ML20134C959
| ML20134C959 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 10/09/1996 |
| From: | Taylor J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Petersen F OMAHA PUBLIC POWER DISTRICT |
| References | |
| NUDOCS 9610110181 | |
| Download: ML20134C959 (13) | |
Text
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, A P Ef og I g IT UNITED STATES j
- r NUCLEAR REGULATORY COMMISSION 1
- WASHINGTON, D.C. 20666 4 001 2
k.....f October 9, 1996 I
- l Mr. F. M. Petersen President and Chief Executive Officer )
l Omaha Public Power District l l 444 South 16th Street Mall l Omaha, NE 68102-2247
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l
SUBJECT:
REQUEST FOR INFORMATION PURSUANT TO 10 CFR 50.54(f) REGARDING ;
l ADEQUACY AND AVAILABILITY OF DESIGN BASES INFORMATION
Dear Mr. Petersen:
The purpose of this letter is to require information that will provide the U.S. Nuclear Regulatory Commission (NRC) added confidence and assurance that l your plant (s) are operated and maintained within the design bases and any 1 deviations are reconciled in a timely manner.
Backaround In the mid- to late 1980s, NRC safety system functional inspections (SSFIs) and safety systems outage modifications inspections (SS0 mis) identified concerns that design bases information was not being properly maintained and plant modifications were being made without the licensee having an understanding of the plant design bases. The NRC's findings heightened the nuclear industry's awareness of the need to improve the adequacy and availability of design documentation, and many licensees voluntarily initiated ,
extensive efforts to improve the design bases information for their plants. ;
To assist the industry in performing design bases improvement programs, the l Nuclear Management and Resources Council (NUMARC)' developed a guidance document, NUMARC 90-12, " Design Basis Program Guidelines." These guidelines were intended to provide a stanpard framework for licensee programs to improve plant design bases information. The NRC staff reviewed the guidelines and provided comments to NUMARC in November 1990. In emphasizing the importance of validating the facility against current design information, the staff
'NUMARC was consolidated into the Nuclear Energy Institute (NEI) on March 23, 1994. 4 l
l 2As discussed in NUMARC 90-12, these programs or efforts would emphasize l collation of design basis information and the supporting design information, not the identification or re-creation of the licensing basis for a plant or the regeneration of missing analyses and calculations.
9610110181 961009 bl I PDR ADOCK 05000285 P PDR f
Mr. F. M. Petersen i l
! stated that the goal of any program should be to establish confidence that the I
' existing facility is in accordance with the current design documents and that any deviatinns will be reconciled. The staff concluded that the NUMARC guidelines would provide worthwhile insights to utilities undertaking design l reconstitution programs and that the guidelines appeared to provide sufficient flexibility for licensees to structure their programs to respond most efficiently to any unique needs and circumstances of a particular licensee.
The staff requested NUMARC to consider making design reconstitution a formal NUMARC initiative and commented that design documents that support technical specification values and that are necessary to support operations or to respond to events should be regenerated if missing. NUMARC subsequently concluded that a formal initiative was not necessary because most of its members were already conducting or evaluating the need to conduct design reconstitution programs, and agreed to forward the guidelines, with the NRC's comments, to its members for use on a voluntary basis.
To provide more information to the industry on this topic and to provide an independent view of the design control issue, the staff conducted a survey of six utilities and one nuclear steam supply system vendor to determine the status of design control problems and the strengths and weaknesses of the sample utility programs. The results were published in February 1991 in NUREG-1397, "An Assessment of Design Control Practices and Design Reconstitution Programs in the Nuclear Industry." The survey observations were as follows:
The need for a design documentation reconstitution program was directly proportional to the age of the plant.
The general intent of the program should be to provide a central location for design bases information, with emphasis on the design intent (the "why" of the design).
The design bases documents should be a top-level directory that defines the current plant configuration.
Reestablishment of design bases without reconstitution of the supporting design documents, as necessary, may not provide a sufficient level of information for future modifications or current plant operation, or to quickly respond to operating events.
Minor changes to the design should be tracked to support the conclusion that the changes in the aggregate do not affect the validity of existing calculations and the ability of a system to perform its design functions. ~
~
. Mr. F. M. Petersen !
- Some common weaknesses of licensee programs identified during the survey
- included the following
- -
= Design reconstitution programs had not identified in advance the documents that are necessary to demonstrate that a structure, a system, or a component will function properly.
- The process for regenerating missing design documentation was not always proceduralized so that it could be handled in a systematic manner.
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- Validation of the content of specific output documentation was not always thoroughly carried out.
l In late 1991, the NRC staff evaluated whether rulemaking, guidance, or a policy statement was needed to address the issue of licensees retaining accurate design bases information. It concluded that the existing regulatory l requirements for design control were adequate; however, it determined that the publication of a policy statement addressing design bases information and publication of a generic letter requesting licensees to describe their design !
reconstitution programs would be beneficial. Additionally, the staff stated '
its intention to continue to evaluate design control adequacy during its performance-based inspections such as SSFIs and SS0 mis. The staff also l expectedthattheenforcempntpolicyguidancetoprovidegreateropportunities !
for enforcement discretion would encourage voluntary identification of past i design, engineering, and installation issues by licensees. With the Commission's approval, the staff proceeded with this approach.
In August 1992, the NRC issued a Commission policy statement " Availability and Adequacy of Design Bases Information at Nuclear Power Plants" (57 FR 35455)
(Attachment 1). This policy statement stressed the importance of maintaining current and accessible design documentation to ensure that (1) plant physical and functional characteristics are maintained and consistent with design bases, (2) systems, structures, and components can perform their intended functions, and (3) the plant is operated in a manner consistent with the design bases. In the policy statement, the Commission recomended that all power reactor licensees assess the accessibility and adequacy of their design bases information and that they be able to show that there is sufficient ,
documentation to conclude that the current facility configuration is i consistent with the design bases. The policy statement outlined the '
additional actions the NRC would take to keep apprised of the. industry's design reconstitution activities previously discussed.
Following review by the Committee To Review Generic Requirements (CRGR) and the Commission, a draft generic letter was issued for public comment on March 24, 1993. The proposed generic letter requested licensees, on a 3
NRC would refrain from imposing civil penalties for violations up to Severity Level If if the violations were identified and corrected as a result of systematic voluntary initiatives.
