ML20207A246
| ML20207A246 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 03/04/1987 |
| From: | Celeste R OHIO, STATE OF |
| To: | Zech L NRC COMMISSION (OCM) |
| Shared Package | |
| ML20207A252 | List: |
| References | |
| NUDOCS 8704240217 | |
| Download: ML20207A246 (35) | |
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STATE OF OHIO OFFICE OF THE GOVERNOR COLUMBUS 43266-0601 RICHARD F. CELESTE GOVERNOR March 4, 1987 Mr. Lando Zech, Chairman Nuclear Regulatory Commission Washington, DC 20555
Dear Chairman Zech:
I am writing to urge the Nuclear Regulatory Commission to investigate and correct quickly, as required, the alleged generic problems at Babcock and Wilcox nuclear plants described in the recent petition of the Union of Concerned Scientists.
As Governor of a State where one Babcock and Wilcox reactor, Davis-Besse, operates, I am deeply concerned about any design problems that may affect public safety.
While I have welcomed the recent NRC review of operating procedures at Davis-Besse, and the upgrading of key systems there, I still urge the swiftest possible review and resolution of the generic problems alleged in the petition.
I note doubts raised by the NRC itself in its response of April 15, 1986, to a letter from the Subcommittee on Energy Conservation and Power of the House Energy and Commerce Committee.
These questioned the adequacy of its computer codes to model complex transients and accidents at Babcock and Wilcox plants.
I am also concerned by the internal NRC memo of April 30, 1986, which describes budget cuts that make the NRC unable to resolve satisfactorily outstanding safety issues at these plants.
Because of these apparent gaps in the NRC's ability to assess these safety issues fully, I ask the Commission to clarify the technical basis on which it is allowing these plants to operate.
I look forward to your early reply.
Best regards, k.
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Richard F.
Celeste Governor RFC/fj D0YZYOld]~
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NUCLEAR REGULATORY COMMISSION E
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s, +....s CHAIRMAN The Honorable Philip R. Sharp, Chairman Subcommittee on Energy and Power Committee on Energy and Commerce United States House of Representatives Washington, DC 20515
Dear Mr. Chairman:
Your letter of March 16, 1987, requested information that would assist the Subcommittee on Energy and Power in their investigation of the implications of the safety of nuclear power olants in light of the recent Surry accident.
Answers to the specific questions in your March 16, 1987, letter are enclosed.
The NRC continues to take an active interest in degradation o# any nuclear power plant equipment which has relevance to the safety of the plant, and we will continue to monitor individual plant performance and overall industry experience in this area.
Where plant specific problems occur, the need for generic action will be assessed and appropriate corrective measures will be taken.
This is a normal part of the NRC's ongoinq monitoring of industry performance.
In this regard the NRC staff will continue to assess the safety imolications of the Surry feedwater pipe failure.
I hope that the information provided will assist your review.
Comnissioner Asselstine disagrees with this response and will provide his views in a separate letter.
Sincerely, AA-t4o lv. M Lando W. Zed 1, Jr.
Enclosure:
Answers to Specific Questions cc:
Rep. Carlos J. Moorhead om,-u n nW
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r QUESTION 1(al.
What codes, standards, specifications and regulatory requirements are applied to the failed.feedwater line and associated equionent (condenser, feedwater pumps, steam turbine, piDelines and components)?
Are these systems classified as nuclear or non-nuclear?
Are they classified as safety or nonsafety related systems?
ANSWER.
The requirements for the construction and inservice inspection of the safety-related systems differ from nonsafety-related systems because safety-related systems are relied upon to provide the capability to prevent or mitigate the consequences of accidents, remove heat from the reactor and maintain it in a safe shutdown condition.
The construction requirements of safety-related systems differ from nonsafety-related systems in the areas of materials inspection and non-destructive examination of piping system weldments, overpressure orotection, and quality assurance, including third party inspection.
For the main steam and feedwater system, the principal difference between the design of the safety-related and nonsafety-related components are that the safety-related systems are required to meet seismic criteria and requirements for dasian quality assurance which complies with 10 CFR 50, Appendix B.
Safety-related portions of these lines are also required to receive inservice inspection and testing under 10 CFR 50.55a(q), which invokesSection XI of the ASME Boiler and Pressure Vessel Code.
Nonsafety-related systems are not required by any NRC standard, or regulatory requirement to receive inservice inspection.
The term non-nuclear is not well defined, but as used by many and in the response below it describes piping not constructed to Section I!! of ASME Boiler and Pressure Vessel Code.
Power plants built prior to the adoption of Section III of the ASME Code were constructed to other standards such as ANSI /ASME R31.1.
The condensate and feedwater systems of PWRs provide feedwater at the required temperature, pressure, and flow rate to the secondary side of the steam generators.
Condensate is pumpad from the main condenser hotwell by the condensate pumps, passes through the low pressure feedwater heaters to the feedwater pumps, and then is cumped through the high pressure feedwater heaters to the secondary side of the steam generators.
That portion of the condensate and feedwater system located within the turbine building and the portion of the feedwater lines between turbine building up to the containment isolation valves located outside the reactor containment building are not classified
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00ESION 11al.
(Continued) safety-related.
The portion of the feedwater system from the
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containment isolation valves located outside the reactor containment buildina up to and including the secondary side of the steam generators are within the nuclear portion of the power plant and are classified safety-related.
An auxiliary feedwater system is connected to th'e main feedwater ~
systen and normally operates during startup, bot standby and shutdown to provide feedwater to the steam generators.
This system also functions as an emergency system for the removal of heat from the primary system when the main feedwater system is not available and for emergency conditions including small loss-of-coolant accidents.
The entire auxiliary feedwater system is classi#ied as a safety-related system.
Regulatory guidance with respect to the auxiliary feedwater system, the main feedwater system, main condensers and condensate system is provided in the following sections of Standard-Review Plan, NUREG-0800, Revision 2, (July 1981) (Attached).
NUREG Section Title 10.4.1 Main Condensers 10.4.7 Condensate and Feedwater
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System 10.4.9 Auxiliary Feadwater System The following Reculatory Guides also provide guidance with respect to Quality Group Classification (applicable codes and standards), Seismic Design requirements, and Quality Assurance requirements for components of nuclear power plants (Attached).
Regulatory Guide Title 1.26, Revision 3, Quality Group Classifications (February 1976) and Standards for Water,
Steam, and Radioactive-Waste-Containino Components of Nuclear Power Plants.
1.29, Revision 3, Seismic Design Classification (September 19781 I
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-r 00ESTION 1(a).
(Continued) During plant operation,Section XI of the ASME Boiler and Pressure Vessel Code, " Rules for Inservice Inspection of Nuclear Power Plant Components" provides guidance on inservice inspection of components and inservice testing of pumps and valves which are safety-related; this'is due to the fact that Surry was constructed prior to the development of ASME Section-III which is applicable to Safety Related Systems today.
The construction codes and standards applicable to the auxiliary feedwater system and' the safety-related portion of the nain feedwater system at Surry Units 1 and 2 are as follows:
1.
Portions of main feedwater piping - USAS B31.1.0 - 1967 supplemented by ASP 1E Code Case N-7.
Auxiliary feedwater piping - USAS B31.1.0 - 1967, 2.
Pumps, such as auxiliary feedwater pumps - manufacturer's standards 3.
Valves - manufacturer's standards and USAS B31.1.0 - 1967 and related standards,' such as B16.5.
The construction codes and standards applicable to the nonsafety-related portions of the condensate and feedwater system at Surry Units 1 and 2 are as follows:
1.
Condensate and feedwater piping - USAS B31.1.0 - 1967 Power Piping Code.
2.
Pressure vessels, such as feedwater heaters - ASME Boiler and Pressure Vessel Code,Section VIII, Pressure Vessels.
3.
Pumps, such as condensate and feedwater pumps, and steam turbines - manufacturer's standards.
4.
Valves - manufacturer's standards and USAS B31.1.0 - 1967 and related standards, such as B16.5.
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NUREG 0800 (Fcrmsrly NUREG-75/087) i A "cg I
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STANDARD REVIEW PLAN
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10.4.1 MAIN CONDENSERS REVIEW RESPONSIBILITIES Primary Power Systems Branch (PSB)
Secondary - None l
1.
AREAS OF REVIEW The main condenser (MC) system is designed to condense and deaerate the exhaust steam from the main turbine and provide a heat sink for the turbine bypass system.
1.
The PSB reviews the performance requirements of the main condenser for both direct and indirect cycle plants during all operating conditions.
Emphasis will be placed on the review of direct cycle facilities with regard to the i
prevention of loss of vacuum, corrosion and/or erosion, and hydrogen buildup.
j 2.
The PSB reviews the design of the MC system with respect to the following:
The means to detect, control and facilitate correction of the leakage of a.
cooling water into the condensate; to detect radioactive leakage into or out of the system; and to preclude accidental releases of radioactive g
materials to the environment in amounts in excess of established limits, b.
Instrumentation and. control features that determine and verify that the MC is operating in a correct mode.
The means provided to deal with flooding from a complete failure of the c.
MC and to preclude damage to safety-related equipment from the flooding, l
d.-
The capability of the MC to withstand the blowdown effects of steam from 1
the turbine bypass system.
In the review of the Main Condenser, the PSB will coordinate other branch evalua-tions that interface with the overall review of the system as follows.
The Rev. 2 - July 1981 USNRC STANDARD REVIEW PLAN Standard review plans are prepared for the guidance of the Office of Nuclear Reactor Regulation staff responsible for the review of applications to construct and operate nuclear power plants. These documents are made available to the public as part of the Commission's policy to inform the nuclear industry and the general public of regulatory procedures and policies. Standard review plans are not substitutes for regulatory guides or the Commission's regulations and compliance with them is not required The standard review plan sections are keyed to the Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants.
Not all sections of the Standard Format have a corresponding review plan.
Published standard review plans will be revised periodically, as appropriate, to accommodate comments and to reflect new informa-tion and experience.
Comments and suggestions for improvement will be considered and should be sent to the U.S. Nuclear Regulatory Commission.
offier, of Nuclear Reactor Regulation. Washington, D.C. 205Ei6.
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' Effluent Treatment Systers Branch evaluates the inventory of radioactive contami-
'nants'in the MC during power operation and during shutdown as part of its primary review responsibility for SRP Section 11.5.
The Materials Engineering Branch, upon request of PSB, evaluates the adequacy of the materials of construction, the methods used to reduce the corrosion and/or erosion of-MC tubes and compo -
nents, the permissible cooling water inleakage, and the allowed time of operation i
with inleakage without affecting condensate /feedwater quality for safe reactor operation. The Auxiliary Systems Branch determines that safety-related systems and structures are protected from the effects of flooding as part of its primary review responsibility for SRP Section 3.4.1.
The procedures and Test Review Branch determines the acceptability of the preoperational and startup tests as part of its primary review responsibility for SRP Section 14.0.
The reviews for fire protection, technical specifications, and quality assurance are coordinated and performed by the Chemical Engineering Branch, Licensing Guidance Branch, and Quality Assurance Branch as part of their primary review responsibility for SRP Sections 9.5.1, 16.0, and 17.0, respectively.
3 For those areas of review identified above as being part of the primary review responsibility of the other branches, the acceptance criteria necessary for the review and their methods of application are contained in the referenced SRP section of corresponding primary branches.
II. ACCEPTANCE CRITERIA Acceptability of the design of the main condenser system, as described in the applicant's safety analysis report (SAR), is based on meeting the requirements 1,
of General Design Criterion 60 (GDC 60) and on the similarity of the design to that of plants previously reviewed and found acceptable.
l The design of the Main Condenser System is acceptable if the integrated design of the system meets the requirements of GDC 60 as related to failures in the 1
design of the system which do not result in excessive releases of radioactivity 1
to the environment.
In addition, GDC 60 is satisfied-if the system is designed such that failures do not cause unacceptable condensate quality, or flooding of areas housing safety-related equipment.
III.
REVIEW PROCEDURES The procedures below are used during the construction permit (CP) review to determine that the design criteria and bases and the preliminary design meet the acceptance criteria given in subsection II.
For the review of operating license (0L) applications, the procedures are used to verify that the initial design criteria and bases have been appropriately implemented in the final design as set forth in the final safety analysis report. The reviewer will select and emphasize material from this SRP section as may be appropriate for a particular case.
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The primary reviewer will coordinate this review with other branches' areas of review as stated'in subsection I.
The primary reviewer obtains and uses such input as required to assure that this review procedure is complete.
1.
The SAR is-reviewed to determine that the system description delineates the main condenser system capabilities including the minimum system heat k
transfer and system flow requirements for normal plant and turbine bypass operation. Measures provided to prevent loss of vacuum, corrosion and/or erosion of MC tubes and components, and hydrogen buildup in the MC are 10.4.1-2 Rev. 2 - April 1981
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reviewed, with particular emphasis onithese measuras for: direct. cycle-
. System performance requirements' are, Lt-
- (boiling water-reactor) plants. reviewed to determine that they ' satisfactorily dation conditions (e.g., leakage, partial loss of vacuum) and describe
.the procedures that are followed to detect.and correct these conditions.-
The SAR is also reviewed to determine that any allowed MC system degraded
. operation doet not'have-an adverse effect on the reactor primary. system or secondary system in the case of pressurized water. reactors.
The reviewer evaluates the MC system design to verify that:
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2.
Means have been provided for detecting, controlling and correcting condenser cooling water leakage into the condensate.
a.
The permissible cooling water inleakage and time of operation with inle.kage are provided to assure that condensate /feedwater. quality
.b.
can be maintained within safe limits.
4
' Measures have been provided to detect radioactive leakage into and out of the MC system and to preclude unacceptable accidental releases
' c.
of radioactivity to the environment from the Jystem.
l-The system is provided with instrumentation and control features that d.
determine and verify that the MC is operating -in a correct mode.
l The reviewer uses engineering judgment and the results of failure modes 3.
and effects analyses to determine that:
The failure of a main condenser and the resulting-flooding lwill not Reference to sections a.
preclude operation of any essential systems.of the SAR describing layout drawings will be necessary, as well as the SAR tabulation of Statements seismic design classifications for structures and systems.
in the SAR that verify that the above conditions are met are acceptable.
The system, in conjunction with the main steam' system, has provisions b.
to detect loss of' condenser vacuum and to effect isolation of the For direct cycle plants, it will be acceptable if the i
detection system in the MC can actuate the main steam isolation valves steam source.
to limit the quantity of steam lost-from the condenser.
Design provisions have been incorporated into the MC that will preclude
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component or tube failures due to steam blowdown from the turbine c.
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bypass' system.
IV.
EVALUATION FINDINGS The reviewer verifies that sufficient information has been provided and his review supports conclusions of the following type, to be included-in the staff's safety evaluation report:
The main condenser system (MC) includes all components and equipment l
from the turbine exhaust to the connections and' interfaces with the i
The scope of review of the main main condensate and other systems.
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condenser system for the included layout drawings, piping and instrumentation diagrams, and descriptive information for the main condenser system and supporting
. systems that are essential to its operation.
Rev. 2 - July 1981 10.4.1-3 i
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,The basis for acceptance of the main condenser system in our review was conformance of the design, design criteria, and design bases to the Commission's regulation as set forth in GDC 60.
The staff concludes that the main condenser system design is acceptable and meets the requirements of GDC 60 with respect to failures in the design of the system which do not result in excessive releases of radioactivity to the environment.
The applicant has met this require-ment by providing radioactive monitors in the system to detect leakage into and out of the main condenser.
V.
IMPLEMENTATION The following is intended to provide guidance to applicants and licensees regarding the NRC staff's plans for using this SRP section.
Except in those cases in which the applicant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, the method described herein will be used by the staff in its evaluation of conformance with Commission regulations.
VI. REFERENCES 1.
10 CFR Part 50, Appendix A, " General Design Criterion 60, " Control of Releases of Radioactive Materials to the Environment."
2.
Regulatory Guide 1.68, " Initial Test Programs for Water-Cooled Reactor Power Plants."
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10.4.7 CONDENSATE AND FEE 0 WATER SYSTEM REVIEW RESPONSIBILITIES Primary - Auxiliary Systems Branch (ASB) l Secondary - None I.
