United States Nuclear Regulatory Commission - Protecting People and the Environment

Request for Information on The Status of Licensee Implementation of Generic Safety Issues Resolved with Imposition of Requirements or Corrective Actions (Generic Letter 90-04)



April 25, 1990

TO:       ALL HOLDERS OF OPERATING LICENSES AND CONSTRUCTION PERMITS FOR
          NUCLEAR POWER REACTORS

SUBJECT:  REQUEST FOR INFORMATION ON THE STATUS OF LICENSEE IMPLEMENTATION 
          OF GENERIC SAFETY ISSUES RESOLVED WITH IMPOSITION OF REQUIREMENTS 
          OR CORRECTIVE ACTIONS (GENERIC LETTER 90-04)

This letter is being issued as part of our continuing effort to establish 
and maintain an accurate and validated implementation status for all 
significant staff-imposed regulatory requirements or corrective actions.  It 
requests that you review and provide documentation of the current 
implementation status of all generic safety issues (GSIs) identified herein 
that apply to your facility.  An important objective of this effort is to 
obtain licensee and staff agreement on the GSI implementation status at each 
reactor facility.

A GSI is a safety concern, as identified and characterized in NUREG-0933, "A 
Prioritization of Generic Safety Issues," that affects the design, 
construction, or operation of all, several, or a class of nuclear power 
plants and may have the potential for safety improvements at such plants.  
This request applies to those GSIs which have been resolved by the staff and 
whose resolutions have involved the promulgation of new or revised 
requirements or guidance to the industry.  The determination of the status 
of other generic activities, such as multiplant actions (MPAs) not 
designated as GSIs, that have imposed requirements on or requested action of 
licensees is not included in this request, but is being tracked separately.

Enclosure 1 is a table of GSIs that we have included in this request.  We 
have provided the GSI number and associated MPA number, where applicable, 
the GSI title and the applicability of the issue to various classes of 
facilities.  You should complete the "Status" column in accordance with the 
guidance that accompanies Enclosure 1.  To assist you, we have provided a 
summary of each GSI and its resolution in Enclosure 2, along with applicable 
documentation references.

As in our previous requests related to the implementation status of the TMI 
Action Plan items (individual letters to licensees, April 1989) and the 
unresolved safety issues (USIs), (GL 89-21, October 1989), implementation is 
considered complete when you have performed all of the actions necessary to 
satisfy the requirements, corrective actions, or assumptions in the staff's 
technical resolution of the GSI.

We request that you respond to this letter by June 29, 1990.  In preparing 
your response we suggest that you coordinate with your NRC Project Manager 
to resolve any questions.

This request is covered by Office of Management and Budget Clearance Number 
3150-0011, which expires on January 31, 1991.  The estimated average number 
of burden hours is 80 person-hours per facility, including searching data 
sources, gathering the data, and preparing the required response.  These 
estimated average burden hours pertain only to the identified 
response-related matters and do not include implementation of the 
recommendations or requirements that resulted from resolution of the GSIs.  
Send comments regarding this burden estimate or any other aspect of this 
collection of information, including 

9004200195
.

                                    - 2 -

suggestions for reducing this burden, to the Information and Records 
Management Branch, Division of Information Support Services, Office of 
Information Resources Management (MNBB-7714), 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.

Please address your response to this generic letter to the U.S. Nuclear 
Regulatory Commission, ATTN:  Document Control Desk, Washington, DC  20555, 
pursuant to 10 CFR Section 50.4 of the NRC's regulations. 

                              Sincerely,



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

Enclosures:
1.  GSI Table
2.  GSI Summaries
3.  List of Most Recently Issued NRC Generic Letters 

.







                                 Enclosure 1




                    Status of Licensee Implementation of
                     Generic Safety Issues Resolved With
              Imposition of Requirements or Corrective Actions
.

FACILITY NAME:                          
DOCKET NO.:                          
LICENSEE:                            


         STATUS OF LICENSEE IMPLEMENTATION OF GENERIC SAFETY ISSUES

       RESOLVED WITH IMPOSITION OF REQUIREMENTS OR CORRECTIVE ACTIONS 



GSI/(MPA No.)       TITLE                                   APPLICABILITY 

40 (B065)           Safety Concerns Associated With         All BWRs 
                    Pipe Breaks In The BWR Scram 
                    System 

41 (B058)           BWR Scram Discharge Volume Systems      All BWRs 

43 (B107)           Reliability Of Air Systems              All Plants

51 (L913)           Improving the Reliability of            All Plants
                    Open-Cycle Service Water Systems 

67.3.3 (A017)       Improved Accident Monitoring            All Plants 

75** (B076)         Item 1.1 - Post-Trip Review             All Plants
                    (Program Description and 
                    Procedure) 

75 (B085)           Item 1.2 - Post-Trip Review -           All Plants
                    Data and Information Capability 


*Please follow attached guidance for completing this column.

**The 16 items listed for GSI 75 all relate to actions derived from the 
generic implications of Salem ATWS events.  Item numbers correspond to 
Generic Letter 83-28 action item numbers.
.

                                    - 2 -


GSI/(MPA No.)       TITLE                                   APPLICABILITY 

75 (B077)           Item 2.1 - Equipment Classi-            All Plants
                    fication and Vendor Interface 
                    (Reactor Trip System Components)

75 (B086)           Item 2.2.1 - Equipment Classifi-        All Plants
                    cation for Safety-Related Components

75 (L003)           Item 2.2.2 - Vendor Interface           All Plants 
                    for Safety-Related Components

75 (B078)           Items 3.1.1 & 3.1.2 - Post -            All Plants
                    Maintenance Testing (Reactor 
                    Trip System Components)

75 (B079)           Item 3.1.3 - Post-Maintenance           All Plants
                    Testing-Changes to Test Require-
                    ments (Reactor Trip System 
                    Components)

75 (B087)           Items 3.2.1 & 3.2.2 - Post-             All Plants
                    Maintenance Testing (All Other 
                    Safety-Related Components)

75 (B088)           Item 3.2.3 - Post-Maintenance           All Plants
                    Testing-Changes to Test Require-
                    ments (All Other Safety-Related 
                    Components)

75 (B080)           Item 4.1 - Reactor Trip System          All Plants
                    Reliability (Vendor-Related
                    Modifications)
.

                                    - 3 -


GSI/(MPA No.)       TITLE                                   APPLICABILITY 

75 (B081)           Items 4.2.1 & 4.2.2 - Reactor           All PWRs
                    Trip System Reliability-
                    Maintenance and Testing 
                    (Preventative Maintenance and
                    Surveillance Program for 
                    Reactor Trip Breakers)

75 (B082)           Item 4.3 - Reactor Trip System          All W and B&W
Reliability - Design Modifications      Plants
                    (Automatic Actuation of Shunt Trip 
                    Attachment for Westinghouse and B&W Plants)

75 (B090)           Item 4.3 - Reactor Trip System          All W & B&W Reliability - Tech Spec Changes         Plants
                    (Automatic Actuation of Shunt 
                    Trip Attachment For Westinghouse 
                    and B&W Plants) 

75 (B091)           Item 4.4 - Reactor Trip System          All B&W Plants
                    Reliability (Improvements in
                    Maintenance and Test Procedures
                    for B&W Plants)

.