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i j .
}, Mr. F. M. Petersen i i voluntary basis, to submit information and schedules for any design bases l programs completed, planned, or being conducted, or a rationale for not implementing such a program. All but one of the commenters concluded that the i
! generic letter was unnecessary and unwarranted. NUMARC responded that it l l believed the NRC's request for descriptions, schedules, and dates would have a '
a negative impact on ongoing design efforts and that NRC's focus on schedules
! would undermine the licensees' ability to manage the activities. In
! SECY-93-292, " Generic Letter on the Availability and Adequacy of Design Bases i
' Information," dated October 21, 1993, the staff recommended that the generic I
letter not be issued.^ihe staff stated that publication of the policy statement and the proposed generic letter conveyed to the industry the Commission's concern and that publication of the generic letter would not further licensees' awareness of the importance of the activities. The staff proposed to continue performing design-related inspections and to gather information and insights as to how well the licensees' design-related programs were being implemented. The Commission issued a staff requirements memorandum that agreed with the staff's proposal.
In response to the findings relating to the regulatory burden of team inspections identified in the 1991 Regulatory Impact Survey, during the past several years the staff has reduced its effort on specific, resource-intensive, design-related team inspections, and followed the issue of accurate and accessible design documentation at plants principally as an element of inspection and followup of operations-related activities. The issuance of the NUMARC guidelines and ongoing industry efforts to improve and maintain design bases information also contributed to this decision.
Current Problem Over the past several months, NRC's findings during inspections and reviews have identified broad programmatic weaknesses that have resulted in design and configuration deficiencies at some plants, which could impact the operability of required equipment, raise unreviewed safety questions, or indicate discrepancies between the plant's updated final safety analysis report (UFSAR) and the as-built or as-modified plant or plant operating procedures. These i inspections and reviews have also highlighted numerous instances in which !
timely and complete implementation of corrective action for known degraded and nonconforming conditions and for past violations of NRC requirements has not been evident. Overall, the NRC staff has found that some licensees h failed to (1) appropriately maintain or adhere to plant design bases,pve
'As described in 10 CFR 50.2, design bases is defined as, " Design bases mean that information which identifies the specific functions to be performed by a structure, system, or component of a facility, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design..." The design bases of a facility, as so defined, is a subset of the licensing basis and is contained in the FSAR. Information developed to implement the design bases is contained in other documents, some of which are docketed and some of which are retained by the licensee.
j
, Mr. F. M. Petersen (2) appropriately maintain or adhere to the plant licensing basis,5
- (3) comply with the tegms and conditions of licenses and NRC regulations, and
- (4) assure that UFSARs properly reflect the facilities. Attachment 2 i provides examqples of some of the deficiencies recently identified by the :
staff. As a consequence of this new information, the NRC believes that the industry's voluntary efforts to improve and maintain design bases information ,
for their plaats, consistent with NUMARC 90-12, the staff's coments on the l industry guidelines, and the Comission policy statement, have not been '
effective in all cases.
The magnitude and scope of the problems that the NRC staff has identified raise concerns about the presence of similar design, configuration, and i
, operability problems and the effectiveness of quality assurance programs at l j other plants. Of particular concern is whether licensee programs to maintain configuration control at plants licensed to operate are sufficient to demonstrate that plant physical and functional characteristics are consistent
- with and are being maintained in accordance with their design bases. The extent of the licensees' failures to maintain control and to identify and correct the failures in a timely manner is of concern because of the potential impact on public health and safety should safety systems not respond to challenges from off-normal and accident conditions.
It is emphasized that the EC's position has been, and continues to be, that
! it is the responsibility of individual licensees to know their licensing basis, to have appropriate documentation that defines their design bases, and
- to have procedures for performing the necessary assessments of plant or i procedure changes required by NRC regulations. Attachments 3 and 4 are a 4
recent exchange of correspondence between J. Colvin, NEI, and j Chairman S. Jackson, NRC, regarding these subjects.
5 The licensing basis for a plant originally consists of that set of information upon which the Commission, in issuing an initial operating
, license, based its comprehensive determination that the design, construction,
- and proposed operation of the facility satisfied the Comission's requirements l and provided reasonable assurance of adequate protection to public health and safety and common defense and security. The licensing basis evolves and is I modified throughout a plant's licensing term as a result of the Comission's continuing regulatory activities, as well as the activities of the licensee.
'The FSAR is required to be included in, and is one portion of, an application for an operating license (OL) for a production or utilization facility.
10 CFR 50.34(b) describes the information which must be included in an FSAR.
The FSAR is the principal document upon which the Comission bases a decision to issue an OL and is, as such, part of the licensing basis of a facility. It is also a basic document used by NRC inspectors to determine whether the facility has been constructed and is operating within the license conditions.
( .
Mr. F. H. Petersen l Action The NRC has concluded that it requires information that can be used to verify compliance with the terms and conditions of your license (s) and NRC regulations, and that the plant UFSAR(s) properly dpscribe the facilities,s as well as to determine if other inspection activities or enforcement action should be taken. Therefore, you are required, pursuant to Section 182(a) of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f), to submit a response to this letter within 120 days of its receipt. Your response must be written and signed under oath or affirmation.
Please submit the original copy of your response to the NRC Document Control Desk, and send a copy to the Director, Office of Nuclear Reactor Regulation and to the appropriate regional administrator. The following information is required for each licensed unit:
(a) Description of engineering design and configuration control processes, including those that implement 10 CFR 50.59, 10 CFR 50.71(e), a'id Appendix B to 10 CFR Part 50; (b) Rationale for concluding that design bases requirements are translated into operating, maintenance, and te:, ting procedures; (c) Rationale for concluding that system, structure, and component configuration and performance are consistent with the design bases; (d) Processes for identification of problems and implementation of corrective actions, including actions to determine the extent of problems, action to prevent recurrence, and reporting to NRC; and l
l 7
A number of design bases inspections are being planned, and your response will be used in the planning process.
sSection VII.B.3 of the NRC Enforcement Policy addresses how old design issues involving past problems in engineering, design, or installation are to be handled from an enforcement standpoint. In a related matter, the Commission recently approved changes that would modify this policy to encourage licensees to undert de voluntary initiatives to identify and correct FSAR noncompliances by (1) the exercise of discretion to refrain from issuing civil penalties for a two-year period where a licensee undertakes a voluntary initiative in this area and (2) the exercise of discretion to escalate the amount of civil penalties for violations associated with departures from the FSAR identified by the NRC subsequent to the two-year voluntary initiative period.
l l' Mr. F. M. Petersen .
