Protecting People and the EnvironmentUNITED STATES NUCLEAR REGULATORY COMMISSION
GL79005
UNITED STATES
NUCLEAR REGULATORY COMMISSION
WASHINGTON, D. C. 20555
January 26, 1979
Docket No.: 50-471
Mr. R. M. Butler
Nuclear Projects Manager
Boston Edison Company
800 Boylston Street
Boston, Massachusetts 02199
Dear Mr. Butler:
SUBJECT: INFORMATION RELATING TO CATEGORIZATION OF RECENT REGULATORY
GUIDES BY THE REGULATORY REQUIREMENTS REVIEW COMMITTEE
PILGRIM STATION, UNIT 2
We have recently advised utilities with plants in the post-CP phase of
the reactor licensing process of the status of NRC staff review and use
of recently-approved regulatory guides, and have indicated how these
guides would be used in the Operating License review of their Final
Safety Analysis Reports. Such information, while not directly
applicable to you at this time, may nonetheless be useful to you for
your future planning. The text of our letter to these utilities is the
following:
"SUBJECT: IMPLEMENTATION OF STAFF REVIEW REQUIREMENTS - (Name of
Plant) - OPERATING LICENSE REVIEW
During the last several years, we have reviewed and approved
several new regulatory guides and branch technical positions or
other modifications to existing staff positions. Our practice is
that substantive changes in staff positions be considered by the
NRC's Regulatory Requirements Review Committee (RRRC) which then
recommends a course of action to the Director, Office of Nuclear
Reactor Regulation (NRR). The recommended action includes an
implementation schedule. The Director's approval then is used by
the NRR staff as review guidance on individual licensing matters.
Some of these actions will affect your application. This letter is
intended to bring you up to date on these changes in staff
positions so that you may consider them in your Final Safety
Analysis Report (FSAR) preparation.
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"The RRRC applies a categorization nomenclature to each of its actions.
(A copy of the summary of RRRC Meeting No. 31 concerning this
categorization is attached as Enclosure 1.) Category 1 matters are
those to be applied to applications in accordance with the
implementation section of the published guide. We have enclosed lists
of actions which are either Category 2 or Category 3, which are defined
as follows:
Category 2: A new position whose applicability is to be determined
on a case-by-case basis. You should describe the extent
to which your design conforms, or you should describe an
acceptable alternate, or you should demonstrate why
conformance is not necessary.
Category 3: Conformance or an acceptable alternative is required. If
you do not conform, or do not have an acceptable
alternate, then staff-approved design revisions will be
required.
"We believe that providing you with a list of the Category 2 and 3
matters approved to date will be useful in your FSAR preparation, and
they will be an essential part of our operating license review.
Enclosure 2 is a list of the Category 2 matters. Enclosure 3 is a list
of the Category 3 matters.
"In addition to the RRRC categories, there also exists an NRR Category
4 list which are those matters not yet reviewed by the RRRC, but which
the Director, NRR,, has deemed to have sufficient attributes to warrant
their being addressed and considered in ongoing reviews. These Matters
will be treated like Category 2 matters until such time as they are
reviewed by the RRRC, and a definite implementation program is
developed. A current list of Category 4 matters is attached (Enclosure
4). These also should be considered in your FSAR.
"In some instances the items in the enclosures may not be applicable to
your application. Also, we recognize that your application may, in some
instances, already conform to the stated staff positions, in your FSAR
you should note such compliance.
"If you have any questions please let us know."
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For your information, I am enclosing a set of the enclosures that
accompanied these individual letters. These enclosures list the present
Category 1-4 matters discussed in the letter.
Sincerely,
Roger S. Boyd, Director
Division of Project Management
Office of Nuclear Reactor
Regulation
Enclosures:
As stated
cc: See next page
.
Mr. R. M. Butler
cc: Mr. William Griffin Charles Corkin, II, Esq.
Project Engineer Assistant Attorney General
Boston Edison Company Commonwealth of Massachusetts
800 Boylston Street One Ashburton Place, 19th Floor
Boston, Massachusetts 02199 Boston, Massachusetts 03105
Dale G. Stoodley, Counsel Henry Herrman, Esq.