                                    - 4 -


GSI/(MPA No.)       TITLE                                   APPLICABILITY 

75 (B092)           Item 4.5.1 - Reactor Trip System        All Plants
                    Reliability-Diverse Trip Features 
                    (System Functional Testing)

75 (B093)           Items 4.5.2 & 4.5.3 - Reactor Trip      All Plants
                    System Reliability - Test Alterna-
                    tives and Intervals (System 
                    Functional Testing)

86 (B084)           Long Range Plan for Dealing             All BWRs 
                    with Stress Corrosion 
                    Cracking in BWR Piping 

93 (B098)           Steam Binding of Auxiliary              All PWRs 
                    Feedwater Pumps

99 (L817)           RCS/RHR Suction Line Valve              All PWRs 
                    Interlock on PWRs 

124                 Auxiliary Feedwater System              AN0-1&2, Rancho 
                    Reliability                             Seco, Prairie
                                                            Island 1&2, 
                                                            Crystal River-3
                                                            Ft. Calhoun 

A-13 (B017)         Snubber Operability Assurance -         All Plants 
                    Hydraulic Snubbers 

.

                                    - 5 -


GSI/(MPA No.)       TITLE                                   APPLICABILITY 

A-13 (B022)         Snubber Operability Assurance -         All Plants 
                    Mechanical Snubbers 

A-16 (D012)         Steam Effects on BWR Core               Oyster Creek 
                    Spray Distribution                      & NMP-1 

A-35 (B023)         Adequacy of Offsite Power               All Plants 
                    Systems 

B-10                Behavior of BWR Mark III                All BWR Mark III
                    Containments                            Plants 

B-36                Develop Design, Testing and             All Plants with 
                    Maintenance Criteria for                OL Applications 
                    Atmosphere Cleanup System               After 4/1/80 
                    Air Filtration and Adsorption 
                    Units for Engineered Safety 
                    Features Systems and for 
                    Normal Ventilation Systems 

B-63 (B045)         Isolation of Low Pressure               All Plants 
                    Systems Connected to the Reactor 
                    Coolant System Pressure 
                    Boundary 

.

                                                            Attachment to
                                                            Enclosure 1



            Guidance For Completing Status Column in Enclosure 1



(1)  Provide a separate entry for each licensed reactor unit.  If the 
     information is identical for multiple units, so state.

(2)  If a GSI is not applicable to a unit(s), enter "NA".

(3)  If a GSI is applicable but no changes were necessary to implement the 
     resolution, enter "NC".  If the GSI implementation was completed prior 
     to issuance of the operating license, enter "NC", as no post-licensing 
     changes were necessary.

(4)  If a GSI is applicable, submittal of information and/or changes were 
     necessary and such submittals were made or changes are complete, enter 
     "C".  Also identify the licensee's document(s) to the NRC which 
     certified completion, and the document date(s).

(5)  If a GSI is applicable and changes are necessary but such changes are 
     not yet fully implemented, enter "I" and the projected month and year 
     of completion.  Provide a brief explanation of the outstanding work in 
     the "Comments" column.

(6)  If implementation guidance for a resolved GSI was issued recently and 
     the licensee is still evaluating the appropriate response, enter "E" 
     and the projected response date.

(7)  The "Comments" column may be used to explain any entry in the "Status" 
     column.
.







                                 Enclosure 2


                       Generic Safety Issue Summaries


     NOTE:  For further details on any of the issues, consult NUREG-0933
.

                                      1

GSI No. 40     (MPA No. B-065)     TITLE:  Safety Concerns Associated With
                                           Pipe Breaks in the BWR Scram 
                                           System

This issue arose from staff concerns related to the possibility of a break 
or leakage in scram discharge volume (SDV) piping which could 
environmentally threaten safety-related equipment, or introduce problems in 
maintaining reactor coolant system inventory.

On April 10, 1981, the NRC staff sent a generic letter to all BWR applicants 
and licensees requesting them to provide their plant-specific responses 
addressing the concerns identified in NUREG-0785.  Subsequently, Generic 
Letters 81-34 and 81-35 were sent to BWR licensees and applicants, 
respectively, wherein it was stated that plant-specific responses conforming 
to the guidance contained in NUREG-0803 would satisfy the request for 
information in the April 10, 1981 letter.

The staff's resulting generic Safety Evaluation Report for this issue was 
transmitted to all BWR applicants and licensees by Generic Letter 86-01.  
The evaluation concluded that through-wall cracks in the SDV piping need not 
be postulated.  In addition, even if a through-wall flaw is initially 
present in the SDV system, it will not propagate into a break under the 
staff-defined piping loads.  Further, leakage from such a flaw will be small 
and, therefore, a harsh environment over large areas of the reactor building 
which could affect redundant safety-related mitigating equipment will not 
result.  Thus, the potentially exposed safety-related equipment need not be 
qualified for operation in a harsh environment associated with an SDV break.

References:

1.   NUREG-0785, "Safety Concerns Associated With Pipe Breaks in the BWR 
     Scram System," U.S. Nuclear Regulatory Commission, April 1981.

2.   Letter to All BWR Licensees from D. Eisenhut, "Safety Concerns 
     Associated with Pipe Breaks in the BWR Scram System," April 10, 1981.

3.   NUREG-0803, "Generic Safety Evaluation Report Regarding Integrity of 
     BWR Scram System Piping," U.S. Nuclear Regulatory Commission, August 
     1981.

4.   Letter to All GE BWR Licensees (Except Humboldt Bay) from D. Eisenhut, 
     "Safety Concerns Associated with Pipe Breaks in the BWR Scram System 
     (Generic Letter 81-34)," August 31, 1981.

5.   Letter to All BWR Applicants for CPs, Holders of CPs, and Applicants 
     for OLs from D. Eisenhut, "Safety Concerns Associated with Pipe Breaks 
     in the BWR Scram System (Generic Letter 81-35)," August 31, 1981.

6.   Letter to All BWR Applicants and Licensees, "Safety Concerns Associated 
     with Pipe Breaks in the BWR Scram System (Generic Letter 86-01),"  
     January 3, 1984.
.

                                      2

GSI No. 41     (MPA No. B-058)     TITLE:  BWR Scram Discharge Volume 
                                           Systems

This issue arose from staff concerns related to the Browns Ferry 3 partial 
scram failure event of June 28, 1980, failures of scram level instruments, 
and subsequent staff evaluations of boiling water reactor (BWR) scram 
discharge volume (SDV) systems.

The staff's resulting generic Safety Evaluation Report of SDV systems was 
transmitted to all BWR licensees and applicants by letter dated December 9, 
1980.  This letter identified both short- and long-term corrective action 
programs.  The short-term actions were covered by Bulletins 80-14 and 80-17, 
as supplemented.  GSI No. 41 addressed the long-term program.

The resolution of this GSI affected all BWRs and addressed the following 
long-term actions:  (1) improvement of the hydraulic coupling between the 
SDV headers and the instrumented volume; (2) improvement of the reliability 
of the float switches in the instrumented volume; (3) modification of the 
instrumented volume to prevent level sensor damage from hydrodynamic forces 
and water hammer during a scram; and (4) submittal of Technical 
Specifications changes appropriate to the modified SDV systems.

A BWR Owner's Group developed criteria to implement the resolution and the 
criteria were endorsed by the staff with addition by the staff of a 
criterion for diverse level instrumentation.  Licensee commitments to 
implement the permanent corrective actions were confirmed by NRC orders 
issued in June 1983.