(e) The overall effectiveness of your current processes and programs in concluding that.the configuration of your plant (s) is consistent with the design bases.
1 In responding to items (a) through (e), indicate whether you have undertaken a
any design review or reconstitution programs, and if not, a rationale for not implementing such a program. If design review or reconstitution programs have been completed or are being conducted, provide a description of the review programs, including identification of the systems, structures, and components (SSCs), and plant-level design attributes (e.g., seismic, high-energy line break, moderate-energy line break). The description should include how the program ensures the correctness and accessibility of the design bases information for your plant and that the design bases remain current. If the program is being conducted but has not been completed, provide an implementation schedule for SSCs and plant-level design attribute reviews, the expected completion date, and method of SSC prioritization used for the review.
This request is covered by the Office of Management and Budget (0MB) clearance number 3150-0011, which expires July 31, 1997. The reporting burden for this collection of information is estimated to average 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> per response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information. Send comments regarding this burden estimate or any other aspect of this collection of information, including suggestions for reducing this burden, to the Information and Records Management Branch (T-6 F33), U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001, and to the Desk Officer, Office of Information and Regulatory Affairs, NE08-10202 (3150-0011), Office of Management and Budget, Washington, D.C. 20503. The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number.
In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and your response will be placed in the NRC Public Document Room (PDR), the Gelman Building, 2120 L Street, N.W., Washington, DC, and in the local public document room (s) for your facility or facilities.
l
, Mr. F. M. Petersen 8-If you have any questions about this matter, please contact the staff members listed below, or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
Sincerely,
/
mes M Executive Director for Operations !
Docket No. 50-285 Attachments: !
- l. Policy Statement on Availability and Adequacy of Design Bases Information at Nuclear Power Plants
- 2. Background Information on Recently Identified Problems
- 3. Letter from J. Colvin (NEI) to Chairman S. Jackson (NRC) j dated 8/2/96 1 4 Letter from Chairman S. Jackson (NRC) to J. Colvin (NEI) dated 8/14/96 Contacts: Kristine M. Thomas, NRR (301) 415-1362 Internet: kmtenrc. gov Eileen M. McKenna, NRR (301) 415-2189 Internet: emm0nrc. gov cc w/atts: See next page
OMAHA PUBLIC POWER DISTRICT FT. CALHOUN STATION, UNIT NO. I cc:
Winston & Strawn ATTN: Perry D. Robinson, Esq.
1400 L Street, N.W.
Washington, DC 20005-3502 Mr. Jack Jensen, Chairman Washington County Board of Supervisors Blair, Nebraska 68008 Mr. Wayne Walker, Resident Inspector U.S. Nuclear Regulatory Commission Post Office Box 309 Fort Calhoun, Nebraska 68023 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Ms. Cheryl Rodgers, LLRW Program Manager Environmental Protection Section Nebraska Department of Health 301 Centennial Mall, South P.O. Box 95007 Lincoln, Nebraska 68509-5007 Mr. James W. Chase, Manager Fort Calhoun Station Post Office Box 399 Fort Calhoun, Nebraska 68023 Mr. Terry L. Patterson Division Nanager - Nuclear Operations Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.
Post Office Box 399 Hwy. 75 - North of Fort Calhoun Fort Calhoun, Nebraska 68023-0399
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35453 Rules and Regulations r d- i a+'a Vol. 57. No 154 Monday. August to.1992 This section of the FEDERAL REGISTER Program Cuidelines." NUMARC 90-12. functional characteristics ere l i
contains regulatory documents having While developing these guidelines. maintained and are consistent with the I
generat apphcabit.ty and legal effect most NUMARC discussed them at several design bases as required by NRC of which are keyed to and codified in public meetings held with ths NRC.The regulation. (2) systems. structures. and the Code of Federal Regulatons, which is staff has concluded the NUMARC components can perform their intended shed under 50 t: ties pursuant to 44 uidelines provide a useful standard functions, and (3) the plant is operated ramework for implementing design in a manner consistent with the design The Code of Federal Regulations is sold by the Supenntendent of Documents. reconstitutiori programs.The staff also bases.The Commission believes the I
Pnces of new books are listed in the agrees no single approach would enable regulatory framework already exists to utilities to best accomplish the address the need for accessible design l first FEDERAL REGISTER issue of each week reconstitution task.The NUMARC bases and control of design information. I guidance oppeared to provide sufficient The availability of current design and flexibility for individual utilities to licensing bases will also expedite the NUCLE AR REGULATORY ,.ructure their programs to respond most license renewal process.
COMMISSION efficiently to their unique needs and The Commission beiieves. as a result j circumstances. of NRC inspections and licensees
- self-10 CFR Part 50 The staff sent comments on the assessments, that all power reactor guidelines to NUMARC on November 9. licensees should assess the accessibility Availability and Adequacy of Design 1990. Commission paper SECY-90-365 and adequacy of their design bases Bases Information at Nuclear Power informed the Commissioners m advance documentation.The results of this self. ,
Plants; Policy Statement about the staff response to NUMARC. i assessment should form the basis for a AoENev: Nuclear Regulatory The staff requested NUMARC licensee's decision whether a design Commission. consider making the design bases effort reconstitution program is necessary and ACTION: Policy statement. a NUM ARC initiative. NUMARC the attributes to be included in the concluded they would not pursue a program.The Commission recognizes sVMM ARY: The Nuclear Regulatory formal initiative. but would forward the the need for a design reconstitution Commission is issuing this policy guidelines to their members to use on a program to be tailored to meet the statement on availability and adequacy voluntary basis. Their reason for not unique needs of a particular utility.The of design information at nuclear power pursuing an initiative was that most of structure and content of the design plants. This policy statement describes their members were already conducting document reconstitution program will be the Commission's expectations and or evaluating the need to conduct design influenced by various factors. such as future actions with regard to the bases reconstitution programs. the utility's organizational structure. the i availability of design information and The Commission's ev0uation of the availability er unavailability of design l I
emphasizes the Commission s view that status of reconstitution programs clearly documentation, and the intended users facilities should nut be modified without indicates the licensees' substantial of the documentation.The Commission a clear understanding of the applicable investment in these programs should expects that after completing a engineering design bases. yield positive safety benefits for a reconstitution program. or as a basis for EFFECTIVE DATE: August 10.1992. majority of sites.The NRC commends concluding that such a program is l FOR FURTHER INFORMATION CONTACT: those licensees that are acting to ensure unnecessary. the licensee will have Eugene V. Imbro. Office of Nuclear technically adequate and accessible current design documents and adequate Reactor Regulation. U.S. Nuclear design bases documentation is technical bases to demonstrate that the Regulatory Commission. Washington. maintained. plant physical and functional However, the Commission is characteristics are consistent with the DC 20555. telephone [301) 504-2907.
