Request for Information Concerning Status of Implementation of Unresolved Safety Issue (USI) Requirements (Generic Letter 89-21)



                              October 19, 1989 



As part of our continuing effort to validate staff understanding regarding 
implementation of significant regulatory issues, the staff is conducting a 
comprehensive review of the implementation status of unresolved safety issues
(USIs).  An important aspect of this effort is to ensure that the licensee and

NRC agree on the status of USI resolution implementation at each facility.  
The purpose of this letter is to request your review and reporting of the 
status of implementation of USIs for which a final technical resolution has 
been achieved and which are applicable to your facility.

To assist you in this effort, I have enclosed a table of USIs for which a 
final technical resolution has been achieved (Enclosure 1).  This table 
indicates other information, such as multiplant action (MPA) number, generic 
letter number, applicability, and reference NUREG number.  For your 
facility, determination of requirements for a particular USI may necessitate 
review of applicable generic letters, NUREG documents, or plant-specific 
correspondence.  For your information, a summary of the resolution of each 
USI is provided in Enclosure 2.

As in the case of our earlier correspondence related to the status of imple-
mentation of TMI Task Action Plan items, implementation should be considered 
complete when activities have been performed necessary to satisfy the require-
ments (or assumptions) made in the staff's technical resolution of the 
particular USI.  If you have not fully completed an item, we ask you to mark 
up the enclosure to reflect your projected implementation date.  You should 
add a short note identifying remaining work (e.g., hardware, procedures, 
training, technical specifications).  More explicit instructions are 
provided as part of Enclosure 1.

Your NRC Project Manager is developing data sheets that identify significant 
plant-specific correspondence between each licensee and the staff relating 
to a particular USI.  Once we have researched agency files we will provide 
this information to your staff.  This will ensure we both have a clear 
record of major actions regarding the USI.  The Project Manager can provide 
additional clarification which may be of assistance to you and will work 
with your staff to identify plant-specific references.

We request that this information be provided within 30 days of receipt of this

letter.  The information we are requesting will be utilized to validate and 
update our existing databases so that we will have an accurate and complete 
understanding of the status of USI implementation at each nuclear power 


Generic Letter 89-21                - 2 -               October 19, 1989

This request is covered by Office of Management and Budget Clearance Number 
3150-0011, which expires December 31, 1989.  The estimated average burden 
hours is 80 person hours per plant, including searching data sources, 
gathering and analyzing the data, and preparing the required letter.  Send 
comments regarding this burden estimate or any other aspect of this 
collection of information, including suggestions for reducing this burden, 
to the Records and Reports Management Branch, Division of Information 
Support Services, Office of Information Resources Management, U.S. Nuclear 
Regulatory Commission, Washington, D.C. 20555; and to the Paperwork 
Reduction Project (3150-0011), Office of Management and Budget, 
Washington, D.C. 20503.  


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

1. USI Table
2. USI Issues Summary
3. List of Most Recently Issued NRC Generic Letters

                                                       ENCLOSURE 1

                              HAS BEEN ACHIEVED

NUMBER           TITLE               REF. DOCUMENT       APPLICABILITY 

A-1       Water Hammer               SECY 84-119              All
                                     NUREG-0927, Rev. 1
                                     NUREG-0993, Rev. 1
                                     NUREG-0737 Item
                                     SRP revisions

A-2/      Asymmetric Blowdown        NUREG-0609               PWR
MPA D-10  Loads on Reactor Primary   GL 84-04, GDC-4
          Coolant Systems

A-3       Westinghouse Steam         NUREG-0844               W-PWR
          Generator Tube Integrity   SECY 86-97
                                     SECY 88-272
                                     GL 85-02
                                     (No requirements)

A-4       CE Steam Generator Tube    NUREG-0844, SECY 86-97   CE-PWR
          Integrity                  SECY 88-272
                                     GL 85-02
                                     (No requirements)

A-5       B&W Steam Generator        NUREG-0844, SECY 86-97   B&W-PWR
Tube Integrity             SECY 88-272
                                     GL 85-02
                                     (No Requirements)

A-6       Mark I Containment         NUREG-0408             Mark I-BWR
          Short-Term Program



                                    - 2 -

NUMBER           TITLE               REF. DOCUMENT        APPLICABILITY 

A-7/      Mark I Long-Term           NUREG-0661             Mark I-BWR
D-01      Program                    NUREG-0661 Suppl. 1
                                     GL 79-57

A-8       Mark II Containment        NUREG-0808             Mark II-BWR
          Pool Dynamic Loads         NUREG-0487, Suppl. 1/2
                                     GDC 16

A-9       Anticipated Transients     NUREG-0460, Vol. 4        All
          Without Scram              10 CFR 50.62

A-10/     BWR Feedwater Nozzle       NUREG-0619                BWR
MPA B-25  Cracking                   Letter from DG Eisenhut
                                     dated 11/13/80
                                     GL 81-11

A-11      Reactor Vessel Material    NUREG-0744, Rev. 1        All
          Toughness                  10 CFR 50.60/

A-12      Fracture Toughness of      NUREG-0577, Rev. 1        PWR
          Steam Generator and        SRP Revision
          Reactor Coolant Pump       5.3.4

A-17      Systems Interactions       Ltr:  DeYoung to          All
                                     licensees - 9/72
                                     NUREG-1174, NUREG-
                                     1229, NUREG/CR-3922,
                                     NUREG/CR-4261, NUREG/
                                     CR-4470, GL 89-18
                                     (No requirements)

A-24/     Qualification of Class     NUREG-0588, Rev. 1        All
MPA B-60  1E Safety-Related          SRP 3.11
          Equipment                  10 CFR 50.49
                                     GL 82-09, GL 84-24
                                     GL 85-15

                                    - 3 -

NUMBER           TITLE               REF. DOCUMENT       APPLICABILITY 

A-26/     Reactor Vessel Pressure    DOR Letters to          PWR
MPA B-04  Transient Protection       Licensees 8/76
                                     SRP 5.2
                                     GL 88-11

A-31      Residual Heat Removal      NUREG-0606           All OLs After
          Shutdown Requirements      RG 1.113,                01/79.
                                     RG 1.139
                                     SRP 5.4.7

A-36/     Control of Heavy Loads     NUREG-0612              All
C-10,     Near Spent Fuel            SRP 9.1.5
C-15                                 GL 81-07, GL 83-42,
                                     GL 85-11
                                     Letter from DG
                                     Eisenhut dated