AREAS OF REVIEW The condensate and feedwater system (CFS) provides feedwater at the require ture, pressure, and flow rate to the reactor for boiling water reactor (BWR)
Condensate and to the steam generators for pressurized water reactor (PWR) plants.
is pumped from the main condenser hotwell by the condensate pumps, pa the low pressure feedwater heaters to the feedwater pumps, and then is pumpe through the high pressure feedwater heaters to the nuclear steam supply syste ASB reviews the CFS from the condenser outlet to the connectio steam supply system and to the heater drain system to assure conforman For indirect cycle plants, there are also
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2, 4, 5, 44, 45 and 46.
interfaces with the secondary water makeup system and the auxiliary feedwate Design Criteria
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The only part of the CFS classified as safety-The CFS is used for normal shutdown.
related, i.e., required for safe shutdown or in the event of postulated accidents is the feedwater piping from the steam generators for PWRs and from the nuclea steam supply system for BWRs, up to and including the outermost containm tion valve.
The ASB reviews the characteristics of the CFS with respect to the capability to supply adequate feedwater to'the nuclear steam supply system as requ 1.
for normal operation and shutdown.
The ASB review determines that an acceptable design has been established fo 2.
The interfaces of the CFS with the auxiliary feedwater system (PWR), the reactor core isolation cooling system (BWR), and the condensate cleanup a.
system with regard to functional design requirements and seismic d classification.
Rev. 2 - July 1981 USNRC STANDARD REVIEW PLAN i
f Standard review plans are prepared for the guidance of the office of Nuclear Reactor Regulat t of the applications to construct and operate nuclear power plants. These documente are made availab d d review Commission's policy to enform the nuclear industry and the general public of reguistory procedure h
i t required. The plans are not substitutes for regulatory guidea or the Commission's regufstions and compliance with Nuclear Power Plants.
standard review plan sections are keyed to the Standard Format and Content of Safety Analysis Report Not all sections of the Standard Format have a corresponding review plan.
i f rme-Published standard review plans will be revised periodically es appropriate. to accommodate co tion and experience.
Commente and suggestions for improvement will be considered and should be sent to th Office of Nuclear Reactor Regulation Washington, D.C. 20566.
r.
l b.
The fe:dwater syste] (PWR), including th2 auxiliary feedwater systcm piping entering the steam generator, with regard to possible fluid flow instabilities (e.g., water hammer) during normal plant operation as well as during upset or accident conditions.
c.
The detection of major system leaks that could affect the functional performance of safety-related equipment.
i 3.
ASB also performs the following reviews under the SRP sections indicated:
(a) Review for flood protection is performed under SRP Section 3.4.1, (b) Review of the protection against internally generated missiles is performed under SRP Section 3.5.1.1, (c) Review of the structures, systems, and components to be protected against externally generated missiles is performed under SRP Section 3.5.2, and (d) Review of high-and moderate-energy pipe breaks is performed under SRP Section 3.6.1.
The ASB will coordinate evaluations performed by other branches that interface with the overall evaluation of the system as follows:
The Reactor Systems Branch (RSB) determines that transients resulting from feedwater flow control malfunctions will not violate the primary system pres-sure boundary integrity criterion as part of its primary review responsibility for SRP Sections 15.1.1 through 15.1.4, and that the loss of normal feedwater flow will not violate the fuel damage criterion or the system pressure boundary integrity criterion as part of its primary review responsibility for SRP Section 15.2.7.
The Power Systems Branch (PSB) evaluates the system power sources with respect to their capability to perform safety-related functions during normal, transient, and accident conditions as part of its primary review responsibility for SRP Section 8.3.1.
The Structural Engineering Branch (SEB) determines the accepta-bility of the design analyses, procedures, and criteria used to establish the ability of seismic Category I structures housing the system and supporting systems to w.ithstand the effects of natural phenomena such as the safe shutdown earthquake (SSE), the probable maximum flood (PMF), and tornado missiles as part of its primary review responsibility for SRP Sections 3.3.1, 3.3.2, 3.5.3, 3.7.1 through 3.7.4, 3.8.4, and 3.8.5.~ The Mechanical Engineering Branch (MEB) determines that the components, piping and structures are designed in accordance with applicable codes and standards as part of its primary review responsibility for SRP Sections 3.9.1 through 3.9.3.
The MEB determines the acceptability of the seismic and quality group classifications for system components as part of its primary review responsibility for SRP Sections 3.2.1 and 3.2.2.
The MEB also reviews the adequacy of the inservice testing program of pumps and valves j
as part of its primary review responsibility for SRP Section 3.9.6.
Upon request, the MEB determines the acceptability of design analyses, procedures, l
and criteria used to establish the adequacy of devices or restraints as they 1
may relate to significant water hammers in system piping and the MEB reviews test programs of components that may be affected by water hammers.
The Materials En Engineering Branch (MTEB) verifies that inservice inspection requirements are met for system components as part of its primary review responsibility for SRP 10.4.7-2 Rev. 2 - July 1981 P
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i ility of the Eaterials of c' Sect, ion 6.6 and, upon request, verifies the compat bThe review for Fire Protection construction with service conditions.
Specifications, and Quality Assurance are coordinated and perform Chemical Engineering Branch, Licensing Guidance Branch, and Qualit Branch as part of their primary review responsibility for SRP 16.0, and 17.0, respectively.
t tion and electrical equipment-
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.the seismic qualification of Category I instrumen a
' and the environmental qualification of mechanical:and electrical equip l
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11, part of its primary review responsibility for SRP Sectio I-
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will review the instrumentation and controls associated with respectively.
l control system (8WR) or steam generator-level control system (PWR).
For those areas of review identified above as being part of the prima f
responsibility of other branches, the acceptance criteria necessary.fo f
d SRP review'and their. methods of application are contained in the re erence sections of the corresponding primary branches.
i II.
ACCEPTANCE CRITERIA Acceptability of the condensate and feedwater sy t
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Listed below t
of General Design Criteria and the positions of regulatory guides.
il are the' specific criteria as they relate to the CFS.
General Design Criterion 2, as related to the system being capable l
Acceptance is based on meeting the l
1.
standing the effects of earthquakes.
i guidance of Regulatory Guide 1.29, Position C.1 for safety-related and Position C.2 for nonsafety-related portions.
General Design Criterion 4, as related to the dynamic effects associa l
l with possible fluid flow instabilities (e.g., water i
J 2.
J Acceptance is based on meeting the guidance co hammers in steam generators with top feedring designs.
l General Design Criterion 5, as related to the capability of shared syste l
and components important to safety to perform required safety functio 3.
i j
General Design Criterion 44, as it relates to:
4.
l The capability to transfer heat loads from the reactor system' to a heat sink under both normal operating.and accident conditions, a.
Redundancy of components so that under accident conditions the sj function can be performed assuming a single active component failu b.
j (This may be coincident with the loss of offsite power for certain i
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events.)
l' The capability to isolate components, subsystems, or piping if r l.
so that the system safety function will be maintained.
c.
General Design Criterion 45, as related to design provisions to perm periodic inservice inspection of system components and equipment.
5.
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G:neral' Design Critorion 46, as related to design provisions to parmit appropriate functional testing of the system and components to assure
' structural integrity and leak-tightness, operability and performance of.
- active components, and capability of the integrated system to function as J
. intended during normal, shutdown, and accident conditions.
III. REVIEW PROCEDURES' f
The procedures below'are used 'during the construction permit (CP) review to determine that the design' criteria and bases and the preliminary design as set
'forth in the preliminary safety analysis report meet.the acceptance criteria given in subsection II of this SRP section.
For the review of operating license (OL) applications, the procedures are used to verify.that the initial design 4
criteria and bases have been appropriately implemented in the final design as set forth in the final safety analysis report.
The primary reviewer will coordinate this review with the areas of review of i
. interfacing. branches as stated in subsection I of this SRP section.
The primary _
reviewer obtains and uses such-inputs as required to assure that this review.
procedure is complete.
The reviewer will select and emphasize material from this'SRP section as may-4 be appropriate for a particular case.
The SAR is reviewed to determine that the system description and diagrams delineate the function of the condensate and feedwater system under normal and abnormal conditions. ~The reviewer verifies the following:
1.
The system has been designed to function as required for all modes of-operation.
The results of failure modes and effects analyses presented in the SAR, if any, are used in making this determination.
2.
The system piping is designed to preclude hydraulic instabilities from occurring in the piping for all modes of operation.
As appropriate, the reviewer evaluates the results of model tests and analyses that are relied on to verify that water hammer will not occur, or proposed tests of the installed system that are intended to verify design adequacy.
Steam generators that use top feed designs are reviewed in accordance with Branch Technical Position ASB 10-2.
3.
The outermost containment isolation valves and all downstream piping to the nuclear steam supply system are designed in accordance with seismic Category I requirements.
The review for seismic design is performed by j
SEB and the review for seismic and quality group classification is per-1 formed by MEB as indicated in subsection I of this SRP section.
1 4.
The CFS design is such that the plant can be safely shut down using the auxiliary feedwater system or the reactor core isolation cooling system, if required.
L 5.
The CFS design, or other plant systems, provide the capability to detect and control leakage from the system.
1 i
6.
The reviewer verifies that the essential portion of the system has been
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designed so that system function will be maintained as required in the event of adverse environmental phenomena or loss of offsite power.
The 4
10.4.7-4 Rev. 2 - July 1981 1
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1 a-review for protection against natural phenomena is performed in I
Chapter ~3 SRP sections.
ing judgment-and the.results of failure modes and effects analyses, to-h determine that the failure of; nonessential portions oflthe. system or of d ds;and located
.other systems not designed to seismic Category I stan ar close to essential portions of the system, or of nons CFS, will not preclude operation of the essential portions of the CFS.
i IV. EVALUATION FINDINGS The reviewer verifies that sufficient information has b safety evaluation report:
The condensate.and feedwater system includes all components and
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equipment from the condenser outlet to the connection with the nuclear steam supply system and to the heater drain system [
Based on the review of the applicant's proposed design criteria, the 4,.
design bases, and safety classification for the safety-related por-l tions of the condensate and feedwater system and the requirements for system performance for all conditions of plant operation, the f
staff concludes that the design of the condensate and feedwater system i
and supporting systems is in conformance with the Commission regula-r tions as set forth in General Design Criterion 2, 4, 5, 44, 45 and
]
This conclusion is based on the following:
46.
The applicant has met the requirements of General Design Criterion 2 with respect to safety-related portions of the system 1.
~
being capable of withstanding the effects of earthquakes by meeting Regulatory Guide 1.29 Position C.1 for the safety-related portions i
i and Position C.2 for the nonsafety-related portions, t
The applicant has met the requirements of General De i
2.
~
possible fluid flow instabilities (e.g., water hammers) by having I
the feedwater system designed in accordance with the guidance ~
contained in Branch Technical Position AS810-2 and thereby eliminating or reducing the possibility of water hammers in this i
system (PWRs only).
I The applicant has met the requirements'of General Design l
Criterion 5 with respect to the capability of shared systems
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3.
l and components important to safety to perform required safety j
We have reviewed the interconnections of the CFS J
functions.
The interconnections are designed so that between each unit.
the capability to mitigate the consequences of an accident in 4
l either unit and achieve safe shutdown in that unit is retained without reducing the capability of the other unit to achieve i'
l The applicant has met the requirements of General Design Criterion 44 with respect to cooling water by providing a l
4.
redundant and isolable system capable of transferring heat loads 1
4 Rev. 2 - July 1981 10.4.7-5 I... -
-. --, ~_ _ ; %. _ -.. _ _ _ _ -,
from the reactor system to a heat sink under both normal opera-ting and accident conditions.
The applicant has demonstrated that the condensate and feedwater system can provide sufficient cooling water to transfer the heat load of the reactor system under normal operating conditions and accident conditions assuming loss of.offsite power and a single failure and that portions of the system can be isolated so that the safety function of the system will not be compromised.
5.
The applicant has met the requirements of General Design Criterion 45 with respect to inspection of cooling water systems by providing a feedwater system design that permits inservice inspection of safety-related components and equipment.
6.
The applicant has met the requirements of General Design Criterion 45 with respect to testing of cooling water systems by providing a feedwater system design that permits operational fcactional testing of the safety related portion of the system and its components.
The staff concludes that the design of the CFS conforms to all applicable GDCs and positions of the regulatory guide cited and is, therefore, acceptable.
V.
IMPt.EMENTATION The following is intended to provide guidance to all applicants and licensees regarding the NRC staff's plans for using this SRP section.
Except in those cases in which the applicant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, the method described herein will be used by the staff in its evaluation of conformance with Commission regulations.
Implementation schedules for conformance to parts of the method discussed herein are contained in the referenced regulatory guide.
VI.
REFERENCES 1.
10 CFR Part 50, Appendix A, General Design Criterion 2, " Design Bases for Protection Against Natural Phenomena."
2.
10 CFR Part 50, Appendix A, General Design Criterion 5, " Sharing of Structures, Systems, and Components."
3.
10 CFR Part 50, Appendix A, General Design Criterion 44, " Cooling Water."
4.
10 CFR Part 50, Appendix A, General Design Criterion 45, " Inspection of Cooling Water System."
5.
10 CFR.Part 50, Appendix A, General Design Criterion 46, " Testing of Cooling Water System."
6.
Regulatory Guide 1.29, " Seismic Design Classification."
l h
7.
Branch Technical Position ASB 10-2, " Design Guidelines for Water Hammer in Steam Generators with Top Feedring Designs."
10.4.7-6 Rev. 2 - July 1981
t BRANCH TECHNICAL POSITION ASB 10-2 DESIGN GU10ELINES FOR WATER HAMMERS IN l
STEAM GENERATORS WITH TOP FEEDRING OESIGNS BACKGROUND Experience over the past decade or so has shown that top-feed steam generators containing feedwater spargers with bottom drain holes incur steam condensation
'nduced water hammers. This type of water hammer has frequently occurred after the feedwater sparger was uncovered (due to some plant transient) and cold auxiliary feedwater flow was subsequently initiated. The initiation of the auxiliary feedwater flow into the steam generator produces a water slug in the sparger or feedwater piping, which is then accelerated by the unbalanced pres-The resultant sures produced by the condensation of a steam pocket in the line.
impulse could be of a sufficient magnitude to cause damage to the steam generator or associated piping. An increasing number of these steam generator water hammer events has been occurring, in some cases resulting in negligible effects, and in others resulting in damage to piping support systems.
Beginning with the Yankee Rowe Nuclear Plant in 1966, more than 25 steam generator water hammers have been reported to the NRC.
These water hammers or flow instabi-lities have involved 15 of the 34 presently operating PWRs, and resulted in four of these plants needing extensive repairs because of permanent pipe deforma-tion and component damage.
In 1973, the most damaging of these water hammer C
incidents occurred at Indian Point 2, resulting in the fracture and bulging of an 18-inch feedwater pipe and thermal deformation of the containment.
In view of the frequency and potential severity of these flow instability events, various b
hardware modifications and operating procedures have been adopted by most U.S.
and foreign utilities. These ad hoc approaches by U.S. industry were based almost entirely on qualitative descriptions of the phenomena, and accordingly a successful means to eliminate water hammer did not emerge.
In 1976, the NRC engaged an engineering firm (CREARE, Inc.) to undertake its own study to determine the cause and effects of past steam generator water ham-The outcome of the program provided information which led to the develop-mers.
ment of procedural and design changes to avoid these incidents.
CREARE utilized j
the results of their review of the available data from past incidents in the
)
development and performance of its own test program which was to perform small l
scale tests (1/10) to qualitatively investigate the phenomena associated with slug formation and acceleration.
From the test results obtained, CREARE was able to show that feedwater type spargers located within the steam generators with top feed and combined with a short length of pipe significantly reduced the overpressure magnitude (factor 5 to 10) vis a vis bottom drain spargers and long pipes.
Thus, these results generally confirmed the results of the European tests to install J-Tubes in top-feed steam generators.
Large-scale testing has so far confirmed the efficacy of reducing water hammers by providing top discharge frem the sparger and short pipe arrangement, as shown by testing at Trojan, Indian Point 2, Farley and St. Lucie in the U.S., and at Doel, Tihange 1, Ringhals 2 and Fessenheim 1 in Europe.
A detailed discussion of most of the above work is published in NUREG-0291, entitled, "An Evaluation of PWR Steam Generator Water Hammer."
10.4.7-7 Rev. 2 - July 1981
% Q ^'
,The staff believes that the evidence to date warrants the establishment of design guidelines for steam generators and the associated piping.
Guidelines have been developed that may be used to reduce the probability of a damaging steam condensation induced water hammer, especially in the Westinghouse and Combustion Engineering PWR designs which use top-feed steam generators.
The B&W once-through steam generators have not experienced damaging water hammer incidents, and therefore are not considered.
BRANCH TECHNICAL POSITION Top-Feed Desians In CP and OL application reviews, the staff requires the applicant to provide the following for steam generators utilizing top feed:
To eliminate or reduce possible water hammer in the feedwater system:
1.