concerned some situations exist where design basis. the systems, structures. l sVPPLEMENT ARv INFORM ATION: NRC liC'.nsees have no' critically examined and components can perform their inspection findings have demc...trated intended functions md the plant is being that some licensees have not adequatelv their design control and configuration
- management processes to identify operated in a mannte consistent with maintained their design bases .
information as required by NRC requisite measures to ensure the plant is the design basis.
NUMARC has developed guidance for regulations. Bnth the problems identified operating within the deugn bases .
envelope. Therefore, the Commission is the conduct of design bases during the NRC inspections and those articulating its expectations with regard reconstitution programs.The guidance identified by licensees have prompted to design information and elaborating on outlines a framework to organize and niost power reactor licensees to initiate, its planned activities to confirm the collate nuclear power plant design bases over the past several years. design information.This information provides integrity of the as. configured plant with bases reconstitution programs. To the rationale for the design bases respect to the plant design bases-implement a reconstitution program. consistent with the definition of design licensees seek to identify missing design Policy Statement documenistion and to selectively bases contained in 10 CFR 50.2.
Posin.on NUMARC 90-12," Design Basis Program regenerate missing documentstion as required. The Commission has concluded that Guidelines." was issued in October 1990 in 1989. Nuclear Utilities Management maintaining current and accessible for voluntary use by NUMARC memb:r and Resources Council. Inc.. [NUMARC) design documentation is important to organizations as a reference point from began developing their " Design Basis ensure that (1) the plant physical and which licensees would review their s .tioo9e noonooxuv. AUG 02-12 nn 2,3 Attachment 1
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35456 Federal Regist:r / Vol. 57. No.154 / Monday. August 10. 1992 / Rules and Regulations existing or planned efforts to collate ensuie their validity for the life of the DEPARTMENT OF THE TREASURY supporting design information. The facility, including any renewal period.
Commission believes NUMARC's in order to ensure the Commission is Office of Thrift Supervlaton approach provides a useful framework appraised of mdustry's activities, the 12 CFR Part 584 and worthwhile insights to those utilities NRC will take the following actions.
undertaking design basis programs. (1)The staff willissue a generic letter [92-1951 The Commission believes a licensee requesting slllicensees to describe the should be able to show that it has programs that are in place to ensure MN 1550-AA38 sufficient documentation, including design information is correct. accessible. Reglatration, Examination and calculations or pre-operational. etartup and maintained current. Those licensees Reporta; Statements, App!! cations, or surveillance test data to conclude the that are not implementing a design Reporte and Notices To Be Flied current facility configuration is reconstitution program will be requested consistent with its design bases. The to provide their rationale for not doing AotNcy: Office of Thrift Supervision.
Commission further believes the design so. lf a reconstitution program is under Treasury, bases must be understood and way, the schedule for implementation AcTiow: Final rule.
documented to support operability and completion will be requested.
determinations and 10 CFR 50.59 SUMMany: The Office of Thrift evaluations that may need to be made (2)The staff will prioritize NRC Supervision (OTS)is hereby amending quickly in respending to plant events. inspections of licensee's managemeni of its regulations pertaining to holding The design bases related information design and configuration using SSF1- company reporting requirements. In should be retnevable within a type techniques based upon responsca updating existing forms to reflect reasonable period of tinie. however,it is to the generic letter and other plant changes necessitated by the Financial not necessary for all design basis specific information known to the NRC. Institutions Reform. Recovery, and documantation to be organized in one Additional staff guidance will be Enforcement Act of1989, the OTS has place. The information used solely to developed. where needed, for the design combined several forms to streamline support the development of a bases aspects of these inspections. the reporting process and ease the modification package would not need t regulat ry burden on savings and loan be able to be retrieved as expeditiously (3IThe NRC systematic assessment of holding companies. In particular, the as information needed to support an hcensu performance (SAZ pmcess reporting requirements set forth in will be modified to explicitly address operability determination. Forms H-(b)3. H-(b)4. H-(b)5 and H-assessment of licensee programs t (b)10 Registration Otatements are now
, in the event the design bases control design bases information that contained in one body of instructions for information is found technically reflect NRC inspection activity in this all Registrants, the H4b)10. In addition.
madequate or not accessible, licensees area and assure consistenl evaluations. the H-(b)11 Annual Report and the H-should consider whether remedial action (b)12 Current Report have been merged is warranted. A methodology should be (41 The staff will continue t into one set of instructions requiring an developed and implemented to ensure encourage self. identification of design annual filing with quarterly updates licensee resources are focused on design bases issues through apylication of the informing the OTS of any changes.The mformation regeneration in a timeframe provisions of the Commysion s commensurate with the safety enforcement policy.The staff will. li-(f) Dividend Notification has been re=tinded. since the requirements significance of the missing or erroneous however. pursue enfo cement actions for contained in the Capital Distributions information. engineering deficiencies whose root regulation are sufficient for the OTS's The Commission also emphasizes it is cause lies in the inadequacy or monitoring and supervision purposes.
very important that modifications to a ynavailability of design bases arrscTive DATE: September 9.1992.
facility be made after a thorough review information and which are identified e m T.
has been conducted and an durmg NRC inspections.