A-39      Determination of SRV       NUREG-0802              BWR
          Pool Dynamic Loads         NUREGs-0763,0783,0802
          and Pressure Transients    NUREG-0661

A-40      Seismic Design             SRP Revisions, NUREG/   All
          Criteria                   CR-4776, NUREG/CR-0054,
                                     NUREG/CR-3480, NUREG/
                                     CR-1582, NUREG/CR-1161,
                                     NUREG-1233, NUREG-4776

A-42/     Pipe Cracks in Boiling     NUREG-0313, Rev. 1      BWR
MPA B-05  Water Reactors             NUREG-0313, Rev. 2
                                     GL 81-03, GL 88-01 


                                    - 4 -

NUMBER           TITLE               REF. DOCUMENT       APPLICABILITY 

A-43      Containment Emergency      NUREG-0510,             All
          Sump Performance           NUREG-0869, Rev. 1
                                     NUREG-0897, R.G.1.82 
                                     (Rev. 0), SRP 6.2.2
                                     GL 85-22
                                     No Requirements

A-44      Station Blackout           RG 1.155                All
                                     10 CFR 50.63

A-45      Shutdown Decay Heat        SECY 88-260             All
          Removal Requirements       NUREG-1289
                                     SECY 88-260
                                     (No requirements)

A-46      Seismic Qualification      NUREG-1030              All
          of Equipment in            NUREG-1211/
          Operating Plants           GL 87-02, GL 87-03

A-47      Safety Implication         NUREG-1217, NUREG-      All
          of Control Systems         1218
                                     GL 89-19

A-48      Hydrogen Control           10 CFR 50.44          All, except 
          Measures and Effects       SECY 89-122           PWRs with
          of Hydrogen Burns                                large dry 
          on Safety Equipment                              containments

A-49      Pressurized Thermal        RGs 1.154, 1.99         PWR
          Shock                      SECY 82-465
                                     SECY 83-288
                                     SECY 81-687
                                     10 CFR 50.61/
                                     GL 88-11


                        GUIDE FOR UPDATING USI STATUS

(1)  Enclosure 1 lists all unresolved safety issues (USIs) for which a final 
     technical resolution has been achieved.  Please review the entire 
     listing for each licensed reactor unit.  Where an item is not 
     applicable for your facility, mark "NA" in the status column.

(2)  Where an item is applicable to your facility, but no changes were 
     necessary, mark "NC" in the status column.

(3)  Where an item is applicable to your facility and changes are complete, 
     mark "C" in the status column and indicate month and year 
     implementation was complete, including reference to any supporting 

(4)  Where an item is applicable to your facility and is not fully 
     implemented, provide your projected implementation date (month and 
     year) and a short note identifying the outstanding item (e.g., 
     hardware, procedures, training, Technical Specifications).  Mark "I" 
     for incomplete.

(5)  Where a USI resolution was only recently issued, such as A-40 and A-47, 
     and you are evaluating your response, identify expected response date 
     and indicate "E" in the status column. 

                                                            ENCLOSURE 2 


1.  USI NO.  A-1         TITLE:  Water Hammer 

This Unresolved Safety Issue (USI) was resolved in March 1984, with the 
publication of NUREG-0927, "Evaluation of Water Hammer in Nuclear Power 
Plants- Technical Findings Relevant to Unresolved Safety Issue A-1."  Also 
on March 15, 1984, the EDO sent the Commissioners SECY 84-119 titled, 
"Resolution of Unresolved Safety Issue A-1, Water Hammer."  

In SECY 84-119, the staff concluded that the frequency and severity of water 
hammer occurrences had been significantly reduced through (a) incorporation 
of design features such as keep-full systems, vacuum breakers, J-tubes, void 
detection systems, and improved venting procedures; (b) proper design of 
feedwater valves and control systems; and (c) increased operator awareness 
and training.  Therefore, the resolution of USI A-1 did not involve any 
hardware or design changes on existing plants.  It did involve Standard 
Review Plan (SRP) changes (forward fits) and a comprehensive set of 
guidelines and criteria to evaluate and upgrade utility training programs 
(per TMI Task Action Plan Item I.A.2.3).  In addition, the assumption was 
made that for BWRs with isolation condensers (ICs) a reactor-vessel high 
water-level feedwater pump trip was in place or being installed.  This was 
necessary because calculated values had postulated an IC failure by water 
hammer that opened a direct pathway to the environment.  

2.  USI NO.  A-2         TITLE:  Asymmetric Blowdown Loads in Reactor
                                 Coolant  System

This USI was resolved in January 1981 with the publication of NUREG-0609, 
"Asymmetric Blowdown Loads on PWR Primary Systems." 

In October 1975, the NRC notified each operating PWR licensee of a potential 
safety problem concerning the fact that asymmetric LOCA loads had not been 
considered in the design of any PWR piping system.  In June 1976 the NRC 
informed each PWR licensee that it was required to reassess the reactor 
vessel support design of its facility.  The staff expanded the scope of the 
problem in January 1978 with a request for additional information to all PWR 
licensees.  NUREG-0609 provided guidance for these analyses.  For operating 
PWRs, Multi-Plant Action (MPA) Item D-10 was established by NRC's Division 
of Licensing for implementation purposes.

During the course of the work on USI A-2, it was demonstrated that there 
were only a very limited number of break locations which could give rise to 
significant loads.  Subsequently, after substantial new technical work, it 
was demonstrated that pipes would leak before break and that new fracture 
mechanics techniques for the analyzing of piping failures assured adequate 
protection against failures in primary system piping in PWRs (Generic Letter 
84-04).  This was reflected in a revision of General Design Criteria (GDC)-4 
(Appendix A to 10 CFR Part 50) published in the Federal Register in final 
form on April 11, 1986, and in a subsequent revision to GDC-4 published in 
the Federal Register 

                                    - 2 -

on July 23, 1986.  In addition, it has also been satisfactorily demonstrated 
in the course of the A-2 effort that there is a very low likelihood of 
simultaneous pipe loading with both LOCA and safety shutdown earthquake 
(SSE) loads.  Therefore, the last revision of GDC-4 represented the final 
technical action of NRC regarding the issue of asymetric blowdown loads 
issue in PWRs primary coolant main loop piping.

3.  USI NO.  A-3,4,5          TITLE:  Steam Generator Tube Integrity 

USIs A-3, 4, and 5, were resolved in September 1988 with the publication of 
NUREG-0844 "NRC Integrated Program for the Resolution of Unresolved Safety 
Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity."  USIs 
A-3, A-4, and A-5 did not result in new generic requirements for industry in 
view of the small potential for reducing risk.

