TRANSMITTED TO DOE/M. WHITAKER 04/19/95

DEFENSE NUCLEAR FACILITIES SAFETY BOARD

September 16, 1994

MEMORANDUM FOR:     G. W. Cunningham, Technical Director

COPIES:        Board Members

FROM:               Daniel G. Ogg, Program Manager, INEL

SUBJECT:       Tritium Testing and Safety Analyses, Idaho National
               Engineering Laboratory, Advanced Test Reactor, Report of Site
               Visit, August 30-September 1, 1994

1. Purpose:  This memorandum documents the results of the DNFSB staff visit to
   the Idaho National Engineering Laboratory (INEL).  The trip focused on a new
   tritium production test being conducted at the Advanced Test Reactor (ATR)
   for the Office of Reconfiguration (DP-25).  Additionally, the staff reviewed
   the design basis of the reactor, excluding the in-pile loop experiments. 
   The review team included DNFSB staff members Daniel Ogg, Joseph Roarty, and
   Sol Pearlstein.

2. Summary:  The tritium production feasibility test at the ATR presents little
   challenge to the safety of the core of the reactor.  The test targets or
   "shadow slugs" are located around the periphery of the core, and the amount
   of tritium to be produced will measure only a few grams.  The integrity of
   the shadow slugs, fabricated at the Savannah River Site (SRS), has been
   poor, with a failure rate of 80 percent.  Measures to prevent a defective
   shadow slug from reaching the ATR are essential to preclude a tritium
   release during the test.

   The DNFSB staff review of the design basis of the ATR disclosed several
   characteristics that indicate the safety margin in the ATR is less than that
   of the SRS K Reactor (after restart) or of commercial reactors.  The
   following observations were noted:

   a.  Irradiation induced growth of beryllium (Be) results in cracking and
       bowing of the ATR reflector such that a complete replacement of core
       internals is required every 6-8 years.  Although this condition could
       constitute a potential safety issue due to the release of Be fragments
       and distortion of the reflector, considerable experience has been
       gained from the operation of the core since 1968.  The Staff
       acknowledges that this problem has long been identified and evaluated
       and believes that continued careful monitoring of the reflector is
       appropriate to mitigate challenges to the safety of the core.

   b.  The thermal design of the ATR can accommodate a primary coolant pipe
       break equivalent to 3 inches in diameter.  This break size is less than
       that in the design basis for the SRS K Reactor.  The K Reactor analysis
       assumed a Double Ended Guillotine Break of a primary system pipe.

   c.  The ability to inspect ATR primary coolant piping is limited as some
       piping is inaccessible.  Intergranular Stress Corrosion Cracking
       (IGSCC) was noted at K Reactor during a comprehensive inspection of the
       reactor vessel and piping, and sections of piping were replaced.  ATR
       operators conduct periodic inspections for leaks and no leaks due to
       IGSCC have been detected throughout the life of the ATR.

   d.  The ATR primary coolant nozzles are located below the reactor core; a
       break in this piping would drain the core.  Further, the location of
       heat exchangers is below the elevation of the core, which nullifies
       natural convection cooling of the core following a loss of pumping
       power.  These conditions are not permitted in commercial pressurized
       water reactors, but did exist in K Reactor.  An additional emergency
       coolant injection loop and a seismically qualified gadolinium-nitrate
       poison injection system were added as safety enhancements prior to the
       restart of K Reactor.  ATR has provided a fire main emergency cooling
       system.

   e.  The maximum power density in the ATR is about twice that of commercial
       reactors or K Reactor.  During a reactivity addition accident, the
       maximum local heat flux exceeds 3x106 Btu/hr-ft2 in ATR.  The DNFSB
       staff believes that investigation of a rate response system and lowered
       trip settings is appropriate to mitigate this type of accident.

3. Background:  The ATR is operated by EG&G for the DOE to test reactor
   materials and to produce radioisotopes and has been in operation since 1968. 
   The EG&G operations staff at ATR completed the third Core Internals Change-
   out (CIC) in July 1994, and commenced a new test cycle on August 22, 1994. 
   While materials testing for the Naval Nuclear Propulsion Program continues
   at the ATR, the current test cycle also includes a new test, sponsored by
   the DOE Office of Reconfiguration (DP-25), called the Tritium Validation and
   Feasibility Study.  The purpose of this test is to validate predicted
   tritium production, demonstrate target survivability, and predict optimum
   tritium production in the ATR.