Boston Edison Company 151 Tremont Street, 27K
800 Boylston Street Boston, Massachusetts 02111
Boston, Massachusetts 02199
Mr. & Mrs. Alan R. Cleeton
George H. Lewald, Esq. 22 Mackintosh Street
Ropes & Gray Franklin, Massachusetts 02038
225 Franklin Street
Boston, Massachusetts 02110 W. M. Sides
Quality Assurance Manager
William S. Abbott Boston Edison Company
Attorney & Counsellor at Law 800 Boylston St.
50 Congress Street, Suite 925 Boston, Massachusetts 02199
Boston, Massachusetts 02109
Mr. R. A. Fortney
B. N. Pushek EDS Nuclear
Bechtel Power Corp. 220 Montgomery Street
P. O. Box 3695 San Francisco, California 94104
San Francisco, California 94119
Edward Luton, Esq., Chairman
John D. Fassett Atomic Safety and Licensing Board
Vice President and General U. S. Nuclear Regulatory Commission
Counsel Washington, D. C. 20555
United Illuminating Company
80 Temple Street Dr. Dixon Callihan
New Haven, Connecticut 06506 Union Carbide Corporation
P. O. Box Y
Mr. R. Newman Oak Ridge, Tennessee 37830
Combustion Engineering, Corp.
1000 Prospect Hill Road Dr. Richard F. Cole
Windsor, Connecticut 06095 Atomic Safety and Licensing Board
U. S. Nuclear Regulatory Commission
W. C. Tallman, President Washington, D. C. 20555
Public Service Company of
New Hampshire
1000 Elm Street
Manchester, New Hampshire 03105
.
Mr. R. M. Butler
cc: Richard S. Salzman, Esq., Chairman
Atomic Safety and Licensing Appeal Board
U. S. Nuclear Regulatory Commission
Washington, D. C. 20555
Dr. John H. Buck
Atomic Safety and Licensing Appeal Board
U. S. Nuclear Regulatory Commission
Washington, D. C. 20555
Michael C. Farrar, Esq.
Atomic Safety and Licensing Appeal Board
U. S. Nuclear Regulatory Commission
Washington, D. C. 20555
.
UNITED STATES
NUCLEAR REGULATORY COMMISSION
WASHINGTON, D. C. 20555
SEP 24 1975
Lee V. Gossick
Executive Director for Operations
REGULATORY REQUIREMENTS REVIEW COMMITTEE MEETING NO. 31,
JULY 11, 1975
1. The Committee discussed issues related to the implementation of
Regulatory Guides on existing plants and the concerns expressed in the
June 24, 1974 memorandum, A. Giambusso to E. G. Case, subject:
REGULATORY GUIDE IMPLEMENTATION, and made the following recommendations
and observations:
a. Approval of new Regulatory Guides and approval of revisions of
existing guides should move forward expeditiously in order that
the provisions of these regulatory guides be available for use as
soon as possible in on-going or future staff reviews of license
applications. The Committee noted that over the recent past, the
approval of proposed regulatory guides whose content is acceptable
for these purposes has experienced significant delays in RRRC
review pending the determination of the applicability of the guide
to existing plants, often requiring significant staff effort. To
avoid these delays, the Committee concluded that, henceforth,
approval of proposed regulatory guides should be uncoupled from
the consideration of their backfit applicability.
b. The implementation section of new regulatory guides should
address, in general, only the applicability of the guide to
applications in the licensing review process using, in so far as
possible, a standard approach of applying the guide to those
applications docketed 8 months after the issuance date of the
guide for comment. Exceptions to this general approach will be
handled on a case-by-case basis.
c. The regulatory position of each approved proposed guide (or
proposed guide revision) will be characterized by the Committee as
to its backfitting potential, by placing it in one of three
categories:
Category 1 - Clearly forward fit only. No further staff
consideration of possible backfitting is required.
ENCLOSURE 1
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Category 2 - Further staff consideration of the need for
backfitting appears to be required for certain identified items of
the regulatory position--these individual issues are such that
existing plants need to be evaluated to determine their status
with regard to these safety issues in order to determine the need
for backfitting.
Category 3 - Clearly backfit. Existing plants should be evaluated
to determine whether identified items of the regulatory position
are resolved in accordance with the guide or by some equivalent
alternative.