Steam generator tube integrity was designated an unresolved safety issue in 
1978 after it became apparent that steam generator tubes were subject to 
widespread degradation, frequent leaks, and occasional ruptures (i.e., gross 
failures).  USI Task Action Plans A-3, A-4, and A-5 were established to 
evaluate the safety significance of these problems for Westinghouse, 
Combustion Engineering, and Babcock & Wilcox steam generators, respectively.  
These studies were later combined into a single effort because PWR vendors 
were all experiencing many of the same problems. 

NUREG-0844 provides a generic risk assessment that indicates that risk from 
steam generator tube rupture (SGTR) events is not a significant contributor 
to the total risk at a given site, nor to the total risk to which the 
general public is routinely exposed.  This finding is considered indicative 
of the effectiveness of licensee programs and regulatory requirements for 
ensuring steam generator tube integrity in accordance with 10 CFR Part 50, 
Appendices A and B. 

NUREG-0844 also identifies a number of staff-recommended actions that can 
further improve the effectiveness of licensee programs in ensuring the 
integrity of steam generator tubes and in mitigating the consequences of a 
SGTR.  As part of the integrated program, the staff issued Generic Letter 
85-02 encouraging licensees of PWRs to upgrade their programs, as necessary, 
to meet the intent of the staff-recommended actions; however, such 
recommended actions do not constitute NRC requirements.  The staff's 
assessment of licensee responses to Generic Letter 85-02 was provided to the 
Commission in SECY 86-97. 

4.  USI NO.  A-6              TITLE:  Mark I Containment Short-Term Program

This USI was resolved in December 1977 with the publication of NUREG-0408, 
"Mark I Containment Short-Term Program Safety Evaluation Report." 

The objectives of the Mark I short-term program were:  (a) to examine the 
containment system of each BWR facility with a Mark I containment design to 
verify that it would maintain its integrity and functional capability when 
subjected to the most probable hydrodynamic loads induced by a postulated 

                                    - 3 -

design-basis LOCA, and (b) to verify that licensed Mark I BWR facilities 
could continue to operate safely, without undue risk to the public health 
and safety until such time as a methodical, comprehensive long-term program 
is conducted.  

The NRC staff used a safety factor of at least two to failure for the 
weakest structural or mechanical component in the Mark I containment system 
in judging that containment integrity and functions would be assured under 
most probable design-basis LOCA-induced hydrodynamic loads.  

As indicated in NUREG-0408, the staff required full implementation of the 
calculation of the hydrodynamic loads and structural analysis as an interim 
measure until complete implementation of the long-term program had been 
achieved.  In NUREG-0408 the staff concluded that the objectives of the 
Short-Term Program had been satisfied, thus documenting the basis for 
resolving this safety issue.  This issue is considered complete for all 
affected BWRs.  

5.  USI NO.  A-7         TITLE:  Mark I Long-Term Program

This USI was resolved in August 1982 with the publication of Supplement 1 to 
NUREG-0661, "Safety Evaluation Report, Mark I Containment Long-Term Program" 
and Standard Review Plan Section  For operating BWRs, MPA D-01 
was established for implementation purposes. 

The focus of this USI was the suppression pool hydrodynamic loads, 
associated with a postulated LOCA, which had not explicitly been included in 
the original Mark I containment design.  The issue was identified during 
large-scale testing of a Mark III containment design.  The staff addressed 
this issue in NUREG-0661, pub-lished in July 1980, and in Supplement 1 to 
NUREG-0661, published in August 1982.

The objective of the long-term program (LTP) was to establish the 
design-basis loads that are appropriate for the anticipated life of each 
Mark I BWR facility and to restore the originally intended design-safety 
margins for each Mark I containment system.  The principal thrust of the LTP 
was the development of generic methods for defining suppression pool 
hydrodynamic loadings and the associated structural assessment techniques 
for the Mark I configuration.  On the basis of experimental and analytical 
programs conducted by the Mark I Owners Group, it was determined that the 
hydrodynamic load definition procedures, with some modifications defined in 
NUREG-0661, provided a conservative estimate of these loading conditions.  
Thus, the requirements associated with this USI were concerned with the 
structural assessment of Mark I containments and related structures to the 
hydrodynamic loads defined by the staff in the LTP.

In January 1981, the staff issued "Orders For Modification of License and 
Grant of Extension of Exemptions" to each licensee of a Mark I plant.  The 
orders required the licensees to assess the suppression pool hydrodynamic 
loads in accordance with General Electric documents and NUREG-0661 on a 
defined schedule.  For some plants, the implementation schedule was extended 
by a subsequent order.  

                                    - 4 -

6.  USI NO.  A-8         TITLE:  Mark II Containment Pool Dynamic Loads

This USI was resolved in August 1981 with the publication of NUREG-0808, 
"Mark II Containment Program Load Evaluation and Acceptance Criteria,"  and 
Standard Review Plan (SRP) Section  The requirement is that the 11 
BWRs having the Mark II containment shall meet the requirements of GDC 16. 

As stated in NUREG-0808, the original design of the Mark II containment 
system considered only those loads normally associated with design-basis 
accidents that were known at the time.  These included pressure and 
temperature loads associated with a LOCA, seismic loads, dead loads, jet 
impingement loads, hydrostatic loads due to water in the suppression 
chamber, overload pressure test loads, and construction loads.  However, 
since the establishment of the original design criteria, additional loading 
conditions were identified that must be considered for the 
pressure-suppression containment-system design.  

In the course of performing large-scale testing of an advanced design 
pressure-suppression containment (Mark III), and during inplant testing of 
Mark I containments, new suppression-pool hydrodynamic loads were identified 
that had not been included explicitly in the original Mark II 
containment-design basis.  These additional loads result from dynamic 
effects of drywell air and steam being rapidly forced into the suppression 
pool during a postulated LOCA and from suppression-pool response to 
safety/relief valve (SRV) operation; these are generally associated with 
plant transient operating conditions.  Because these new hydrodynamic loads 
had not been considered, the NRC staff determined that a detailed 
reevaluation of the Mark II containment system was required.  

The issuance of NUREG-0808, NUREG-0802, "Safety Relief Valve Quencher Loads: 
Evaluation for BWR Mark II and III Containments," and NUREG-0487, "Mark II 
Containment Lead Plant Program Load Evaluation and Acceptance Criteria," 
documented acceptable methods for calculating the hydrodymanic loads 
associated with plant transient conditions.  Specifically, the loads 
referenced in these NRC staff reports, as modified by the acceptance 
criteria, constituted the resolution of USI A-8.  SRP Section 6.2.1 has been 
modified to reflect the applicability of these reports to Mark II 
containment evaluations.  

