ML20040A935
| ML20040A935 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 01/11/1982 |
| From: | Bridenbaugh D, George Minor CALIFORNIA, STATE OF, MHB TECHNICAL ASSOCIATES |
| To: | |
| Shared Package | |
| ML20040A930 | List: |
| References | |
| NUDOCS 8201220320 | |
| Download: ML20040A935 (22) | |
Text
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l i
UNITED STATES OF AMERICA
^
NUCLEAR REGULATORY COMMISSION i
y BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
)
In the Matter of
)
)
PACIFIC GAS AND ELECTRIC COMPANY
)
Docket Nos. 50-275 0.L.
)
50-323 0.L.
(Diablo Canyon Nuclear Power
)
Plant, Units 1 and 2)
)
)
l PREPARED DIRECT TESTIMONY OF 1
DALE G.
BRIDENBAUGH AND GREGORY C. MINOR I
ON BEHALF OF GOVERNOR EDMUND G.
BROWN JR.
I i
]
REGARDING CONTENTION 10 January ll, 1982 t
I 82 012@ o 32o
PREPARED DIRECT TESTIMONY i
OF DALE G.
BRIDENBAUGH AND GREGORY C. MINOR REGARDING CONTENTION 10 I
i I.
INTRODUCTION 1.
My name is Dale G.
Bridenbaugh.
I am a Professional Nuclear Engineer, licensed by the State of California, technical consultant, co-founder and president of MHB Technical Associates, l
technical consultants on energy and environment, with offices at 1723 Hamilton Avenue, Suite K, San Jose, California.
I have participated as an expert witness in licensing proceedings before the U.S. Nuclear Regulatory Commission (NRC); have served as a consultant to the NRC; have testified at the request of the Advisory Committee on Reactor Safeguards; have appeared before various committees of the U.S.
Congress; and testified in various state licensing and regulatory proceedings.
I received a Bachelor of Science in Mechanical Engineering from the South Dakota School of Mines and Technology in 1953.
From June, 1953, until February, 1976, I worked as an engineer and manager with the General Electric Company on a wide variety of most of the aspects of power genera -
tion equipment design, manufacture and operation.
During the last 10 of those 22 years, I was in management positions in the General Electric Nuclear Energy Division where I had the responsi-bility for managing the monitoring of operation of nuclear
. power plants, for the implementation of solutions to nuclear plant operational problems, and for the development of a master performance improvement plan aimed at bringing about the long term improvement of power reactor performance.
2.
In my capacity as technical consultant'with MHB Technical Associates, I have provided technical advice to various govern-mental bodies and individual groups on subjects related to the design and operation of commercial nuclear power plants.
As examples of this work, in 1978 I served as a consultant to the United States Nuclear Regulatory Commission to review the NRC plan for research to improve the safety of light water nuclear
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power plants, and have served in various consulting capacities to the United States General Accounting Office, the states of i
California, Illinois, Massachusetts, New Jersey, Pennsylvania, to Suffolk County, New York, and to the governments of Sweden and Norway, all in the evaluation of nuclear plants or programs.
A statement of my qualifications and professional experience is I
i appended to this testimony as Attachment A.
I 3.
My name is Gregory C. Minor.
A statement of my qualifications and experience has previously been provided to this Board as part of my testimony on Contention 1 and in Attachment B to that testimony.
. II.
STATEMENT OF CONTENTION
?
4.
The purpose of our testimony is to respond to Contention 10 as admitted by the Board as follows:~1/
The Staff recognizes that pressurizer heaters and associated controls are necessary to maintain natural circulation at hot stand-by conditions.
Therefore, this equipment should be classified as ' components important to safety' and required to meet all applicable safety-grade design criteria, including but not limited to diversity (GDC 22),
seismic and environmental qualification (GDC 2 and 4), automatic initiation (GDC 20), separation and independence (GDC 3 and 22), quality assurance (GDC 1), adequate, reliable on-site power supplies (GDC 17) and the single failure criterion.
The i
Applicant's proposal to connect two out of four '
emergency power supplies does not provide an equivalent or acceptable level of protection.
The results of our review of some of the important matters en-compassed by this Contention are summarized in the following paragraphs.
III.
DISCUSSION OF ISSUES III.A.:
Background and Summary of Position I
5.
The essence of Contention 10 is that the pressurizer heaters, including the associated heater controls, should be l
-1/
ASLB Memorandum and Order, September 30, 1981.
On September 21, 1981, the Commission directed the Licensing Board to include in the full power proceeding Joint Intervenors' low power Contention 10.
formally classified as " components important to safety" and, accordingly, be designed, ma.n'ufactured, and constructed with all the care that should be afforded such components.
6.
The origin of this Contention is the experience of the Three Mile Island accident and the subsequent reviews performed l
to consider its significance.
This accident, along with its extended recovery period, demonstrated the need to reconsider the safety classifications and design practices for nuclear i
!x systems and components.