From time to time, for a specific guide, there will probably be
some variation among these categories or even within a category,
and these three broad category characterizations will be qualified
as required to meet a particular situation.
d. It is not intended that the Committee categorization appear in the
guide itself. The purpose of the categorization is to indicate
those items of the regulatory position for which the Committee can
make a specific backfit recommendation without additional staff
work (Categories 1 and 3), and to indicate those items for which
additional staff work is required in order to determine backfit
considerations (Category 2).
e. The Committee recommends that for approved guides in Category 2,
staff efforts be initiated in parallel with the process leading to
publication of the guide in order that specific backfit
requirements for existing plants be determined within a reasonable
period of time after publication of the guide.
f. The Committee observed that more attention needs to be given to
the identification of acceptable alternatives to the positions
outlined in the guides in order to provide additional options and
flexibility to applicants and licensees, with the possible
benefits of additional innovation and exploration in the solution
of safety issues.
2. The Committee reviewed the proposed Regulatory Guide 1.XX: THERMAL
OVERLOAD PROTECTION FOR MOTORS ON MOTOR-OPERATED VALVES and recommended
approval. This guide was characterized by the Committee as Category 1
- no backfitting, with the stipulation that as an appropriate occasion
presented itself in conjunction with the review of some particular
aspect of existing plants, the thermal overload protection provisions
be audited.
ENCLOSURE 1 (CONT'D)
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3. The Committee reviewed the proposed Regulatory Guide I.XX: INSTRUMENT
SPANS AND SETPOINTS and recommended approval subject to the following
comment:
Paragraph 5 of Section C (page 4 of the proposed Guide) should be
reworded in light of Committee comments, to the satisfaction of
the Director, Office of Standards Development. This guide was
characterized by the Committee as Category 1 - no backfit.
4. The Committee reviewed Proposed Regulatory Guide 1.97: INSTRUMENTATION
FOR LIGHT WATER COOLED NUCLEAR POWER PLANTS TO ASSESS PLANT CONDITIONS
DURING AND FOLLOWING AN ACCIDENT and deferred further consideration to
a later meeting in order to permit incorporation of recent comments by
the Division of Technical Review.
Edson G. Case, Chairman
Regulatory Requirements Review
Committee
ENCLOSURE 1 (CONT'D)
.
September 15, 1978
CATEGORY 2 MATTERS
Document
Number Revision Date Title
RG 1.27 2 1/76 Ultimate Heat Sink for Nuclear
Power Plants
RG 1.52 1 7/76 Design, Testing, and
Maintenance
Criteria for Engineered-Safety-
Feature Atmosphere Cleanup
System Air Filtration and
Adsorption Units of Light Water
Cooled Nuclear Power Plants
(Revision 2 has been published
but the changes from Revision
1 to Revision 2 may, but need
not, be considered.
RG 1.59 2 8/77 Design Basis Floods for Nuclear
Power Plants
RG 1.63 2 7/78 Electric Penetration Assemblies
in Containment Structures for
Light Water Cooled. Nuclear
Power Plants
RG 1.91 1 2/78 Evaluation of Explosions
Postulated to Occur on
Transportation Routes Near
Nuclear Power Plant Sites
RG 1.102 1 9/76 Flood Protection for Nuclear
Power Plants
RG 1.105 1 11/76 Instrument Setpoints
RG 1.108 1 8/77 Periodic Testing of Diesel
Generator Units Used as Onsite
Electric Power Systems at
Nuclear Power Plants
RG 1.115 1 7/77 Protection Against
Low-Trajectory Turbine
Missiles
RG 1.117 1 4/78 Tornado Design Classification
RG 1.124 1 1/78 Service Limits and Loading
Combinations for Class 1 Linear
Type Component Supports
RG 1.130 0 7/77 Design Limits and Loading
Combinations for Class 1 Plate-
and Shell-Type-Component
Supports
(Continued)
ENCLOSURE 2
.