Prevent or delay water draining from the feedring following a drop in steam generator water level by means such as J-Tubes.
2.
Minimize the volume of feedwater piping external to the steam generator which could pocket steam using the shortest possible (less than seven feet) horizontal run of inlet piping to the steam generator feedring.
3.
Perform tests acceptable to NRC to verify that unacceptable feedwater hammer will not occur using the plant operating procedures for normal and emergency restoration of steam generator water level following loss of normal feedwater and possible draining of the feedring.
Provide the procedures for these tests for approval before conducting the tests.
1 i
?
L 10.4.7-8 Rev. 2 - July 1981
,ev n w w..
(Fcrm:rly NUREG 75/0871 p** **ossq\\
U.S. NUCLEAR REGULATORY COMMISSION y
(Pp#) STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION
\\e v
,o eee l
s 10.4.9 AUXILIARY FEEDWATER SYSTEM (PWR)
REVIEW RESPONSIBILITIES Primary - Auxiliary Systems Branch (ASB)
Secondary - None I.
AREAS OF REVIEW The auxiliary feedwater system (AFWS) normally operates during startup, hot standby and shutdown as the feedwater system for pressurized water reactor (PWR) plants.
In conjunction with a seismic Category I water source, it also functions as an emergency system for the removal of heat from the primary system when the main feedwater system is not available for emergency conditions including small LOCA cases. The AFWS operates over a time period sufficient either to hold the plant at hot standby for several hours or to cool down the primary system, at a rate not to exceed limits specified in technical specifications, to temperature and pressure levels at which the low pressure decay heat removal system can operate.
The design of the AFWS should meet the requirements of General Design Criteria 2, 4, 5, 19, 34, 44, C.
45 and 46.
The ASB reviews the AFWS from the condensate storage tank (normal operation),
or the seismic Category I water supply including valving and cross-connects (emergency operation), to the connections with the steam generators, which are made either through a connection to the main feedwater piping or through separate auxiliary feedwater piping connected directly to the steam generators.
All inter-connections and cross-connections are included in the review.
The review also includes AFWS components, e.g., pumps, valves, and piping, with respect to their functional performance as affected by adverse environ-mental occurrences, abnormal operational requirements, and off-normal conditions, e.g., small breaks in the primary system or the loss of offsite power.
The system is reviewed to determine that a single malfunction, a failure of a component, or the loss of a cooling source does not reduce the safety-relat.d.
Rev. 2 - July 1981 USNRC STANDARD REVIEW PLAN starderd review plans are prepared for the guldence of the Office of Nuclear Reactor Regulation staff resp applications to construct and operate nuctear power plants. These documents are made eveilable to the p Commission's pokcy to inform the nuclear industry and the general pubhc of regulatory procedures and pohcies.
plans are not substitutes for regulatory guides or the Commession's regulations and comphence with th standard review plan sections are keyed to the standard Format and Content of Safety Analysis Reports f o Not ett sections of the Standard Format have a corresponding review plan.
Pubhshed standard review plans witi be revised periodically, as appropriate, to accommodate comrnents an tion and esperience.
Comments and suggestions for improvement will be considered and should be sent to the U.s Nucteer Reg office of Nuclear Reactor Regulation, Washington, D C. 20565.
-. e -
functional performance cap"bilities of the system.
The ASB review assures that:
1.
System components and piping have sufficient physical separation or shielding to protect the essential portions of the system from the effects of internally and externally generated missiles.
This review is performed according to SRP Section 3.5.1.1 for internally generated missiles and Sections 3.5.1.4 and 3.5.2 for externally generated missiles.
2 The system is protected against the effects of pipe whip and jet impinge-ment that may result from high or moderate energy piping breaks or cracks This review is performed according to SRP Section 3.6.1.
3.
The failure of non essential equipment or components does not affect essential functions of the system.
4.
The system is capable of withstanding a single active failure.
5.
The system possesses diversity in motive power sources such that system performance requirements may be met with either of the assigned power sources, e.g., a system with an a-c subsystem and a redundant steam /d-c subsystem.
6.
The system design precludes the occurrence of fluid flow instabilities, i
e.g., water hammer, in system inlet piping during normal plant operation or during upset or accident conditions (see SRP Section 10.4.7).
7.
high water levels (adequate flood protection considering j
maximum flood). This review is performed according to SRP Section 3.4.1.
4 8.
The capability exists to detect, collect, and control system leakage and to isolate portions of the system in case of excessive leakage or component malfunctions.
9.
Provisions are made for operational testing.
10.
Instrumentation and control features are provided to verify the system is operating in a correct mode.
11.
The system is capable of automatically initiating auxiliary feedwater t
flow upon receipt of a system actuation signal.
12.
The system satisfies the recommendations of Regulatory Guide 1.62 with respect to the system capability to manually initiate protective action by the auxiliary feedwater system.
i 13.
The system design possesses the capability to automatically terminate auxiliary feedwater flow to a depressurized steam generator, and to automatically provide feedwater to the intact steam generator.
i Or as an alternative if it is shown that the intact steam generator wi'11 recieve the minimum required flow without isolation of the depressurized steam generator and containment design pressure is not exceeded, then operator action may be relied upon to isolate the depressurized steam generator, b,
10.4.9-2 Rev. 2- - July 1981 r em.
The syste:t possesses sufficient auxiliary feedwater flow capacity so,tha i-Upon request from AS8, the Reactor-14.
a cold shutdown can be achieved.
z' Systems Branch-(RSB).will verify that the system meets the minimum
-requirements for decay heat removal.
I The applicant's proposed technical specifications are such as to assure the continued reif ability of the AFWS during plant operation; i.e., the 15.
limiting conditions for' operation and the surveillance testing requirem l
are specified and are consistent with the Standard Technical Spec
.the ASB verifies that the system design meets t long l
16.
[
term recommendations identified in NUREGS-0611 and -0635 recommendations will apply to all PWRs.
An AFWS reliability analysis is performed in accordance with Item II of NUREG-0737 using the methodology defined by Appendix III and A I
17.
of Appendix X in NUREG-0611 and NUREG-0635 to determine lity and major contributors to AFW system failure under v l-main feedwater transients.that the requirements and guidance of II.E i
l-The reviewer verifies that the-system design has the capability.to pe I
operation at hot shutdown for at-1 east four hours followed by co f
' 18.
the RHR cut-in temperature from the control room using only safety gr 4
equipment and assuming the worst case single active failure in ac with Branch Technical Postion RSB 5-1.
l Coordinated reviews are performed by other branches and the results u the ASB to complete the overall evaluation of the system..
reviews are as follows. essential components of the reactor coolant or em and the AFWS that are required for operation during norma i
t
)
accident conditions.
The associated time intervals available for cooling various components.
Structural Engineering Branch (SE8) determines the acceptability of th l
analyses, procedures, and criteria used to establist the ability of seis 1
Category I structures housing the system and supporting systems to i
the effects of natural phenomena such as the s i
l through 3.3.1,3.3.2,3.5.3,(3.7.1 review responsibility for SRP Sections d
MEB) detere nes The Mechanical Engineering Branch j
3.7.4, 3.8.4 and 3.8.5.
that the components piping and structures are d f'
The MEB also determines the accept-for SRP Sections 3.9.1 through 3.9.3.
abilityoftheseismicandqualitygroupclassIficatIonsforsystemcomponent as part of its primary review responsibility for SRP Sections 3.2.1 The MEB also reviews the adequacy of the inservic l
The Materials Engineering Branch (MTEB) verifies that inservice inspec J
requirements are met for system components as part of its p and The review i
responsibility for SRP Section 6.6lityofthematerialsofcons i
for Fire Protection, Technical Specifications coordinatedandperformedbytheChemicalEngIneeringBranch, Licensing'ie J
i Guidance Branch and Quality Assurance Branch as part of their primary rev li Rev. 2- - July 1981 10.4.9-3 l
l i
d responsibility for SRP Sections 9.5.1, 16.0 and 17.0, respectively.
The Equipment Qualification Branch (EQB) reviews the seismic qualification of i
Category I instrumentation and electrical equipatnt and the environmental qualification of mechanical and electrical equipment as part of its primary review responsibility for SRP Sections 3.10 and 3.11, respectively.
The ICSB and Power System Branch (PSB) evaluate system controls, instrumentation, and i
power sources with respect to capability, capacity, and reliability during nor-mal and emergency conditions as part of their primary review responsibility i
L for SRP Sections 7.1 and 7.3 through 7.5 for ICSB and Section 8.3 for PSB.
l For those areas of review identified above as being reviewed as part of the primary responsibility of other branches, the acceptance criteria and their methods of application are contained in the SRP sections corresponding to those branches.
5 II.
ACCEPTANCE CRITERIA-Acceptability'of the design of the auxiliary feedwater system, as described in i
the applicant s safety analysis report (SAR), is based on specific general design criteria and regulatory guides.
Listed below are the specific criteria used in this SRP section as they relate to the AFWS.
1.
General Design Criterion 2, as related to structures housing the system and the system itself being capable of withstanding the effects of earth-i' quakes. Acceptability is based on meeting position C.1 of Regulatory Guide 1.29 for safety-related portions and position C.2 for nonsafety-related portions.
1 2.
General Design Criterion 4, with respect to structures housing the system i
and the system itself being capable of withstanding the effects of external i
missilesandinternallygeneratedmissiles,pipewhip,andjetimpingement g
forces associated with pipe breaks.
The basis for acceptance for meeting-
}
this criterion is set forth in the SRP Section 3.5 and 3.6 series.
I 1
3.
General Design Criterion 5, as related to the capability of shared sys-I tems and components important to safety to perform required safety functions.
l 4.
General Design Criterion 19, as related to the design capability of system instrumentation and controls for prompt hot shutdown of the reactor and potential capability for subsequent cold shutdown. Acceptance is based on meeting Branch Technical Position RSB 5-1 with regards to cold shutdown from the control room using only safety grade equipment.
I 4
n i
5.
General Design Criteria 34 and 44, to assure:
l 1
l The capability to transfer heat loads from the reactor system to a a.
j heat sink under both normal operating and accident conditions.
i b.
Redundancy of components so that under accident conditions the i
safety function can be performed assuming a single active component failure.
certaineve(Thismaybecoincidentwiththelossofoffsitepowerfor 1
j nts.) Branch Technical Position ASB 10-1 as it relates i
{
to AFW pump drive and power supply diversity shall be used in meeting these criteria, b
c.
The capability to isolate components subsystems, or piping if required t
sothatthesystemsafetyfunctionwllibemaintained, t
10.4.9-4 Rev. 2- - July 1981 4'
'______-_..__.__*r.__,_.
_,.J
~. -
I
.in meeting these criteria, the recommendations of NUREG-0611 and 0635-shall also be met.
An acceptable AFWS should have an unreliability in the range-of 10 4 to 10 5.per demand based on an analysis using methods and data presented in NUREG-0611 and NUREG-0635. Compensating factors
.such as other methods of accomplishing the safety functions'of the AFWS or:other reliable methods for cooling the reactor core during abnormal
- conditions may be considered to justify a larger unavailability of the i
i i
AFWS.
General Design Criterion 45, as related to design provisions made to 6.
. permit periodic inservice inspection of system components.and equipment.'
General Design Criterion 46, as related to design provisions made to 7.
permit appropriate functional testing of the system and components to i
assure structural integrity and leak-tightness, operability and perform-ance of active components, and capability of.the integrated system.to l
function as intended during normal, shutdown, and accident conditions, j
In meeting this criteria the technical specifications should specify that.
the monthly AFWS pum) test shall be performed on a staggered test basis to reduce the likeliNood of leaving more than one pump in a-test mode 4
following the tests.
III. REVIEW PROCEDURES The procedures below are used during the construction permit (CP) review to l
determine that the design criteria and bases and the preliminary design as set j
forth in the preliminary safety analysis report meet the acceptance criteria For operating license (OL) appitcations, the procedures i
given in subsection'II.
are utilized to verify that the initial design criteria and bases have been-(
appropriately implemented in the final design as set forth in the final safety analysis report.
The procedures for OL applications also include a determina-1 I
tion that the content and intent of the technical specifications prepared by.
the applicant are in agreement with the requirements for system testing minimum performance and surveillance developed as a result of the staff,s i
review.
Upon request from the primary reviewer,-the coordinating review branches will t
provide input for the areas of review stated in subsection I.
The primary l
reviewer obtains and uses such input as required to assure that this review j
procedure is complete.
For the purpose of this SRP section, a typical system is assumed which has redundant auxiliary feedwater trains, with a 50% capacity motor-driven pump in j
each train feeding directly to the steam generators, and a 100% capacity steam turbine-driven pump able to supply either of the redundant trains.
The pumping l
capacity should permit the system to hold the plant at hot standby and subse-quently to cool down the reactor at specified cooldown rates.
The 50% capacity i
pump is assumed to have sufficient capacity for decay heat removal following any accident or transient although cooldown to RHR cut in temperature may take longer than design. This requirement should also be met for conditions involving a small break area loss-of-coolant accident (LOCA) or a pipe break outside containment.
For cases where there are variations from the typical arrangement, therevieweradjuststhereviewprocedurestosuitthedesign. However, the system design is required to meet the acceptance criteria given in subsection II.
i i
10.4.9-5 Rev. 2- - July 1981 i
i
.----~.._.._._.-..__.--_..,.---,___,.....,-.._.,_---,___,_.~._.,_._m..__
~ _
1.
The SAR is reviewed to determine that the system description and piping-
'I and instrumentation diagrams (P& ids). identify the AFWS equipment and
-. arrangement that is used for. normal operation and for safe plant shutdown 1
(essential) operation.
The system P& ids, layout drawings, and component-j descriptions and characteristics are then reviewed to verify that:
1 c.
a.
Minimum performance requirements.for the system are sufficient for the various functions of the AFWS.
b.
Essential portions of the AFWS are isolable from non-essential. portions, i.
so that system performance is not impaired in the event of a failure of a non essential component.'
4
~.
. Component and system descriptions in the SAR include appropriate c
seismic and quality group classifications, and the P& ids indicate-any points of change in piping quality group classification. The review for seismic design is performed by the SE8 and the review for a
i seismic and quality group classification is performed by the MEB as i-indicated in Subsection I of this SRP section..
1 d.
Design provisions have been made that permit appropriate inservice i
inspection and functional testing of system components important.to I
safety.
It is acceptable if the.SAR information delineates a testing j
and inspection program if the system drawings show the necessary recirculation loops around pumps or isolation valves as may be i
required by this program, c
2.
The reviewer verifies that the system safety function will be maintained j
as required, in the event of adverse environmental phenomena, breaks or cracks in fluid system piping outside containment, system component i
i failures, loss of an onsite motive power source, or loss of offsite i
power. The reviewer uses engineering judgment and the results of failure modes and effects analyses to determine that:
The failure of portions of the system or of other systems not designed a.
I to seismic Category I standards and located close to essential portions of the system, or of nonseismic Category I. structures that house, support, or are close to essential portions of the AFWS, will -
not preclude operation of the essential portions of the AFWS.
Refe,rence to SAR sections describing site features and the general 1
arrangement and layout drawings will be necessary, as well as the i
SAR tabulation of seismic design classifications for structures and i
systems.
b.
The essential p "tions of the AFWS are protected from the effects of floods, hurricanes, tornadoes, and internally or externally generated t
i missiles.
Flood protection and missile protection criteria are i
discussed and evaluated in detail under the SRP Section 3 series.
1 The location and design of the system, structures, and pump rooms i
(cubicles) are reviewed to determine that the degree of protection provided is adequate. A statement to the effect that the system is located in a seismic Category I structure that is tornado missile and flood protected}c Category I cubicles or rooms that will withstand (:
or the components of the system will be located in individual seism the effects of both flooding and missiles is acceptable.
i, l
l 10.4.9-6 Rev. 2- - July 1981 i
l'
- The essential portions of the system are protected froa the effects 4
of high and moderate energy line breaks.
c.
to assure that no high or moderate energy piping systems are close to essential portions of the AFWS, or that protection from theThe mea effects of failure will be provided.
protection will generally be given in Section 3.6 3.6.1..
Essential components and subsystems necessary for safe shutdown can The SAR function as required in the event of loss of offsite power.
d.
is reviewed to see that for each AFWS component or subsystem af by the loss of offsite power, system flow and heat transfer cap Statements in the SAR and the results of meet minimum requirements. failure modes and effects analyses are the system meets these requirements.
The system is designed with adequate redundancy to accommodate a This single active component failure without loss of funct e.
tank (or other primary source) to the AFW pump suctions.
