Initiation of The Individual Plant Examination For Severe Accident Vulnerabilities -10 CFR 50.54(f) (Generic Letter No. 88-20, Supplement No. 1)

                                 August 29, 1989

TO:       ALL LICENSEES HOLDING OPERATING LICENSES AND CONSTRUCTION PERMITS 
          FOR NUCLEAR POWER REACTOR FACILITIES 

SUBJECT:  INITIATION OF THE INDIVIDUAL PLANT EXAMINATION FOR SEVERE ACCIDENT 
          VULNERABILITIES-10 CFR 50.54(f) - GENERIC LETTER NO. 88-20, 
          SUPPLEMENT NO. 1

This letter announces the availability of NUREG-1335, "Individual Plant 
Examination: Submittal Guidance," (enclosed) and initiation of the Individual 
Plant Examination (IPE) process.  In accordance with Generic Letter No. 88-20,
licensees are requested to submit within 60 days from the date of the Federal 
Register notice announcing the availability of the enclosed guidance document,
their proposed programs for completing their IPEs.  The proposed programs 
should be submitted to the U.S. Nuclear Regulatory Commission, Document 
Control Desk, Washington, DC 20555, and should: 

1.  Identify the method and approach selected for performing the IPE, 
2.  Describe the method to be used, if it has not been previously submitted 
    for staff review (the description may be referenced), and 
3.  Identify the milestones and schedules for performing the IPE and submit-
    ting the results to the NRC. 


NUREG-1335 was published in draft form in January 1989 and issued for public 
comment.  All comments received, including those made during the IPE Workshop 
on February 28 through March 2, 1989, and staff responses to them, may be 
found in Appendix C of NUREG-1335.  Licensees may find it useful in preparing 
their initial responses to review two options discussed on the matters of 
internal flooding and submittal format in Appendix C, in response to comments 
5.1 and 11.3 respectively. 

In accordance with a recent Commission decision on staff recommendations for 
enhancements to BWR Mark I plants, the staff plans to communicate directly 
with each licensee who possesses a Mark I plant on the matter of a hardened 
vent path.  A summary of the staff's conclusions and recommendations for other
potential Mark I enhancements is given in the enclosure hereto, for 
consideration in each Mark I licensee's IPE.  Additional information is 
contained in SECY 89-017, "Mark I Containment Performance Improvement 
Program," dated January 23, 1989.  The staff expects to issue conclusions and 
recommendations for all other plants and containment types in about 6 months 
for similar consideration in IPEs. 

Regulatory Basis 

Generic Letter 88-20 was issued pursuant to 10 CFR 50.54(f).  A copy of the 
10 CFR 50.54(f) evaluation which justified issuance of Generic Letter 88-20 






8908300001
.


Generic Letter 88-20,              - 2 -                    August 29, 1989
Supplement No. 1


is in the Public Document Room.  This supplement does not change the scope of 
Generic Letter 88-20.  Therefore, there is no additional burden associated 
with this letter, and an OMB clearance number is not required.  

                                   Sincerely, 



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

Enclosures:  
1.  NUREG-1335, "Individual 
    Plant Examination: 
    Submittal Guidance," 
    August 1989 
2.  Mark I Containment 
    Performance Improvements 
3.  List of Most Recently Issued
    Generic Letters

.


                                   Enclosure 2

                   Mark I Containment Performance Improvements

The NRC staff has identified certain containment performance improvements that
would likely reduce the vulnerability of the Mark I containment to severe 
accident challenges (Ref. 1 and 2).  The Commission expects that licensees of 
Mark I plants will seriously consider these improvements during their 
Individual Plant Examinations.  It should be noted that these improvements 
should be considered in addition to improvements that stem from the evaluation
and implementation of the hardened vent. 

(a)  Alternate Water Supply for Drywell Spray/Vessel Injection: 

An important improvement would be to employ a backup or alternate supply of 
water and a pumping capability that is independent of normal and emergency AC 
power.  By connecting this source to the low pressure residual heat removal 
system (RHR) system as well as to the existing drywell sprays, water could be 
delivered either into the reactor vessel or to the drywell, by use of an 
appropriate valving arrangement. 

An alternate source of water injection into the reactor vessel would greatly 
reduce the likelihood of core melt due to station blackout or loss of 
long-term decay heat removal, as well as provide significant accident 
management capability. 

Water for the drywell sprays would also provide significant mitigative 
capability to cool core debris, to cool the containment steel shell to delay 
or prevent its failure, and scrub airborne particulate fission products from 
the atmosphere. 

A review of some BWR Mark I facilities indicates that most plants have one or 
more diesel driven pumps which could be used to provide an alternate water 
supply.  The flow rate using this backup water system may be significantly 
less than the design flow rate for drywell sprays.  The potential benefits of 
modifying the spray headers to assure a spray were compared to having water 
run out of the spray nozzles.  Fission product removal in the small crowded 
volume in which the sprays would be effective was judged to be small compared 
with the benefit of having a water pool on top of the core debris.  

(b)  Enhanced Reactor Pressure Vessel (RPV) Depressurization 
     System Reliability: 

The Automatic Depressurization System (ADS) consists of relief valves which 
can be manually operated to depressurize the reactor coolant system.  
Actuation of the ADS valves requires DC power and pneumatic 
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                                      - 2 -


supply.  In an extended station blackout after station batteries have been 
depleted, the ADS would not be available and the reactor would be 
re-pressurized.  With enhanced RPV depressurization system reliability, 
depressurization of the reactor coolant system would have a greater degree of 
assurance.  Together with a low pressure alternate source of water injection 
into the reactor vessel, the major benefit of enhanced RPV depressurization 
reliability would be to provide an additional source of core cooling which 
could significantly reduce the likelihood of high pressure severe accidents, 
such as from the short-term station blackout. 

Another important benefit is in the area of accident mitigation.  Reduced 
reactor pressure would greatly reduce the possibility of core debris being 
expelled under high pressure, given a core melt and failure of the reactor 
pressure vessel.  Enhanced RPV depressurization system reliability would also 
delay containment failure and reduce the quantity and type of fission products
ultimately released to the environment.  In order to increase reliability of 
the RPV depressurization system, assurance of electrical power beyond the 
requirements of existing regulations may be necessary.  Performance of the 
cables needs to be reviewed for temperature capability during severe accidents
as well as the capacity of the pneumatic supply.  

(c) Emergency Procedures and Training: 

NRC has recently reviewed and approved Revision 4 of the BWR Owners Group EPGs
(General Electric Topical Report NEDO-31331, BWR Owner's Group "Emergency 
Procedure Guidelines, Revision 4," March 1987). 

Revision 4 to the BWR Owners Group EPG is a significant improvement over 
earlier versions in that they continue to be based on symptoms, they have been
simplified, and all open items from previous versions have been resolved.  The
BWR EPGs extend well beyond the design bases and include many actions 
appropriate for severe accident management. 

The improvement to EPGs is only as good as the plant-specific EOP 
implementation and the training that operators receive on use of the improved 
procedures.  The NRC staff encourages licensees to implement Revision 4 of the
EPGs and recognize the need for proper implementation and training of 
operators. 

1.   E. Claiborne et al., "Cost Analysis for Potential BWR Mark I 
     Containment Improvements," Science and Engineering Associates Inc., 
     NUREG/CR-5278, SEA 87-253-07-A:1, January 1989.

2.   Wagner, K. C. et al., "An Overview of BWR Mark I Containment 
     Venting Implications, Addendum 1:  An Evaluation of Potential Mark I 
     Containment Improvements, NUREG/CR-5225 Addendum 1, July 1989.