References:

1.   Letter to All BWR Licensees, "BWR Scram Discharge System," December 9, 
     1980.

2.   IE Bulletin No. 80-14, "Degradation of BWR Scram Discharge Volume 
     Capability," June 12, 1980.

3.   IE Bulletin No. 80-17, "Failure of 76 of 185 Control Rods to Fully 
     Insert During a Scram at a BWR," July 3, 1980.

4.   IE Bulletin No. 80-17, Supplement 1, July 18, 1980.

5.   IE Bulletin No. 80-17, Supplement 2, "Failures Revealed by Testing 
     Subsequent to Failure of Control Rods to Insert During a Scram at a 
     BWR," July 22, 1980.



GSI No. 43     (MPA No. B-107)     TITLE:  Reliability of Air Systems

This issue arose from staff concerns related to the Three Mile Island 
accident and subsequent air-operated equipment failures at other plants.  
Some of these 
.

                                      3

equipment failures are described in Information Notice (IN) 87-28 and IN 
87-28, Supplement 1.

The staff's generic Safety Evaluation Report, NUREG-1275, V.2, was provided 
to all licensees and applicants by IN 87-28, Supplement 1.  Generic Letter 
88-14 identified requested corrective actions.  These actions consisted of 
three types of verification and a discussion of a program for maintaining 
air quality.  The three types of verification included:  (1) test 
verification of air quality, (2) verification of adequate maintenance 
practices, emergency procedures, and training, and (3) verification of 
design and failure modes.  Responses concerning implementation of these 
actions were to be submitted within 180 days with allowances made for 
implementation of actions requiring outages to complete.

References:

1.   NRC Information Notice No. 87-28, "Air Systems Problems at U.S. Light 
     Water Reactors," June 22, 1987.

2.   NRC Information Notice No. 87-28, Supplement 1, December 28, 1987.

3.   NUREG-1275, "Operating Experience Feedback Report - Air Systems 
     Problems," U.S. Nuclear Regulatory Commission, Vol. 2, December 1987.

4.   NRC Letter to All Holders of Operating Licenses or Construction Permits 
     for Nuclear Power Plants, "Instrument Air Supply Systems Problems 
     Affecting Safety-Related Equipment (Generic Letter 88-14)," August 8, 
     1988.



GSI No. 51     (MPA No. L-913)     TITLE:  Improving the Reliability of
                                           Open-Cycle Service Water Systems

This issue arose from operating experience and studies related to Bulletin 
81-03 which led the NRC to question the compliance of the service water 
systems with the requirements of GDC 44, 45, 46 and Appendix B to 10 CFR 
Part 50.

The resolution of GSI No. 51, along with implementation of AEOD and Region 
II recommendations, affected all plants and addressed the following actions:  
(1) reduce flow blockage problems from biofouling, (2) conduct a heat 
transfer testing program on safety-related heat exchangers in open-cycle 
systems, (3) establish a routine inspection and maintenance program for 
open-cycle system piping and components, (4) confirm that the service water 
system will perform its intended function in accordance with the licensing 
basis for the plant; and (5) confirm the adequacy of relevant maintenance 
practices, operating and emergency procedures, and training.

.

                                      4

Generic Letter 89-13 requested licensees to advise the staff whether they 
have established programs to implement the above five actions resulting from 
the resolution of GSI No. 51, or equally effective alternative courses of 

action.  The Generic Letter also requested licensees to confirm to the staff 
that all recommended actions or equivalent alternatives have been 
implemented. 

References:

1.   NRC Bulletin No. 81-03, "Flow Blockage of Cooling Water to Safety 
     System Components by Corbicula sp. (Asiatic Clam) and Mytilus sp. 
     (Mussel), " April 10, 1981.

2.   NRC Letter to All Holders of Operating Licenses or Construction Permits 
     for Nuclear Power Plants, "Service Water System Problems Affecting 
     Safety-Related Equipment (Generic Letter 89-13)," July 18, 1989.



GSI No. 67.3.3   (MPA A-017)       TITLE:  Improved Accident Monitoring

This issue addresses compliance with Regulatory Guide 1.97.  NUREG-0737, 
"Clarification of TMI Action Plan Requirements," was issued in 1980, 
followed by Supplement 1 (issued as Generic Letter 82-33) in December 1982.  
Supplement 1 requested proposed schedules for implementing the provisions of 
Revision 2 to Regulatory Guide 1.97.  In addition, licensees and applicants 
were requested to submit details, for staff review, of how they would comply 
with the provisions of Regulatory Guide 1.97, Rev. 2, and to identify any 
exceptions to or deviations from these provisions.

Based on licensee responses to Supplement 1, confirmatory orders were issued 
to operating plants in 1985.  For license applications still under review, 
implementation would be addressed as part of the licensing process.

References:

1.   Regulatory Guide 1.97, Revision 2, "Instrumentation for 
     Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs 
     Conditions During and Following an Accident," U.S. Nuclear Regulatory 
     Commission, December 1980.

2.   NRC Letter to Licensees of Operating Reactors, Applicants for Operating 
     Licenses, and Holders of Construction Permits, "Supplement 1 to 
     NUREG-0737 - Requirements for Emergency Response Capability (Generic 
     Letter No. 82-33)," December 17, 1982.

.

                                      5

GSI No. 75     (17 Individual MPAs)     TITLE:  Generic Implications of ATWS
                                                Events at the Salem Nuclear
                                                Plant

This issue arose from staff concerns resulting from analysis of events that 
occurred at the Salem Nuclear Power Plant on February 22 and 25, 1983.  The 
analysis of the events revealed that a total loss of automatic scram 
capability (an anticipated transient without scram, or ATWS event) had 
occurred each time.  The relatively mild transients, coupled with the rapid 
manual shutdown of the reactor by the operators both times, turned these 
potentially serious events into little more than routine reactor shutdowns.  
However, the implications of these events vis a vis scram system reliability 
were considered to be extremely safety-significant.

The study of these events resulted in the issuance of NUREG-1000 and Generic 
Letter 83-28.  The Generic Letter contained a number of items and sub-items 
addressing those aspects of GSI 75 which have been resolved, each requesting
specified actions of all or identified categories of licensees and 
applicants.

It should be noted that two aspects of GSI 75 have not yet been fully 
resolved and thus are not included herein.  One of these was not addressed 
in GL 83-28 and involves possible revisions to Reg. Guide 1.33, "QA Program 
Requirements (Operations)" to contain more detailed guidance for operational 
QA programs.  The second relates to Items 4.2.3 and 4.2.4 of GL 83-28 which 
address life testing and replacement of reactor trip system components.  The 
staff is currently reassessing the methods for establishing reactor trip 
reliability and may issue a future generic communication on these items.

The 16 sub-issues of GSI 75, described below, consist of items and sub-items 
from GL 83-28 in accordance with how they were grouped into Multi-plant 
Actions (MPAs) by the staff for tracking purposes.  Each sub-issue relates 
to a single MPA and may contain more than one sub-item from GL 83-28.

References:

1.   NRC Letter to All Licensees of Operating Reactors, Applicants for 
     Operating Licenses, and Holders of Construction Permits, "Required 
     Actions Based on Generic Implications of Salem ATWS Events (Generic 
     Letter 83-28)," July 8, 1983.

2.   NUREG-1000, Volume 1, "Generic Implications of ATWS Events at the Salem 
     Nuclear Power Plant," U.S. Nuclear Regulatory Commission, April 1983.

