Resolution of the Issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(F) (Generic Letter 92-04)

August 19, 1992                             


          10 CFR 50.54(F) (GENERIC LETTER NO. 92-04)


The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to
request information regarding the adequacy of and corrective actions for
Boiling Water Reactor (BWR) water level instrumentation with respect to the
effects of noncondensible gases on system operation.  

Background and Safety Considerations

As discussed in NRC Information Notice No. 92-54 "Level Instrumentation
Inaccuracies Caused by Rapid Depressurization," the staff is concerned that
noncondensible gases may become dissolved in the reference leg of BWR water
level instrumentation and can lead to a false high level indication after a
rapid depressurization event.  The dissolved gases which accumulate over time
during normal operation can rapidly come out of solution during
depressurization and displace water from the reference leg.  A reduced
reference leg level will result in a false high level indication.  This is
important to safety because water level signals are used for actuating
automatic safety systems and for guidance to operators during and after an

On July 29, 1992, the NRC staff held a public meeting with the Regulatory
Response Group (RRG) of the Boiling Water Reactor Owners Group (BWROG) to
discuss the effect of inaccuracies in the reactor vessel level instrumentation
system in BWRs.  During the meeting, the BWROG and its consultant, General
Electric Company (GE), presented the results of analyses assessing the safety
implications of the postulated error in level indication.  The analyses
consisted of two basic parts:  (1) an assessment of the mechanism and
potential magnitude of errors in the level instruments and (2) a review of the
relevant licensing basis transients and accidents to determine the effect of
this error on plant response, including post-accident operator actions. 

The BWROG analyses indicated that significant errors in level indication can
occur as a result of degassing the instrument reference leg if noncondensible
gas is dissolved in the reference leg and if the reactor abruptly
depressurizes below 450 psig.  


Generic Letter 92-04             - 2 -             August 19, 1992

The NRC staff reviewed the BWROG analyses and selected design basis accident
scenarios which lead to a lowering of the reactor vessel water level and has
concluded that automatic safety systems will be actuated at pressures well
above 450 psig, even for postulated worst-case noncondensible gas
concentrations in the reference legs.  Therefore, the NRC is confident that
all emergency cooling systems will initiate as they were designed to do.  In
addition, the BWROG discussed diverse signals which would also initiate ECCS
for reactor water level lowering events.  The NRC staff reviewed the backup
systems and concluded that the ECCS would be initiated by diverse signals as
analyzed by the BWROG.

After ECCS actuation, reactor water level indication is used by the operators
for long term actions (i.e., maintaining adequate reactor water level and
ensuring adequate core cooling).  Operators would not utilize only reactor
vessel level indications to determine accident mitigation actions but would
also utilize other indications such as containment pressure, temperature, and
humidity to determine accident mitigation strategies.  Additionally, events
characterized by gradual depressurization would lead to a reduced error in the
indicated level.  There are two or four reference leg columns in each plant,
depending on plant design.  The amount of noncondensible gases dissolved in
each depends primarily upon system leakage and geometry.  Because of this, a
common mode, common magnitude level indication error is unlikely.  Operators
would therefore see a mismatch in indicated level alerting them to a level
indication problem.  Finally, emergency procedure guidelines (EPGs) state that
when reactor vessel water level is indeterminate, operators should flood the
reactor vessel using at least one pump guided by the unaffected diverse
instrumentation (i.e., high containment pressure indication).  Reactor
operators are trained to deal with these situations should they occur.
Upon reviewing the information provided by the BWROG and the staff's
assessment, the staff concluded that interim plant operation is acceptable. 
The bases for the staff's conclusion are as follows:  1) the level
instrumentation is expected to initiate safety systems prior to a significant
depressurization of the reactor; 2) emergency procedures which are currently
in place in conjunction with operator training are expected to result in
adequate operator actions; and 3) an abrupt depressurization event resulting
in a common mode, common magnitude level indication error is unlikely.

For longer term operation however, the staff considers potential water level
instrumentation inaccuracies an important issue because level indication has
safety and control functions in all


Generic Letter 92-04             - 3 -             August 19, 1992

modes of BWR operation.  Furthermore, since the analyses provided are of a
generic nature and the magnitude of possible errors depends strongly upon
plant-specific factors such as system leakage and geometry, it is important
that the analyses be reviewed promptly by all individual licensees.

Basis for Compliance Determination

The level errors that could result from the effects of noncondensible gas may
prevent the level instrumentation systems in BWRs from satisfying the
following regulations:  

   (1) General Design Criterion (GDC) 13, "Instrumentation and 
       control," which requires that "Instrumentation shall 
       be provided to monitor variables and systems over their 
       anticipated ranges for normal operation, for 
       anticipated operational occurrences, and for accident 
       conditions as appropriate to assure adequate safety." 
       Existing instrumentation may not accurately monitor 
       reactor vessel water level under accident conditions.  

   (2) GDC 21, "Protection system reliability and testability,"
       which requires that "The protection system shall be  
       designed for high functional reliability...commensurate
       with the safety function to be performed."  The
       instrumentation may not be reliable under rapid
       depressurization conditions.  