4. Discussion:

   a.  Tritium Demonstration Test:  Lithium targets, originally fabricated for
       the K Reactor at SRS, are being irradiated in the ATR.  The objective
       is to demonstrate that modest but sufficient amounts of tritium can be
       produced in the ATR and that such production could extend the date, by
       approximately two years, when DOE must decide its long-term method for
       producing tritium.  The targets will be placed in ATR for two cycles,
       40 days each.  At the end of that time the irradiated targets will be
       returned to SRS for extraction of tritium. The integrity of the lithium
       targets is important to avoid release of tritium in the ATR system.  At
       SRS, an 80 percent failure rate was observed for the test specimens
       examined.  Only targets showing no defects were sent to the ATR for
       irradiation.  If this program is to continue, strict quality assurance
       procedures must be used to select or manufacture targets for
       irradiation in the ATR.

   b.  ATR Design Review:  The ATR contains a number of similarities to the
       Savannah River Site K Reactor that allow an approximate comparison to
       be made between the thermal design margin of each reactor.  The cores
       are both designed with curved aluminum plates, and for down flow of the
       coolant, and operate at relatively low temperature and pressure.  The
       local power density in ATR is significantly higher than that of the K
       Reactor with heat fluxes about two times higher.

       The K Reactor has, as a design basis accident, a Double Ended
       Guillotine Break of a large diameter primary coolant pipe, while the
       ATR design is limited to an equivalent 3-inch diameter primary coolant
       break.

       Another approximate comparison of the thermal design margin in ATR and
       K Reactor is provided by examining the heat flux reached during a
       design basis reactivity addition accident.  In ATR, the steady state
       power of 250 MW increases to 460 MW, and a local heat flux of 1.8 x 106
       Btu/hr-ft2 increases to a value of over 3 x 106 Btu/hr-ft2.  In K
       Reactor, the peak heat flux is about one-half of these values.  A
       reevaluation of the ATR thermal design, based on modern methodologies
       (including single fault assumptions as required by DOE Order 5480.30,
       Nuclear Reactor Safety Design Criteria), is appropriate to identify
       ways (e.g. lower scram trip, rate protection) to mitigate overpower
       accidents.

   c.  ATR Aging Effects:  The ATR has operated for about 25 years and is
       potentially vulnerable to IGSCC of the 304L primary coolant piping. 
       Such a condition was noted in Savannah River Site  reactors, and a
       section of the primary coolant piping was replaced in K Reactor. 
       Reactor vessel repair was also required in C Reactor.

       ATR conducts periodic in-service inspections and a walk-down following
       each start-up to confirm the absence of leaks.  None has been found to
       date; however, it is noted that some primary coolant piping is
       inaccessible and cannot be directly inspected.

       The ATR organization does not include a materials/metallurgy
       specialist.  Such an individual is available on an on-call basis.  In
       view of the safety dependence on materials integrity, additional
       support in this area may be warranted.

       The DNFSB staff believes that the ATR organization can be aided
       significantly by a stronger tie with the Savannah River Site
       organization.  Although ATR has some awareness of SRS activities,
       contacts have been limited.  The K Reactor Safety Analysis Report (SAR)
       completed in 1992 is particularly relevant as are several issues
       (corrosion, water-hammer, corrosion-erosion, etc.) which have been
       investigated by personnel at the Defense Waste Processing Facility, F-
       Canyon, and the Savannah River Technology Center.

   d.  ATR Beryllium Reflector Cracking:  Irradiation induced growth of
       beryllium (Be) results in cracking and distortion of the ATR reflector,
       such that a complete replacement of core internals is required every 6-
       8 years.  This condition constitutes a potential safety issue as the
       release of Be fragments and/or the distortion of the reflector could
       block a coolant channel or interfere with movement of a safety rod. 
       This problem was identified early in the life of the ATR and, upon
       evaluation, led to the current practice of periodic replacement of the
       reflector.  The DNFSB staff believes that continued monitoring of the
       reflector is appropriate to prevent challenges to the safety of the
       core.