3.   NUREG-1000, Volume 2, August 1983.



(MPA No. B-076)          TITLE:  Item 1.1- Post-Trip Review (Program 
                                 Description and Procedure)

The resolution of this item, applicable to all plants, requests that 
licensees and applicants describe their programs for ensuring that 
unscheduled reactor shutdowns are analyzed and a determination made that the 
plant can be restarted safely.

.

                                      6

As a minimum, each licensee is requested to describe:  (1) the criteria for 
determining the acceptability of restart, (2) the responsibilities and 
authorities of personnel who perform the review and analysis, (3) the 
necessary qualifications and training for the responsible personnel, (4) the 
sources of plant information necessary to conduct the review and analysis, 
(5) the methods and criteria for comparing the event information with known 
or expected plant behavior, (6) the criteria for determining the need for an 
independent assessment of an event and guidelines on the preservation of 
physical evidence to support independent analysis of the event, and (7) the 
systematic safety assessment procedures compiled from (1) to (6) which are 
used in conducting the evaluation by the staff.



(MPA No. B-085)  TITLE: Item 1.2 - Post-Trip Review - Data and Information
                        Capability

Item 1.2 requests that licensees and applicants have the capability to 
record, recall, and display data and information to permit diagnosing the 
causes of unscheduled reactor shutdowns and the proper functioning of 
safety-related equipment during these events using systematic safety 
assessment procedures.  The data and information are to be displayed in a 
form that is user-friendly and reflects human factors considerations.  It 
further requests licensees and applicants to prepare and submit a report 
which describes and justifies the adequacy of their equipment for diagnosing 
an unscheduled reactor shutdown.  Submittals are to be reviewed by the staff 
to determine whether adequate data and information will be available to 
support the systematic assessment of unscheduled reactor shutdowns.



(MPA No. B-077)  TITLE:  Item 2.1 - Equipment Classification and Vendor
                         Interface (Reactor Trip System Components)

Item 2.1 addresses components whose functioning is required to trip the 
reactor and requests all licensees and applicants to describe their program 
to assure that all such components are identified as "safety-related" in 
documents, procedures and information handling systems used to control 
safety-related activities in the plant.  In addition, the item requests that 
a vendor interface program be established, implemented and maintained for 
such components to ensure that relevant vendor information is complete, 
current and controlled throughout the plant lifetime, that it is 
appropriately referenced or incorporated in plant instructions and 
procedures, and that it include periodic communication with the vendor.  The 
licensees' submittals are to be reviewed by the staff to determine their 
adequacy.



(MPA No. B-086)  TITLE:  Item 2.2.1 - Equipment Classification for 
                         Safety-Related Components)

Item 2.2.1 addresses all other safety-related components and requests all 
licensees and applicants to describe their program used to classify such 
components.  The classification program is necessary to ensure that all such 
components are identified as "safety-related" in documents, procedures and 

.

                                      7

information handling systems used to control safety-related activities in 
the plant, and must include periodic communication with the vendor.  The 
staff is to review the licensees' submittals to determine their adequacy.

This MPA originally addressed vendor interface programs for safety-related 
components in addition to component classification, as identified in GL 
83-28.  The original vendor interface program guidelines were modified and 
superseded by way of GL 90-03 on March 20, 1990.  A new MPA was established 
to track implementation of the revised guidelines.  They are discussed 
separately below.

Additional Reference:

1.   NRC Letter to All Power Reactor Licensees and Applicants, "Relaxation 
     of Staff Position in Generic Letter 83-28, Item 2.2 Part 2 'Vendor 
     Interface for Safety-Related Components' (Generic Letter No. 90-03)," 
     March 20, 1990.



(MPA No. L-003)     TITLE:  Item 2.2.2 - Vendor Interface for Safety-Related
                            Components

The original needs for vendor interface programs for safety-related 
components were specified in GL 83-23 and licensee implementation was being 
tracked via MPA No. B-086, together with equipment classification.  GL 90-03 
was issued on March 20, 1990 which relaxes and supersedes the original 
vendor interface program guidance based upon industry initiatives and 
experience.  The revised interface program with the NSSS vendor covers all 
safety-related components within the NSSS scope of supply and is to conform 
with the Vendor Equipment Technical Information Program (VETIP) as described 
in the Nuclear Utility Task Action Committee Report, INPO 84-010 issued in 
March 1984.  A program of periodic contact with non-NSSS vendors of other 
key safety-related components is also specified.

Additional References:

1.   NRC Letter to All Power Reactor Licensees and Applicants, "Relaxation 
     of Staff Position in Generic Letter 83-28, Item 2.2 Part 2 'Vendor 
     Interface for Safety-Related Components' (Generic Letter No. 90-03)," 
     March 20, 1990.

2.   INPO 84-010, "Vendor Equipment Technical Information Program," Nuclear 
     Utility Task Action Committee, March 1984.



(MPA No. B-078)  TITLE:  Items 3.1.1 and 3.1.2 - Post-Maintenance Testing
                         (Reactor Trip System Components)

Items 3.1.1 and 3.1.2 concern post-maintenance testing procedures and vendor 
recommendations for reactor trip system components.  Licensees and 
applicants 
.

                                      8

are to review their test and maintenance procedures and Technical 
Specifications to assure that they require post-maintenance operability 
testing of safety-related components in the reactor trip system and that 
such testing demonstrates that the equipment is capable of performing its 
safety functions prior to returning it to service.  Licensees and applicants 
are also to review applicable vendor and engineering recommendations to 
ensure that any appropriate test guidance is included in the test and 
maintenance procedures or in the Technical Specifications, where required.  
The results of these reviews are to be submitted for staff evaluation.



(MPA No. B-079)     TITLE:  Item 3.1.3 - Post-Maintenance Testing - Changes
                            to Test Requirements (Reactor Trip System 
                            Components)

Item 3.1.3 requests identification of any applicable post-maintenance test 
requirements in existing Technical Specifications for reactor trip system 
components which can be demonstrated to degrade rather than enhance safety.  
Licensees and applicants are to perform the required reviews and notify the 
staff of their findings.  Appropriate changes to these test requirements, 
with supporting justification, are to be submitted for staff approval.  



(MPA No. B-087)     TITLE:  Items 3.2.1 and 3.2.2 - Post-Maintenance Testing
                            (All Other Safety-Related Components)

Items 3.2.1 and 3.2.2 concern post-maintenance testing procedures and vendor 
recommendations for all safety-related components other than the reactor 
trip system components.  Licensees and applicants are to review their test 
and maintenance procedures and Technical Specifications to assure that they 
require post-maintenance operability testing of all safety-related 
components (non-reactor trip system components) and that such testing 
demonstrates that the equipment is capable of performing its safety 
functions prior to returning it to service.  Licensees and applicants are 
also to review applicable vendor and engineering recommendations to ensure 
that any appropriate guidance is included in the test and maintenance 
procedures or in the Technical Specifications, where required.  The results 
of these reviews are to be submitted for staff evaluation.



(MPA No. B-088)     TITLE:  Item 3.2.3 - Post-Maintenance Testing - Changes
                            to Test Requirements (All Other Safety-Related 
                            Components)

Item 3.2.3 requests identification of any applicable post-maintenance test 
requirements in existing Technical Specifications for safety-related 
components which can be demonstrated to degrade rather than enhance safety.  
Licensees and 
.

                                      9

applicants are to perform the required reviews and notify the staff of their 
findings.  Appropriate changes to these test requirements, with supporting 
justification, are to be submitted for staff approval.  