Implementation is believed to be complete for all Mark II BWRS.  As part of 
the licensing process, the staff required that the applicants utilize the 
new calculation methodology defined in the reference documents before a full 
power license was issued.  

7.  USI NO.  A-9         TITLE:  Anticipated Transient Without Scram (ATWS)
                                 per 10 CFR 50.62


This USI was resolved in June 1984 with the publication of a final rule 
(10 CFR 50.62) to require improvements in plants to reduce the likelihood of 
failure of the reactor protection system (RPS) to shut down the reactor 
following anticipated transients and to mitigate the consequences of an 
anticipated transient without scram (ATWS) event.  

                                    - 5 -

The rule includes the following design-related requirements:  50.62(C)(1), 
diverse and independent auxiliary feedwater initiation and turbine trip for 
all PWRs; 50.62(C)(2), diverse scram systems for CE and B&W reactors; 
50.62(C)(3) alternate rod injection (ARI) for BWRs; 50.62(C)(4); standby 
liquid control system (SLCS) for BWRs; and 50.62(C)(5), automatic trip of 
recirculation pumps under conditions indicative of an ATWS for BWRs.  
Information requirements and an implementation schedule are also specified.

8.  USI NO.  A-10        TITLE:  BWR Feedwater Nozzle Cracking

This issue was resolved in November 1980 with the publication of NUREG-0619, 
"BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking."  
MPA B-25 was established by NRC's Division of Licensing for implementation 

Inspections of operating BWRs conducted up to April 1978 revealed cracks in 
the feedwater nozzles of 20 reactor vessels.  It was determined that 
cracking was due to high-cycle fatigue caused by fluctuations in water 
temperature within the vessel in the nozzle region.  

By letter dated November 13, 1980, Darrell G. Eisenhut provided licensees 
with a copy of NUREG-0619.  The letter stated that NUREG-0619 provided the 
resolution of the staff's generic technical activity USI A-10, which 
resulted from the inservice discovery of cracking in feedwater nozzles and 
control rod drive return line nozzles.  NUREG-0619 describes the technical 
issues, General Electric and staff studies and analyses, and the staff's 
positions and requirements.  Licensees were required to respond, pursuant to 
10 CFR 50.54(f), that they would meet implementation dates indicated in 

Generic Letter 81-11 was subsequently issued to provide technical 
clarification to the November 13, 1980 letter, to clarify that it had been 
sent to PWR licensees for information only, and that no response was 
required from PWR licensees. 

9.  USI NO.  A-11        TITLE:  Reactor Vessel Materials Toughness

This USI was resolved in October 1982 with the publication of NUREG-0744, 
"Pressure Vessel Material Fracture Toughness.".  NUREG-0744 was issued by 
Generic Letter 82-26 and provided only a methodology to satisfy the require-
ments of 10 CFR Part 50, Appendix G.  No licensee response to Generic Letter 
82-26 was required.  

Because of the remote possibility that nuclear reactor pressure vessels 
designed to the ASME Boiler and Pressure Vessel Code would fail, the design 
of nuclear facilities does not provide protection against reactor vessel 
failure.  Prevention of reactor vessel failure depends primarily on 
maintaining the reactor vessel material fracture toughness at levels that 
will resist brittle fracture during plant operation.  At service times and 
operating conditions typical of current operating plants, reactor vessel 
fracture toughness properties provide adequate margins of safety against 
vessel failure; however, 

                                    - 6 -

as plants accumulate more and more service time, neutron irradiation reduces 
the material fracture toughness and initial safety margins.  

Appendix G to 10 CFR Part 50 requires that the Charpy upper shelf energy 
throughout the life of the vessel be no less than 50 ft-lb unless it is 
demonstrated that lower values will provide margins of safety against 
failure equivalent to those provided by Appendix G of the ASME code.  USI 
A-11 was initiated to address the staff's concern that some vessels were 
projected to have beltline materials with Charpy upper shelf energy less 
than 50 ft-lb.

NUREG-0744 provides a method for evaluating reactor vessel materials when 
their Charpy upper shelf energy is predicted to fall below 50 ft-lb.  Plants 
will use the prescribed method when analysis of irradiation damage predicts 
that the charpy upper shelf energy is below 50 ft-lb. 

10. USI NO.  A-12        TITLE:  Potential of Low Fracture Toughness and 
                                 Lamellar Tearing in PWR Steam Generator and
                                 Reactor Coolant Pump Supports

This USI was resolved in October 1983 with the publication of NUREG-0577, 
"Potential of Low Fracture Toughness and Lamellar Tearing in PWR Steam 
Generator and Reactor Coolant Pump Supports."  The resolution contained no 
backfit requirements; it only applied to plants with a new construction 
permit issued after October 1983.  Standard Review Plan Section 5.3.4 was 
issued at the same time this USI was resolved. 

The concern in this USI, as the title indicates, was the potential of low 
fracture toughness of some materials selected for fabrication of steam 
generator (SG) and reactor coolant pump (RCP) supports in operating PWRs.  
Lamellar tearing was also of concern.  Fracture toughness is a measure of a 
material's resistance to fracture in the presence of a previously existing 
crack.  Generally, a material is considered to have adequate fracture 
toughness if it can withstand loading to its design limit in the presence of 
detectable flaws under stated conditions of stress and temperature.

The modifications to address this USI could involve maintaining minimum 
temperature around the supports above its fracture transition temperature, 
or total replacement of existing SG and RCP supports with supports 
fabricated of material grade which has a higher Charpy upper shelf energy 
and a lower transition temperature.  Analysis performed for the resolution 
of this USI determined that, even with the failure of the SG and RCP 
supports, the amount of incremental release of radioactivity would not be 
sufficiently high enough to justify any modification in terms of increasing 
the toughness of these supports.  This conclusion is based on a value-impact 
analysis documented in Appendix C of NUREG-0577.

11. USI NO.  A-17        TITLE:  Systems Interactions in Nuclear Power

Generic Letter (GL) 89-18, dated September 6, 1989, was sent to all power 
reactor licensees and constitutes the resolution of USI A-17.  The generic 
letter did not require any licensee actions.