Michael P. Scott. Program Manager, understanding of the applicable (202) 900 -5748. Supervision Policy.
underlying design bases has been Paperwork Raduct. ion Act Statement Office of Thrift Supervision.1700 C gained in order to ensure appreoriate This final polic) statement does not Street. NW., Wa shington. DC 20552.
design margins are preserved. contain a new or amended information SUPPLEMENTARY INF ORM ATION:
collection requirement subject to the future Aetions ackgmund Paperwork Reduction Act of1980(44 U.S C. 3501 et seq.). Existing The OTS is today issuing a final rule The Commission will continue to requirements were approved by the amending its holding company reporting inspect routinely the adequacy of design Office of Management and Budget requirements.This amendment affects control program effectiveness. The approval number 315M)011. the registration, annual, and current Commission concludes that ensuring the reporting requirements.
design bases and configuration of a Dated at Rockville. Af arytend, this 4th day .
facility are well understood and Registration Stolements of August. W92.
controlled in plant documents will also As previously structured, holding ensure that those parts of the current 8'or the Nuclear Regulatory Com nission. companies were required to choose from licensing bases of most safety four separate registra, tion statements.
significance are understood and Samuel l. Chilk. These separate statements were controlled. Other aspects of the current Secretary of the Commissroa. originally deemed necessary to licensing bases, such as emergency preparedness and security plans, should IFR Doc. 92-1s895 riled a-7-s2. e 45 a il "C " ** #
companies (i e., companies that become also be appropriately examined to amo coc4 rsso ot-" savings and loan hold:ng companies as 5-310999 0002(00XO7-AUG 92-12 00 32) 47m FMT (1 A 'Lol 4 't102
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BACKGROUND INFORMATION ON RECENTLY 3DENTIFIED PROBLEMS Over the past several months, design and engineering information has been obtained that indicates that design bases at certain plants have not been appropriately maintained or adhered to. Specific examples follow:
Millstone Units 1. 2. and 3 An NRC inspection team recently found examples in which design bases information and the Updated Final Safety Analysis Report (UFSAR) did not agree with the as-built plant, operational procedures, and maintenance practices.
The team found inconsistencies that required analyses, procedure changes, and design changes to resolve. For example, the Millstone Unit 3 operating procedures required isolation for the turbine-driven auxiliary feedwater pump during certain plant conditions, in conflict with technical specification requirements for operability. The team found that certain protective relays at Millstone Unit 3 were not set in accordance with the design bases information. This required re-analyses and resetting of certain relays.
Based on the team's findings, the licensee initiated design changes to correct nonconfoming conditions between the UFSAR and the as-built plant, including changes to the design of the Millstone Unit 2 reactor protection system to meet the design bases with respect to physical separation of redundant channels and changes to the design of the Millstone Unit 2 (post-loss-of-coolant accident (LOCA)) hydrogen monitors to meet the design bases for single failure vulnerabilities.
i Haddam Neck An NRC inspection team found examples in which the design bases information and the UFSAR did not agree with the as-built plant, operational procedures, and maintenance practices. The team identified a number of deficiencies in engineering calculations and analyses that were relied upon to ensure the adequacy of the design of key safety systems. Deficiencies were identified in the calculations and analyses supporting the station batteries, emergency diesel generators, containment cooling system, and other key safety systems.
In some cases, the inspection findings were resolved by revising the calculations and analyses. .In other cases, procedure and design changes were required to resolve the issues. For example, the team identified that the design bases calculations supporting the size of the station batteries were inconsistent with the design bases stated in the FSAR. Field measurements and design modifications were required to resolve this issue.
Other issues were identified by the NRC and the licensee following the issuance of this special team inspection report that led the licensee to enter a refueling outage earlier than originally scheduled. Discrepancies included inadequate configuration management of the containment sump design and as-built conditions; a lack of detailed analysis and technical justification for the reliance on post-accident back pressure inside the containment to assure adequate net positive suction head for the residual heat removal pumps; inadequate inspection and verification of the sump as-built and material Attachment 2
conditions; and the lack of aggressive action in response to generic *
. cosaunications of industry events, which contributed to an inadequate l operability determination regarding the sump screen design and mesh size.
These issues impacted the operability of the emergency core cooling systems l (ECCSs) under certain postulated design basis events.
t Maine Yankee On January 10, 1996, the NRC issued a Confirmatory Order Suspending Authority J for and Limiting Power Operation and Containment Pressure and a Demand for
!- Information to the Maine Yankee Atomic Power Company. The order was based, in I part, on the NRC's determination that Maine Yankee did not apply a computer i code that was proposed to demonstrate compliance with the ECCS requirements of
! 10 CFR 50.46 in a manner that conforms to the requirements of 10 CFR Part 50, Appendix K, nor to the conditions specified in the staff's safety evaluation
! . dated January 30, 1989. Specifically, the licensee did not demonstrate that 4 the RELAP5YA code will reliably calculate the peak cladding temperature for i all break sizes in the small-break LOCA spectrum for Maine Yankee, nor has the
! licensee submitted the justification for the code options selected and other
- justifications and sensitivity studies to satisfy conditions in the staff's
- safety evaluation.
' In addition, the licensee assumed an initial containment pressure of 2.0 psig for calculating peak design-basis accident pressure, even though the plant's technical specifications allow a maximum operating pressure in containment of j 3.0 psig. Assuming an initial containment pressure of 3.0 psig results in a 1 calculated peak accident pressure in excess of the containment design pressure l l described in the UFSAR. _
1 i Refuelina Practices Survey j
) In a survey of licensee refueling practices conducted during the spring of ;
j 1996, the NRC identified deficiencies in the management of design bases
! assumptions. Many plants were found to have aspects of their design bases j that were only loosely proceduralized or not proceduralized at all. Typical of this kind of discrepancy was the identification of a lack of procedures for
- controlling the assumptions regarding hold-up time before beginning fuel
- transfer. The NRC found a number of instances in which other design bases assumptions were not captur'ed in procedures. In addition, it was necessary for licensees at 12 sites (23 units) to upgrade procedures to directly implement the design bases assumptions. In other cases, the licensee
! performed engineering analyses, documented pursuant to 10 CFR 50.59, as
! necessary, to ensure that the planned activities would not exceed design hases j assumptions. .