(MPA No. B-080)     TITLE:  Item 4.1 - Reactor Trip System Reliability
                            (Vendor-Related Modifications)

The resolution of this item, applicable to all plants, requests that each 
licensee review all vendor-recommended reactor trip breaker modifications to 
verify that either:  (1) each modification has been implemented, or (2) a 
written evaluation of the technical reasons for not implementing a 
modification exists.  Submittals were to be made by all 
licensees/applicants.  For those plants that were licensed at the time, the 
submittals were to be reviewed by the cognizant regions and Safety 
Evaluations were issued by NRR.  For plants licensed since 1983, Item 4.1 
was to be included as part of the licensing review and the results reported 
in the licensing SER or in one of the supplements.



(MPA No. B-081)     TITLE:  Items 4.2.1 and 4.2.2 - Reactor Trip System
                            Reliability - Maintenance and Testing 
                            (Preventative Maintenance and Surveillance 
                            Program for Reactor Trip Breakers)

Item 4.2.1 addresses development of a planned program of periodic 
maintenance, including lubrication, housekeeping and other items recommended 
by the equipment suppliers.  Item 4.2.2 addresses development and 
implementation of a program for trending of parameters which affect breaker 
operation and are measured during testing in order to predict performance 
degradation.  All PWR licensees and applicants were to provide descriptions 
of their programs for staff review.



(MPA No. B-082)     TITLE:  Item 4.3 - Reactor Trip System Reliability-
                            Design Modifications (Automatic Actuation of
                            Shunt Trip Attachment for Westinghouse and
                            B&W Plants)

This portion of Item 4.3 requests all licensees and applicants with 
Westinghouse and B&W plants to modify their reactor trip systems to provide 
automatic actuation of the breaker shunt trip attachments.  The staff was to
review the submittals and issue SERs for all affected plants.  



(MPA No. B-090)     TITLE:  Item 4.3 - Reactor Trip System Reliability - 
                            Technical Specification Changes (Automatic 
                            Actuation of Shunt Trip Attachment for 
                            Westinghouse and B&W Plants)

This portion of Item 4.3 requests submittal of Technical Specifications 
changes addressing the implementation of automatic actuation of the breaker 
.

                                     10

shunt trip attachments on all Westinghouse and B&W plants (see previous 
discussion on MPA No. B-082).  The staff developed model Technical 
Specifications and transmitted them to affected licensees and applicants in 
Generic Letter 85-09.  The staff was to review the submittals and issue 
license amendments and/or SERs for all affected plants.

Additional Reference:

1.   NRC Letter to All Westinghouse Pressurized Water Reactor Licensees and 
     Applicants, "Technical Specifications for Generic Letter 83-28, Item 
     4.3 (Generic Letter 85-09)," dated May 23, 1985.



(MPA No. B-091)     TITLE:  Item 4.4 - Reactor Trip System Reliability
                            (Improvements in Maintenance and Test Procedures
                            for B&W Plants)

Item 4.4 requests B&W reactor licensees and applicants to apply 
safety-related maintenance and test procedures to the diverse reactor trip 
feature provided by interrupting power to control rods through the silicon 
controlled rectifiers (SCRs).  Specifically, licensees and applicants are 
requested to submit for staff review:  (1) confirmation that procedures 
which comply with all requirements of safety-related procedures are being 
used to maintain and test the SCRs, (2) a brief description of the 
procedures used to conduct periodic surveillance, testing and maintenance of 
the SCR diverse reactor trip feature; such tests should verify that the SCRs 
under test have degated and opened the power supply circuit to the control 
rod holding coils, and (3) Technical Specifications changes which include 
requirements for safety-related surveillance and tests of the SCRs to be 
performed at intervals commensurate with existing test intervals for other 
safety-related portions of the reactor trip system or alternatively, show 
that these requirements are in the existing Technical Specifications.



(MPA No. B-092)     TITLE:  Item 4.5.1 - Reactor Trip System Reliability-
                            Diverse Trip Features (System Functional 
                            Testing)

Item 4.5.1 requests that licensees perform on-line functional testing of the 
reactor trip system, including independent testing of the diverse trip 
features.  The diverse trip features to be tested include the breaker 
undervoltage and shunt trip features on Westinghouse, B&W and CE plants; the 
circuitry used for power interruption with the silicon controlled rectifiers 
on B&W plants; and the scram pilot valves and backup scram valves (including 
all initiating circuitry) on GE plants.

Licensees were requested to confirm that the required on-line functional 
surveillance testing is being performed for the diverse trip features of the 
plant.

Some licensees do not test backup scram valves on-line, because such testing 
would result in a reactor scram.  In such cases the NRC allows scram valves 
to be tested during each refueling outage to avoid unnecessary reactor 
scrams and challenges to the reactor protection system.  Conformance with 
this item is verified by follow-up inspections.
.

                                     11

(MPA No. B-093)     TITLE:  Items 4.5.2 & 4.5.3 - Reactor Trip System
                            Reliability - Test Alternatives and Intervals
                            (System Functional Testing)

Item 4.5.2 requests licensees and applicants to certify whether their plants 
are designed to permit on-line functional testing of the reactor trip system 
(RTS).  For plants not designed to permit such testing, licensees are 
requested to commit to design modifications which would permit such testing 
and provide an implementation schedule, or to provide justification for not 
implementing on-line testing capability.  The staff will consider 
alternatives to on-line testing where special circumstances exist and where 
the objective of high reliability can be met by other means.

Item 4.5.3 requests licensees and applicants to confirm that on-line 
functional testing of the RTS is being performed and that existing test 
intervals required by their Technical Specifications are adequate for 
achieving high RTS reliability.  All four vendors submitted topical reports 
which presented analyses demonstrating that current test intervals provide 
high reliability.  Based on staff review of the Owner's Group topical 
reports, the contractors' independent analyses, and the generic safety 
evaluation findings in NUREG-0460, the staff concluded that the existing 
intervals, as recommended in the topical reports, for on-line functional 
testing are consistent with achieving high RTS availability at all operating 
reactors.  Licensees and applicants are to submit a description of how they 
are implementing the provisions of their Owner's Group topical report.

Additional References:

1.   Topical Report WCAP-10271, "Evaluation of the Surveillance Frequencies 
     and Out of Service Times for the Reactor Protection Systems," 1985. 
     
2.   Topical Report WCAP-10271, Supplement 1. 

3.   NECD-30844, "BWR Owner's Group Response to NRC Generic Letter 83-28, 
     Item 4.5.3," January 1985.

4.   NECD-30851P, "Technical Specification Improvement Analyses for BWR 
     Reactor Protection System," May 1985.

5.   CE NPSD-277, "Reactor Protection System Test Interval Evaluation, Task 
     486," December 1984.

6.   BAW-10167, "Justification for Increasing the Reactor Trip System 
     On-Line Test Interval," May 1986.

7.   BAW-10167, Supplement 1, February 1988.

.

                                     12

GSI No. 86     (MPA No. B-084)     TITLE:  Long Range Plan for Dealing With
                                           Stress Corrosion Cracking in BWR
                                           Piping

This issue arose from inspections conducted at several boiling water 
reactors (BWRs) which revealed intergranular stress corrosion cracking 
(IGSCC) in large-diameter recirculation and residual heat removal piping.  
These inspections were conducted pursuant to IE Bulletins 82-03, 82-03 
Revision 1, and 83-02 and the NRC August 26, 1983 Orders.  The Commission 
concluded that the results of these inspections mandated an ongoing program 
for similar reinspections at all operating BWRs.