                                    - 7 -

GL 89-18 had two enclosures which (a) outlined the bases for the resolution 
of USI A-17, and (b) provided five general lessons learned from the review 
of the overall systems interaction issue.  The staff anticipated that 
licensees would review this information in other programs, such as the 
Individual Plant Examination (IPE) for Severe Accident Vulnerabilities.  
Specifically, the staff expected that insights concerning water intrusion 
and flooding from internal sources, as described in the appendix to 
NUREG-1174, would be considered in the IPE program.  Also considered in the 
resolution of this USI was the expectation that licensees would continue to 
review information on events at operating nuclear power plants in accordance 
with the requirements of TMI Task Action Plan Item I.C.5 (NUREG-0737).

12. USI NO.  A-24        TITLE:  Qualification of Class 1E Equipment

This USI was resolved in July 1981 with the publication of NUREG-0588, 
Revision 1, "Interim Staff Position on Environmental Qualification of 
Safety-Related Electrical Equipment."  Part I of the report is the original 
NUREG-0588 that was issued for comment; that report, in conjunction with the 
Division of Operating Reactor (DOR) Guidelines, was endorsed by a Commission 
Memorandum and Order as the interim position on this subject until "final" 
positions were established in rule making.  On January 21, 1983 the 
Commission amended 10 CFR 50.49 (the rule), effective February 22, 1983, to 
codify existing qualification methods in national standards, regulatory 
guides, and certain NRC publications, including NUREG-0588.

The rule is based on the DOR Guidelines and NUREG-0588.  These provide 
guidance on (a) how to establish environmental service conditions, (b) how 
to select methods which are considered appropriate for qualifying the 
equipment in different areas of the plant, and (c) such other areas as 
margin, aging, and documentation.  NUREG-0588 does not address all areas of 
qualification; it does supplement, in selected areas, the provisions of the 
1971 and 1974 versions of IEEE Standard 323.  The rule recognizes previous 
qualification efforts completed as a result of Commission Memorandum and 
Order CLI-80-21 and also reflects different versions IEEE 323, dependent on 
the date of the construction permit Safety Evaluation Report (SER).  
Therefore, plant-specific requirements may vary in accordance with the rule.

In summary, the resolution of A-24 is embodied in 10 CFR 50.49.  A measure 
of whether each licensee has implemented the resolution of A-24 may 
therefore be found in the determination of compliance with 10 CFR 50.49.  
This was addressed by 72 SERs for operating plants issued shortly after 
publication of the rule and subsequently in operating license reviews 
pursuant to Standard Review Plan Section 3.11.  This was further addressed 
by the first-round environmental qualification inspections conducted by the 

13. USI NO.  A-26        TITLE:  Reactor Vessel Pressure Transient

This USI was resolved in September 1978 with the publication of NUREG-0224, 
"Reactor Vessel Pressure Transient Protection for PWRs," and Standard Review 
Plan Section 5.2.  The licensees of all operating PWRs were requested to 

                                    - 8 -

provide an overpressure prevention system that could be used whenever the 
plants were in startup or shutdown conditions. The issue affected all 
operating and future plants, and the staff established MPA B-04 for 
implementing the solution at operating PWRs.

Since 1972, there have been numerous reported incidents of pressure 
transients in PWRs where technical specification pressure and temperature 
limits have been exceeded.  The majority of these events occurred while the 
reactors were in a solid-water condition during startup or shutdown and at 
relatively low reactor vessel temperatures.  Since the reactor vessels have 
less toughness at lower temperatures, they are more susceptible to brittle 
fracture under these conditions than at normal operating temperatures.  In 
light of the frequency of the reported transients and the associated 
potential for vessel damage, the NRC staff concluded that measures should be 
taken to minimize the number of future transients and reduce their severity.  

Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor 
Vessel Materials and its Impact on Plant Operations," was published July 12, 
1988.  This generic letter provides guidance regarding review of pressure-
temperature limits and indicates that licensees may have to revise low-
temperature-overpressure protection setpoints. 

14. USI NO.  A-31        TITLE:  Residual Heat Removal Shutdown Requirements

This USI was resolved in May 1978 with the publication of Standard Review 
Plan (SRP) Section 5.4.7.   Only those plants expected to receive an 
operating license after January 1, 1979 were affected by this resolution.  
The USI involved establishment of criteria for the design and operation of 
systems necessary to take a power reactor from normal operating conditions 
to cold shutdown.  

SRP Section 5.4.7 stated that, for purposes of implementation, plants would 
be divided into three classes:  Class 1 would require full compliance for 
Construction Permit (CP) or Preliminary Design Approval (PDA) applications 
which were docketed on or after January 1, 1978.  Class 2 required a partial 
implementation for all plants for which CP or PDA applications were docketed 
before January 1, 1978, and for which an Operating License (OL) issuance was 
expected on or after January 1, 1979.  Class 3 affected all operating 
reactors and all other plants for which issuance of the OL was expected 
before January 1, 1979.  The extent to which Class 3 plants would require 
implementation was based on the combined staff review of related plant 
features.  In general, the outcome of these evaluations were that only 
plants receiving an OL after January 1, 1979 were affected by this USI 
resolution, and there were no backfits to operating plants that had received 
an operating license before January 1, 1979.

15. USI NO.  A-36        TITLE:  Control of Heavy Loads, Phases I & II

This USI was resolved in July 1980 with the publication of NUREG-0612, 
"Control of Heavy Loads at Nuclear Power Plants," and Standard Review Plan 
(SRP) Section 9.1.5.  The staff established MPAs C-10 and C-15 for the 
implementation of Phases I and II, respectively, of the resolution of this 
issue at operating plants. 

                                    - 9 -

In nuclear power plants, heavy loads may be handled in several plant areas.  
If these loads were to drop in certain locations in the plant, they may 
impact spent fuel, fuel in the core, or equipment that may be required to 
achieve safe shutdown and continue decay heat removal.  USI A-36 was 
established to systematically examine staff licensing criteria and the 
adequacy of measures in effect at operating plants, and to recommend 
necessary changes to ensure the safe handling of heavy loads.  The 
guidelines proposed in NUREG-0612 include definition of safe load paths, use 
of load handling procedures, training of crane operators, guidelines on 
slings and special lifting devices, periodic inspection and maintenance for 
the crane, as well as various alternatives.  