In particular, the inoperability of the s
reactor coolant pumps and the low pressure decay heat removal systems emphasized the importance of the ability to remove heat from the reactor via natural circulation and required associated systems.
Thus, the NRC Lessons Learned Task Force found that
" maintenance of natural circulation capability is important to safety."-2/ The pressurizer heater system is the normal and pre-ferred system for this capability.
In addition, the pressurizer heaters must also maintain physical integrity for the reactor coolant pressure boundary to be maintained.
While it may be possible to maintain natural circulation at hot standby condi-tions without use of the pressurizer heater and associated controls, such operation may be difficult to control and is contrary to the normal and emergency plant operating procedures.
In this regard, PG&E's response No. 45, dated October 26, 1981, 2/
1;UREG-0 57 8, p. A-2.
= _.. - -
m i,
i to Joint Intervenors' Second Set of Interrogatories provided f
a list of emergency operating procedures that include the i
use of pressurizer heaters.
We have reviewed these. procedures-j and find that " alternate" (to the use of the pressurizer heaters) pressure control methods are not specified for t'he operators' i
use.
These procedures thus appear to place total reliance on l
I automatic or manual operation of the heaters.
We therefore conclude that the heater system has been improperly classified, i
i_
or the procedures have been inadequately prepared in failing to provide safety-related backup systems, or both may be at fault:
Further, plant safety may be affected by many things, not the least of which is the need to minimize the number of challenges 4
to the total system integrity and to optimize the operability.
[
and controlability of the systems used in the mitigation or control of abnormal events.
The logical response to the informa-tion gained from the TMI-2 accident, in our opinion, is to c1hssify the pressurizer heater system as important to safety (safety-related) so as to ensure its operability for response i
to accidents or transient conditions.
7.
It is important to place in proper context the intended meaning of the phrase " components important to safety."
Contention 10'was formally accepted by the ASLB on September 30, 1981.
On November 20, 1981, Harold R.
Denton, Director of the NRC's Office of. Nuclear Reactor Regulation, issued a Memorandum to clarify the 1
1 l
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i use of safety classification terms.~3/ This Memorandum stated that safety classification terms had not been consistently applied by the NRC Staff, and that three terms, "important to safety,"
" safety-grade," and " safety-related," have been used inconsistently or interchangeably.
Mr. Denton's Memorandum goes on to identify the recommended usage of these terms.
This should serve to make these terms more definitive when used in future licensing; however, our understanding of the usage intended in the Contention 10 language is that the pressurizer heaters and controls should be classified as " safety-related" (as defined in the Denton Memorandum) and should, therefore, be subject to the general requirements of the General Design Criteria (GCD) and that applicability of various GDC's should be judged by the guidance of 10 C.F.R.
~4/
Part 100, Appendix A.
1 l
-3/
Memorandum from H.
R.
Denton to All NRR Personnel, November 20, 1931,
Subject:
" Standard Definitions for Commonly-Used Safety Classification Terms," Attachment B hereto.
i
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Our review of the Diablo Canyon pressurizer heater documenta-4/
tion affirms our view that precise safety classification terminology is necessary and significant.
The NRC Staff believes the pressurizer heaters are considered " components important to safety" with respect to their pressure control function.
NRC Staff Motion for Summary Disposition of Contentions 10 and 12, p.
6.-
The Applicant believes these components are not required to be classified as "important to safety."
Pacific Gas and Electric Company's Motion for Summary Disposition, December 21, 1981, p. 4.
The Staff's position would lead to the belief that at least some of the General Design Criteria have been applied, whereas the Applicant's position would indicate that none of the GDC apply (other than those that admittedly apply to the RCPB pressure retaining capability and to the breakers which can be used to connect the heaters to the onsite emergency power system).
Applicant's response to NUREG-0578 states that (Cont'd on next page) j
1
- i i
1 III.B.:
Importance of Pressurizer Heaters i
i 8.
The pressurizer heater system used at the.Diablo Canyon i
plant provides an important function, namely, the ability to control primary coolant pressure under various conditions.
Not i
only is the system used during normal power operation, but is especially needed for control of pressure and of natural circula-tion capability in the hot standby mode.
The NRC Staff's recom-mendations emanating from the TMI reviews recognize that maintenance of safe plant conditions depends on maintenance of i
pressure control in the primary system for the associated main-
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~
tenance of natural circulation capability.
The Staff, therefore, t
i recommended upgrading the pressurizer heaters and associated i
l 1
4/
(Cont'd) equipment identified as non-safety-grade will not be quali-j fied for the Hosgri event, implying that the heaters, therefore, are not seismically qualified.
Pacific Gas and Electric Company Response to NUREG-0578, April 21, 1980, l
p.
III-B-5.
The Westinghouse specification under which
}
the pressurizer heaters were procured seems to confirm that only the coolant boundary GDC's were applied.