Generic Letter 84-11 requested all BWR licensees and applicants to submit, 
for staff review, their plans and surveillance measures relative to the 
staff positions set forth in the Generic Letter and to commit to develop and 
implement an acceptable program to detect potential IGSCC.

Inspections conducted pursuant to GL 84-11 disclosed a significant number of 
cracks in BWR piping.  The staff concluded that augmented inspections and 
licensee actions beyond those in GL 84-11 were warranted.  Generic Letter 
88-01 was subsequently issued describing the staff's revised positions on 
what were acceptable actions that licensees/applicants should take to 
minimize the potential for IGSCC.  The staff positions in GL 88-01 
superseded those in GL 84-11 and are beyond the scope of this GSI.

References:

1.   IE Bulletin 82-03 "Stress Corrosion Cracking in Thick-Wall, Large 
     Diameter, Stainless Steel, Recirculation System Piping at BWR Plants," 
     U. S. Nuclear Regulatory Commission, October 14, 1982. 
     
2.   IE Bulletin 82-03, Revision 1, October 28, 1982.

3.   IE Bulletin 83-02, "Stress Corrosion Cracking in Large Diameter 
     Stainless Steel Recirculation Systems Piping at BWR Plants," U.S. 
     Nuclear Regulatory Commission, March 4, 1983.

4.   NRC Letter to All Licensees of Operating Reactors, Applicants for 
     Operating License, and Holders of Construction Permits for Boiling 
     Water Reactors, "Inspections of BWR Stainless Steel Piping," (Generic 
     Letter 84-11), April 19, 1984.

5.   NUREG-0313, "Technical Report on Material Selection and Processing 
     Guidelines for BWR Coolant Pressure Boundary Piping," U.S. Nuclear 
     Regulatory Commission, July 1977; Rev. 1, July 1980; Rev. 2, January 
     1988.

6.   NRC Letter to All Licensees of Operating Boiling Water Reactors (BWRs), 
     and Holders of Construction Permits, "NRC Position on IGSCC in BWR 
     Austenitic Stainless Steel Piping (Generic Letter 88-01)," January 25, 
     1988.
.

                                     13

GSI No. 93     (MPA No. B-098)          TITLE:  Steam Binding of Auxiliary
                                                Feedwater Pumps

The issue concerns the potential disabling of auxiliary feedwater pumps by 
steam binding caused by back-leakage of main feedwater past the isolation 
check valves.  IE Bulletin 85-01, issued October 29, 1985, requested that 
certain licensees implement procedures for monitoring the auxiliary 
feedwater piping temperatures for indications of possible back-leakage and 
for restoring the pumps to operable status if steam binding were to occur.

Generic Letter 88-03, issued February 17, 1988, stated that the plants that 
received Bulletin 85-01 should continue following the Bulletin's 
recommendations, and requested that these recommendations be followed on all 
PWR's.

References:

1.   IE Bulletin No. 85-01, "Steam Binding of Auxiliary Feedwater Pumps," 
     U.S. Nuclear Regulatory Commission, October 29, 1985.

2.   NRC Letter to All Licensees, Applicants for Operating Licenses, and 
     Holders of Constructions Permits for Pressurized Water Reactors, 
     "Resolution of Generic Safety Issue 93, 'Steam Binding of Auxiliary 
     Feedwater Pumps' (Generic Letter 88-03)," February 17, 1988.



GSI No. 99     (MPA No. L-817)     TITLE:  RCS/RHR Suction Line Valve
                                           Interlock on PWRs

This issue concerns the inadvertent closing of RHR suction valves when the 
RHR system is in use.

Interlocks are provided on these valves to ensure that a double barrier 
(i.e., two closed valves) is maintained between the RCS and RHR systems when 
the plant is at normal operating conditions.  However, the loss of one 
instrument bus or disturbance of one logic channel will result in the 
automatic closure of one of the RHR suction line isolation valves.  Such 
closure gives rise to the potential for RHR pump damage and loss of decay 
heat removal capability if the RHR pump is not interlocked with the RHR 
suction valves.

The scope of this issue was broadened in June 1986 to include the less 
frequent but higher risk mode of failure associated with mid-loop operation.  
Generic Letter 87-12 addressed this concern.

Generic Letter 88-17 superseded GL 87-12 and requested responses regarding 
licensee plans with respect to operation on shutdown cooling.  This letter 
requested expeditious licensee actions in the areas of:  (1) training of 
operators before entering a reduced inventory condition, (2) implementation 
of procedures and administrative controls related to decay heat removal, (3) 
temperature and level indications, and (4) alternate means of adding water 
to the RCS.  Further, 
.

                                     14

GL 88-17 identified a number of programmed enhancements to be developed in 
the following six areas:  (1) instrumentation, (2) procedures, (3) 
equipment, (4) analyses, (5) Technical Specifications, and (6) RCS 
perturbations.

References:

1.   NRC Letter to All Licensees of Operating PWRs and Holders of 
     Construction Permits for PWRs, "Loss of Residual Heat Removal (RHR) 
     While the Reactor Coolant System (RCS) is Partially Filled (Generic 
     Letter 87-12)," July 9, 1987.

2.   NRC Letter to All Holders of Operating Licenses or Construction Permits 
     for Pressurized Water Reactors (PWRs), "Loss of Decay Heat Removal 
     (Generic Letter No. 88-17), 10 CFR 50.54(f)," October 17, 1988.

3.   NUREG/CR-5015, "Improved Reliability of Residual Heat Removal 
     Capability in PWRs as Related to Resolution of Generic Issue 99," U.S. 
     Nuclear Regulatory Commission, May 1988.



GSI No. 124      TITLE:  Auxiliary Feedwater System Reliability

This issue was initially established after implementation of upgrades to the 
auxiliary feedwater (AFW) systems in all PWR plants, under TMI Action Plan 
Clarification, NUREG-0737, Items II.E.1.1 and I.E.1.2, in order to determine 
if further improvements in AFW system reliability were necessary.  
NUREG-0737, Items II.E.1.1 and II.E.1.2 addressed implementation of 
recommendations for improving AFW system reliability identified in 
NUREG-0611 and -0635.

Based on evaluation of AFW system reliability studies for various plants, 
the staff subsequently determined that three-pump AFW systems demonstrated 
significantly greater reliability than did two-pump systems and, therefore, 
limited this issue to those two-pump plants for which the licensee had not 
committed to add a third means of delivering water to the steam generators 
for post-transient/accident decay heat removal.  The affected plants are 
ANO-1 & 2, Rancho Seco, Prairie Island 1 & 2, Crystal River-3 and Ft. 
Calhoun.

The staff performed plant-specific reviews of the reliability of the AFW 
systems in the above plants, including assessments of the system design, 
operating experience, and emergency procedures.  From these reviews, the 
staff determined whether additional means of secondary decay heat removal 
capability was necessary.  No further hardware modification was determined 
to be required for ANO-1 and Prairie Island 1 & 2 on the basis of the 
startup feedwater pump and AFW system sharing capability, respectively.  The 
licensees for Rancho Seco, Crystal River-3 and Ft. Calhoun committed to 
install additional means of secondary decay heat removal, thereby resolving 
the issue.  The staff issued a plant-specific backfit analysis for ANO-2 
requiring the addition of a third train of secondary decay heat removal.  
Implementation of the modifications to these plants is proceeding in 
accordance with schedules agreed to by the staff.
.