By Generic Letters dated December 22, 1980, and February 3, 1981 (Generic 
Letter 81-07), all utilities were requested to evaluate their plants against 
the guidance of NUREG-0612 and to provide their submittals in two parts: 
Phase I (six month response) and Phase II (nine month response).  Phase I 
responses were to address Section 5.1.1 of NUREG-0612 which covered the 
following areas:

     1.   Definition of safe load paths
     2.   Development of load handling procedures
     3.   Periodic inspection and testing of cranes
     4.   Qualifications, training and specified conduct of operators
     5.   Special lifting devices should satisfy the guidelines of ANSI 
     6.   Lifting devices that are not specially designed should be 
          installed and used in accordance with the guidelines of ANSI B30.9
     7.   Design of cranes to ANSI B30.2 or CMAA-70

Phase II responses were to address Sections 5.1.2 thru 5.1.6 of NUREG-0612 
which covered the need for electrical interlocks/mechanical stops, or 
alternatively, single-failure-proof cranes or load drop analyses in the 
spent fuel pool area (PWR), containment building (PWR), reactor building 
(BWR), other areas and the specific guidelines for single-failure-proof 
handling systems.

As stated in Generic Letter 85-11, "Completion of Phase II of 'Control of 
Heavy Loads at Nuclear Power Plants' - NUREG-0612," all licensees have 
completed the requirement to perform a review and submit a Phase I and a 
Phase II report.  Based on the improvements in heavy loads handling obtained 
from implementation of NUREG-0612 (Phase I), further action was not required 
to reduce the risks associated with the handling of heavy loads. Therefore, 
a detailed Phase II review of heavy loads was not necessary and Phase II was 
considered completed.  

While not a requirement, NRC encouraged the implementation of any actions 
identified in Phase II regarding the handling of heavy loads that were 
considered appropriate. 

16. USI NO.  A-39        TITLE:  Determination of Safety Relief Valve Pool
                                 Dynamic Loads and Temperature Limits

This USI was resolved with the publication of Standard Review Plan (SRP) 
Section, in October 1982. In addition, NUREGs 0763, 0783 and 0802 
were issued for Mark I, Mark II, and Mark III containments, respectively. 

BWR plants are equipped with safety/relief valves (SRVs) to protect the 

                                   - 10 -

from overpressurization.  Plant operational transients, such as turbine 
trips, will actuate the SRV.  Once the SRV opens, the air column within the 
partially submerged discharge line is compressed by the high-pressure steam 
released from the reactor.  The compressed air discharged into the 
suppression pool produces high-pressure bubbles.  Oscillatory expansion and 
contraction of these bubbles create hydrodynamic loads on the containment 
structures, piping, and equipment inside containment.  

NUREG-0802 presents the results of the staff's evaluation of SRV loads.  The 
evaluation, however, is limited to the quencher devices used in Mark II and 
III containments.  With respect to Mark I containments, the SRV acceptance 
criteria are presented in NUREG-0661, "Safety Evaluation Report, Mark I 
Containment and Long-Term Program," and are dealt with as part of USI A-7. 

SRP Section addresses the applicable review criteria, since all 
Mark II and III containment designs are understood to have completed their 
operating license (OL) reviews subsequent to resolution of this USI and 
reflection of the resolution in the SRP.  

17. USI NO.  A-40                  TITLE:  Seismic Design Criteria

The staff has resolved USI A-40 as documented in NUREG/CR-5347, "Recommenda-
tions for Resolution of Public Comments on USI A-40," issued in June 1989, 
and NUREG-1233, "Regulatory Analysis for USI A-40," issued in September 

For plants not covered under the scope of USI A-46, "Seismic Qualification 
of Equipment in Operating Plants," the staff concluded that tanks in plants 
that were subject to licensing review by the staff after 1984 had been 
reviewed to current requirements and found acceptable.  For tanks in plants 
reviewed during 1980-1984, the staff identified four plant sites (six units) 
that were not explicitly reviewed to current requirements.  The four plants 
(Callaway 1/2, Wolf Creek, Shearon Harris 1, and Watts Bar 1/2) are being 
handled on a plant-specific basis. 

USI A-40 originated in 1977.  The basic objectives were (a) to study the 
seismic design criteria, (b) to quantify the conservatism associated with 
the criteria, and (c) to recommend modifications to the Standard Review Plan 
(SRP) if changes are justified.  Lawrence Livermore National Laboratory 
(LLNL) completed the study and published its findings in NUREG/CR-1161, 
"Recommended Revisions to USNRC - Seismic Design Criteria," dated May 1980.  
The report recommended specific changes to the Standard Review Plan (SRP).  
NRC staff reviewed the report and developed some other changes that would 
reflect the present state of seismic design practices.  The resulting SRP 
changes were issued for public comment in June 1988, and the final SRP 
changes are to be published in October 1989. 

The major SRP changes consist of (a) clarification of development of site 
specific spectra, (b) justification for use of single synthetic time-history 
by power spectral density function, (c) location and reductions of input 
ground motion for soil structure interaction, and (d) design of above-ground 
vertical tanks.  Except for item (d), these items do not constitute any 
additional requirements for current licenses and applications, and thus, no 
backfitting is being required for these items.  However, the revised 
provisions could be used for margin studies and reevaluations or individual 
plant examination for external events (IPEEE). 

                                   - 11 -

The participant utilities in the Seismic Qualification Utility Group (SQUG) 
agreed to implement the changed criteria for flexible vertical tanks for 
their plants.  For the four plants where this issue has to be resolved on an 
individual basis a 10 CFR 50.54(f) request-for-information letter has been 
sent to the affected utilities.  If the information received indicates that 
large above-ground vertical tanks do not meet the new criteria, 
plant-specific backfits will be considered. 

18. USI NO.  A-42             TITLE:  Pipe Cracks in Boiling Water Reactors

This USI was resolved in February 1981 with the publication of NUREG-0313, 
Revision 1, "Technical Report on Material Selection and Processing 
Guidelines for BWR Coolant Pressure Boundary Piping."  That NUREG document 
was issued to all holders of BWR operating licenses or construction permits 
and to all applicants for BWR operating licenses.  The staff established MPA 
B-05 for implementation of the resolution at operating plants. 

Pipes have cracked in the heat-affected zones of welds in primary system 
piping in BWRs since mid-1960.  These cracks have occurred mainly in Type 
304 stainless steel, which is the type used in most operating BWRs.  The 
major problem is recognized to be intergranular stress corrosion cracking 
(IGSCC) of austenitic stainless steel components that have been made 
susceptible to this failure by being "sensitized," either by post-weld heat 
treatment or by sensitization of a narrow heat affected zone near welds.