Furnished with Applicant Pacific Gas and Electric Company's Supple-mental Response to Joint Intervenors' Second Set of Inter-2 rogatories, December 23, 1981, Immersion Heater Spec. 393A701.
The specification provides r.o design requirement on the l
radiation exposure the unit must withstand (the specific i
concern is the insulating boot at the electrical connection),
nor does it address ceismic loadings.
No information is given on heater sheath supports along the length of the heater l
(the heater rods are approximately eight feet long and are 7/8" in diameter).
These omissions provide little assurance 1
that these important aspects have been adequately considered so as to produce a reliable source of pressure control.
. controls to achieve greater reliability.-5/.The NRC Staff's
- Motion for Summary:Dispositio;n states that pressurizer heaters are required to maintain system pressure at the hot standby 6/
condition.
PG&E claims that heaters are not required for hot standby pressure control and natural circulation.-7/ We agree with the Staf f that the heaters should be used for this function.
The basis of this position is that this is the normal control mode, that the procedures specify this mode, and that it is difficult for the operators to follow a different and infrequently used procedure under stressful conditions.
9.
PG&E's intended reliance on the pressurizer heaters is indicated by frequent mer' ion of them in the Diablo Canyon Emergency Operating Procedures.
No less than'nine such procedures call for the use of the pressurizer heater system.~8/ PG&E claims 5/
NUREG-0578, NRR Lessons Learned Task Force Short-Term Recommendations, page A-2.
-6/
NRC Staff Motion for Summary Disposition of Contentions 10 and 12, page 5.
7/
Affidavit of John B.
Hoch, page 1, a part of Pacific Gas and Electric Company's Motion for Summary Disposition, December 21, 1981.
8/
Applicant Pacific Gas and Electric Company's Answers to t
Governor Edmund G. Brown Jr.'s Second Set of Interrogatories, L
page 47.
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t
i that alternate means (to the heater system) for pressure control are available; howe'ver, none of the cited emergency operating procedures specifically direct the operator how to proceed with i
r.lternatives if the heater system becomes unavailable.
(See Paragraph 11 for further discussion of procedural inadequacies.)
10.
The NRC Staff states that primary system pressure control is not a prerequisite for natural circulation as the Westinghouse i
design will provide natural circulation as long as adequate water is provided to the secondary side of the steam generators, even if the primary coolant pressure decays to bring the system to a i
saturated condition.
Applicant and NRC Staff also cite test data obtained at the Sequoyah Nuclear Plant that supports the i
f claim that the Diablo Canyon primary system pressure will decay i
at about 100 psig per hour if the pressurizer heaters are lost.
It has not yet been demonstrated, however, that these character-istics are true at Diablo Canyon.
Further, the Applicant has provided no directions in the Emergency Operating Procedures as to how the characteristics would be utilized to assure proper operation.
If it is the Applicant's intent to rely upon these claimed reactor characteristics, they should be demonstrated and necessary operator action (s) should be fully described in the procedures.
III.C.:
Deficiencies of Present Pressurizer Heater System 11.
The purpose of the pressurizer heater system upgrading j
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required by NUREG-0578 (and 0737) is to assure that primary I
coolant pressure control willibe available when needed.
The 4
time when this need is the greatest is during or following
~
i transient and/or accident conditions.
Emergency Operating i
F j
Procedure OP-13, Malfunction of Reactor Pressure Control System, I
is intended to provide guidance on how to maintain primary i
4 pressure control when the pressure control devices malfunction.
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^
This procedure only assumes control channel failure or failure to
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deenergize and therefore provides corrective action by placing the system in manual control.
No guidance is given as to how to l
- i proceed to " feed and bleed" or the other " alternate control methods" claimed by the Applicant.
}
i.
Circulation of Reactor Coolant, has as a basic assumption that i
I offsite power and the heaters are available, making it incomplete 1
L for certain accident sequences.
1 i
12.
PG&E appears to be in a paradoxical situation.
On the one hand, PG&E has argued that the pressurizer heaters are not i
i required for natural circulation; rather, other methods are l
i I
l available to ensure that this important cooling mechanism occurs.
However, in the Diablo Canyon Emergency Procedures (OP-13 and 23),
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no other methods are provided for the operators' use.
- Thus, 4
l in our opinion, at a minimum, either the heaters should be up-graded to safety grade or the other methods which presumably rely on safety grade systems should be specified.
Since the 4
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It also is not clear that the Staff has adequately l
t assessed the potential delays-and disruption to area accessibility i
inherent in a confusing post-accident situation.
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III.D.:
Impact of Upgrading the Safety Clari-fication l
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14.
The possibility of upgrading all of the pressurizer
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heater system components to a " safety-related" classification j
i has been considered in the past and was, in fact, recomrended l
I by one of the major NRC groups assembled to review the TMI i
accident.