Diversity in pump motive power sources and essential instrumenta j
The diverse system I
and control power sources has been provided.
f.
including pump (s), controls and valves should be independent of offsite and onsite AC power sources in accordance with the guidelin of Branch Technical Position ASB 10-1.
The system is designed with adequate instrumentation to automa l
initiate auxiliary feedwater flow to the steam generators uponThe init g.
receipt of an actuation signal.
4 all auxiliary feedwater pumps and supporting sys l
The system is also feedwater pumps to the steam generator ( 0,
designed with the capability to manually initiate the necessa orifice rather than instrumentation) or terminat protective actions.
flow to a depressurized steam generator, and to assure that the minimum required flow is directed to the intact s If a flow limiter is used then in subsection I of this SRP section.
it must be demonstrated that sufficient flow still g by the AFW flow to the depressurized generator.
The AFWS is designed with sufficient flow capacity so that the system can remove residual heat over the entire range of reactor j.
ti
. operation and cool the plant to the decay heat removal system cu This review is performed by RSB upon request as temperature.
indicated in subsection I of this SRP section.
The reviewer verifies that the design has features to meet the generic For additional short term recom-4 3.
recommendations of NUREG-0611 and 0635.
mendation No. 2 regarding AFW pump endurance tests, a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> te Rev. 2- - July 1981 1
10.4.9-7 i
c
-m
-....-_,_.__-..___.._y
i l
acceptable rather than the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. test specified in the NUREGS.
The ASB 4
reviewer coordinates with the ICSB reviewer to assure that the instrumen tation and control system aspects of these recommendations are met by the i
system design, r
t 4.
The reviewer verifies that an AFWS reliability evaluation has been per-formed in accordance with item II.E.1.1 of NUREG-0737.
i' analysis is reviewed to determine the potential. for AFW system failureThe reliabi under various loss of main feedwater transients, I
i
{
- IV.
EVALUATION FINDINGS i
The reviewer verifies that sufficient information has been provided and his j
review supports conclusions of the following type, to be -included in the staff's. safety evaluation report:
The auxiliary feedwater system' includes all components and equipment from the condensate storage tank (normal operation) or the seismic Category I t
emergency water supply (including valves and cross connections) to the i
connection with the steam generators.
The AFWS is designed to seismic l
Category I requirements since system operation is necessary to mitigate the consequences of an accident.
Category I, tornado protected si This includes an automatic, seismic i
i Based on the review of the appli, ply'of water to the AFW pump suction.
cant s proposed design criteria, design j
bases and safety classification for the auxiliary feedwater system, and t
system performance requirements during normal abnormal, and accident j
conditions, the staff concludes that the desig,n of the auxiliary feed-t water system and supporting systems is acceptable and meets the Commission's
[
regulations as set forth in General Design Criteria 2, 4, 5, 19, 34, 44, 45, and 46.
j This conclusion is based on the following:
1.
The AFW system design meets the requirements of General Desi Criterion 2withrespecttoprotectionagainsttheeffectsofn l
earthquakes since the safety related portions are designed to seismic Category I requirements in accordance with position C.1 of Regulatory Guide 1.29 and the nonsafety-related portions are designed in accordance with position C.2 of Regulatory Guide 1.29.
2.
The AFW system design meets the requirements of General-Design l
Criterion 4 with res breaks and missiles.pect to protection against the effects of pipe-Acceptance was based on locating the AFW system pumps and trains in individual cubicles which separate redundant components and are prutected against the effects of tornado missiles. ' Refer to the Chapter 3 sections of this report
]
for a description of how this protection is accomplished.
i i'
i 3.
The AFW system is designed in accordance with the requirements of General Design Criterion 5 with respect to sharing of structures systems and components.
component including a pipe break and single active failure will no j
i i
prevent the safe shutdown and cooldown of either unit (together or i
singularly).
4.
The system design meets the re ufrements of General Design Criterion 19 b
j as related to the design capab 11ty of system instrumentation and i
I
}
10.4.9-8 i
Rev. 2- - July 1981
,.rewp-n-m,---=+,--
s,-4
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memm--,-n.-nnwem-r-mm.,,..,
w^
controls for prompt hot shutdown of the reactor and potential capabil-ity for subsequent cold shutdown since the design meets the requirement' of Branch Technical Position RS8 5-1 which requires the capability to bring primary plant temperature to the RHR cut-in point following four hours at hot standby from the control room using only safety grade equipment and assuming any single active failure.
5.
The system' design meets the requirements of Genral Design Criteria 34 and 44 since it has the capability to transfer heat loads, including decay heat from the reactor, during normal operating and accident conditions assuming any single active failure.
The system has_ suit-able redundancy such that-it can withstand a pipe break and single active failure and still perform its_ safety function.
The system
- deign also has sufficient diversity such that it meets the require--
ments of Branch Technical Position AS8 10-1.
In meeting these General Design Criteria the applicant has also met the generic
- recommendations identified in NUREGS-0611 and -0635 and has performed
- a reliability analysis in accordance with NUREG-0737, ites II.E.1.1.
The results of the reliability analyses were acceptable since it was shown that the AFWS has an unreliability in the range of-10 4 to 10 5 per demand.
6.
The pumps, valves, heat exchangers and piping of the system, to the extent practicable are designed and located to facilitate periodic inspectionasrequIredbyGeneralDesignCriterion45.
This is accomplished by providing adequate accessability to conduct the required examinations.
7.
To meet the requirements of General Design Criterion 46, the auxiliary feedwater system is designed to include the capability for testing through the full operational sequence that brings the system into
+
N operation for reactor shutdown and for loss-of-coolant accidents, including operation of applicable portions of the protection system and the transfer between normal and emergency buses.
V.
IMPLEMENTATION The following is intended to provide guidance to applicants and licensees
~
regarding the NRC staff's plans for using this SRP section.
Except in those cases in which the applicant proposes an acceptable alterna-tive method for complying with specified portions of the Commission's regula-tions, the method described herein will be used by the staff in its evaluation of conformance with Commission regulations.
Implementation schedules for conformance to part of the method discussed herein are contained in the referenced regulatory-guides and NUREGs.
VI.
REFERENCES 1.
10 CFR Part 50, Appendix A, General Design Criterion 2 " Design Bases for Protection Against Natural Phenomena."
2.
10 CFR Part 50, Appendix A, General Design Criterion 4 " Environmental and Missile Design Bases.
10.4.9-9 Rev. 2- - July 1981 sep
3'.
10 CFR Part 50, Appendix A, General Design Criterion 5, " Sharing of Structures, Systems, and Components."
4.
10 CFR Part 50, Appendix A, General Design Criterion 19, " Control Room."
5.
10 CFR Part 50, Appendix A, General Design Criterion 34, " Decay Heat Removal."
6.
10 CFR Part 50, Appendix A, General Design Criterion 44, " Cooling Water."
7.
10 CFR Part 50, Appendix A, General Design Criterion 45, " Inspection of Cooling Water System."
8.
10 CFR Part 50, Appendix A, General Design Criterion 46, " Testing of Cooling Water System."
1 9.
Regulatory Guide 1.29, " Seismic Design Classification."
10.
Branch Technical Position RSB 5-1, " Design Requirements of the Residual Heat Removal System," attached to SRP Section 5.4.7.
11.
Branch Technical Position ASB 10-1, " Design Guidelines for Auxiliary Feedwater System Pump Drive and Power Supply Diversit Water Reactor Plants," attached to this SRP section. y for Pressurized
- 12. N'JREG-0611 " Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse - Designed Operating Plants,"
January 1980.
13.
NUREG-0635 " Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Combustion Engineering - Designed Operating Plants," January 1980.
14.
NUREG-0737 " Clarification of THI Action Plan Requirements," November 1980.
k 10.4.9-10 Rev. 2- - July 1981
P l
l
+
f-BRANCH TECHNICAL ~ POSITION ASB 10-1 DESIGN GUIDELINES FOR AUXILIARY FEEDWATER SYSTEM PUMP DRIVE AND i
POWER SUPPLY DIVERSITY FOR PRESSURIZED WATER -
t REACTOR PLANTS l
4 A.
8ACKGROUND Heatremovalfrompressurizedwaterreactorbantsfollowingreactortripand l
I a loss of. offsite power is accomplished by t operation of several systems Similar capability L
including the secondary system via the steam relief system.
is required to mitigate the consequences of certain postulated piping breaks.
Such heat removal involves heat transfer from the reactor to the steam generators, i
resulting in the production of. steam which is then released to the atmosphere.
t In this process it becomes necessary to supply makeup water to the steam This is accomplished by the use of an auxiliary feedwater system, generators.
which generally consists of redundant components that are powered by both' i.
electrical and steam-driven sources, I
The auxiliary feedwater system functions as an engineered safety system because i
it is the only source of. makeup water to the steam generators for decay heat i
(
removal when the main feedwater system becomes inoperable.
It must, therefore, using the principles of redundancy and l
\\
'be designed to operate when needed diversityinordertoassurethatItcanfunctionunderpostulatedaccident l
Themajorityofcurrentsystemsarepoweredbyelectricalor 1
I conditions.
steam-driven sources. Operating experience demonstrates that each type of motivepowercanbesubjecttoafailureofthedrivingcomponentitself,its i
The effects of such source of energy, or the associated control. system.
failures can be minimized by the utilization of diverse systems that include l
energy sources of at least two different and distinct types.
j The provision of several independent flow paths for the auxiliary feedwater system serves to preclude the possibility.of a complete loss of function due l
to a single event, either occurring alone, or in conjunction with the failure The auxiliary feedwater system is categorized as a i
of an active component.
high energy system because either that section of Ifne which connects to the l
mainfeedwaterpipIngorthesteamgeneratorispressurizedduringplant l
operation or else the entire system is pressurized when in use during startup, l
1
?
hot standby, and shutdown.
i The staff believes that it is necessary to establish design guidelines for the i
i and in this regard has developed guidelines that auxiliary feedwater system,inimes diversity acceptable for auxiliary feedwater i
may be used to select the m system pump drives and power supplies.
I 10.4.9-11 Rev. 2- - July 1981 l-i
~~,---:~=,,_
B., BRANCH TECHNICAL POSITION i
1.
The auxiliary feedwater system should consist of at least two full-capacity, independent systems that include diverse power sources.
2.
Other powered components of the auxiliary feedwater system should also i
use the concept of separate and multiple sources of motive energy. An example of the required diversity would be two separate auxiliary feedwater trains, each capable of removing the afterheat load of the reactor system, having one separate train powered from either of two a-c sources and the other train wholly powered by steam and d-c electric power.
3.
The piping arrangement, both intake and discharge, for each train should be designed to permit the pumps to sumply feedwater to any combination of steam generators.
This arrangement sWuld take into account pipe failure, active component failure, power supply failure, or control system failure that could prevent system function.
One arrangement that would be accept-able is crossover piping containing valves that can be operated by remote manual control from the control room, using the power diversity principle for the valve operators and actuation systems, i
4.
The auxiliary feedwater system should be designed with suitable redundancy I
to offset the consequences of any single active component failure; however, 1
each train need not contain redundant active components.
I 5.
When considering a high energy line break, the system should be so arranged i
I as to assure the capability to supply necessary emergency feedwater to
)
the steam generators, despite the postulated rupture of any high energy section of the system, assuming a concurrent single active failure.
C.
REFERENCES i
None
)
i l
L 1
I kn :
i, k
10.4.9-12 Rev. 2- - July 1981
.,,_.m.-.
_..m..,,
U.S. NUCLEAR REGULATORY COMMISSION
""*'"3
+
February 1976 REGULATORYGU DE OFFICE OF STANDARDS DEVELOPMENT REGULATORY GulDE 1.29 3
OUALITY GROUP CLASSlFICATIONS AND STANDARDS FOR WATER., STEAM., AND R ADIOACTIVE. WASTE.CONTAINING COMPONENTS OF NUCLEAR POWER PLANTS A. INTRODUCTION quahey groups, A thro s for assignmg components to these oups d the specific General Design Critenon 1, "Quahty Standards and quahiy :andards app h quahty group. The Records," of Appendix A. " General Design Crttena for mitial portien of t s
nbed in 650.55: of Nuclear Power Plants," to 10 CFR Part 50,"lJcensmg 10 CFR Part 5 r
s that components of the of Production and Utahzation Facdities, requires that reactor coo i re oundary be designed, fabn.
structures, systems, and components important io safety cated, ere a
d to the highest available be designed, fabricated, erected, and tested to quahty national a s corresponds to the quauty standards comrnensurate with the importance of the stan safety funcuons to be performed. Section 50.55a, or quahty Group A of the NRC
" Codes and Standards," of 10 CFR Part 50 reqmres that e desenbes a method for determining s te components of the reactor coolant pressure boundary be ty standards for the remammg safety.
u co designed, fabricated, erected, and tested m accordance ponents contaming radioactive materis'.
steam, i.e., quahty Group B, C, and 0 with the requirements for Class l' components of ments. Other systems not co ered by this guide, Section 111 of the ASME Boiler and Pressure Venel i as instrument and service air, diesel engme and its b-or equivalent quahty standards. This guide desc nerators and auxihary support systems, diesel fuel, es a quably classification system related to specified al emergency and normal ventilauon, fuel handhng, and standards that may be used to deter standards acceptable to the NRC staff radioactive waste rnanagement systems,8 should be l' q
r sati ing designed, fabricated, erected, and tested to quahty General Design Cnterion i for other ty.
ted standards commensurate with the safety function to be components contaimng water, steam, o ettve perforrned. Evaluation to estabhsh the quahty group matenal m hght. water <ooled nuclear power plants.
classification of these other systems should include consideration of the guidance provided in regulatory positions C.1 and C.2 of thn guide.
C. REGULATORY POSITION After reviewing er appbcations for con.
struction permits da tmg hcenses and after dis.
- 1. The Group B quahty standards givenin Table i of cuisions withfep of professional societies this gwde should be apphed to water. and steam.
and indust t
ff has developed a quality contairung pressure vessels, heat exchangers (other than classificatio for safety.related components con.
turbmes and condensers), storage tanks, pipmg, pumps,
taming water,
, or radioactive material in water.
cooled nuclear po er plants The system consists of four and valves that are either part of th teactor coolant pressure boundary defined in l50.2(v) but excluded t
' t ditions priot to 1971 of the A5MI. Botier and Pressure Vesiet
' specific guidance on the quabt, group classification of radio. t I
Code, Section lit. " Nuclear Power Plant Components." une the term Ctast A in b u or Clsto t.
acuve waste management systems is under development l
- Lanet Indacate substanuve chaneen ftom previous astue.
U5NRC RE OUL AiOR V GUi0sf, t.=a.m..a.s.
=,*. na,s t.
.a
- ,.;;f;;;,c;--~
'u~ra a e = "~ *~u a. e
'. ~,*::.'.' *.'.:::..,":.*:...........................
- ~e-*
.............,,,~..e lll.*;::."::'.'.':.',':. '".*.'"':
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...................................e....
...a.....cu.................
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........~.......................................-.......
.-.~....*.~a-....
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l' rom the' requirements of $ 50.55a pursuant to foot-radioacuve waste contairung pressure vessels. heat ex-8 note 2 of that section or not part of the reactor coolant changers (other than turbines and condensers), storage pressure boundary but part of:
tanks, piping, pumps, and valves not part of the reactor coolant pressure boundary or included in quahty Group
- a. Systems or portions of systems
- important to B but part of:
safety that are designed for (1) emergency core cochng.,
(2) postaccident containment heat removal, or (3) post-
- a. Coobng water and auxthary feedwater systenu accident fission product removal-or portions of'these systerns* important to safety that are designed for (l) emergency core coohng. (2) post.
