Home NRC Library Document Collections Generic Communications Generic Letters 1990

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
.

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.
.

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
.

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.

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.

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.

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

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:

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
.

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
.

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
.

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.
.

(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.

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.
.

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,
.

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.
.

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.
.

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.

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.
.

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
.

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
.

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.
.

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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