The recommendations presented included:
l 1
l The pre'ssurizer heater system should be j
classified as safety grade which would j
assure emergency power availability and r
protection from failures due to environ-I mental conditions. 11/
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This recommendation, if followed, would have required full ad-i herence to all applicable safety requirements and qualification
?
l of the components to appropriate-seismic and environmental con-f ditions.
There are no reasons to believe that such upgrading l
t could not be done (from a " state-of-the-art" standpoint).
15.
If safety classification upgrading were to be required, f
I t
r the pressurizer heater system should become more reliable.
Plant i
i safety would be improved by the minimization of the number of I
--11/
Memorandum for J. M. Allan, NRC, from R.
D. Martin, NRC,
" Operations Team Recommendations," October 10, 1979, p.
t 23 (emphasis added).
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challenges to the system and by the optimization of the operabil-ity and controllability of sy# stems used in the mit.igation or
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control of abnormal events.
The NRR Lessons Learned Task Force 4
l found that " maintenance of natural circulation capability is 4
important to safety."
Pressurizer heaters are the preferred components for this capability.
It is our opinion that such upgrading would impose more of the safety design criteria on this system and its operability.
GDC 20 requires, for example, that s
the protection system shall be designed "to initiate the opera-tion of systems important to safety."
Standard Review Plan l
Table 7-1 extends the applicability of GDC 20 to all instrumenta-tion and control functions important to safety.--12/
PG&E's January 26, 1981 response to Full Power License Requirements describes the manual procedure necessary for transferring the presserizer heater power supply onto the ESF buses.
This requires the dispatch of an operator to a location three floors down in the Auxiliary Building and verbal confirmation that such action i
~~13/
has been teken.
This procedure does not meet the automatic initiation requirements of GDC 20.
None of the pressurizer heater system, other than the breakers, switches and portion of the bus
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connection cables identified in Response 1, has been qualified in i
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12,/
NUREG 75/087, Section 7, Table 7-1.
13/
See Philip A. Crane to Frank J. Miraglia, January 26, 1981, p.
II.3-14.
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accordance with GDC.2 (seismic and environmental qualification),
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GDC 22 (protection' system independence, separation"), or GDC 3 i
(fire protection).
Since these components have not been classi-fied as important to safety, the requirements of GDC 1 (Quality j
i standards and records) does not appear to have been applied.
~
l IV.
CONCLUSION 1
i 16.
The discussion in Part III above indicates a number of l
reasons why the pressurizer heater system components should be 1
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classified as safety-related components.
It also indicates 1
f some of the benefits to be obtained by such classification, i
We therefore conclude that this action should be taken at the i
Diablo Canyon plant.
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ATTACHMENT A ATTACHMENT A EXPERIENCE AND QUALIFICATIONS OF DALE G. BRIDENBAUGH Dale G. Bridenbaugh is a Professional Nuclear Engineer. licensed by the State of California (license NU 973), president of MHB Tech-nical Associates, and a member,of the American Nuclear Society.
Bridenbaugh received a B.S.
in ' Mechanical Engineering from the South Dakota School of Mines and Technology in 1953.
He has been intimate-ly involved with.the commercial nuclear power program since 1958, when he was first assigned in a supervisory capacity for the General Electric Company in the construction of the Dresden Nuclear Power Station near Chicago, Illinois.
Subsequent to that assignment, he has accumulated over.20 years of nuclear experience,, including re-sponsible management of_ positions in construction, startup, Electric's op.eration, maintenance, and prod ~ct improvement planning with General u
nuclear program.
Included in his background experience is the man-agement of the design, construction, and checkout of the first mo-bile test facility assembled by the General Electric Company for the on-site testing, under simulated environmental operating conditions, of various nuclear system safety and relief valves.
Bridenbaugh has been involved with Pacific Gas and Electric's (PGSE) nuclear plant programs since 1966, when he was assigned re-sponsibility for liason with utilities on all operating nuclear plant matters.
This included the ongoing engineering effort by General Electric in support of PGGE's Humboldt Bay No. 3 nuclear unit.
After the formation of MHB Technical Associates in 1976, he has participated in the review and licensing process of the Dia-blo Canyon plang, presenting testimony before the ASLB in 1976 on expected plant capacity factors.
Bridenbaugh has analyzed the operations of numerous nuclear plants in his previous and present positions.
He evaluated the re-sponse of the Sacramento Municipal Utility District Rancho Seco Plant to equipment and operating procedure recommendations made as a result of the TMI accident.
Results of this evaluation were pre-sented in direct testimony on behalf of the California Energy Com-mission in a hearing before the ASLB on Rancho Seco in 1980.
He has testified on similar matters before the ASLB on the Black Fox (Oklahoma) case.
He has also served as a consultant to the NRC on the review of the NRC safety improvement program and on the safety goals assessment program.
Bridenbaugh has testified on nuclear safety, reliability, anC economic matters before the NRC (Commission and ASLB), before the Joint Committee on Atomic Energy of the United States Congress, and i
before the energy and utility commissions of Ohio, New York, New Jer-sey, Massachusetts, and California.