- b. Systems or portions of systems
- important to accident contamment heat removal. (3) postaccident safety that are designed for (1) reactor shutdown or containment atmosphere cleanup, or (4) residual heat (2) rendual heat removal.
removal from the reactor and from the spent fuel storage pool (mcludmg ptsmary and secondary coohng systems).
c, Those portions of the steam systems of boihng Poruons of these systerm that are required for their water reactors exten&ng from the outermost contam.
safety functions and that (1)do not operate during any ment isolauon valve up to but notincludmg the turbme mode of normal reactor operation and (2)cannot be stop and bypens valves' and connected piping up to and tested adequately should be classified as Group B.
including the first valve that is either normally closed or scapable of automatic closure during all modes of normal
- b. Coohng water and seal water systems or por-reactor operation. Alternatively, for boihng water re.
tions of these systems
- important to safety that are actors containing a shutoff valve (in addition to the two designed for functioning of components and systems containment isolation valves) in the main steam une and tmportant to safety, such as reactor coolant pumps, in the main feedwater kne, Group B quahty standards aesels, and controt room.
should be apphed to those portions of the steam and feedwater systems extending from the outermost con-
- c. Systems or portions of systems
- that are tainment isolauon valves up to and includmg the shutoff connected to the reactor coolant pressure boundary and valve or the first valve that is either normally closed or are capable of bemg isolated from that boundary during capable of automatic closure during all modes of normal all modes of normal reactor operauon by two va!"
reactor operanon-each of which is either normally closed or capable of s I automaus closure.'
i
- d. Those poruons of the steam and feedwater
/ /
systems of pressunzed water reactors extending from
- d. Systems, other than radioactive waste manage-l and inclueng the secondary side of steam generators up ment systems,8 not covered by items 2.a. through 2.c.
to and inclueng the outermost contamrnent isolation above that contain or may contain ra&oactive material valves and connected pipmg up to andincluding the first and whose postulated fadure would result in conserva-valve (inclu&ng a safety or relief valve) that is etther ovely calculated potential offsite doses (using meteorol-i l
normally closed or capable of automatic closure dunng ogy as recommended by Regulatory Guide 1.3, all modes of' normal reactor operation.
" Assumptions Used for Evaluating the Potential Ra&o-loWeal Conequences of a Loss of Coolant Accident for
- e. Systems or poroons of systems' that are B ihng Water Rea: tors," and Regulatory Guide 14 cunnected to the reactor coolant pressure boundary and "A8sumptions Used for Evaluatmg the Potennal Ra&o.
are not capable of bems isolated from the boundary I gical Consequences of a Loss of Coolant Accident for dunng all modes of normal reactor operadon by two Pressunted Water Reactors") that exceed 0.5 rem to the valves, each of which is either normally closed or espeble whole body or its equivalent to any part of the body.
of automatic cicsure.
For those systems located in Seismic Category I struc-tur". nly 8insk c mPent fah nad k assud
- 2. The Group C quality standards given in Table I of
@owm, no cre&t for s'at mauc is lad n kom other this guide should be appbed to water, steam, and components in the system or for treatment of released
' Group A quahty standards that are reqused for pressure.
material should be taken unless the isolauon or treat-contauwsg components of the reactor toolant preneure bound.
ment capabthty is designed to the appropriate setsmic
,ery are spectised in Secuan 50.55e of 10 CFR Part 50.
and quahty group standards and can withstand loss of The system boundary includes those poruons of the eystem offsite power and a single failure of an active compo-required to secomphth the specif' sed safety function and connected piping up to and indudvig the fust valve Omduding a nent.)
safety or rehef velve) that is either normauy closed or repable sf automatic dosure when the safety function is reqwwed.
8 The turbine stop valve and the turbine bypeas velve, although
- Components in sinuent Lanes may be claiasfied as Group D j
not induded en quabty group B, should be subsected to a provided they are capeble of being snotated from the reactor 1
quabty assurance program at a level genetsu equrvelent to coolant preneure boundary by an additional velve which has i
r quahty group D.
hieh lest tight integnty.
y 1
I l.26 2 9
- 3. The Group D quahty standards given in Table 1 of This guide reflects current NRC staff practice. There-this guide should be applied to water and steam-fore, except in those caws in wiuch the appbcant contairung components not part of the reactor coolant proposes an acceptable alternahve method for cumply-pressure boundary or included in quahty Groups B or C ing with specsfied portions of the Comtrussion's regula.
but part of systems or poruons of systems that contain tions, the method described herein is being and ws!!
or may contain radioactiw material.
continue to be used in the enluation of submittals for operating license or construction permit appbcauons D. IMPLEMENTATION until this guide is revised as a result of suggestions from the pubhc or additional staff teview.
The purpose of this section is to provide information to appbcants regarding the NRC staff's plans for using this regulatory guide.
TABLE 1 QUALITY STANDA MOS Cbmponents Quehty B Quakty C Quebtr D Pressure Vessels ASME Boiler and Pressure ASME Boiler and Pressure ASME Boiler and Pressure vessel Code, Section 111, Vessel Code, Section 111, Vessel Code, Section Vill,
Nuclear Power Plant Com.
" Nuclear Power Plant Com-Division 1 ponents," a. b, c Class 2 ponents,"a.b.c Class 3 Piping As above As above ANSI B31.t 0 Power Peping Pumps As above As above Manufacturers standards
(
Valves As above As above ANSI B31.1.0 Atmospheric As above As above API 650, AWWA D 100, or Storage Tanks ANSI B 96.1 015 psig As above As above API 620 Storage Tanks
- $ee Section 50 55e for guidance with regard to the Code and Addenda to be opphed.
bASME Code N-symbol news not tw opphed.
'The scocific appbcabihty of code cases will be covered separately en other regulatory guides or en Commission reg appropriate. Apphtents pr0 Doling she une Of Code Cases not Covered by guides or regulations should dernonstrate that en acceptele levee OfQuality and safety would be echeeved.
e i
l i
I; l.26 3 r
I
RevlCIIn 3 a a%
U.S. NUCLEAR REGULATORY COMMISSION September 1978
%) REGULATORY GUIDE
- e.e*
OFFICE OF STANDARDS DEVELOPMENT REGULATORY GUIDE 1.29 j
SEISMIC DESIGN CLASSIFICATION A.
INTRODUCTION tures of light-water-cooled nuclear power plants that should be designed to withstand the effects of the General Design Criterion 2. " Design Bases for SSE. The Advisory Committee on Reactor Protection Against Natural Phenomena," of Appen.
Safeguards has been consulted regarding this guide dix A, " General Design Criteria for Nuclear Power and has concurred in the regulatory position.
Plants," to 10 CFR Part 50, " Domestic Licensing of Production and Utilization Facilities," requires B.
DISCUSSION that nuclear power plant structures, systems, and After reviewing a number of applications for con-components important to safety be designed to with-struction permits and operating licenses for boiling stand the effects of earthquakes without loss of capa-and pressurized water nuclest power plants, the NRC bility to perform their safety functions.
staff has developed a seismic design classification Appendix B, " Quality Assurance Criteria for Nu-system for identifying those plant features that should clear Power Plants and Fuel Reprocessing Plants," to be designed to withstand the effects of the SSE.
10 CFR Part 50 establishes quality assurance re-Those structures, systems, and components that quirements for the design, construction, and opera.
should be designed to remain functional if the SSE tion of nuclear power plant structures, systems, and occurs have been designated as Seismic Category 1.
components that prevent or mitigate the consequences of postulated accidents that could cause undue risk to C.
REGULATORY POSITION l
the health and safety of the public. The pertinent re-
- 1. The following structures, systems, and compo-C., quirements of Appendix B apply to all activities af-nents of a nuclear power plant, including their foun-fecting the safety related functions of those struc-dations and supports, are designated as Seismic Cate-tures, systems, and components.
Sory I and should be designed to withstand the effects Appendix A, " Seismic and Geologic Siting of the SSE and remain functional. The pertinent qual-Criteria for Nuclear Power Plants," to 10 CFR Part sty assurance requirements of Appendix B to 10 CFR 100, " Reactor Site Criteria," requires that all nu-Part 50 should be applied to all activities affecting
)
clear power plants be designed so that, if the Safe the safety-related functions of these structures, sys-Shutdown Earthquake (SSE) occurs, certain struc-tems, and components, tures, systems, and components remain functional-
- a. The reactor coolant pressure boundary.
These plant features are those necessary to ensure (1) the integrity of the reactor coolant pressure boundary,
- b. The reactor core and reactor vessel internals.
(2) the capability to shut down the reactor and main-
- c. Systems' or portions of systems that are re.
tain it in a safe shutdown condition, or (3) the capa-quitec' for (1) emergency core cooling, (2) postacci-bility to prevent or mitigate the consequences of ac-cidents that could result in potential offsite exposures e Lines endicate substantive changes from previous issue.
~
comparable to the guideline exposures of 10 CFR
'The syster1 boundary includes those portions of the system re-Part 100*
quired to accomplish the specified safety function and connected Piping up to and including the first valve (including a safety or This guide describes a method acceptable to the relief $atie) that is either normally closed or capable of automatic NRC staff for identifying and classifying those fea-closure hen the safety function is required.
USNRC REOut.ArORY OUlDES Co*" wats enound to sent to ~ secretary e the commies.on. U s Nucmar 2"ic e i 83*"- " * ""'"*" "* '" ^""'* """""*
aegato,, c.id. to.~to deime.te t.cnni.es e.d D, ~ 9 t. e
.-~,mes.on.e.
are N to d. esc,,.,e end,en.e 8..
~ to,e~,,s o.,a, c s em Nac s
,m ier ting
. tic
~
ras goi.es -e e ed. ~,.w-o ion brou d c.
.g...
-e w
~m,v Guides.o oeuds end.s$or regu 1 Pow Reactors "e e'ca and test a-t-s 6 Pro. ducts istor are r e
t e cons com nr -.-
notreou
.,, d~ent ero,n ~se eet out in the guidse ed be acceptabie et 9%ey provide e besas for the findings 3 Fveis and Metenais 8.cd.tes S Occupational Heait#1 e
t the eeuence or continuance of a permet or bconee by the (n.worwnental nd Set A
rust end Financial Review Raouests for sengte copes of sesued guides (which may be reproducedt or for
(
Comments and suggestions for au n in these guides e encoureged et piecement on en evtoma,te d,sireweion not #ar ngie copurs of fvto e.ou, des r
til times and ou. des we be re.eed. es soproprote. to accommodate cornments in specrte asiviewens shou d be mede in earte 's to the U $ Nucie-Reg torv r
,"*el.O'." M*o* *,d t rTrol"l'r#,o**L" "t" '. ' " " "
%":':"nior *%;"s'?"oric,ent T"t;o,'"*
" * ' " * ' ~
- i (ySOD$ i. * ~
w
-.. ~.-
a aL I
~ ' dent containment heat removal, or (3) postaccident v
L
- n. The control room, including its associated containment atmosphere cleanup (e.g., hydrogen re.-
equipment and all equipment needed to maintain the moval system),
control room within safe habitability-limits for
'W
- d. Systems t ~or portions of systems that are re-personnel and safe environmental limits for vital i
quired for (1) reactor shutdown, (2) residual heat re-
- l,;P,,,,.
moval, or (3) cooling the spent fuel storage pool.
i
. e. Those portions of the steam systems of boil.
. o. Primary and secondary reactor containment.
j ing water reactors extending from the outermost con.
P. Systems,8 other than radioactive waste man-tainment isolation valve up to but not including the agement systems,2 not covered by items 1.a through turbine stop valve, and connected piping of 2%
1.o above that contain or may contain radioactive ma-g l
' inches or larger nominal pipe size up to and including terial and whose postulated failure would result m -
the first valve that is either normally closed or capa-conservatively calculated potential offsite doses (us.
i ble of automatic closure during all modes of normal ing meteorology as recommended in Reguletory
'}
']'
reactor operation. The turbine stop valve should_
Guide 1.3, 'fAssumptions-Used for Evaluatmg the be designed to withstand the SSE and maintain its
. Potential Radiological Consequences of a Loss of integrity.
. Coolant Accident for Boiling Water Reactors," and
- f. Those portions of the steam and feedwater Regulatory. Guide 1.4, " Assumptions Used for i
systems of pressurized water reactors extending from Evaluating the Potential Radiological Consequences and including the secondary side of steam generators of a Loss of Coolant Accident for Pressurized Water '
i up to and including the outermost containtnent isola-Reactors") that are more than 0.5 rem to the whole tion valves, and connected piping of 2% inches or body or its equivalent to any part of the body.'
larger nominal pipe size up to and including the first valve (including a safety or relief valve) that is either
- q. The Class IE electric systems, including the -
normally closed or capable of automatic closure dur.
auxiliary systems for the onsite electric power ing all modes of normal reactor operation.
supplies, that provide the emergency electric power needed for functioning of plant features included in 1
- 3. Cooling water, component cooling, and items 1.a through 1.p above.
]g I
auxiliary feedwater systems' or portions of these sys-tems, including the intake structures, that are re-2.' Those po.. sis of structures, systems, or com-4 quired for (1) emergency core cooling (2) postacci-ponents whos -
stinued function is not required but j
dent containment heat removal, (3) postaccident con.
whose failure vuld reduce the functioning of any N
tainment atmosphere cleanup, (4) residual heat re-plant feature included in items 1.a through 1.q above
)
moval from the reactor, or (5) cooling the spent fuel to an unacceptable safety level or could result in in-i storage pool.
capacitating injury to occupants of the control room k
should be designed and constructed so that the SSE
- h. Cooling water and seal water systems 8 or would not cause such failure.8
{
portions of these systems that are required for func.
tioning of reactor coolant system components impor-
- 3. S CmWIW gimusM
[
tant to safety, such as reactor coolant pumps.
extend to the first seismic restraint beyond the de-
- i. Systems' or portions of systems that are re-fined boundaries. Those portions of structures, sys-A 4
quired to supply fuel for emergency equipment.
tems, or components that form interfaces between Seismic Category I and non-Seismic Category I fea-
-j. All electric and mechanical devices and cir-tures should be designed to Seismic Category I cuitry between the process and the input terminals of
9" *"*
- the actuator systems involved in generating signals that initiate protective action.
4, Ilie pertinent quality assurance requirements of Appendix B to 10 CFR Part 50 should be applied to I
- k. Systems' or portions of systems that are re.
all activities affecting the safety.related functions of i
quired for (1) monitoring of systems important to those portions of structures, systems, and compo-safety and (2) actuation of systems important to nents covered under Regulate-Positions 2 and 3 safety.
abover
- 1. The spent fuel storage pool structure, incluJ-ing the fuel ra kc s.
sSpecific guidance on seismic requirements for radioactive waste
- m. The reactivity control systems, e.g., control-manasement systems is under denlopment.
i tods, control rod drives'and boron injection Wherever practical, structures and equipment whose failure sy5 tem, could possibly cause such injuries should be retocated or sepa.
l' N
rated to the entent required to eliminate this possibilityc i
i 9
1.29-2 i
c 3,..
_v
.,,,.._-.r_
..m.
D.
IMPLEMENTATION cint proposes an acceptble alternative m:thod for complying with specified portions of the Commis-The purpose of this section is to provide informa.
sion's regulations, the method described herein is tion to applicants regarding the NRC staff's plans for being and will continue to be used in the evaluation using this regulatory guide.
of submittals for operating license or construction permit applications until this guide is revised as a re-This guide reflects current NRC staff practice.
sult of suggestions from the public or additional staff Therefore, except in those cases in which the appli-review.
(
9 h
i d
i 1.29-3 4
Y s
mes -
,=
r v
00ESTION 1(b).
Are these requirements different than those applicable to_other portions of the feedwater and steam lines that are closer to the steam generators and reactor vessel? 'If so, whv are they,.and do you.think this distinction is appropriate in view of what occurred in the.Surry Plant accident?
What.is the safety
,fustification for the-difference?
ANSWER The construction codes and standards and regulatory requirements applied to the safety-related and nonsafety-related portions of the feedwater and steam systems on PWR's are described in response to~0uestion 1(a) above.
Section XI of the Code currently does not contain a requirement to explicitly measure wall thickness to detect thinning.
Weldments are inspected by non-destructive examinations to determine if indications are within allowable limits.
The requirements noted above for the safety-related portions are appropriate since these lines (i.e., auxiliary feedwater) are relied upon to mitigate accidents resulting from failures in other lines.
In general, this distinction appears to.still.be appropriate.
However, the erosion-corrosion degradation and other failure mechanisms in nonsafety-related systems are under review to determine how on-site non-radiological injuries or fatalities which result from failures in such lines should be dealt with by the NRC.
If the results of such review indicate that NRC should play a nore active role in protecting on-site personnel from non-radiological hazards, such systems could receive more attention in initial design and specific requirements to detect wall thinning-and the inspection of weldment for flaws at critical locations 'could. be added as part of an inservice inspection program.
1 0 lJ E S T I O N l'( c ).
If a failure in the feedwater pioing occurred at a similar-location, e.g., between the condenser and feedwater_ piping in a Boiling Water Peactor' nuclear power. plant, could radioactivity be released to the environment?
ANSWER.
Yes.
If a feedwater pipe break occurred outside of containment, some fraction of the radioactivitv in the water in the feedwater piping and main condenser would be released to the turbine building.
Isolation valves in the feedwater system would prevent backflow of water from the reactor vessel.
Since the turbine building is not designed as a containment structure, activity released to the turbine building is assumed to immediately enter the environment.