                                     15

References:

1.   NUREG-0611, "Generic Evaluation of Feedwater Transients and Small-Break 
     Loss-of-Coolant Accidents in Westinghouse Designed Operating Plants," 
     U.S. Nuclear Regulatory Commission, January 1980.

2.   NUREG-0635, "Generic Evaluation of Feedwater Transients and Small-Break 
     Loss-of-Coolant Accidents in Combustion Engineering Designed Operating 
     Plants," U.S. Nuclear Regulatory Commission, January 1980.

3.   NUREG-0737, "Clarification of TMI Action Plan Requirements," U.S. 
     Nuclear Regulatory Commission, November 1980.



GSI No. A-13   (MPA No. B-017)     TITLE:  Snubber Operability Assurance-
                                           Hydraulic Snubbers

This issue concerns operability of hydraulic snubbers which is required to 
assure that the structural integrity of the reactor coolant system is 
maintained during and following a seismic or other event initiating dynamic 
loads.  Operating experience in the 1970's indicated the need for changes, 
clarifications and improvements in snubber Technical Specifications.  These 
changes provided for:  (1) precluding use of an arbitrary snubber capacity 
as a limit for inservice test requirements, (2) elimination of the 
requirement that seal material be approved by NRC, (3) implementation of a 
monitoring program to assure snubber reliability, (4) development and 
implementation of clearly defined inservice test requirements, and (5) 
permissible in-place inservice testing.

By letter dated November 20, 1980, the NRC requested that all power reactor 
licensees (except Systematic Evaluation Program (SEP) licensees) incorporate 
the above changes in plant-specific Technical Specifications.  A similar 
request was sent to SEP licensees on March 23, 1981.  Also, revisions to the 
Standard Technical Specifications (W, GE, CE and BW) incorporated the 
appropriate Technical Specifications to address these changes for NTOLs.

References

1.   NRC Letter to All Power Reactor Licensees (Except SEP Licensees), 
     "Technical Specification Revisions for Snubber Surveillance," November 
     20, 1980.

2.   NRC Letter to all SEP Power Reactor Licensees, (Except SEP Licensees), 
     "Technical Specification Revisions for Snubber Surveillance," March 23, 
     1981.
.

                                     16

GSI No. A-13   (MPA No. B-022)     TITLE:  Snubber Operability Assurance-
                                           Mechanical Snubbers

This aspect of the issue addresses mechanical snubbers.  In the mid 1970's, 
several deficiencies were noted in the Technical Specifications for assuring 
snubber reliability.  Also, mechanical snubbers were not included in the 
Technical Specifications surveillance requirements.  Many licensees used 
mechanical snubbers as original equipment and others requested to replace 
their hydraulic snubbers with mechanical ones to simplify or avoid inservice 
surveillance.  The most likely failure for an unsurveilled mechanical 
snubber is permanent lock-up which can be harmful to plant systems during 
normal operations and during seismic events initiating dynamic loads.  
Therefore, changes were needed which would:  (1) include mechanical snubbers 
in the surveillance program, (2) preclude use of an arbitrary snubber 
capacity as a limit for inservice test requirements, (3) implement a 
monitoring program to assure snubber reliability, (4) develop and implement 
clearly defined test requirements, and (5) permit in-place inservice 
testing.

By letter dated November 20, 1980, the NRC requested that all power reactor 
licensees (except Systematic Evaluation Program (SEP) licensees) incorporate 
the above changes in plant-specific Technical Specifications.  A similar 
request was sent to SEP licensees on March 23, 1981.  Also, revisions to the 
Standard Technical Specifications (W, GE, CE and BW) incorporated the 
appropriate Technical Specifications to address these changes for NTOLs.

References:

1.   NRC Letter to All Power Reactor Licensees (Except SEP Licensees), 
     "Technical Specification Revisions for Snubber Surveillance," November 
     20, 1980.

2.   NRC Letter to All SEP Power Reactor Licensees, "Technical Specification 
     Revisions for Snubber Surveillance," March 23, 1981.



GSI No. A-16   (MPA No. D-012)     TITLE:  Steam Effects on BWR Core
                                           Spray Distribution

This issue arose from tests which showed that the presence of steam and/or 
increased pressure in and above the upper core region of BWRs could 
adversely affect the distribution of flow from certain types of core spray 
nozzles following a LOCA.  The distribution that had been assumed in BWR 
LOCA analyses was based on tests of core spray nozzles conducted by GE in an 
air (non-steam) environment.

In response to staff concerns regarding the core spray performance, GE took 
the lead for resolving the issue generically.  This GSI was established for 
staff review of generic actions.  However, because of design differences 
among the various BWR product lines, resolution of this issue has taken 
different forms for different classes of BWR plants.  Each of the different 
resolution paths is summarized below.
.

                                     17

1.   BWR/1 (Big Rock Point) - The licensee performed a test with the 
     installed core spray system which demonstrated the adequacy of the 
     spray flow distribution.  The staff found the test results acceptable 
     and concurred in the licensee's resolution of this issue in 1979.

2.   BWR/2 (Oyster Creek and Nine Mile Point 1) - Core spray is less 
     important in plants with jet pumps because these plants are designed to 
     reflood to 2/3 core height during a LOCA and cooling over that 2/3 
     height is effective.  Non-jet pump plants of the BWR/2 design do not 
     reflood for large breaks below the core, and must rely on adequate 
     spray flow to each assembly and steam cooling to avoid fuel melt.  
     Because of this concern, letters were issued to the two BWR/2 licensees 
     requesting that they justify the spray cooling (heat transfer) 
     coefficients assumed in their ECCS analyses.  Multi-plant Action (MPA) 
     D-012 was established for the review of the core spray issue on these 
     two BWR/2 plants, the only plants which were affected by GSI A-16.  In 
     both cases, the licensees, with assistance from GE, were able to show 
     to the staff's satisfaction that even when including the effects of a 
     steam environment on core spray distribution, the degraded distribution 
     of core spray along with steam cooling was adequate to ensure that clad 
     temperature limits specified in 10 CFR 50.46 would not be exceeded.

3.   BWR/3/4/5 - In reviewing the core spray distribution issue for these 
     jet pump plant class designs, the staff found that as long as the 
     reflood water level could be maintained with injection from the core 
     spray system, the distribution of core spray over the top of the core 
     was not a significant factor in achieving adequate core cooling.  For 
     this reason the core spray distribution issue was resolved generically 
     for these plants and no actions were requested of these licensees.

4.   BWR/6 - To resolve the core spray distribution issue for the BWR/6 
     design, GE performed full-scale tests of the BWR/6 core spray sparger 
     in a steam environment.  The staff inspected the GE test facility, 
     reviewed the test results and concluded that the BWR/6 core spray 
     design was adequate.  This resolved the core spray distribution issue 
     for the BWR/6 design.

References: 

1.   Letter to I.R. Finfrock, Jr., JCP&L Co. from George Lear, NRC; 
     December 10, 1976.

2.   Letter to NMPC from NRC dated December 10, 1976.



GSI No. A-35   (MPA No. B-023)     TITLE:  Adequacy of Offsite Power Systems

This issue arose from a July 1976 degraded grid voltage condition which 
occurred at Millstone 2 and which resulted in blown fuses in certain 
engineered safety feature equipment.  As a result, the staff determined that 
a potential 
.