He has also served as consultant to private and governmental bodies in Pennsylvania, Massachussetts, New York, Illinois, Texas, Oklahoma, and Oregon, as well as in Sweden, Italy, and Australia.
Additional information on the professional qualification of Dale G. Bridenbaugh is set forth in the following:
A-1
ATTACtiMENT A l
P RO FES SI ON AL O UALI FI C ATIONS OF DALE C.
B RI DENB AUGH DALE G.
B RI DENB AUGH 1723 Hamilton Avenue Suite K San Jose, CA 95125 (408) 266-2716
,E XP E RI E N CE :
19 7 6 - P RESENT President MHB Technical Associates. San Jose, California.
Co-founder and partner of technical con s ult in g firm.
Specialists in energy consulting to governmental and other groups interested in evaluation of nucient plant safety and licensing.
Consultant in this capacity to state agencies in California, New York, Illi-nois, Ucw J e r s ey,,
P en n s y lv an ia, Oklahoma and Minnesota and to the Norwegian Nuclear Power Committee, Swedish Nuclear Inspectorate, and various other organizations and environmental groups.
Per-formed extensive safety analysis for Swedish Energy Commission and contributed to the Union of Concerned S cientis t's Review of W AS H-14 0 0.
Consultant to the U.S.
NRC LWR S af ety Improvement l
Program. performed Cost Analysis of Spent Fuel Disposal for the Natural Resources Defense Council, and contributed to the Depart-i ment of Energy LWR Safety Improvement Program f or Sandia Labora-tories.
Served as expert witness in NRC and s tate utility commic sion hearings.
)
l 1976 - ( FEB RU ARY - AUGUS T) i Consultant, Project Survival, Palo Alto, California.
Volun teer work on Nuclear S af eguards Initiative campaigns,in I
California, Oregon, W a s h in g t on, Arizona, and Colorado.
Numerous presentations on nuclear power and alternative energy options to l
civic, government, and college groups.
Also resource person for j
public service presentations on radio and television.
l 1973 - 1976 i
t Manager, Performance Evaluation and Improvement, General Electric Company - Nuclear Energy Division, San Jose, California.
Managed seventeen technical and seven clerical personnel with responsibility for establishment and management of systems to monitor and measure Boiling Water Reactor equipment and system operational performance.
Integrated General Electric resources in customer plant modifications, coordinated correction of causes of forced outages and of efforts to improve reliability and per-formance of BWR systems.
A-2
a 4
J '17 3 - 19 7 6 (Contd) 1 kesponsible for development of Division Mas ter P erf ormance Improvement Plan as well as for numerous Staff special assign-ments on long-range studies.
Was on special assignment for the management of two different ad, hoc proj ects formed to resolve unique technical problems.
1972 - 1973 Manager, Product Service, General Electric Company - Nuclear Energy Division, San Jose, California.
Managed group of twenty-one technical and four clerical personnel.
Prime responsibility was to direct interface and liaison personnel involved in corrective actions required under contract warranties.
Also in charge of refueling and service planning, performance analysis, and service communication functions supporting all com-
]
pleted commercial nuclear power reactors supplied by General Electric, both domestic and overseas (Spain, Germany, Italy, Japan, India, and Switzerland).
t968 - 1972 Manager, Product Service. General Electric Company - Nuclear Energy Division, S an J ose, California.
Managed sixteen technical and six clerical personnel with the responsibility for all customer contact, planning and execution of work required after the customer acceptance of department-supplied plants and /or equipmen t.
This included quotation, sale and delivery of spare and renewal parts.
Sales volume of parts increased from $1,000,000 in 1968 to over $3,000,000 in 1972.
1966 - 1968 M ar. a g e r, Complaint and Warranty S ervic e, General Electric Company Nuclear Energy Division, San Jose, California.
Managed group of six persons with the responsibility for customer contacts, planning and execution of work required after customer acceptance of department-supplied plants and/or equipment--both domestic and overseas.
1963 - 1966 Field En g ineering Supervisor, General Electric Company, Installation and Service E n g in e e rin g Departm en t, Los Angeles, California.
Supervised approximately eight field representatives with responsi-bility f or General Electric steam and gas turbine installation and main tenance work in S outhernlCalif ornia, Arizona, and Southern Nevada.
During this period was responsible for the installation of eight different central station steam turbine generator units, plus much maintenance activity.
Work included customer contact, prepa-ration of qtotations, and contract negotiations.
J
d 1956 - 1963 Field En g inee r, General Electric Company, Installation and Service Engineering Department, Chicaco, Illinois.
Supervised installation and malntenance of steam turbines of all sizes.
Supervised crews of from ten to more than one hundred men, depending on the job.
Worked primarily with large utilities but had significant work with steel, petroleum and other process industries.
Had four years of experience at construction, startup, trouble-shooting and refueling of the first large-scale commercial nuclear power unit.