However, as discussed in the response to Question 1(c)(i), no significant radiolooical consequences would be-expected in the surrounding area.
l
_ QUESTION 1(c)(1).
If so, how much radioactivity could.be released and what would be the consequences to the surrounding area?-
ANSWER.
Accident analyses demonstrate that no fuel damage would occur as a consequence of a feedwater pipe break outside of containment because other engineered safety systems would supply the core with coolant _ water.
While we don't have a specific calculation of the occupational exposure for a comparative pipe failure in the turbine building of a BWR, the amount of radioactivity available for release to the environment is limited to the radioactivity of the water in the condenser and the feedwater piping which is controlled by_the plant Technical Specifications.
i
_ Tables 1 and 2 (attached) show the estimated radioactivity release and thyroid dose consecuences.for a typical boiling water reactor assuming that_the activity released _.from a feedwater line break immediately enters the environment.
The 1
calculated doses are for an individual at the site boundary are far below the guideline values of 10 CFR Part 100, and therefore no significant radiological consequences would be expected in the surrounding area.
TABLE 1 i
FEEDWATER'LINE BREAK (REALISTIC ANALYSIS)
ACTIVITY RELEASE.TO ENVIRONMENT, Ci Isotope Activity I-131 2.64E-2 I-132 1.54E-1 I-133 1.14E-1 I-134 1.97E-1 1-135 1.14E-1 Total equivalent 5.35E-2 I-131 TABLE 2 FEEDWATER LINE BREAK RADIOLOGICAL EFFECTS 10 CFR Part 100 Thyroid, rem Guideline Values Site boundary (2-hour dose) 1.73E-3 300 rem to the thyroid i
Low population zone
<<1*
300 rem,to the (30 day dose) thyroid
- Estimated based upon calculated 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose =-1.32E-4
~
h
=
Qt!ESTION 1(c)(ii)..How are these areas of'the feedwater and steam: lines classified-in Boiling. Water Reactors?
ANSWER.
Regulatory' Guide (R.G.) 1.26, Standard Review Plan (SRP) Section-3.2.2, (Attached) and 10 CFR 50.55-provides.the staff's criteria for classifying the main steam line and the feedwater.line from the reactor up to and includino the outermost isolation valve as-
~
Quality Group A (ASME Section III, Class 1).
R.G.
1.26 also classified the main steam line up to but not including the turbine stop valve and. bypass valves.as Quality Group B (ASME Section III, Class 2) (See Table A.1,.SRP 3.2.2).
Alternatively, for BWRs containing a shut-off valve (_in addition.to the two containment isolation valves) in the main steam line and in the main feedwater line, Quality Group B. standards should be applied to those portions of the steam and feedwater systems extending from the outermost containment. isolation valve up to and including the. shutoff valve-(See SRP 3.2.2).
Weldments in the steam and feedwater systems that are classified as Quality Groups A.and B are subject to periodic inservice inspection in accordance with Section XI of the ASME Code per 10 CFR 50.55(a)(o).
s l
Ocu cmu-ww (Fermarly NUREG 75/08D
'/ce r t co,~\\
U.S. NUCLEAR REGULATORY COMMISSION iP i STANDARD REVIEW PLAN i
OFFICE OF NUCLEAR REACTOR REGULATION k..s(o 3.2.2 SYSTEM QUALITY GROUP CLASSIFICATION REVIEW RESPONSIBILITIES Primary - Mechanical Engineering Branch (MEB)
Secondary - None I.
AREAS OF REVIEW Nuclear power plant systems and components important to safety should be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety function to be performed.
The MEB reviews the applicant's classification system for pressure-retaining components such as pressure vessels, heat exchangers, storage tanks, pumps, piping, and valves in fluid systems important to safety, and Where required, specific information or assistance may be safety functions.
required from the ICSB to review electrical and instrumentation systems needed This review which is
/
for functioning of plant features important to safety.
(
coordinated with each branch that has primary review responsibility for these plant features is performed for both construction permit (CP) and operatingstructures i
license (0L) applications.
Excluded from this review are:
parts of mechanical components such as shafts, seals, impellers, packing, and gaskets; fuel, electrical, and instrumentation systems, electrical valve actuation devices, and pump motors.
The applicant presents data in his safety analysis report (SAR) in the form of a table which identifies the fluid systems important to safety; the system components such as pressure vessels, heat exchangers, storage tanks, pumps, piping, and valves; the associated quality group classification, ASME Code and In addition, the applicant code class; and the quality assurance requirements.
presents on suitable piping and instrumentation diagrams the system quality group classifications.
~
Rev. 1 - July 1981 USNRC STANDARD REVIEW PLAN star.dard review plans are prepared for the guidance of the office of Nuclear Reactor Regulatio f he applications to construct and operate nuclear power plants. These docum i d The plans are not substitutes for regulatory guides or the Commission's regulations and compliance with standard review plan sections are keyed to the Standard Format and Content of safety Analysis R Not all sections of the Standard Format have a corresponding review plan.
Published standard review plans will be revised periodically, as appropriate, to accommodate com tion and superience, Comments and suggestions for improvement will be considered and should be sent to the U.S. Nu Office of Nuclear Reactor Regulation. Washington, D C. 20566.
- - ~
,--a e
,m a
Tho ME8 also performs the following reviGws for the SRP = sections indicated:
1.
Determines the acceptability of-theiseismic cla'ssification of-system-components.i'n accordance with SRP Section 3.2.1. =The information may be combined with-the information in this' SRP section which may result in cross referencing rather than repetition of the information, i
2.
Verifies that systems and components important to-'sa'fety that are designated as Quality Groups A, B, C, or D items are constructed.in e
accordance with the regulatory guides, industry codes and standards that-are referenced in SRP Sections 3.2.1, 3.9.1 through 3.9.3,'and 1
3.
Determines the a'dequacy of the inservice t'esting program for pumps and valves in accordance with SRP Section 3.9.6.
II~ ' ACCEPTANCE CRITERIA Acceptance criteria is based on meeting the relevant requirements of the following regulations:
10 CFR Part 50, Appendix A, General Design Criterio'n 1 and.10 CFR Part 50, 5 50.55a, as they. relate to the requirement-that structures, systems,'and components important to safety shall be designed, fabricated, erected, and i
tested.to quality standards commensurate with the importance of the safety
~
l function to be performed.
To meet the requirements of General Design Criterion 1 and 10 CFR Part 50, 5 50.55a, the following regulatory guide is used:
Regulatory Guide 1.26, " Quality Group' Classification and Standards."
}
This guide describes an acceptable method for determining quality i
standards for Quality Group B, C, and D water-and steam-containing components important to safety of water-cooled nuclear power plants.
III. REVIEW PROCEDURES Selection and emphasis of various aspects of the areas covered by this SRP section will be made by the reviewer on each case. The. judgement on the areas
~
to be given attention during the review is to be based on an inspection of the material presented, the similarity of the material to that recently reviewed j
j on other plants, and whether items of special safety significance are involved.
Section 50.55a of 10 CFR Part 50 identifies those ASME Section III, Code Class 1 components of light-water-cooled reactors important to safety which are part of the reactor coolant pressure boundary.
These components are i
designated in Regulatory Guide 1.26 as Quality Group A.
In addition, l
Regulatory Guide 1.26 identifies, on a functional basis, water-and steam-containing components of those systems important to safety that are Quality 3 -
Groups B and C.
Quality Group D applies to water-and steam-containing components of systems that are less important to safety. 'An applicant may use the NRC Group Classification system identified in Regulatory Guide 1.26 or, alternately, the corresponding ANS classification system of Safety Classes which can be cross-referenced with the classification groups in Regulatory:
Guide 1.26.
There are also systems of light-water-cooled reactors important i
to safety that are not identified in Regulatory Guide 1.26 and which the staff considers should be classified Quality Group C.
Examples of these systems i
l 3.2.2-2 Rev. 1 - July.1981 z- :
m
, diesel fuel oil.storaga and transfer system; diesel engine cooling water-system, diesel engine lubrication system, diesel engine starting. system,.
,are:
' diesel engine combustion air intake and exhaust system,' and instrument and service air systems required to perform a safety function; and certain Gas treatment systems which are considered as ventilation plant systems.
engineered safeguards systems should be classified Quality Group B.
The information supplied _in the application identifying fluid systems f
important to safety is reviewed for completeness, and the quality group 1
4 classification,~ ASME Code and code. class, and quality assurance requirements of each individual major component are checked for compliance with the above i
The various modes of system operation are checked to assure that criteria.
the assigned NRC quality groups are acceptable.
i e
The piping and instrumentation diagrams are reviewe boundaries for systems;important to safety.
is checked to assure the accuracy of the assigned qua l
l Changes in quality group classification-are permitted and sample lines.
normally only at valve locations, with the valve assigned the permitted only when it can be demonstrated that the s sification.
boundary.
The following fluid systems important to safety for pressurized water reactor (PWR) and boiling water reactor (BWR) plants are reviewed by the MEB with.
regard to quality group classification.
FLUID SYSTEMS IMPORTANT TO SAFETY FOR PWR PLANTS Reactor Coolant System Emergency Core Cooling System 1:
l Containment Spray System Chemical and Volume Control System Boron Thermal Regeneration System,2 1
Boron Recycle Systen :
t i
Residual Heat Removal System s
Component Cooling Water System 2
Spent Fuel Pool Cooling and Cleanup System i
8 Sampling System 2-Service Water System Compressed Air System,2 i
Emergency Diesel Engine Fuel Oil Storage and Transfer System Emergency Diesel Engine Cooling Water System l
Emergency Diesel Engine Starting System i
Emergency Diesel Engine Lubrication System Emergency Diesel Engine Combustion Air Intake and Exhaust System l
i 8
Main Steam System 8
Feedwater System Auxiliary Feedwater System l
i.
8 Steam Generator Blowdown System
' Containment Cooling System Containment Purge System i
Rev. 1 - July 1981 3.2.2-3 1
.f
- - - -.,--.-.:=.
- . e i..
~
L V; Features Rooms:ntilagtfon Systems for Areas;such asiControl Room and Engineered Safety'
~
Combustible Gas. Control' System j_
. Condensate. Storage System 2
[
FLUID SYSTEMS IMPORTANT TO SAFETY FOR'8WR PLANTS keactor Recirculation System
- Main Steam System (up to but. not including the turbine)
Feedwater System (up to' outermost containment isolation valve or shutoff valve, as. applicable)-
Relief Valve Discharge Piping 4
Control Rod Drive Hydraulic System 2 Standby Liquid Control System
} _
. Reactor Water Cleanup System Fuel Pool Cooling.and Cleanup System 2 l
Sampling System 3 Residual Heat Removal System
~ High Pressure Core Spray System-l-
Low Pressure Core Spray System
{
Reactor Core Isolation Cooling System i
RHR Service Water System Emergency Equipment Servi'ce Water System
{
Compressed Air System
.2 t
i Emergency Diesel Engine Fuel Oil Storage and Transfer System Emergency Diesel Engine Cooling Water System 4
l Emergency Diesel Engine Starting System Emergency Diesel Engine Lubrication. System Emergency Diesel Engine Combustion Air Intake and Exhaust System Standby Gas Treatment System 3
4 Combustible Gas Control System i
Containment Cooling System i
Main Steam Isolation Valve Leakage Control System l
l Condensate and Refueling Water Storage System.
2 Ventilation Systems for Areas such as Control Room and Engineered Safety Features Rooms
\\
i Clarification of the Quality Group Classification provided_in Regulatory Guide 1.26 and applicable to those portions of BWR main steam _and feedwater systems (other than the reactor coolant pressure boundary) on the turbine side of the containment isolation valves, are given in Appendices A and 8,_ attached j
4 to this SRP section.
F Additional guidance on the quality group classification of= systems and
~
i components important to safety for a typical PWR plant is given in Appendix C l
attached to this SRP section.
Similarly, additional guidance en~the quality group classification of systems and components important to safety for a typical i.
_ BWR plant is given in Appendix D attached to this SRP section. Appendices C-and D, in part, identify individual system components including appropriate interconnecting piping and valves,.by quality group and the applicable code and 3
i 20n some plants this' system may be non-safety-related, providing it complies ~
with.the requirements of Regulatory Guide 1.26.
2 Portions of the system that perform a safety-related function.
~
j 3 Portions of the system to outermost containment isolation valve.
3.2.2-4 Rev.1 - July.1981 1
4
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4 b
- code class. (Table!3.2.21~ attached to this SRP.section provides a summary of-'
- l
- s the" construction Codes and Standards for components;of water-cooled nuclear
~
power plants and is-based on the NRC; quality group classification system.in:
In the event an applicant intends to take exception to Regulatory Guide -1;26 i
and has not provided adequate. justification for his proposed quality group
~
classification, questions _are prepa' red by.the staff which'may-require 4
additional documentation or an analysis to. establish an acceptable. basis;for his proposed quality group classification'.
Staff comments may also be prepared requesting clarification, in order to' assure a clear understanding of-the quality group; classifications assigned to'a system by the applicant.
Exceptions and alternatives-to the specified quality group classifications of Regulatory Guide 1.26 are unacceptable unless " equivalent quality level" is justified..In such cases, justification can be demonstrated if:
the' component is classified to meet,the requirements of a higher group clas--
'sification than specified in Regulatory. Guide l.26 or alternative design rules l
^
are based on the use of a more conservative design;tthe extent of component nondestructive examination is equal to or. greater than required by the specified code; and the quality assurance requirements of Appendix 8, 10 CFR Part 50 are met.-
If the" staff's questions are not resolved in a satisfactory manner, a staff position is taken requiring conformance to Regulatory Guide 1.26.
IV.
EVALUATION FINDINGS The staff's review should verify that adequate and sufficient informatio'n is l
r contained in the SAR and amendments to arrive at a conclusion of the following.
J
(
type, which is to be included in the staff's safety evaluation report:
Pressure-retaining components of. fluid systems important to safety such as pressure vessels,-heat exchangers, storage tanks,. pumps, j
piping and valves have been classified Quality Group A,'B, C, or D.
and have been identified in an acceptable manner in Table 3.'X.X and on system piping and instrumentation diagrams in the'.SAR.
These components-have been constructed to quality standards commensurate with the importance of the safety function-to be performed.
The review of Quality Group A and B (ASME Section III, Class.1 and 2) reactor coolant pressure boundary components is discussed in Sec -
tion 5.2.1.1 of the SER.
Other Quality Group B components of systems j
identified in Position C.1.a through C.1.e of Regulatory' Guide 1.26 are constructed to ASME Section III, Class 2.
Components in systems identified in Postion C.2.a thro' ugh C.2.d of Regulatory Guide -1.26 J
ace constructed to Quality Grup.C' standards,.ASME Section III, Class 3.
Components in systems identified in Position C.3 of R'egulatory Guide 1.26 are constructed to Quality Group D standards such as, ASME Section VIII and ANSI B31.1.
s-The staff concludes that pressure-retaining components of fluid systems important to safety have been properly classified as Quality Group-A, B, C, or D items and meets the requirements of General Design Criterion 1, " Quality Standards and Records." This conclusion is based on the applicant having met the requirements of General Design Criterion.1 by having properly classified these 3.2.2-5 Rev. 1 -. July 1981 L
.a..
pressure-retaining coJponents important to safety Quality Group A, i B, C', or D in accordance with the positions of Regulatory Guide 1.26, " Quality Group Classifications and Standards," and by our conclusion that the identified pressure-retaining components are those necessary (1) to prevent or mitigate the consequences of accidents and malfunctions originating within the reactor coolant pressure boundary, (2) to permit shutdown of the reactor and maintain it in a safe shutdown condition, and (3) to contain radioactive materials.
V.
IMPLEMENTATION The following is intended to provide guidance to applicants and licensees regarding the NRC staff's plan for using this SRP section.
Except in those cases in which the applicant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, the method described herein will be used by the staff in its evaluation of conformance with Commission regulations.
Implementation schedules for conformance to parts of the method discussed herein are contained in the referenced Regulatory Guide.
V.
REFERENCES 1.
10 CFR Part 50, Appendix A, General Design Criterion.1, " Quality Standards and Records."
l 2.
-Regulatory Guide 1.26, " Quality Group Classifications and Standards."
)
3.
ASME Boiler and Pressure Vessel Code, 1980 Edition,Section III, " Nuclear Power Plant Components," American Society of Mechanical Engineers (1980).
l 4.
ASME Boiler and Pressure Vessel Code,1980 Edition,Section VIII, Division 1, " Pressure Vessels," American Society of Mechanical Engineers (1980).
5.