"Safe ends" that have been highly sensitized by furnace heat treatment while 
attached to vessels during fabrication were found to be susceptible to IGSCC 
in the late 1960s.  Most of the furnace-sensitized safe ends in older plants 
have been removed or clad with a protective material, and only a few BWRs 
still have furnace-sensitized safe ends in use.  Most of these, however, are 
in smaller diameter lines.

Cracks reported before 1975 occurred primarily in 4-inch-diameter 
recirculation loop bypass lines and in 10-inch-diameter core spray lines.  
Cracking is most often detected during inservice inspections using 
ultrasonic test techniques.  Some piping cracks have been discovered as a 
result of primary coolant leaks.

NUREG-0313, Revision 1 provided the NRC staff's revised acceptable methods 
for reducing the IGSCC susceptibility of BWR code class 1, 2, and 3 pressure 
boundary piping of sizes identified above and safe ends.  In addition, it 
provided the requirements for augmented inservice inspection of piping with 
nonconforming materials.

As a result of further IGSCC degradations in larger piping, the staff 
provided licensees with additional requirements in several NRC 
communications (i.e., Bulletins 82-03, 83-2, and 84-11).  The long-term 
resolution of IGSCC in BWR piping (including the scope of A-42) was provided 
in NUREG-0313, Revision 2 which was transmitted to all holders of BWR 
operating licenses via Generic Letter 88-01. 

                                   - 12 -

19. USI NO.  A-43        TITLE:  Containment Emergency Sump Performance

The resolution of this USI was presented to the Commission in October 1985 
in SECY-85-349.  NUREG-0897, Revision 1, "Containment Emergency Sump 
Performance," presents the results of the staff's technical findings.  These 
findings established a need to revise current licensing guidance on these 
matters.  RG 1.82 Revision 0 and Standard Review Plan Section 6.2.2, 
"Containment Heat Removal Systems" were revised to reflect this new 
guidance.  No licensee actions were required.

Initially, an issue existed concerning the availability of adequate 
recirculation cooling water following a loss-of-coolant accident (LOCA) when 
long-term recirculation of cooling water from the PWR containment sump, or 
the BWR residual heat removal system (RHR) suction intake, must be initiated 
and maintained to prevent core melt.  

The technical concerns evaluated under USI A-43 were:  (a) post-LOCA adverse 
conditions resulting from potential vortex formation and air ingestion and 
subsequent pump failure, (b) blockage of sump screens with LOCA generated 
insulation debris causing inadequate net positive suction head (NPSH) on 
pumps, and (c) RHR and containment spray pumps inoperability due to possible 
air, debris, or particulate ingestion on pump seal and bearing systems.  

This revised guidance applies only to future construction permits, 
preliminary design approvals, final design approvals, standardized designs, 
and applications for licenses to manufacture.  The staff performed a 
regulatory analysis to determine if this new guidance should be applied to 
operating plants.  The results of this analysis were reported in NUREG-0869 
Revision 1, "USI A-43 Regulatory Analysis," issued in October 1985.  The 
staff concluded that the regulatory analysis does not support any new 
generic requirements for present licensees to perform debris assessments.  

20. USI NO.  A-44        TITLE:  Station Blackout

This USI was resolved in June 1988 with the publication of a new rule (10 
CFR 50.63) and Regulatory Guide 1.155. 

Station blackout means the loss of offsite ac power to the essential and 
nonessential electrical buses concurrent with turbine trip and the 
unavailability of the redundant onsite emergency ac power systems.  
WASH-1400 showed that station blackout could be an important risk 
contributor, and operating experience has indicated that the reliability of 
ac power systems might be less than originally anticipated.  For these 
reasons station blackout was designated as a USI in 1980.  A proposed rule 
was published for comment on March 21, 1986.  A final rule, 10 CFR 50.63, 
was published on June 21, 1988 and became effective on July 21, 1988.  
Regulatory Guide 1.155 was issued at the same time as the rule and 
references an industry guidance document, NUMARC-8700.  In order to comply 
with the A-44 resolution, licensees will be required to:  

                                   - 13 -

o    maintain onsite emergency ac power supply reliability above a minimum 

o    develop procedures and training for recovery from a station blackout

o    determine the duration of a station blackout that the plant should be 
     able to withstand

o    use an alternate qualified ac power source, if available, to cope with 
     a station blackout

o    evaluate the plant's actual capability to withstand and recover from a 
     station blackout

o    backfit hardware modifications if necessary to improve coping ability

Section 50.63(c)(1) of the rule required each licensee to submit a response 
including the results of a coping analysis within 270 days from issuance of 
an operating license or the effective date of the rule, whichever is later.

21. USI NO.  A-45        TITLE:  Shutdown Decay Heat Removal Requirements

USI A-45 was resolved by SECY 88-260, "Shutdown Decay Heat Removal Require-
ments (USI-A-45)," issued September 13, 1988, without imposing any new 
licensing requirements other than the Individual Plant Examination (IPE), as 
described below.  At the same time the staff issued NUREG-1289, "Regulatory 
and Backfit Analysis:  USI A-45."  Since all of the significant USI A-45 
results have been found to be highly plant specific, the Commission decided 
it was not appropriate to propose a single generic corrective action to be 
applied uniformly to all plants.  

The Commission is currently implementing the Severe Accident Policy (50 FR 
32138) and will require all plants presently operating or under construction 
to undergo a systematic examination termed the IPE.  The reason for this 
examination is to identify any plant-specific vulnerabilities to severe 
accidents.  The IPE analysis intends to examine and understand the plant 
emergency procedures, design, operations, maintenance, and surveillance, in 
order to identify vulnerabilities.  The analysis will examine both the decay 
heat removal systems and those systems used for other related functions.  
This includes CE plants without power-operated relief valves. 

NRC has decided to subsume A-45 into the IPE program as the most effective 
way of achieving resolution of specific plant concerns associated with A-45. 

22. USI NO.  A-46        TITLE:  Seismic Qualification of Equipment in
                                 Operating Plants

USI A-46 was resolved with the issuance of GL 87-02 on February 19, 1987, 
which endorsed the approach of using the seismic and test experience data 
proposed by the Seismic Qualification Utility Group (SQUG) and Electric 
Power Research Institute (EPRI).  This approach was endorsed by the Senior 
Seismic Review and Advisory Panel (SSRAP) and approved by the NRC staff. 