CATEGORY 2 MATTERS (CONT'D)
Continued
Document
Number Revision Date Title
RG 1.137 0 1/78 Fuel Oil Systems for Standby
Diesel Generators (Paragraph
C.2)
RG 8.8 2 3/77 Information Relevant to
Ensuring that Occupational
Radiation Exposures at Nuclear
Power Stations Will be as Low
as is Reasonably, Achievable
(Nuclear Power Reactors)
BTP ASB Guidelines for Fire Protection
9.5-1 1 for Nuclear Power Plants (See
Implementation Section,
Section D)
BTP MTEB 5-7 4/77 Material Selection and
Processing Guidelines for BWR
Coolant Pressure Boundary
Piping
RG 1.141 0 4/78 Containment Isolation
Provisions for Fluid Systems
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ENCLOSURE 2 (CONT'D)
.
September 15, 1973
CATEGORY 3 MATTERS
Document
Number Revision Date Title
RG 1.99 1 4/77 Effects of Residual Elements on
Predicted Radiation Damage to
Reactor Vessel materials
(Paragraphs C.1 and C.2.
RG 1.101 1 3/77 Emergency Planning for Nuclear
Power Plants
RG 1.114 1 11/76 Guidance on Being Operator at
the Controls of a Nuclear
Power Plant
RG 1.121 0 8/76 Bases for Plugging Degraded PWR
Steam Generator Tubes
RG 1.127 1 3/78 Inspection of Water-Control
Structures Associated with
Nuclear Power Plants
RSB 5-1 1 1/78 Branch Technical Position:
Design Requirements of the
Residual Heat Removal System
RSB 5-2 0 3/73 Branch Technical Position:
Reactor Coolant System
Overpressurization Protection
(Draft copy attached)
RG 1.97 1 8/77 Instrumentation for Light Water
Cooled Nuclear Power Plants to
Assess Plant Conditions During
and Following an Accident
(Paragraph C.3 - with
additional guidance on
paragraph C.3.d to be provided
later)
RG 1.68.2 1 7/78 Initial Startup Test Program to
Demonstrate Remote Shutdown
Capability for Water-Cooled
Nuclear Power Plants
RG 1.56 1 7/78 Maintenance of Water Purity in
Boiling Water Reactors
Attachment:
BTP RSB 5-2 (Draft)
ENCLOSURE 3
.
BRANCH TECHNICAL POSITION RSB 5-2
OVERPRESSURIZATION PROTECTION OF PRESSURIZED WATER REACTORS
WHILE OPERATING AT LOW TEMPERATURES
A. Background
General Design Criterion 15 of Appendix A. 10 CFR 50, requires that
"the Reactor Coolant System and associated auxiliary, control, and
protection systems shall be designed with sufficient margin to assure
that the design conditions of the reactor coolant pressure boundary are
not exceeded during any condition of normal operation, including
anticipated operational occurrences."
Anticipated operational occurrences, as defined in Appendix A of 10 CFR
50, are "those conditions of normal operation which are expected to
occur one or more times during the life of the nuclear power unit and
include but are not limited to loss of power to all recirculation
pumps, tripping of the turbine generator set, isolation of the main
condenser, and loss of all offsite power.
Appendix G of 10 CFR 50 provides the fracture toughness requirements
for reactor pressure vessels under all conditions. To assure that the
Appendix G limits of the reactor coolant pressure boundary are not
exceeded during any anticipated operational occurrences, Technical
Specification pressure-temperature limits are provided for operating
the plant.
The primary concern of this position is that during startup and
shutdown conditions at low temperature, especially in a water-solid
condition, the reactor coolant system pressure might exceed the reactor
vessel pressure-temperature limitations in the Technical Specifications
established for protection against brittle fracture. This inadvertent
overpressurization could be generated by any one of a variety of mal-
functions or operator errors. Many incidents have occurred in operating
plants as described in Reference 1.
Additional discussion on the background of this position is contained
in Reference 1.
ENCL 3 (CONT)
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B. Branch Position
1. A system should be designed and installed which will prevent
exceeding the applicable Technical Specifications and Appendix G
limits for the reactor coolant system while operation at low
temperatures. The system should be capable of relieving pressure
during all anticipated overpressurization events at a rate
sufficient to satisfy the Technical Specification limits,
particularly while the reactor coolant system is in a water-solid
condition.
2. The system must be able to perform its function assuming any
single active component failure. Analyses using appropriate
calculational techniques must be provided which demonstrate that
the system will provide the required pressure relief capacity
assuming the most limiting single active failure. The cause for
initiation of the event, e.g., operator error, component
malfunction, will not be considered as the single active failure.
