates the ECCS
     injection, but the condensate carries only a fraction of the boron
     concentration.  Thus, the downcomer boron concentration can be
     substantially below (perhaps one-half) the borated water storage
     tank concentration.

     With aggressive core designs, the critical boron concentrations at
     reduced coolant temperatures can range from 1000 to 1500 ppm.  For
     such a design, a rapid flooding of the core with coolant of a boron
     concentration of 600 or 700 ppm could lead to a criticality
     incident.  So long as the RCS is in a separated condition, boiling
     with steam flowing through the hot legs to the steam generators, the
     core is concentrating boron and whatever low boron concentration
     makeup is provided mixes with highly concentrated liquid in the core
     and poses no problem.  However, if bulk circulation of the RCS is
     reinitiated, the highly concentrated core liquid may be displaced by
     downcomer or steam generator liquid of a low boron concentration and
     a reactivity event initiated.  Such a possibility exists if natural
     or forced circulation is reinitiated.

     The potential for a criticality concern depends on the rate of
     concentration in the core, the rate of condensation in the steam
     generators, the rate of downcomer deboration, and the amount of
     mixing between concentrated and low boration coolant during
     circulation initiation.  Theoretically, it is possible for low
     boration levels to accumulate in the steam generators.  However, an
     examination of existing SBLOCA calculations shows that the
     accumulation process may take one to two hours.  Within that time,
     core concentrations will have elevated substantially.  Natural
     convection currents within the core are expected to provide
     substantial mixing to assist in mitigating the consequence of a
     circulation restart.  Even if the core does experience a reactivity
     insertion sufficient to become critical, unless core boiling has
     been suppressed, thermal effects will be likely to produce
     sufficient reactivity feedback to control the power excursion.

                                   -3-

     The first step in the evaluation of this concern by the B&W Owners
     Group was to evaluate selected boundary conditions in an effort to
     demonstrate that the conditions under which a deboration criticality
     concern exists would not be generated by SBLOCA or the other
     possible events.  To this end, several important parameters of the
     concern have been determined.  The carryover of boron in the steam
     flow is one.  However, because the possible scenarios are highly
     dependent on operator response to the accident, it is possible that
     the plant will evolve to a condition under which a substantial
     amount of low boration coolant is present in the steam generators,
     cold legs, and downcomer.  Since unfavorable initial conditions can
     exist, an evaluation has been initiated to quantify the effects of
     limiting conditions during a restart of natural circulation.  The
     expected result of this evaluation is that thermal reactivity
     feedback during the return to power limits the power level to
     reasonable values and that a restart of natural circulation does not
     pose a concern for the safety of the plant or the eventual cooldown
     and recovery of stable plant conditions.

3.   Evaluation of Safety Significance

     Deboration events are an active issue under consideration by the
     NRC.  NUREG/CR-6266 presents an evaluation of two potential
     deboration events caused by continued injection of boron-free
     coolant.  Although the evaluation results were problematic, they
     were limited to a determination of possible reactivity insertions
     based on simplistic neutronics.  This, in combination with the low
     probability of the initiating events, indicates a need for continued
     study which, it is believed, will demonstrate that these events are
     not of consequence.

     The deboration events postulated by this safety concern are a subset
     of the possible events being considered by the NRC.  As presented,
     this concern elevates to one of safety significance only if
     reactivity insertions are sufficient to create prompt critical
     conditions

                                   -4-

     for a significant duration.  That result is not expected.  The B&W
     Owners Group position is that this PSC will not evolve to a safety
     issue.  However, more evaluations are considered necessary prior to
     disposition of the concern.

     Calculations performed in support of the CE System 80**+ SAR and
     accepted in the NRC's SER of System 80**+ are similar to those being
     conducted by the B&W Owners Group and show that the power excursion
     would be limited.  The CE calculations are described starting on
     page 15-27 of the SER.  The event is essentially the same as
     postulated herein except that the plant is an RSG PWR.  Two
     conclusions are evident:

     1)   The excursion as modelled by CE (highly conservative approach)
          allowed the plant to return to critical and even prompt
          critical during natural circulation restart.

     2)   With the simple reactivity model (only doppler feedback) used
          by CE, a stable power fraction of 10 percent was computed.  The
          use of thermal feedback and transient boron concentrations
          should substantially reduce the predicted power and should
          demonstrate a fairly rapid return to subcritical conditions.

     Thus, the B&W Owners Group is confident that there will be no
     significant safety hazard demonstrated as a result of the
     evaluations of this preliminary concern.  However, there is a
     possibility that operational considerations will be involved.
     Because of the need to provide operations with the best possible
     advice and guidance, the evaluations being conducted by the B&W
     Owners Group are based on best estimate conditions.

4.   Conclusions

     A Preliminary Safety Concern on the possible buildup of coolant with
     a low boron concentration in the RCS is under evaluation by the B&W
     Owners Group.  To date, it has not been possible to eliminate
     conclusively the evolution of the postulated unfavorable

                                   -5-

     initiating conditions.  Evaluations performed by industry and
     accepted by the NRC have shown that the consequences of the
     postulated event is limited.  The concern is not expected to evolve
     to one of safety significance, but does require continued evaluation
     to establish a basis for this expectation and to develop operational
     guidance for dealing with the possibility of these conditions
     developing.

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