       In view of this situation, it is appropriate to recall an experience
       which occurred at the Rochester Gas and Electric Plant (GINNA) where a
       loose part led to the rupture of a steam generator tube and loss of
       primary coolant.  It is to be noted that many commercial nuclear power
       plant operators responded by installing a Loose Parts Monitoring
       System.  This equipment was also installed at the K Reactor.

   e.  Reactivity Analysis:  The principal reactivity analysis tool used at
       ATR is PDQ, a 4-energy group, 2- and 3-dimensional fine mesh diffusion
       theory code.  This code is widely used in the design of thermal
       reactors and is also used for burnup calculations.  However, the use of
       PDQ rectilinear coordinates to describe the cylindrical shape of cells
       and reflectors found in ATR can introduce jagged boundaries.

       Data sets of temperature dependent 4-group cross sections were obtained
       from a variety of sources and, therefore, do not constitute an easily
       documentable reference set of data, but the ATR group is working
       towards deriving data from a standard reference data set.  It is also
       not clear that 4 energy groups are adequate to describe the Be hardened
       spectra which differ from those found in conventional light water
       reactors.  PDQ calculations yielded keff=0.985 for measured critical
       loadings.  This tendency to underestimate criticality is non-
       conservative.  The analysis group was urged to work toward improving
       its understanding of physics and methods to reduce the discrepancy
       between calculated and measured absolute criticalities to perhaps
       within 0.5 percent or at least understand the biases that exist and
       where they should be applied.  Recently, the Monte Carlo code, MCNP,
       with continuous energy and combinatorial geometry treatments has been
       implemented and can be used to explore these factors.

       The calculation of incremental reactivity changes, e.g., temperature
       defect, burnup, sample and control worths, is quite accurate. 
       Calculated and measured flux distributions agreed to within a few
       percent generally and within 8 percent at boundaries.  Furthermore,
       reliance on calculations to establish safety margins is alleviated by
       use of the ATR critical facility (ATRC).  This full scale look-a-like
       facility is used to verify design changes, target perturbations, and
       fuel loadings that combine new and spent fuel.  The ATRC is considered
       a vital factor in the ATR program.  At a cost of $300K per year, it is
       a small investment within the $44M ATR program and is cost effective in
       obtaining safety information and savings in ATR time.

   f.  Core Internals Change-out (CIC) and Equipment Upgrades:  The CICs at 6-
       8 year intervals reflect the application of good ALARA principles.  The
       last change-out resulted in an total worker dose of 25 person-rem
       compared to 60 person-rem for the previous change-out.  It has been
       noted that during a CIC there is no appreciable decay in the radiation
       background from the permanent part of the ATR.  This suggests that some
       radioactive isotopes are formed with half-lives comparable with or
       longer than the few months time required to complete the change-out. 
       If the half-lives are comparable with the 6-8 year interval between
       change-outs, the background radiation could continue to increase making
       change-outs more difficult with increasing ATR age.  The longest lived
       gamma activity from the stainless steel components is expected to be
       from 60Co (t1/2=5.3 years).  It might be prudent to analyze whether
       there are known impurities or minor constituents that, if activated,
       might increase the radiation background so as to limit the useful life
       of the core.  No radiation background measurements from previous
       change-outs are available for comparison.

       It was also noted that during the life of the beryllium reflectors,
       about 100 grams of tritium is expected to be formed through the 9Be
       (n,t) reaction.  The feasibility of extracting tritium from the
       reflector was not discussed.

       There is a program to systematically replace aging components.  The
       yearly capital improvement budget is about $500K.  Within the last few
       years, the reactor control system and many electrical components have
       been replaced.  An upgrade to the battery room ventilation system is
       also planned.  Currently, the emergency batteries are charged one at a
       time because the ventilation system is judged by EG&G to be inadequate
       for the removal of hydrogen generated during a two-battery charge.

5. Future Staff Reviews:  Future activity relative to the production of tritium
   will be closely followed by the DNFSB staff.  Other reviews of ATR will
   continue on a periodic basis at a frequency of approximately three per year.