The analysis should assume the most limiting allowable operating
conditions and systems configuration at the time of the postulated
cause of the overpressure event. All potential overpressurization
events must be considered when establishing the worst case event.
Some events may be prevented by protective interlocks or by
locking out power. These events should be reviewed on an
individual basis. If the interlock/power lockout is acceptable, it
can be excluded from the analyses provided the controls to prevent
the event are in the plant Technical Specifications.
3. The system must meet the design requirements of IEEE 279 (see
Implementation). The system may be manually enabled, however, the
electrical instrumentation and control system must provide alarms
to alert the operator to:
a. properly enable the system at the correct plant condition
during cooldown,
b. indicate if a pressure transient is occurring.
4. To assure operational readiness, the overpressure protection
system must be tested in the following manner:
a. A test must be performed to assure operability of the system
electronics prior to each shutdown.
b. A test for valve operability must as a minimum be conducted
as specified in the ASME Code Section XI.
c. Subsequent to system, valve, or electronics maintenance, a
test on that portion(s) of the system must be performed prior
to declaring the system operational.
ENCL 3 (CONT)
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5. The system must meet the design requirements of regulatory Guide
1.26, "Quality Group Classifications and Standards for Water-,
Steam-, and Radioactive-Waste-Containing Components of Nuclear
Power Plants" and Section III of the ASME Code.
6. The overpressure protection system must be designed to function
during an Operating Basis Earthquake. It must not compromise the
design criteria of any other safety-grade system with which it
would interface, such that the requirements of Regulatory Guide
1.29, "Seismic Design Classification" are met.
7. The overpressure protection system must not depend on the
availability of offsite power to perform its function.
8. Overpressure protection systems which take credit for an active
component(s) to mitigate the consequences of an overpressurization
event must include additional analyses considering inadvertent
system initiation/actuation or provide justification to show that
existing analyses bound such an event.
C. Implementation
The Branch Technical Position, as specified in Section B, will be used
in the review of all Preliminary Design Approval (PDA), Final Design
Approval (FDA), Manufacturing License (ML), Operating License (OL), and
Construction Permit (CP) applications involving plant designs
incorporating pressurized water reactors. All aspects of the position
will be applicable to all applications, including CP applications
utilizing the replication option of the Commission's standardization
program, that are docketed after March 14, 1978. All aspects of the
position, with the exception of reasonable and justified deviations
from IEEE 279 requirements, will be applicable to CP, OL, ML, PDA, and
FDA applications docketed prior to March 14, 1978 but for which the
licensing action has not been completed as of March 14, 1978. Holders
of appropriate PDA's will be informed by letter that all aspects of the
position with the exception of IEEE 279 will be applicable to their
approved standard designs and that such designs should be modified, as
necessary, to conform to the position. Staff approval of proposed
modifications can be applied for either by application by the
PDA-holder on the PDA-docket or by each CP applicant referencing the
standard design on its docket.
The following guidelines may be used, if necessary, to alleviate
impacts on licensing schedules for plants involved in licensing
proceedings nearing completion on March 14, 1978:
ENCL 3 (CONT)
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1. Those applicants issued an OL during the Period between March 14,
1978 and a date 12 months thereafter may merely commit to meeting
the position prior to OL issuance but shall, by license condition,
be required to install all required staff-approved modifications
prior to plant startup following the first scheduled refueling
outage.
2. Those applicants issued an OL beyond March 14, 1979 shall install
all required staff-approved modifications prior to initial plant
startup.
3. Those applicants issued a CP, PDA, or ML during the period between
March 14, 1978 and a date 6 months thereafter may merely commit to
meeting the position but shall, by license condition, be required
to amend the application, within 6 months of the date of issuance
of the CP, PDA, or ML, to include a description of the proposed
modifications and the bases for their design, and a request for
staff approval.
4. Those applicants issued a CP, PDA, or ML after September 14, 1978
shall have staff approval of proposed modifications prior to
issuance of the CP, PDA. or ML.
D. References
1. NUREG-0138, Staff Discussion of Fifteen Technical Issues Listed in
Attachment to November 3, 1976 Memorandum from Director, NRR, to
NRR Staff.
ENCL 3 (CONT)