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,i August 2.199G I J
The Honorable Shirley A. Jackson Chairman U.S. Nuclear Regulatory Commission Washington, DC 20555 0001 l
Dear Chairman Jackson:
On July 23.1996, the Nuclear Strategic Issues Advisory Committee (NSIAC) held a special meeting to discuss industry actions for ensuring the licensing basis of nuclear plants is being maintained consistent with the regulations. The chief nuclear officers of the nuclear utilities approved an initiative to provide additional assurance and confidence that existing programs are adequate to ensure that: ,
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. Licensees are operating their plants in conformance with their licensing basis:
. Liceasees are adequately maintaining their licensing basis:
. There are no differences between operating practices and the licensing basis that could result in a significant public health and safety concern; and
. Degraded or non conforming conditions are captured on tracking systems
- and resolved in a timely manner.
Under the initiative. each licensee will conduct an assessment of the programs in place to reaffirm that plants are operated in conformance with their licensing basis.
The program assecsment is accomplished by sampling (1) FSAR information. (2) programs in place for processing changes to procedures and the plant that may impact the FSAR, and (3) changes that may not be governed by licensee programs.
Many licensees have already commenced or recently completed similar program assessments.
Differences identified through the sampling process will be evaluated to assess the overall effectiveness and adequacy of programs. Programmatic enhancements will be made if needed. The overall significance of any differences will be characterized to determine if additional sampling is warranted.
Differences that represent degraded or nonconforming conditions will be captured on a nacking system and resolved in a timely manner. Each licensee will notify the NRC of any differences that are subject to reportim requirements per the regulations.
Abo &9&O-t3Sb-- O Pa '
assara m s 3
o l The Honorable Shirlev A. Jackson '
n i August 2.1996 s
i Page 2 The results of the tndmdual plant assessments will be provided to NEI for l
j compdation in an industry summary report. The summary report wid be provided to the NRC for information '
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! The industry has developed a drait guidehne to assist utilities in performing the programmat.c assessment. The guidehne will be Snalized subsequent to resolution of the issues discussed below. Enclosure 1 provides a copy of NEI 96 05 (Draft)
Revision D. "Guidehnes for Assessing Programs for hiaintaining the Licensing
- Basis. ' for your information.
There are a number of related I'ssues that rec >uire a mutual agreement between the
! industry and the NRC before the industrv can proceed with the initiative in a >
! consistent manner These include:
i . The legal standing of the Updated Final Safety Analysis Report and the NRC 5 Safety Evaluation Reports. ;
. The scope of what constitutes the current bcensing basis and
! . The adequacy of NSAC 125 for performing 10 CFR 50.59 safety evaluations The industry is committed to completing the initiative within six months of the 4
resolution of these issues. Initial progress in resching these issues was accomphshed dunng the NRC pubbe meeting also held on July 23.1996. We appreciate very much the participation of your senior staffin this meeting and the opportunity to discuss these issues. Expedited closure on these fundamentalissues
' is needed to minimize confusion within the NRC repons and the plants and focus
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our resources on addressing the programmatic issues.
We look forward to a continuing dialogue with the Commission and NRC staff to achieve timely resolution of this matter.
Sincerely. !
- f -
(Nq Joe F Colvin l
Ynclosure c Commissioner Kenneth r Rogers. NRC Commissioner Gretn J. Dieus. NRC Sir lamn .1\ Taylor NRC
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NEI 96-05 (DRAFT) .
Revision D l i
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NUCLEAR ENERGY INSTITUTE ,
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GUIDELINES FOR ASSESSING PROGRAMS FOR -
MAINTAINING THE LICENSING BASIS l
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JULY 25,1996 q grot 93t?*2- ', 9 f
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1.0 INTRODUCTION
Licensees have implemented various programs / processes over the period of thetr licenses for ensuring that the licensing basis is known, is maintained current and is accurately described in the FSAR and associated referenced documents. The NRC.
through a review described in SECY 94 066, determined that licensees were adequately saintaining the licensing basis of their facilities. As a result of events j
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that have occurred over the last several months.the NRC has increased attention on licensee compliance with the plant's hcensing basis. Differences can occur due to problems in those processes / programs intended to effect changes in the license i,
basis, missing programmatic controls to translate changed operational activities into the licensing basis, or a lack of awareness oflicensing basis features.
Licensees employ various programmatic mechanisms for identifying, evaluating j and/or processing changes in plant activities that could affect the licensing basis or that are intended to make changes to the licensmg basis. These include: !
. Final Safety Analysis Report (FSAR) updates per 10 CFR 50.71(el l
! = Safety Evaluations per 10 CFR 50 59 i
e Changes to Quahty Assurance. Security and Emergency Plans per 4 10 CFR 50.54
. License amendments per 10 CFR 50.90
. ASME Code rehef requests per 10 CFR 50.55a e rocedure changes (not subject to 50.59)
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. Design changes (not subject to 50.59) 1
. Regulatory commitment changes
. Degraded or non conforming conditions 3
. Con 5guration management changes (e.g., operator workarounds, extended i operation of equipment in manual mode rather than automatic operation modes, and Technical Speci6 cation positions / interpretations) 1
l 2.0 PURPOSE AND SCOPE 2.1 Purpose This document provides guidance for performing a self assessment of the adequacy l of programmatic controls for maintaining the licensing basis in order to identify missing or incorrectly applied programmatic elements that can lead to licensing basis differences.
2.2 Scope The assessmert approach consists of a data gathering phase and an evaluation phase. The data gathering phase employs a three tiered sampling technique. The l l
first tier involves in process sampling for the programmatic elements intended to effect a change in the licensing basis. Since programmatic elements may be missing or incorrectly applied. the next two sampling tiers involve a search for differences by samphng for potential operational changes that could be made without procedure changes. and sampling backwards by comparing selected FSAR statements with operational practices. The purpose of the data gathering phase is ;
to identtfy a set of potential differences between the operational practices and the licensmg basis.
The evaluation phase determines if potential differences are valid. Vahd differences are categorized in order to draw conclusions about the adequacy of particular programmatic controls for maintaining the licensing basis. The signtScance of the findings are used to develop recommendations for programmatic enhancements.
It is recognized that many bcensees have performed safety system functional inspections (SSFIs) and other reviews that have assessed the adequacy of their programs to maintain the licensing basis of their plants and to ensure tha accuracy ofinformation in the FSAR. These completed activities can be credited, as appropriate. in meeting portions of the assessment described below.