   (3) GDC 22, "Protection system independence," which requires
       that "The protection system shall be designed to assure
       that the effects of natural phenomena, and of normal
       operating, maintenance, testing, and postulated
       accident not result in loss of the
       protection function."  The natural phenomena of        
       degassing may cause a loss of the reactor vessel water
       level indication function following a rapid 

   (4) Section 50.55a(h) of Title 10 of the Code of Federal
       Regulations (10 CFR 50.55a(h)), which requires 
       that protection systems, for those plants with             
       construction permits issued after January 1, 1971, 
       shall meet the requirements stated in editions of the 
       Institute of Electrical and Electronics Engineers           
       Standard "Criteria for Protection Systems for Nuclear        
       Power Generating Stations" (IEEE-279).  Section 4.20 of      
       IEEE-279 requires that "The protection system shall be
       designed to provide the operator with accurate, 


Generic Letter 92-04             - 4 -             August 19, 1992

       complete, and timely information pertinent to its own 
       status and to generating station safety."  The water
       level instrumentation for the reactor vessel may not be
       accurate after a rapid depressurization event.

Requested Actions

1.    In light of potential errors resulting from the effects of   
      noncondensible gas, each licensee should determine:  

      a.  The impact of potential level indication errors on automatic safety
          system response during all licensing basis transients and accidents;

      b.  The impact of potential level indication errors on operator's short
          and long term actions during and after all licensing basis accidents
          and transients;

      c.  The impact of potential level indication errors on operator actions
          prescribed in emergency operating procedures or other affected
          procedures not covered in (b).

2.    Based upon the results of (1), above, each licensee should notify the
      NRC of short term actions taken, such as:

      a.  Periodic monitoring of level instrumentation system leakage; and, 

      b.  Implementation of procedures and operator training to assure that
          potential level errors will not result in improper operator actions.

3.    Each licensee should provide its plans and schedule for corrective
      actions, including any proposed hardware modifications necessary to
      ensure the level instrumentation system design is of high functional
      reliability for long term operation.  Since this instrumentation plays
      an important role in plant safety and is required for both normal and
      accident conditions, the staff recommends that each utility implement
      its longer term actions to assure a level instrumentation system of high
      functional reliability at the first opportunity but prior to starting up
      after the next refueling outage commencing 3 months after the date of
      this letter.   

Generic Letter 92-04             - 5 -             August 19, 1992

Required Information

Because of the importance of plant-specific aspects of this issue and the
potential magnitude of the errors, the staff requires, pursuant to 10 CFR
50.54(f) and Section 182 of the Atomic Energy Act, that you provide a response
to this letter by September 27, 1992.  

Merely committing to evaluate the safety significance as part of the
individual plant examination (IPE) program is not an acceptable alternative to
the actions described herein, since the licensee should resolve this issue as
a matter of compliance.

Backfit Discussion

In accordance with NRC procedures, the actions requested herein are considered
a backfit to assure that facilities are in compliance with existing regulatory
requirements discussed above.  Thus, a backfit analysis is not required by 10
CFR 50.109(a)(4)(i), and the staff performed a documented evaluation as
discussed in 10 CFR 50.109(a)(6).  The documented evaluation is provided in
the preceding discussions.

Burden Information

This request is covered by Office of Management and Budget Clearance Number
3150-0011, which expires May 31, 1994.  The estimated average number of burden
hours is 200 person hours for each licensee response, including the time
required to assess the questions, search data sources, gather and analyze the
data, and prepare the required response.  These estimated average burden hours
pertain only to the identified response-related matters and do not include the
time for actual implementation of the requested actions.  Comments on the
accuracy of this estimate and suggestions to reduce the burden may be directed
to Ronald Minsk, Office of Information and Regulatory Affairs (3150-0011),
NEOB-3019, Office of Management and Budget, Washington, D.C. 20503 and to the
U.S. Nuclear Regulatory Commission, Information and Records Management Branch,
Division of Information Support Services, Office of Information and Resources
Management, Washington, D.C. 20555.

Although no specific request or requirement is intended, the following
information would be helpful to the NRC in evaluating the cost of complying
with this generic letter:

(1)   the licensee staff time and costs to perform requested inspections,
      corrective actions, and associated testing;

(2)   the licensee staff's time and costs to prepare the requested reports and

Generic Letter 92-04             - 6 -             August 19, 1992

(3)   the additional short-term costs incurred as a result of the inspection
      findings such as the costs of the corrective actions or the costs of
      down time; and

(4)   an estimate of the additional long-term costs which will be incurred in
      the future as a result of implementing commitments such as the estimated
      costs of conducting future inspections or increased maintenance.

Please address your response to this generic letter to the U.S. Nuclear
Regulatory Commission, Attn:  Document Control Desk, Washington, D.C.  20555
pursuant to 10 CFR 50.4(a) of the NRC's regulations.


                              ORIGINAL SIGNED BY

                              James G. Partlow
                              Associate Director for Projects
                              Office of Nuclear Reactor Regulation

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Technical Contact:      Timothy E. Collins, NRR
                        (301) 504-2897

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