1955 - 1956 Encineering Trainine Procram. General Electric Company, Erie, Pennsylvania, and S chene ctady, New Y ork.
Training assignments in plant facilities d e s ig n' 'an d in steam tu r bin e t e s tir. g at two General Electric Factory identions.
1953 - 1955 United States Army - Ordnance School, Aberdeen, Maryland.
Instructor - Heavy Artillery Repair.
Taught classroom and shop disassembly of artillery pieces.
1953 Engineering Training Program, General Electric Company, Evendale, Ohio.
Training assignment with Aircraft Gas Turbine Department.
yDUCATION & AFFILIATIONS:
1953, S outh Dakota S chool of Mines and Technology, BSME Rapid City, South Dakota, Upper k of class.
P rof es sional Nuclear Enginee r - Calif ornia.
Certificate No. 0973.
Member - American Nucicar Society.
Various Company Training Courses during career including Profes-sional Busines s Management, Kepner Tregoe Decision Making, Effective Presentation, and numerous technical seminars.
i A-4
HONORS & AWARDS:
Sigma Tau - Renorary Engineering Fraternity.
General Managers Award, General Electric Company.
(
i e
)
i PERS ONAL DATA:
Born November 20, 1931, Millcr, South Dakota.
Married, three children 6'2",
190 lbs., health - excellent Honorable discharge f rom United S tates Army Hobbies:
S kilin g, h ikin g, work with Cub and Boy Scout Groups.
PUBLICATIONS & TESTIMONY:
I 1.
Op e r a t in g and Maintenance Experience, presented at Twelfth Annual S eminar f or Electric Utility Executives, Pebble Beach, California, October 1972, published in General Electric NEDC-10697, December 1972.
2.
Maintenance and In-Service Inspection, presented at IAEA Symposium on Experience From Operating and Fueling of luclear P ower Plan ts, B rid en b au.g h, Lloyd & Turner, Vienna, Austria, October, 1973.
3.
Operating and Main tenanc e Experience, presented at Thirteenth Annual S eminar f or Electric Utility Executives, Pebble Beach, California, November, 1973, published in General Electric NEDO-20222, January. 1974.
4.
Improving Plant Availability, presented at Thirteenth Annual S emin ar for Electric Utility Executives, Pebble Beach, Cali-fornia, November 1973, published in General Electric NEDO-20222, J anu ary, 1974.
5.
Application of Plant Outage Experience to Improve Plant Per-formance, Bridenbaugh and Burdsall, American Power Conference, Chicago, Illinois, April 14, 1974.
6.
Nuclear Valve Testing Cuts Cost, Time, Electrical World, j
October, 15, 1974.
7.
The Risks of Nuclear Power Reactors:
A Review of the NRC Reactor S af ety S tudy W ASH-14 00, Kendall, Hubbard, Minor &
Bridenbaugh, et al, for the Union of Concerned S cientists,
August, 1977.
A-5
t 8.
Swedish Reactor Safety Studv:
B ars ebh*ck Risk As ses smen t,
MHF Technical Associates, January, 1978.
(P ublished by the Swedish Department of Industry as Document Ds1 1978:1) 9.
Testimony of D.G.
Bridenbaugh, R.B.
- Hubbard, G.C.
Minor to the Calif ornia S tate Assembly Committee on Resources, Land Use, and Energy, March 8, L976.
10.
Testimony of D.G.
Bridenbaugh, R.B.
Hubbard, and G.C.
Minor before the United S tates Congress, Joint Committee on Atomic Energy, February 18, 19 7 6, W as hin g ton, DC'(Published by the Union of Concerned S cientis ts, Cambridge, Massachusetts.)
11.
Testimony by D.G. Bridenbaugh before the California Energy Commission, entitled, Initiation of Catastrophic Accidents at Diablo Canyon, Hearings on Emergency Planning, Avila Beach, California, November 4, 1976.
12.
Testimony by D.G.
Bridenbaugh before the U.S. Nuclear Regula-tory Commission, subj ec t : Diablo Canyon Nuclear Plant Perfor-mance, Atomic S af e ty and Licensing B oard Hearings, December, 1976.
13.
Testimony by D.G. Bridenbaugh before the Cafifornia Etergy Commission, subject: Interim Spent Fuel S torage Considerations, March 10, 1977.
14.
Testimony by D.G.
Bridenbaugh before the New York S tate Public Service Co= mission Siting Board Hearings concerning the James-port Nuclear Power S tation, subj ec t : Effect of Technical and Safety Deficiencies on Nuclear Plant Cost and Reliability, April, 1977.
15.
Testimony by D.G.
Bridenbaugh before the Calif ornia S tate Energy Commission, subj ect:
Decommissioning of Pres surize l Water Reactors, Sundesert Nuclear Plan t Hearings, June 9, 1977.
16.
Testimony by D.G. Bridenbaugh before the California State Energy Commission, subject: Economic Relationships of Decommissioning, Sundesert Nuclear Plant, for the Natural
' Resources Defense Council, July 15, 1977.