The assessment methodolocy described below represents only one way of l
performing the assessment of the adequacy and effectiveness of programs for l maintaining the licensing basis. Although the methodology contains a reasonable amount of detail,it is not intended to be prescriptive. Other approaches that l provide an equivalent scope of review can also meet the intent of this assessment.
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3.0 ASSESSMENT METHODOLOGY 3.1 Data-Gathening Phase -
The data gathering phase of the assessment employs three samphng techniques.
Each sampling technique is discussed separately 3.1.1 FSAR Sampling *
- Extract the relevant FSAR sections for the system (s) being evaluated a Highlight those FSAR statements that meet the following criteria:
0 Descriptive phrases regarding frequencies for tests. calibrations. etc.
- Configuration descriptions Descriptions of system operation in different modes te.g., normal. abnormal. accident / emergency)
Operating limits
- Descriptive functional performance statements
- NOTE The intent is to identify a range of FSAR statements associated with operational practices that could be changed and. in the presence of programmatic weaknesses. are not adequately reDected in the FSAR.
. Compare the highhghted FSAR statements with current operational practices using iiidividuals cognizant of the operation of the system and associated engineering problems. Document the following:
0 Is the FSAR statement accurate with respect to operational practices (procedures. operating philosophy, standing orders, workarounds. etc.)?
If not. identify the differences.
O Is the FSAR statement clearly understood or in need of clarification?
3.1.2 Programmatic Sampling
. Select three unrelated examples of each of the following types of changes:*
50.59 ( at least one procedure chance evaluationi 3
> 50.54 (one each from QA. EP and Security >
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- Outstanding corrective action fbr a matarial condition greater than one year old i
- Operating procedure change not evaluated under 50.59 l I
l O FSAR change request )
0 50.90/50.55a change and associated SER (where NRC approval has been received) 0 Regulatory commitment addition or change 0 Design change not evaluated under 50.59
- NOTE: The changes selected should have been completed in sufficient time to have been reDected (if required)in the most recent licensing basis update of the FSAR. QA program etc.
. Review the change to determine if the change should have appeared in the beensing basis and. if so, if the licensing basis accurately reflects the change.
For changes that are not accurately reDected in the licensing basis, determine the programmatic step (or missing programmatic step) that would have ensured its accurate incorporation into the licensing basis.
. Document the results of the programmatic sampling.
3.1.3 Sampling for potential changes that may occur separate from programmatic or procedure changes
. Compile the following:
0 Workaround bst 0 Operations standing orders ,
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0 Technical Speci6 cations positions / interpretations (if any) I
- Sample of documents that may show potentially routine "NA'ing" of procedure steps (e.c.. last completed startup procedure. recently completed system operatine instructions. shift supervisor loc entries for deviations >
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l : List of equipment being operated in manual 1
- Old* temporary alterations d
j > Old" non conformances 4
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- Old* tag. outs i * "Old" refers to current items that should be reDected (if required) in the last
- FSAR update per 10 CFR 50.71(e) d
- Select a sample from each of the above categories based on the number of stems compiled.
. Review the item selected to determine if the item represents a change to the l
heensing basis and. If so, if the licensing basis accurately reDects the change.
jl For changes that are not accurately reDected in the licensing basis, j determine the programmatic step (or missing programmatic step) that would have ensured its accurate incorporation into the licensing basis.
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- Document the results of the samphng.
l 3.2 Evaluation Phase At this point in the assessment. the changes to operating practices that should f have been reDected in the beensing basis have been identt6ed. Those not i
accurately reDected in the licensing basis are differences that must be resolved and categorized.
The evaluation phase cannot be precisely structured. However, the following j broad steps should be undertaken.
! . Categorize differences in accordance with the following criteria:
! O Proeram/ process which should'have ensured licensmg basis accuracy 4
j 0 SigntScance i
j = Safety signincant I - Regulatory signincant (i.e. noncomphance or missed commitment)
- '.aw .stentneance ie.c., the orianal information wr.s not required to l
be included in the FSAR. or the information could not have been
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rebed upon by the NRC in reachmg a safety conclusion)
[ NOTE: Section V of SDL-GC 9012. " Design Basis Program Guidelines." provides guidance for assessing the safety signincance of differences and for determining if operability and/or reporting issues exist.)
? 50.59 was correctly / incorrectly applied
- Licensing basis information has always been inaccurate (i.e., there was no change in operating practice)
- Identify areas of programmatic weakness or missing programmatic controls.
If uncertain, pursue additional assessment investigation for the programtnatic area in question.
. Characterize overall signincance of the 6rdings. Based on this characterization determine the need to coaden the assessment scope to obtain the necessary con 6dence that the programs are adequate and are being effectively implemented
- If appbcable. generate necessary quahty de6ciency documents and identtfy any reportable situations.
. Prepare recommendations to address programmatic weaknesses.
- Document the results and brief appropriate management.
4.0 NETREPORT 4.1 Purpose of Report NEI will compile the overall industry results to assess the composite adequacy and effectiveness of programs designed to maintain the licensing basis of tha plants. The overallindustry results will be reported to the NRC by NEI.
4.2 Report Format The report format follows the steps contained in Section 3.0 of this report.
Part 1 - FSAR Sampline
. Identifv the FSAR systems reviewed under Section 51.1
. Indicate the number of FSAR differences identi6ed. i
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Part 2 - Programmatic Samoling Indicate the total number ofitems reviewed. l Indicate the number ofitems where the change was not accurately reDected in the beensing basis.
Part 3 - Samoline for notential changes that mav occur separate from programmatic or orocedure chanres l
i Indicate the total number ofitems reviewed. l Indicate the number ofitems where the enange wn not accurately reDected in the licensing basis if required.
Part 4 - Evaluation Phase Indicate the number of differences that were characterized as being safety significant.
Indicate the number of differences that were characterized as being regulatory signt5 cant.
Indicate the number of differences where 50.59 was incorrectly applied.
Indicate the number of differences where the licensing basis information has always been inaccurate.
Provide a brief summary of the programmatic deSciencies identi5ed, if any.
Indicate whether the assessment scope is being broadened based on the identtBed differences.
Describe any departures from the methodology of this guideline.
5.0 DEFINITIONS 5.1 Regulatory Commitment NEI " Guideline for Managing NRC Commitments." endorsed by NRC in SECY-95-300. defines a regulatory commitment as follows:
"Rerulatorv Commitment means an explicit statement to take a spectSc action agreed to or volunteered by a licensee that has been submitted in wnting on the docket to the Commission."