                                     18

existed for supplying both safety and non-safety equipment with voltages 
outside the design range, which could render the equipment inoperable.

Letters were sent to licensees in June 1977 which requested installation of 
degraded voltage relays designed to separate the safety buses from offsite 
power whenever the degraded voltage condition existed for more than about 10 
seconds.  Licensees were also requested to propose Technical Specifications 
with LCOs and surveillance requirements for these relays and associated 
instrumentation.  Some licensees chose instead to institute procedures for 
manual actions in the event of these degraded voltage conditions.  The 
Regions reviewed these procedures and eventually found them to be 
acceptable.

Some licensees resolved this issue in conjunction with MPA B-048, "Adequacy 
of Station Distribution Voltage," which was initiated by the letter to all 
power reactor licensees (except Humboldt Bay) on August 8, 1979.  MPA B-048 
requested licensees to reanalyze their plants to ensure that safety-related 
equipment was not subjected to voltages outside design limitations when the 
grid voltage was at its maximum and minimum levels.  After performing these 
analyses, licensees were then to perform a test to measure station voltages 
at various places in the plant to verify the accuracy of the calculations.  
As a result of this review, many licensees made tap changes to transformers 
to optimize station distribution voltages.  These tap changes often affected 
MPA-B023 calculations and caused changes to the undervoltage relay 
setpoints.

The changes imposed by resolution of this issue were incorporated into 
licensing reviews after 1977 through Branch Technical Position PSB-1 and, 
subsequently, a 1981 revision to SRP 8.3.1, Appendix A.

References:

1.   NRC Letter to Northeast Nuclear Energy Company, "Millstone Nuclear 
     Power Station Unit Nos. 1 and 2," June 2, 1977.  
     
2.   NRC Letter to All Power Reactor Licensees (Except Humboldt Bay), 
     "Adequacy of Station Electric Distribution Systems Voltages," August 8, 
     1979.

3.   Branch Technical Position PSB-1, "Adequacy of Station Electric 
     Distribution Voltages," July 1981.



GSI No. B-10   TITLE:  Behavior of BWR Mark III Containments

This GSI involved completion of the staff evaluation of the Mark III 
containment loads and documentation of the method used to validate the 
analytical models and assumptions needed to predict the containment pressure 
loads in the event of a LOCA.  The BWR Mark III containment design differed 
from the previously-reviewed Mark I and Mark II designs.  As a result, staff 
.

                                     19

acceptance criteria were required for the various pool dynamic loads 
associated with this new design.

The Mark III suppression pool dynamic loads were reviewed by the staff at 
the CP stage for the Grand Gulf Nuclear Station and at the preliminary 
design analysis (PDA) stage for GESSAR-238NI.  The information available was 
deemed sufficient to adequately define the pool dynamic loads for those Mark 
III nuclear plants under review for CPs.  Since the issuance of the 
GESSAR-238NI SER in Decmber 1975, GE has conducted further tests and 
analyses to confirm and refine the original load definitions.  The GESSAR-II 
FDA application provides the finalized pool dynamic load definition for Mark 
III containments and associated piping and is the basic document used for 
review by the staff.

The staff has published the results of its generic review in NUREG-0978.  
Revision 6 to the SRP Section 6.2.1.1.C states that the acceptability of 
pool dynamic loads for plants with Mark III containments is based on 
conformance with the NRC acceptance criteria identified in Appendix C of 
NUREG-0978.  The plant-specific design for all Mark III plants was reviewed 
at the time of licensing, using this NUREG as the staff's acceptance 
criteria and the results were to be documented in the SER of each Mark III 
plant.

References:

1.   NUREG-0978, "Mark III LOCA-Related Hydrodynamic Load Definition," U.S. 
     Nuclear Regulatory Commission, February 1984.

2.   NUREG-0471, "Generic Task Problem Descriptions (Categories B, C and D 
     Tasks)," U.S. Nuclear Regulatory Commission, June 1978.



GSI No. B-36   TITLE:  Develop Design, Testing and Maintenance Criteria for
                       Atmosphere Cleanup System Air Filtration and 
                       Adsorption Units for Engineered Safety Features 
                       Systems and for Normal Ventilation Systems

This issue is concerned with the implementation of criteria for the design, 
testing and maintenance of air filtration and adsorption equipment.  The 
criteria were published in Revision 2 of Regulatory Guide 1.52 and in 
Revision 1 of Regulatory Guide 1.140.

The major changes in the criteria for this type of equipment, applicable to 
all plants with operating license applications after April 1, 1980, from 
previous requirements were in the provision of redundant protection against 
particulate release resulting from a HEPA filter failure, the requirement 
that equipment be designed for the expected range of temperature and other 
environmental conditions such as radiation, the use of both heating and 
cooling for humidity control, the use of type-tested fan motors, automatic 
initiation, testing of carbon and carbon performance requirements, provision 
of adequate drains, and access requirements and physical external clearances 
for removal and replacement of internals.  These revised criteria also 
superseded those in ORNL-NSIC-65.
.

                                     20

References:

1.   Regulatory Guide 1.52, Rev. 2, "Design, Testing and Maintenance 
     Criteria for Post-Accident Engineered Safety-Feature Atmosphere Cleanup 
     System Air Filtration and Adsorption Units of Light-Water-Cooled 
     Nuclear Power Plants," March 1978.

2.   Regulatory Guide 1.140, Rev. 1, "Design, Testing and Maintenance 
     Criteria for Normal Ventilation Exhaust System Air Filtration and 
     Adsorption Units of Light-Water-Cooled Nuclear Power Plants," October 
     1979.



GSI No. B-63   (MPA No. B-045)     TITLE:  Isolation of Low Pressure Systems
                                           Connected to the Reactor Coolant 
                                           System Pressure Boundary

This issue resulted from staff concerns regarding the potential failure of 
valves comprising the pressure isolation barrier between the reactor coolant 
system (RCS) and interfacing low-pressure systems.  Such a failure could 
result in overpressurization and attendant rupture of low-pressure piping 
and/or components, with a loss of coolant outside containment.  The Reactor 
Safety Study (WASH-1400) identified the intersystem loss-of-coolant accident 
(ISLOCA) in PWRs as a significant contributor to risk from core melt.  The 
study focused on two specific pressure isolation configurations consisting 
of two in-series check valves, with or without an open motor-operated valve 
in series.  This accident scenario was designated as Event V.

Concerns regarding Event V, as well as the staff's position that valve 
closure integrity could be improved by testing, led to the issuance of a 
Generic Letter entitled "LWR Primary Coolant System Pressure Isolation 
Valves," dated February 23, 1980, which requested a response from all 
licensees specifying whether their facilities contained the Event V 
configurations.

For the 34 facilities (32 PWRs, 2 BWRs) responding affirmatively, orders 
were issued on April 20, 1981 imposing certain corrective actions, including 
implementation of periodic testing of the identified Event V pressure 
isolation valves (PIVs) and Technical Specifications addressing surveillance 
and limiting conditions of operation for these PIVs.

References:

1.   WASH-1400 (NUREG-75/014), "Reactor Safety Study, An Assessment of 
     Accident Risks in U.S. Commercial Nuclear Power Plants," U.S. Nuclear 
     Regulatory Commission, October 1975.

2.   NRC Letter to All LWR Licensees, "LWR Primary Coolant System Pressure 
     Isolation Valves," February 23, 1980.  
.ENDEND
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