17.
Testimony by D.G.
Bridenbaugh before the Vermont S tate Board of Health, subj ect : Operation of Vermont Yankee Nuclear Plant and Its Impact on Public Health at.d Safety, October 6, 1977.
18.
Testimony by D.G. Bridenbaugh before the U.S.
Nuclear Regula-tory Commission, Atomic Saf ety and Licens ing B oard, subject:
Deficiencies in Safety Evaluation of Non-Seismic Issues, Lack of a Definitive Finding of S af ety, Diablo Canyon Nuclear Units October 18, 1977, Avila Beach, California.
A-6
19.
Testimony by D.G. Bridenbaugh before the Norwegian Commission on Nuclear P ower, subj e c t: Reactor Safety / Risk. October 26, 1977.
20.
Testimony by D.G.
B ridenbaugh bef ore the Louisiana S tate Legislature Committee on Natural Resources, subject: Nuclear P ower Plan t De ficien cies Impac tin g oa Safety & Reliability, Baton Rouge, Louisiana, February 13, 1978.
21.
Spent Fuel Disposal Corts, report prepared by D.G.
Bridenbaugb for the Natural Resources Defense Council (NRDC), August 31, 1978.
22.
Testimony by D.G. Bridenbaugh, G.C. Minor, and R.B. Hubbard before the Atomic Saf ety and Licensing B oard, in the matter of the Black Fox Nucler: Power S tation Construction Permit Hearings, September 25, 1978, Tulsa, Oklahoma.
23.
Testimony of D.G.
B ridenbaugh and R.B.
Hubbard before the Louisiana Public S ervice Commission, Nuclear Plant and Power Generation Costs, November 19, 1978, Baton Rouge, Louis ian a.
24.
Testimony by D.G.
Bridenbaugh before the City Council and Electric Utility Commission of Austin, Texas, Design, Con-struction, and Operating Experience of Nuclear Generating Facilities, Dece=ber 5, 1978, Austin, Texas.
25.
Testimony by D.G.
Bridenbaugh for the Commonucalth of Massachusetts, Department of Public Utilities, Impact of Unresolved Safety Issues, Generic Deficiencies, and Three Mile Island-Initiated Modifications on P ower Generation Cost, at the P rop os ed Pilgrim-2 N uclear Plan t, June 8,
1979.
26.
Imuroving the Safety of LU R P o w e r Plants, MHB Technical Associates, prepared for U.S.
Dept. of Energy, Sandia Laboratories, September 28, 1979.
27.
BWR Pipe and Nozzle Cracks, MHB Technical Associates, for the Swedish Nuclear Power Inspectorate (SKI), October, 1979.
28.
Testimony of D.G.
Bridenbaugh and G.C. Minor before the Atomic Safety and Licen s in g B oard, in the matter of Sacramento Municipal Utility District, Rancho Seco Nuclear Generating Station f ollowin g TMI-2 accident, subject:
Operator Training and Human Factors En g in e e rin s, for the California Energy Commission, February 11, 1980.
29.
Italian Reactor Safety Study:
Caorso Risk As s es smen t, MHB Technical Associates, for Friends of the Earth, Italy, March, 1980.
l 30.
Decontamination of Krypton-85 from Three Mile Island Nuclear j
- Plant, H.
- Kendall, R.
Pollard, 6 D.G.
B r id enb au gh, et al, l
The Union of Concerned Scientists, delivered to tre Governor of Pennsylvania, May 15, 1980.
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31.
Testimony by D.G. B ridenbaugh bef ore the N ew Jersey Board of Public Utilities, on behalf of New Jersey Public Advocate's Office, Division of Rate Counsel,
- An aly s is of 1979 Salem-1 Refueline Outace, August, 1980.
32.
Position S tatement, Proposed Rulemaking on the Storage and Dis posal o f Nuclear Was te,' Join t Crose-S ta tament of Position of the New England Coalition on Nuclear Pollution and the Natural Resources Defense Council, September, 1980.
33.
Testimony by D.G.
Bridenbaugh and Gregory C.
Minor, before the New f ork S tate Public Service Commission, In the Matter 4
of Long Island Lighting Company Temporary Rate Case, prepared for the Shoreham Opponents Coalition, September 22, 1980, 4
Shoreham Nuclear Plant Construction Schedule.
34.
Supplemental Testimony by D.G. Bridenbaugh before the New Jersey Boad of Public Utilities, on behalf ~of New Jersey 4
Public Advocate's Office, Division of Rate Counsel, An aly s is of 1979 Salem-1 Refueling Outage, December, 1980.