                                   - 14 -

The scope of the review was narrowed to equipment required to bring each 
affected plant to hot shutdown and maintain it there for a minimum of 72 
hours.  The review includes a walkthrough of each plant which is required to 
inspect equipment.  Evaluation of equipment will include:  (a) adequacy of 
equipment anchorage; (b) functional capability of essential relays; 

(c) outliers and deficiencies (i.e., equipment with non-standard 
configurations); and (d) seismic systems interation. 

As an outgrowth of the Systematic Evaluation Program (SEP), the need was 
identified for reassessing design criteria and methods for the seismic 
qualification of mechanical equipment and electrical equipment.  Therefore, 
the seismic qualification of the equipment in operating plants must be 
reassessed to ensure the ability to bring the plant to a safe shutdown 
condition when subject to a seismic event.  The objective of this issue was 
to establish an explicit set of guidelines that could be used to judge the 
adequacy of the seismic qualification of mechanical and electrical equipment 
at operating plants in lieu of attempting to backfit current design criteria 
for new plants.  

Generic Letter 87-02 with associated guidance, required all affected 
utilities to evaluate the seismic adequacy of their plants.  The specific 
requirements and approach for implementation are being developed jointly by 
SQUG and the staff on a generic basis before individual member utilities 
proceed with plant-specific implementation.  

23. USI NO.  A-47        TITLE:  Safety Implication of Control Systems in
                                 LWR Nuclear Power Plants

USI A-47 was resolved September 20, 1989, with the publication of Generic 
Letter (GL) 88-19.

The generic letter states:  

     "The staff has concluded that all PWR plants should provide 
     automatic steam generator overfill protection, all BWR plants 
     should provide automatic reactor vessel overfill protection, 
     and that plant procedures and technical specifications for all 
     plants should include provisions to verify periodically the 
     operability of the overfill protection and to assure that 
     automatic overfill protection is available to mitigate main 
     feedwater overfeed events during reactor power operation.  
     Also, the system design and setpoints should be selected with 
     the objective of minimizing inadvertent trips of the main 
     feedwater system during plant startup, normal operation, and 
     protection system surveillance.  The Technical Specifications 
     recommendations are consistent with the criteria and the risk 
     considerations of the Commission Interim Policy Statement on 
     Technical Specification Improvement.  In addition, the staff 
     recommends that all BWR recipients reassess and modify, if 
     needed, their 

                                   - 15 -

     operating procedures and operator training to assure that the 
     operators can mitigate reactor vessel overfill events that may 
     occur via the condensate booster pumps during reduced system 
     pressure operation."

Also, page 2 of the generic letter provides for additional actions for CE and 
B&W plants.  The generic letter provides amplifying guidance for licensees.

The generic letter requires that licensees provide NRC with their schedule and

commitments within 180 days of the letter's date.  The implementation schedule

for actions on which commitments are made should be prior to startup after the

first refueling outage, but no later than the second refueling outage, 
beginning 9 months after receipt of the letter.

24. USI NO.  A-48        TITLE:  Hydrogen Control Measures and Effects of
                                 Hydrogen Burns on Safety Equipment

The NRC staff concluded April 19, 1989, that USI A-48 is resolved, as stated 
in SECY 89-122. 

USI A-48 was initiated as a result of the large amount of hydrogen generated 
and burned within containment during the Three Mile Island (TMI) accident.  
This issue covers hydrogen control measures for recoverable degraded core 
accidents for all BWRs and those PWRs with ice condenser containments.  
Extensive research in this area has led to significant revision of the Com-
mission's hydrogen control regulations, given in 10 CFR 50.44, published 
December 2, 1981.

10 CFR 50.44 requires inerting of BWR Mark I and Mark II containments as a 
method for hydrogen control.  The BWR Mark I and Mark II reactor containments 
have operated for a number of years with an inerted atmosphere (by addition of

an inert gas, such as nitrogen) which effectively precludes combustion of any 
hydrogen generated.  USI A-48 with respect to BWR Mark I and II containments 
is not only resolved but understood to be fully implemented in the affected 

The rule for BWRs with Mark III containments and PWRs with ice condenser 
containments was published on January 25, 1985.  The rule required that these 
plants be provided with a means for controlling the quantity of hydrogen 
produced, but did not specify the control method.  In addition, the task 
action plan for USI A-48 provided for plant-specific reviews of lead plants 
for reactors with Mark III and ice condenser containments.  Sequoyah was 
chosen as the lead plant for ice condenser containments and Grand Gulf for 
Mark III containments.  Both of the lead plant licensees chose to install 
igniter-type systems which would burn the hydrogen before it reached 
threatening concentrations within the containment.  Final design igniter 
systems have been installed not only in both lead plants, Sequoyah and Grand 
Gulf, but in all other ice condenser and Mark III plants as well.  The staff's

safety evaluations of the final analyses required to be submitted by these 
licensees by the rule are scheduled for completion in 1989.

                                   - 16 -

Large dry PWR containments were excluded from USI A-48 because they have a 
greater ability to accommodate the large quantities of hydrogen associated 
with a recoverable degraded core accident than the smaller Mark I, II, III and

ice condenser containments.  However, this issue has continued to be 
considered and, in 1989, hydrogen control for large dry PWR containments was 
identified as a high-priority Generic Issue (GI) 121.  The resolution of GI 
121 is being actively pursued in close coordination with more recent research 

25. USI NO.  A-49        TITLE:  Pressurized Thermal Shock

The final rule (10 CFR 50.61) on pressurized thermal shock (PTS) was approved 
by the Commission in July 1985.  Regulatory Guide 1.154, "Format and Content 
of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for PWRs,"

was later published in February 1987.  Thus, this issue was resolved and new 
requirements were established, applicable to PWRs only.  The rule required 
that each operating reactor meet the screening criteria provided in the rule 
or provide supplemental analysis to demonstrate that PTS is not a concern for 
the facility.

Neutron irradiation of reactor pressure vessel weld and plate materials 
decreases the fracture toughness of the materials.  The fracture toughness 
sensitivity to radiation-induced change is increased by the presence of 
certain materials such as copper.  Decreased fracture toughness makes it more 
likely that, if a severe overcooling event occurs followed by or concurrent 
with high vessel pressure, and if a small crack is present on the vessel's 
inner surface, that crack could grow to a size that might threaten vessel 

Severe pressurized overcooling events are improbable since they require 
multiple failures and improper operator performance.  However, certain 
precursor events have happened that could have potentially threatened vessel 
integrity if additional failures had occurred and/or if the vessel had been 
more highly irradiated.  Therefore, the possibility of vessel failure due to a

severe pressurized overcooling event cannot be ruled out.  

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