ANSI /ASME B31.1-1980, " Power Piping," American National Standards Institute (1980).
6.
API Standard 620, Sixth Edition, " Recommended Rules for Design and Construction of Large, Welded, Low-Pressure Storage Tanks," American Petroleum Institute (1977).
I 7.
API Standard 650, Sixth Edition, Revision 1, " Welded Steel Tanks for Oil Storage," American Petroleum Institute (1978).
8.
AWWA D100-79, "AWWA Standard for Steel Tanks-Standpipes, Reservoirs, and Elevated Tanks for Water Storage," A.merican Water Works Association (1979).
9.
ANSI B96.1-1980, " Specification for Welded Aluminum-Alloy Field-Erected Storage Tanks," American National Standards Institute (1980).
10.
Appendix A, " Classification of Main Steam Components Other Than the Reactor Coolant Pressure Boundary for BWR Plants," attached to this SRP section.
3.2.2-6 Rev. 1 - July 1981
-.w
dix B,_L" Classification of BWR/6. Main: Steam and Fosdwater Components RP Other.Than.the Reactor Coolant Pressure Boundary," attached to.this S r.11.
Appen
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1 p
4 f
r 1
).
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a i
Rev.- l'- July 1981-3.2.2-7 i
ms
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TABLE 3.2.2-1 SIM4ARY OF CONSTRUCTION: CODES AND STANDARDS FOR COMPONENTS OF WATER-COOLED NUCLEAR POWER PLANTS BY NRC QUALITY CLASSIFICATION SYSTEM 2 4
NRC Quality Classification System Components Quality Group A Quality Group B-Quality Group C Quality Group D Pressure Vessels ASME Boiler and Pressure ASME Boiler and Pressure ASPF Boiler and Pressure ASME Boiler and Pressure Vessel Code,Section III, Vessel Code,Section III, Vessel Code,Section III, Vessel Code,Section VIII, Division 1. Subsection N8 Division 1, Subsection NC Division 1, Subsection NO Division 1.
-Class 1, Nuclear Power
-Class 2, Nuclear Power
-Class 3, Nuclear Power 3
3 3
Plant Components
'4 Plant Components
'4 Plant Components
'4 Piping As above As above As above ANSI B31.1 Power Piping Pumps As above As above As above Manufacturers standards.
[ Valves As above As above As above ANSI B31.1 Power Piping
~
and ANSI B16.34 Atmospheric Not applicable As above As above API-650, AWA D100, or Storage Tanks ANSI B96.1 0-15 psig Storage Not applicable As above As above API-620 Tanks Supports As above except As above except.
As above except.
Manufacturers standards Subsection NF Subsection NF Subsection NF Metal Containment Not applicable As above except Not applicable Not applicable Components Subsection NE, Class MC Core Support Not applicable As above except Not applicable Not applicable Structures Subsection NG
1 NOTES:
1As defined in Subsubarticle NCA-1110 of Section III, of the ASME Boiler and Pressure Vessel Code, construction is an all-inclusive term comprising materials, design, fabrication, examination, testing, inspection, and certification required in the manufacture and installation of components.
2As defined in Regulatory Guide 1.26, the NRC Quality Classification System identifies on a functional basis components of fluid systems by Quality Groups A, B, C, and D.
3See Section 50.55a, " Codes and Standards," of 10 CFR Part 50 for guidance with regard to the Code Edition and Addenda to be applied
.I 4The specific applicability of ASME Code Cases is covered separately in SRP Section 5.2.1.2, Regulatory Guides 1.84 and 1.85, or in Commission regula-tions, where appropriate. Applicants proposing the use of ASME Code Cases not covered by these SRP and Regulatory Guides should receive approval from the Commission prior to their use and should demonstrate that an acceptable level of quality and safety would be achieved.
1 i
1 3.2.2-9 Rev. 1 - July 1981
APPENDIX A*
I CLASSIFICATION OF MAIN STEAM COMP 0NENTS OTHER THAN THE REACTOR COOLANT PRESSURE BOUN9ARY FOR BWR PLANTS A.
BACKGROUND A pipe classification of "D + QA" for main steam line components of BWR plants was proposed by the General Electric Company in 1971 as an alternative to Quality Group B and has been accepted by the staff in a number of licensing case reviews.
However, we have recently identified a number of potential problems _which are applicable to main steam lines of BWR plants.
These problems relate to postu-lated breaks in high-energy fluid containing lines outside the containment.
The criteria pertaining to protection required.for structures, systems, and components outside containment from the effects of postulated pipe breaks, as contained in the Director of Licensing's letter to utilities dated July 12, 1973, reference.ASME Section III, Class 2, which corresponds to NRC Quality Group B.
The recent. ASME Code Section XI revision contains in-service inspection requirements for Class 2 components.
Steam lines classified as "D + QA" could be interpreted to be exempt from these inspection requirements.
Such l
interpretations would be contrary to the intent of the code and inconsistent with requirements of the NRC_ Codes and Standards rule, Section 50.55a of 10 CFR Part 50.
Furthermore, the applicability of. the following NRC Regulatory Guides, l
Section III, Class 2 components is not always clearly identified or implemented
)
Standard Review Plan section, and Regulations, as they relate to ASME in case applications wherever "D + QA" classification is adopted:
1.
SRP Secticn 3.9.3, "ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures."
2.
Regulatory Guide 1.26, " Quality Group Classifications and Standards."
3.
10 CFR Part 50, 6 50.55a, " Codes and Standards for Nuclear Power Plants."
~
4.
10 CFR Part 50, Appendix B, " Quality Assurance Criteria for Nuclear Power Plants."
In view of the foregoing, we find it necessary to clarify the quality group classification criteria for main steam components for BWR plants.'
t B.
BRANCH TECHNICAL POSITION The main steam line components of BWR plants should conform to the criteria listed in the attached Table A-1 of SRP Section 3.2.2.
g A
Formally BTP RSB No. 3-1 l
3.2.2-10 Rev. 1 - July 1981 i
l
t C.
. REFERENCES
.c 1.
Letter of March 22,.1973, J. A. Hinds.to'J.'M.'Hendrie.
-2.
Letters of August 13,'1973 and November.26, 1973, J. M.-Hendrie to J. A.
Hinds.
Ta'ble A-1 l
CLASSIFICATION REOUIREMENTS FOR MAIN STEAM COMPONENTS OTHER THAN THE FEACTOR COOLANT PRES 5URE BOUNDARY -
Classification Item System or Component Quality Group
^
~
~
1.
Main Steam Line from 2nd Isolation-
-8 Valve to: Turbine Stop Valve.
2.
Main Steam Line Branch Lines to B
First Valve.
3.
Main Turbine Bypass Line to B
Bypass Valve.
4.
First Valve in Branch Lines B
Connected to Either Main. Steam Lines or Turbine Bypass Lines.
5.
a.
Turbine Stop Valves, Turbine' D + QA1
[
Control Valves, and Turbine'.
or a
Bypass Valves.
Certification 2 i
b.
Main Steam Leads from Turbine D + QA1,3 Control Valves to Turbine Casing.
'or 4
Certification 2
'The following requirements shall be met in addition to the Quality Group D-i requirements:
i 1.
All cost pressure-retaining parts of a size and configuration for which volumetric examination methods are effective shall be examined i
by radiographic methods by qualified personnel. -Ultrasonic examination to equivalent standards may be used as as-alternate to radiographic methods.
2.
Examination procedures and acceptance standards shall be at least equivalent to those specified as supplementary types of examination in ANSI B31.1-1973, Par. 136.4.
2The following qualification shall be met with respect to the certification requirements:
1.
The manufacturer of the turbine stop valves, turbine control valves, turbine bypass valves, and main steam leads from turbine control 4
~
3.2.2-11 Rev. 1 - July 1981 i
.T,-~
l Table A-1 (cont'd) valves to'the turbine casing shall utilize quality control procedures
. equivalent to.those defined in General Electric Publication GEZ-4982A, " General Electric Large Steam Turbine - Generator Quality i
Control Program."
2.
A certification shall be obtained from the manufacturer of these valves and steam leads that the quality control program so defined has been accomplished.
3The following requirements shall be met in addition to the Quality Group'D requirements:
'1. '.All longitudinal and circumferential butt weld joints shall be radiographed (or ultrasonically tested to equivalent standards).
Where size or configuration does not permit effective volumetric examination, magnetic particle or liquid penetrant examination may be substituted.
Examination procedures and acceptance standards shall be at least equivalent to those specified as supplementary types of-examinations, Paragraph 136.4 in ANSI B31.1-1973.
2Property "ANSI code" (as page type) with input value "ANSI B31.1-1973.</br></br>2" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..
All fillet and socket welds shall be examined by either magnetic -
particle or liquid penetrant methods. All structural attachment welds to pressure retaining materials shall be examined by either magnetic particle or liquid penetrant methods.
Examination procedures and acceptance standards shall be at least equivalent to those specified as supplementary types of examinations, Paragraph 136.4 in ANSI B31.1-1973.
3Property "ANSI code" (as page type) with input value "ANSI B31.1-1973.</br></br>3" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..
All inspection records shall be maintained for the life of the plant.
These records shall include data pertaining to qualification of inspection personnel, examination procedures, and examination results.
3.2.2-12 Rev. 1 - July 1981
APPENDIX B CLASSIFICATION OF BWR/6 MAIN STEAM AND FEEDWATER COMPONENTS.
'OTHER THAN THE REACTOR COOLANT' PRESSURE BOUNDARY A.
BACKGROUND At various times.the NRC staff has discussed with the General Electric Company the subject of appropriate classification requirements in boiling water reactor (BWR) plants for main steam system components.
These discussions have included consideration of components that are (a) not classified ~as safety-related items but are located. downstream of the isolation valves, (b) not specifically designed to seismic Category I standards, and (c) not housed in Seismic Category I. structures.
To date, BWR plant reviews have resulted in'various approaches for different individual applications. While these different approaches have resulted in acceptable levels of safety in each case, they have required time-consuming case-by-case reviews.
The GESSAR (PDA) BWR/6 application which was reviewed l
as part of our standardization program, includes this portion of the BWR plant.
In the course of the GESSAR'PDA review, we have identified a systematic basis for classification of such components that will result in an acceptable and uniform design basis for the main steam lines (MSL) and feedwater lines (MFL) in BWR/6 plants.
B.
BRANCH TECHNICAL POSITION The main steam and feedwater system components of BWR/6 plants should be classified in accordance with SRP Section 3.2.2, Appendix A, or alternately, y
in accordance with the attached Table B-1 of SRP Section 3.2.2.
The classifi-cations indicated are consistent with the guidelines currently specified in Regulatory Guide.l.26 and Regulatory Guide 1.29.
As an additional requirement, a suitable interface restraint should be provided at the point of departure from the Class I structure where the interface exists between the safety and nonsafety-related portions of the MSL and MFL.
A sketch i's attached (Figure B-1) to clarify the specified alternate l
classification system.
C.
REFERENCES 1.
Letter of April 19, 1974, J. M. Hendrie to J. A. Hinds.
f Formally BTP RSB No. 3-2 l
A 3.2.2-13 Rev. 1 - July 1981 3
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i Table B-1 1
CLASSIFICATION REQUIREMENTS FOR BWR/6 MAIN STEAM AND FEEDWATER SYSTEM COMPONENT 5 OTHER THAN THE REACTOR COOLANT PRES 5URE BOUNDARY QUALITY GROUP Item SYSTEM OR COMPONENT CLASSIFICATION 1.
Main Steam Line (MSL) from second isolation valve to and B
including shutoff valve.
2.
Branch lines of MSL between the second isolation valve and the B
MSL shutoff valve, from branch point at MSL to and including the first valve in the branch line.
3.
Main feedwater line (MFL) from second isolation valve and B
including shutoff valve.
4.
Branch lines of MFL between the second isolation valve and the B
MFL shutoff valve, from the branch point at MFL to and including the first valve in the branch line.
5.
Main steam line piping between the MSL shutoff valve and the D (1) turbine main stop valve.
6.
Turbine bypass piping.
D 7.
Branch lines of the MSL between the MSL shutoff valve and the D
j turbine main stop valve.
8.
Turbine valves, turbine control valves, turbine bypass valves, D (1,2) and main steam leads from the turbine control valves to the or turbine casing.
Certification (3) 9.
Feedwater system components beyond the MFL shutoff valve.
D
(
(1) All inspection records shall be maintained for the life of the plant.
These records shall include data pertaining to qualification of inspection i
personnel, examination procedures, and examination results.
i (2) All cast pressure retaining parts of a size and configuration for which j
volumetric methods are effective shall be examined by radiographic methods by qualified personnel., Ultrasonic examination to equivalent standards may be used as an alternate to radiographic methods.
Examination procedures and acceptance standards shall be at least equivalent to those defined in Paragraph 136.4, " Examination Methods of Welds - Non-Boiler External Piping," ANSI B31.1-1973.
A 3.2.2-14 Rev. 1 - July 1981
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Table B-11(cont'd)
- The following qualifications shall be met with respect-to the (3) certification requirements:
The manufacturer of the turbine stop valves, turbine control valves, 1.
turbine bypass valves, and main steam leads from turbine control valves to the turbine casing shall utilize quality control procedures
. equivalent to those defined in General Electric Publication GEZ-4982A,
" General Electric Large Steam Turbine-Generator Quality Control Program."
A certification shall be obtained from the manufacturer of these 2.
valves and steam leads that the quality control program so defined has been accomplished.
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a 3.2.2-15 Rev.1 - July 1981 c
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4 CONTAINMENT OUALITY GROUP D OUALITY OUALITY OUALITY GROUP D GROUP A GROUP P TIFICATION
=
=
=
=
==
=
=
=
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TURBINE BUILDING h
AUXILIARY BUILDING
_ BRANCH-I LINE
-TURBINE MAIN STOP VALVE BRANCH LINE 6
g A
-TURBINE CONTROL VALVE D
[ MAIN STEAM LEADS STEAM LINF N
FEEDWATER LINE TURBINE is GE'JERATOR
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BRANCH LINE 3
J CONDENSER SHUTOFF VALVES J
TURBINE BY-PASS VALVE ISOLATION VALVES INTERFACE j RESTRAINTS SEISMIC CATEGORY I NON-SEISMIC CATEGORY I.
=
=
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STRUCTURES, SYSTEMS & COMPONENTS STRUCTURE, SYSTEMS & COMPONENTS
=
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QUALITY GROUP D
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Figure B-1 NRC Quality Group and Seismic Category Classifications Applicable to Power Conversion System Components in BWR/6 Plants.
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Appendix C-i '
t PWR Plants Classification of Systems and Components-In Course of Preparation Classification of Structures In Course-of Preparation s
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2 3.2.2-17 Rev. 1 - July 1981 1
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-1 Appendix-D BWR Plants Classification of Systems and Components In Course of Preparation J
Classification of Structures In Course of Preparation f
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3.2.2-18 Rev. 1 - July 1981
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QUESTION 1(c)(iii).
In view of the Surry--accident Jdo you think that the classifications of these areas of_the power plant (including the steam turbine, cor. denser and feedwater
. pumps) are appropriate?
ANSWER.
The present-classificationHof the steam and feedwater lines is appropriate in general.
Consideration will be given to incorporating the experience from Surry into the,ASME Section XI inspection requirements for those_ portions of the~ systems classified as Quality Groups A'and B.
Inspections of Ouality Group D systems have not been reouired since failures of these-systems are considered in the facility design and the consequences to public health and safety from a radiation exposure are well within 10 CFR Part 100.
However, as discussed in the response to Questions 1(guidelines.
b) and 1(d), a determination will be made on whether nonsafety-related piping should also have inspection requirements.
ee.
~ H QUESTION 1(dl. What additional requirements could be applied-to the feedwater lines, steam lines, steam turbine, feedwater pumps, condenser and related equipment to improve the safety of-nuclear plant operation?
ANSWER.
Consideration will be oiven to periodically monitoring the pipe wall thickness in feedwater lines and other-lines (both safety-related and nonsafety-related).
Although-the concept _is simple, such requirements would have to be specified with care to avoid testing literally miles of line.
Current inservice inspection of safety-related piping according to'Section XI of the.ASME Boiler and Presdure Vessel Code is done in-the-vicinity of butt welds which have been sites for cracking-in piping.
An analogous deternination must be made for the case of wall thinning by an erosion-corrosion mechanism.
Inspection done to date under programs either instituted prior to.the Surry event or conducted in response to it, have revealed wall: thinning of varying degrees in several piping systems in some-plants.
That pattern must be evaluated in order to prepare a meaningful inspection program.