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ATTACriMENT B
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UNITED STATES NUCLEAR REGULATORY COMMISSION
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WASM NGTON. D. C. 20$55 S
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i MEMORANDUM FOR:
All NRR Personnel FROM:
Harold R. Denton, Director
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Office of Nuclear Reactor Regulation-
SUBJECT:
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STANDARD DEFINITIONS FOR COMMDNLY-USED SAFETY CLASSIFICATION TERMS' Litigation of one of the principal issues in the TMI-l Restart Hearing brought i
to light the fact that there is not complete consistency among all elements of
~
the NRR staff,tE the application of safety classification terms used frequently in the conduct',of NRR's safety review and licensing activities.
More specifi-cally, it appears that terms "important to safety," " safety grade," and " safety-i related" have been used at times interchangeably, or in ways not completely consistent with the definitions and usage of such terms in the regulations, and which d'o not fully reflect the intent of the regulations or current licensing practice.
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Efforts have been underway for some months now to develop guidance for the i
consistent usage of these terms.
These efforts have included:
(a) review of a large number of Reg Guides and SRP's, in conjunction with parts of the regula-tions upon which they are based, for consistency in the application of safety i
classification terminology, (2) extensive discussions among cognizant NRR, RES 1
j (Stds. Devel.) and ELD representatives regarding proper interpretation and application of such terms, including consideration of alternative " standard" definitions and (3) consu'.tation with the cognizant ACRS Subcommittee regarding these matters, and consideration by the full ACRS as well.
As a result of these efforts, I am endorsing and prescribing for use by all NRR personnel the standard definitions set forth in the enclosure to this letter.
It should be noted that in connection with long-term efforts to develop means for ranking reactor plant systems with respect to degree of importance to safety, and i
in connection with related efforts to develop a graded Q.A. approach in reactor i
licensing, the gener,al question of safety classifications and safety classification terminologies will be reexamined; and this could result in changes to the defini-tions set forth in the enclosure or perhaps in development of a completely new j
scheme in this regard.
For the time being, however, the definitions in the en-closure should be considered " standard" and should be applied consistently by all NRR personnel in all aspects of our safety review and licensing activities and l
should be appropriately reflected in our regulatory guidance documents.
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' All 'NRR Personnel 2-3 t
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f t is expected that minor editorial revisions will have to be made to some existing Reg Guides and SRP's in order to make their wording consistent with these definitions'[ You should review the regulatory guidance documents within your purview in this regard and recommend the necessary changes; it is not j
expected tnat this will involve extensive revision efforts.
I'want to make j
clear that my interut here is only in establishing consistency in the language used by all cognizant groups within NRR in expressing our technical requirements.
It is not my intention by this action to dictate new technical requirements, to modify existing technical requirements, or to broaden the existing scope of 4
NRR licensing review.
j
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Harold R. Denton, Director Office of Nuclear Reactor Regulation
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Enclosure :
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Definition of Terms g
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DEFINITION-OF TERMS' Imoortant to Saf $y Definition - From 10 CFR 50, Appenbix A (General Design Ci-iteria) - see first e
paragraph of " Introduction."
"Those structures, systems, and components that providt. easonable assurance i
that the facility can be operated without undue risk to the h.:alth and safety of the public~."
- n Encompasses the broad class of plant features, covered (not necessarily e
exclicitly) in the General Design Criteria, that contribute in important way to safe operation and protection of the ptblic in all phases and. aspects of facility operation-(i.e., normal operation and transient control as well 2
t as accident mitigation).
Includes Safety-Grade (or Safety-Related) as a subset.
e Safety N1ated Definition - From 10 CFR 100, Appendix A - see sections III.(c), VI.a'.(1), ant' e
VI.b.(3).
Those structure, systems, or components designed to remain functional for the SSE (also termed ' safety features') necessary to assure reouired safety functions, i.e. :
' (1 )
the integrit.' of the reactor coolant pressure boundary; (2)' the capability to shut down the reactor and maintain it in a safe
. shutdown condition; or (3) the capability to prevent or mitigate the consequences of accidents which could result in potential off-site exposures comparable to the guideline exposures of this part.
Subset of "Important to Safety" e
Regulatory Guide 1.29 providesan LWR-oeneric, function-oriented listing of e
" safety-related" structures, systems, and components needed to provide or perform' required safety functions.
Additional information (e.g., NSSS type, BOP design A-E, etc.) is needed to generate the complete listing of safety-related SSC's for any soecific facility.
Note: The terin " safety-related" also appears in 10 CFR 50, Appendix B (Q.A. Program Requirements); however, in that context it is framed in somewhat different language than its definition in 10 CFR 100, Appendix A.
That difference in language between the two appendices has contributed to confusion and misunderstanding regarding the exact meaning of " safety-related":and its relationship to "important to safety" and " safety-grade."
A revision to the language of Appendix B has been proposed 'to clarify this situation and remove any ambiguity in the meaning of these terms.
b.
I Sa fe ty-Gra de
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e Ter: not used explicitly in regulations but widely used/ applied by staff
[
and industry in saf2ty review process.
s Equivalent so " Safety-Related," i.e., both terms apply to the same subset
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of the broad class "Important to S6fety."
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