5.2 Licensing Basis 10 CFP. Part 54 dennes the current " licensing basis as follows
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- " Current licensing basis C.B is the set of NRC requirements applicable
- to a spectne plant and a licensee's written commitments for ensuring compliance with and operation within applicable NRC requirements and the plant speciSc design basis (including all modincations and additions j l
l to such commitments over the life of the license) that are docketed and in effect. The CLB includes the NRC requirements contained in 10 CFR parts 2.19, 20, 21. 30, 40, 50, 51, 5 4. 55, 70. 72. 73, and 100 and l appendices thereto: orders: license conditions; exemptions: and technical l 1
spectScations. It also includes the plant speciSc design basis information deSned in 10 CFR 50.2 as documented in the most recent l
Snal safety analysis report (FSAR) as required by 10 CFR 50.71 and the licensee's commitments remaining in effect that were made in docketed ;
licensing correspondence such as licensee responses to NRC bulletins, generic letters, and enforcement actions, as well as licensee commitments documented in NRC safety evaluations or licensee event reports.
NOTE Responses to NRC bulletins. generic letters and licensee event :
reports contain sorne commitments that are outside the scope of the CLB l in that they are not necessarv to ensure comphance with applicable NRC l requirements (e.g., rules. regulations, licenses and orders or to maintain )
the plant spectSc desten basis.
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[r* *s c *g UNITED STATES a
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NUCLE AR REGULATORY COMMISSION j #'
j usmuotoN o e resss-ooot August 14, 1995
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CHAlmMAN Mr. Joe F. Colvin President and Chief Executive Officer Nuclear Energy Institute Suite 400 1776 1 Street, N.W.
Washington, D.C. 20006-3708
Dear Mr. Colvin:
I am responding to your letter of August 2, 1996, concerning industry actions for assessing programs in place to reaffirm that nuclear power plants are operated in conformance with their licensing basis. Your letter also identified three issues that, in your view, require mutual agreement between the industry and the Nuclear Regulatory Commission (NRC) before the industry would proceed with the initiative: (1) the legal standing of the Updated Final Safety Analysis Report (FSAR) and the NRC's Safety Evaluation Reports; (2) the scope of what constitutes the current licensing basis; and (3) the adequacy of NSAC-125, " Guidelines for 10 CFR 50.59 Safety Evaluations," for performing 10 CFR 50.59 evaluations.
We believe that resolution of the issues you identified is not a prerequisite to reviewing, on a retrospective basis, whether existing programs are sufficient to ensure that licensees know their licensing bases, whether licensing bases have been properly maintatued, and whether licensing basesThe are accurately described in each facility's updated FSAR or other documents.
significant issues recently identified by licensee reviews and NRC staff inspections relate to failures to address cegraded and nonconforming conditions properly, failures to perform reviews required by 10 CFR 50.59 before making changes to facilities, and failures to update facility Final In our view, Safety Analysis Reports in accordance with 10 CFR 50.71(e).
industry initiatives can proceed notwithstanding that ongoing NRC activities under the 10 CFR 50.59 Action Plan will consider, in a broad sense, issues such as those you raise.
Existing regulations and guidance are sufficient to conduct a retrospective review for conformance to existing regulatory requirements. These include:
(1) NRC regulations 10 CFR 50.2, 50.34, 50.54, 50.59, 50.71(e), 50.72, 50.73 and Appendix B; (2) NRC's policy statement - Availability and Adequacy 10,of1992; Design Bases Information at Nuclear Power Plants, 57 FR 35455 August (3) " Design Bases Program Guidelines," NUMARC 90-12 and NRC letter dated November 9, 1990; and (4) Generic letter 91-18, "Information To Licensees Regarding Two NRC Inspection Manual Sections On Resolution Of Degraded And Nonconforming Conditions And Operability."
Attachment 4
'%f66 N W d b \ -3p p
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2-The staff 15 concerned, however, that the proposed initiative may not be of sufficient scope and depth to identify tha types of design and operability problems recently identified at several operating plants. Specifically. It is not sufficient to perform a process / procedural based review. An in-depth ;
vertical slice review of actual design basis documentation and comparison of l "as built" and "as operated" safety systems is more appropriate. ,
l For example, reviews similar to safety system functional inspections (Inspection Procedure 9380), " Safety System Functional Inspections"), may be used to evaluate a licensee's program effectiveness to maintain the licensing and design bases. These reviews should include: (1) an in-depth review of selected systems' design and design basis since issuance of the facility operating license: (2) risk- and safety-cased criteria for selection of syste-s (or review; and, (3) a method to ensure that licensee problem '
identification and corrective action on the selected systems are representative and consistent with ctter systems. The in-depth review should exa ine: (1) engineering design and configuration control; (2) verification of as d uilt and as-modified conditions; (3) translation of the design bases requirements into operating procedures, maintenance, and testing; (4) verification of system performance through review of test records and 0::serwattons of selected testing: (5) proposed and implemented corrective actions for licensee-1dentified design deficiencies; and, (6) modifications ade to the systems since initial licensing.
In snort, the NRC position has been, and is, that It is the responsibility of individual licensees to know their licensing basis, to have appropriate docu-entation that defines their design basis, and to have procedures for performing the necessary assessments of pinnt or procedure changes required by NRC reoulations. Until such time that any regulation changes are made as a result of NRC action plan activities, we will continue inspection and I oversight activities related to the design basis to ensure compliance with l
ex15 ting regulattons.
Sincerely, k
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Distribution for 50.54(f) letter to utility CE0s dated: October 9, 1996 Docket File PUBLIC J. Taylor ED0 Rdg File F. Miraglia/A. Thadani R. Zimmerman DRPW/DRPE/DRPM Directors DRPW/DRPE/DRPM Deputy Directors J. Lieberman J. Goldberg W. Dean G. Tracy B. McCabe J. Mitchell D. Matthews E. McKenna K. Thomas !
OPA l DRPW/DRPE PDs l NRR PMs NRR LAs PD Rdg File R.W. Cooper, RI E. Merschoff, RII ,
i W.L. Axelson, RIII J. Dyer, RIV K. Perkins, RIV/WCF0
.