In addition, as discussed in the response to question 1(b), a determination will be made on whether non-safety-related piping should also be included in'such programs.
The above addresses only the issues related to erosion and erosion-corrosion in safety-related piping systems carrying single phase fluids or two phase (steam water).
The broad auestion of what additional requirements could be applied to other equipment (nonsafety-related piping,-steam turbine, feedwater pumps, condenser, etc.) whose failure does not have direct consequances on the release of radioactive material must be reviewed further.
Design basis accident reviews, assuming failure of this equipment, indicate that the plants'can reach safe shutdown in the event of such failures.
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e QUESTION-1(e).;Does theLCommission1 plan'to'make any. changes in.
- its regulatory _ requirements for Surr.y or other nuclear power plants in order to implement' lessons f.
. learned from the Surry. accident?
ANSWER.
At.this time. the issue of regulatory changes for equipment'inL
-1
-this 'part of the ~ plant is still under, study.-
The feedwater line which. failed:at Surry Unit.2 is-not safety-related_and is not.
covered-by detailed.NRC_ construction and periodic examination requirements.
We are reviewing.the; implications of accidents which have minimun or no_ radiological. impact._ Based on the-outcome of.that review,-the Commission may find:it advisable to make changes in the regulatory-requirements.
See also response to Questions 1(b) and 1(d) above..
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'0UESTION 2.-
The NRC team report. cited erosion-corrosion induced thinning of the pipe metal as the cause of the1 failure at theESurry Station.
Do the design, construction, maintenance and integrity monitoring codes, standards, lor other regulations applied to nuclear power plants adequately orovide for
-finding or make allowances for deterioration of plant components and piping in. service?
If.not, does' the. Commission plan any regulatory ' changes to incorporate these' factors in plant design,
~
inspection and maintenance requirements?
ANSWER.
1 The ASME Standards typfcally used for piping construction in nuclear plants are:Section III of the Boiler and Pressure Vessel Code and Power Piping and ANSI /ASME B31.1, of-the-ASME Code for Pressure Piping, B31.
ANSI /ASME B31.1'does explicitly call for consideration of both erosion and corrosion in the design process.
Neither of the above standards provides'for requirements or-offers explicit guidance'on the important parameters to consider to avoid erosion-corrosion in initial designs.
Inspection of the nonsafety-related portion of the feedwater line during operation for either wall thinning or ~weldment flaws is not a requirement of B31.1.
Currently, there are no standards covering periodic monitoring for such piping.
Inspection requirements for safety-related piping are contained in Section XI of the ASME Boiler and Pressure Vessel Code.
Section XI of the Code currently does not contain a requirement to' explicitly measure wall thickness to detect thinning.
ASME has been formally requested by NRC to review its Codes and Standards applicable to both fossil and nuclear. plants for appropriate changes to address the erosion and erosion-corrosion processes, both in the initial design process and later in j
integrity monitoring.
Regarding the plans by the "hC to make regulatory changes in
. design, inspection and n $r erioce requirements, please refer to
'the response to Questio)
'?)
Also, any regulatorv action by the NRC will consider ac. tons.adertaken by the nuclear industry and the national consensus standards.
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QUESTION 3.
The two Surry Station nuclear units are very similar in design, nuclear reactor system and age.
~
The units also " share" some support and auxiliary functions.
- 00ESTION 3(a). In view of this dependency, can you explain why (Init I was not' shut down immediately when the failure occurred in Unit 2? 'Whose responsibility was it to decide-whether or not to shut down immefiately?
In your view, should Unit:1 have been shut down immediately?
ANSWER.
'The feedwater pipe rupture Occurred at approximately 7:20 p.m. on.
December 9, 1986.
At the time both units had been operating at 100% power.
Following an automatic scram because of closure of a main stean trip valve, the rupture of an 18-inch feedwater pipe occurred.
After coping with the immediate actions necessary to-mitigate the accident and account-for and assist injured personnel, the licensee placed the Unit 2'on a cooldown ramp of 50 F.per hour at 6:00 p.m. The unit achieved cold shutdown at 7:00 a.m. nn December 10, 1986.
Inspection of'the failed pipe took place on the morning of December 10 after scaffolding had been erected to reach the failed oipe sections.
It was at this time that it became apparent that accelerated pipe wall thinning had occurred, and that due to the similarity of design and-configuration between Units 1 and 2, Unit 1 could have the same problem.
The root cause of the #ailure appeared to be an erosion / corrosion mechanism that was not fully understood.
When the Augmented Inspection Team (AIT) arrived on site at 9:30 p.m. on December 9, considerations regarding shut down of Unit I were discussed with the licensee.
Immediately.after the pipe rupture event and throughout the evening of~ December 9, the plant staff was engaged in recovering from the accident and in bringing Unit 2 to a cold. shutdown.
At that time, the licensee did not want to undertake the added burden of the shutdown of the other unit, and reasoned that maintaining Unit 1 in a stable operational mode was the safer course of ~ action.
In light of the fact that the root cause of failure had not been determined and Unit I was operating normally, the AIT, with Regional concurrence, agreed with the licensee's conclusion.
Since single phase systems (containing water but no vapor) have historically not been susceptible to the types of failure
- mechanisms found in wet steam systems, it was thought that some unique flaw in the material might have caused the rupture.
The licensee had taken the prudent action to rope off the similar l
m 4
QUESTION 3(a).
(Continued). areas in Unit l 'and had stationed security personnel to' prohibit general entry into the area.
In addition,. safety systems in Unit I were not affected by the accident 1and no dependency on Unit 2 systems existed for Unit 1 safe shutdown.
. Shortly after the inspection of the Unit 2 ruptured piping, when it was determined that general thinning of the pipe wall could also.have occurred in Unit 1, the licensee decided to shut down-Unit 1 at 12:30 p.m. on December 10.
The unit was placed on a.
power ramp-down at 5:30 p.m. and achieved cold shutdown in~the~
morning of December 11.
In summary, the NRC believes the actions taken by'the licensee were prudent and actions of a more immediate nature were not warranted.
It is the licensee's responsibility to provide the hasis for continued reactor operation.
The NRC reviews such bases.
If we disagree with the licensee's decision regarding continued operations, an order.by the NRC to shut down would be appropriate.
~*
00ESTION 3(b).
Should the NPC issue any new guidance for such situations?
. ANSWER.
The Commission does not currentiv contemplate issuing-new guidance which would change the basic responsibilities of the licensee or the NRC's role in. event response as a result of the Surry incident.
For a laroe-variety of events the NRC is informed immediately.
The Incident Response Center at Headquarters. and the appropriate Region is activated to monitor the licensee's actions during the event.
If appropriate, a_ team of experts is sent to the affected plant site.
The potential interactions between units at a site have to be considered on a case-by-case basis.
The actions taken by the licensee following the Surry Unit 2. event were appropriate and would.not be a basis for developing additional guidance.
The operation of each unit is controlled 1by the Technical Specifications applicable to a particular unit.
The. Technical-Specifications are developed with the intent that the plant should not be allowed to operate when it is considered to be in a potentially-unsafe condition.
It is not desirable to develop generic guidance for shutting down the plant as each situation is unique and-it is dif ficult to foresee every situation in developing the generic guidance.
1
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LOUESTION 4
. Changes in~the; control room: ventilation system-were being; implemented while-the-plant was running at-the time.of the accident.
The:NRC inspection-
- team reports. conclude _-.that:the. modification: work resulted in the: control room-beingfflooded with potentially lethal carbon dioxide: gas.
- QUESTION 4(a).. Are NRC regulations adequateifor modifications being performed.while: plants are. operating?
Were.
these rules being-observed c.a t the' time of-the -
accident?
ANSWER.
The modifications being performed atDSurry were to a nonsafety-related system not described in-the. Final Safety Analysis. Report-(FSAR).
Where modifications-involve a-changezin the, facility as described.in the FSAR, 10 CFR'50.59-requires-that.
.an evaluation be performed by the licensee prior to' making the modifications to determine that the modifications do not-involve an unreviewed safety question or change.in the techn'ical l
specifications.
If the modification activities render.a safety _
system incapable of performing its intended. design safety function, the technical specifications require;that the system be-declared inoperable.
Depending-upon.the'particular safety system which has been declared. inoperable, the technical = specifications may require the plant to be shutdown within a soecified time-period.
Given this regulatory framework, NRC. believes that, with proper implementation of the technical specifications and 10 CFR 50.59, the regulations are adequate with respect to modifications for safety systems.
The ventilation system changes.that.were being made at the time of the accident were not in~a system described in the FSAR, nor.
did they directly impact the integrity of the control-room envelope.
Consequently, these changes-did not require an analysis under 10 CFR 50.59, nor were they a violation of the.
technical specifications.
Therefore, the'l.icensee was~in compliance with the regulations-for the' control rocm
~
modifications which were being performed.
As discussed in the answer to Question 4(b)Lthe operators were' not'following the procedures implicitly -- see1 answer to' Question 4(b) for the details.
/
s
- 00ESTION 4(b). Dofyou' feel that differentlprocedures.should?have-been used?' Is the_ Commission considering=any.
regulatory changes to prevent' ongoing modification work from compromising, operational safety?~
LANSWER.-
The accident resulted in the disabling of the key' card security system for the control-roon: doors.
The actions following the; accident and.the plant; response to it required-a considerable number of entries'into the control room.
The licensee apparently' believed.the actions to blockLthe doors open and' post a guard were proper.
The doors-remained open until-the operators
-discovered that there was a release-of carbon-dioxide and.that the doors were allowing.the carbon dioxide to reach the-cont'rol room.
Using hindsight, the control room envelope could possibly have been better maintained by having the guards maintain the control room doors closed and opening'them only to allow access.
as required for the accident mitigation actions.- The operators'
~
decision to close the control. room door and establish the safety grade pressurization and ventilation systems,1once the problem was discovered, was-the proper action to restore habitability..
The Commission is not considering any regulatory changes to prevent ongoing modification work from compromising operational safety.
Compliance with plant technical' specifications and 10 CFR-50.59 are considered adequate to ensure that required safety systems'are maintained operable at all times.
?
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QUESTION S.
The NRC inspection team reports. indicate the action was initiated by an improperly maintained valve.
QUESTION Sa.
Does it seem appropriate that the plant was allowed to operate with-this valve not functionina properly?
Are there adequate inspection requirements for such valves?
ANSWER.
As a point of clarification, the NRC inspection team report did not describe the valve as being improperly maintained;.it was improperly assembled during overhaul in November-1986.
After the overhaul the valve was tested in accordance with the Technical Specifications to verify its ability to close within the required time.
This valve is required to close in the event of a main steam line break in order to isolate the steam generator from the break location.
Even in its improperly assembled state, the valve would and did close properly as confirmed by testing in order to carry out this safety function.
The problem was that because of its improper assembly it was susceptible to premature closure with normal steam flow in the pipe.
There are adequate inspection requirements for such valves.
The Technical Specifications require a closure test (valve must close within five seconds) and a test to assure that the valve disc is free to move.
These tests verify that the valve will be able to carry out its safety function of rapid closure to nitigate the effects of a steam line break.
Even in its improperly assembled state, the valve was capable of carrying out this function.
The deficiencies lie in the procedure for valve assembly, and the inspection required after assembly.
The procedure was inadequate in that it did not prevent, nor did the post-maintenance testing discover, the improper assembly of the valve.
In addition, the maintenance procedure used to overhaul the valve was not correctly followed and non-routine work associated with the overhaul not adequately documented.
Because of these deficiencies, the licensee was issued a Notice of Violation.
The deficiencies have been corrected.
W'
QUESTION 5(b).
Does the Commission plan any regulatory changes as -a result of the maintenance deficiencies discovered during the investigation of this accident?
ANSWER.
The Commission does not plan any changes-as a result'of the-maintenance deficiencies discovered during the investigation'of this accident'.
The Commission recognizes the importance of-maintenance in the safe operation of nuclear power plants and has-focused added attention on these practices-in the industry.
The Commission 11s considering directing the staff to develop a policy statement that would emphasize the Commission's concern with overall industry maintenance performance and to indicate that, to the extent the self-improvement initiatives are effective and implemented n an. industry-wide basis, the NRC o
would defer development of new maintenance requirements.
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%'="' %,.'f # ' O Masfjington, BC 20515 nxomCo March 16, 1987 The Honorable Lando W. Zech Chairman Nuclear Regulatory Ccamission 1717 H Street Washington, DC 20555
Dear Chairman Zech:
The Subcommittee on Energy and Power is investigating the implications for the safety of nuclese power plants of the recent Surry accident. In particu-lar, we are concerned that (1) despite the designation of the f ailed feedwater line as "a nonsafety related system," a similar f ailure in a Boiling Water Reactor could result in the release of radioactive steam outside the contain-ment structure; and (2) standards established for new nuclear power plants and inspection procedures for operational plants may not adequately take into account the possibility of deterioration of materials.
We are requesting your response to the following questions:
1.
The NRC Augmented Inspection Team Reports Nos. 50-280/86-42 and 50-281/86-42 (NRC team reports) indicate that the failure at the Surry Station was caused by service induced deterioration of the feedwater suction line between the condenser and the feedwater pump.
(a) What codes, standards, specifications and regulatory requirements are applied to the f ailed feedwater line and associated equipment (condenser, feedwater pumps, steam turbine, pipelines and components)? Are these systems classified as nuclear or non-nuclear? Are they classified as safety or nonsafety related systems?
(b) Are these requirements different than those applicable to other por-tions of the feedwater and steam lines that are closer to the steam gen-erators and reactor vessel? If so, why are they, and do you think this distinction is appropriate in view of what occurred in the Surry Plant accident? What is the safety justification for the differences?
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- The Honorable Lando W.' Zech
-2
- March'16,'1987 (c) 'If a f ailure in -the feedwater piping occurred at' a similar ilocation, y e.g;,' between the condenser and f eedwater_ piping in a ~ Boiling Water Reactor :
nuclear power plant,' could radioactive material be~ released outside Lthe.
containment?
(i) fIf so, how muchL could be released ' and what would 'be' the. consequences -
to.the surrounding area?
(ii) How are these areas of the.feedwater and stesa lines classified 'in Boiling Water Reactors?
(iii) In. view of the Surry accident, do you think that the classifica '
tions of these areas of the power plant-(including the steam turbine, condenser'and feedwater pumps) are: appropriate?-
(d) What additiona1 ' requirements could be applied 'to"the feedwaternlines, steam lines, steam turbine, feedwater_ pumps, condenser-and-related equip-
-ment to. improve the safety of nuclear plant' operation?
(e) Does the Commission plan to make ~ any: changes in its regulatory require-ments for Surry or other nuclear power plants-in order to-implement lessons learned from the Surry_ accident?'
2.
The NRC team report cited erosion / corrosion induced thinning of.the' pipe.
metal as the cause of the failure at the Surry Station. Do the design, construction, maintenance and integrity monitoring codes, standards, or other regulations applied to nuclear power plants adequately provide fo'r-finding or make allowances for deterioration-of plant components and. piping-in service? If not, does the Commission plan any regulatory changes to-incorporate these factors in plant design, inspection and maintenance requirements?
3.
The two Surry Station nuclear-units are very similar in-design, nuclear reactor system and age. The units also " share" some support and auxiliary functions.
(a) In view of this dependency, can you explain why Unit 1'was not shut down immediately when the failure occurred in Unit 2?- Whose responsibility was it to decide whether or not to shut it down immediately? In your view, should Unit I have been shut down immediately?
(b) Should the NRC issue any new regulatory guidance for such situations?
- 1 i
4.
Changes in the control room ventilation system were being implemented while the plant was running and at the time of the accident. The NRC inspection team reports conclude that the modification' work resulted in the control' room being flooded with potentially lethal carbon dioxide gas.
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The Honorable Lando W..Zech
'^3-March 16, 1987, (a) Are NRC regulations adequate for modifications being performed while
. plants are operating? ' Were these rules being observed at the time of the -
accident?
(b) DoLyou feel that different-procedures'should have been used?. Is the
~
. Commission considering any regulatory 1 changes to prevent ongoing modifica-tion' work from compromising operational safety?
5.
The NRC inspection team reports indicate the accident was initiated by an
- improperly _ maintained valve.
-(a) Does it seem' appropriate'.that the plant was allowed to operate with this valve not' functioning properly? Are there adequate inspection -
requirements for such. valves?
(b) Does the Connaission plan any regulatory changes as a result of the maintenance deficiencies. discovered during the-investigation of-this accident?
Thank you for your assistance with this investigation. We would appreciate having your response no later than April 10.
Sincerely.
pp Phil'
. Sharp Chairman PRS:bh I
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