ML19343B079

Affidavit of RB Hubbard & Gc Minor Re Mod of CP License Until Resolution of All Outstanding Safety Problems Presently Applicable to Facility.Supporting Exhibit Encl

ML19343B079
Person / Time
Site:	Byron
Issue date:	11/21/1980
From:	Cherry M, Hubbard R, George Minor CHERRY, M.M./CHERRY, FLYNN & KANTER, LEAGUE OF WOMEN VOTERS OF ROCKFORD, IL, MHB TECHNICAL ASSOCIATES
To:
Shared Package
ML19343B074	List: ML19343B072 ML19343B077 ML19343B079
References
NUDOCS 8011250333
Download: ML19343B079 (114)

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U NITED ST ATES O F AM E RIC A ,

NUCLEAR REGULATORY C O M MISSIO N In The. Matter Of ).

) Request For Action C O M M O N 'J E A LT R E DISO N C O M P A N Y )

) Pursuant To

37ton Statdon )

) 10 C.F.R. Sees. 2.202,

) 2.206

. (Units N o.1 and N o. 2) )

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A FFID A VIT O F RIC H A R D 3. H U B B A R D AND GREGORY C. MINO R i

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Submitted by t

, Myron M. Cherry Pater Flynn CHERRY 5 FLYNN S utta 'sS01 .,

One IB M Plaza .l

Chicago, Illinofs 60611 (312) 56 5-117 7 l

l Counsel for  !

l I.eague of Ro sen Votars Of Rockford, Illincis 80112503 3, m .

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TABLE OF CONTENTS P_agg I. INTRODUCTION ................... 1 1.1. Affiants and Qualifications . . . . . . . . 1 L 2. The Issues . . . . . . . . . . . . . . . . 3 L3. This Affidavit . . . . . . . . . . . . . . 7 H. BACKGROUND AND CONTEXT . . . . . . . . . . . . 8 2.L The Byron Facility . . . . . . . . . . . . 8 2.2. The NRC's Generic Safety Issue Lists . . . . . . . . . . . . . 9 Table 2.2-1 . . . . . . . . . . . . 12 2.3. The NRC's Post-T'" Safety Issues List . . . . . . . . . . . . . 1S 2.4. Summary . . . . . . . . . . . . . . . . 23 HI. SIX UNRESOLVED SAFETY ISSUES . . . . . . . . . . 23 3.1. Systems Interaction . . . . . . . . . . . 23 3.2. Steam Generator Tube Integrity . . . . . . 31 3.3. Equipment Qualification and Deterioration . . . . . . . . . . - . . 35 3.4. Evaluation of Potential Accidents and Corrective Measures . . . . . . . 40 Table 3.4-1 . . . . . . . . . . . . 50 3.5. Conformance to Current Regulatory Practices . . . . . . . . . . . . . . 53 Table 3.5-1 . . . . . . . . . . . . 57 Table 3.5-2 . . . . . . . . . . . . 63 3.6. Open Generic Issues . . . . . . . . . . . 64

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, IV. - C O N CI. U SIO N S . . . . . . . . . . . . . . . . . . . ' 66 4

4.L The Issues Are I:nportant . . . . . . . . . . . 66 i

4. 2. The Issues are Unresolved  !

- For Syron . . . . . . . . . . . . . . . . 67 j

4. 3. Ongoing Construction Worsens Matters . . . . . . . . 67

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4.4.- Construction Should Be Suspended . . . . . . . . . . . . . . . 68 1

a j - R EFEREN C ES . . . ..................... 70 J

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U NIT E D ST A T ES O F A M E RIC A NUCLEAR REGULATORY C O M MISSIO N In The Matter Of )

) Request F6r A ction C O M M O N iJ E A LTH EDISO N COMPANY )

) Pursuant To 3yron Statdan )

) 10 C.F.R. Secs. 2.202, (Units No.1 and No. 2) ) 2.216 A F FID A VIT O F RIC H A R D 3.HU3 BARD AND GRECORY C. MIN O R ST A TE O F C ALIF O R NIA )

) S S.

COUNTY OF S ANTA CLARA )

RIC H A R D 3. HUBSARD and G R E G 0 R Y C. v.INOR, being first duly sworn, state under oath as follows:

I. IN T R O D U C TIO N 1.1. A ffiants and Qum1Mications 1.1.1. A f fiant Richard 3. R ubbard is a Professional Quality Engineer licensed by the State of California, a technical consultant, and a founder (in 1976) and vice president of M HB Technical Associates, a partne: ship engaged in the business of technical consulting on energy and environ mental issues and having its principal office at 1723 Ha milton Avenue, San Jose, California 95125.

M r.

Hubbard holds a 3.S. in Electrical Engineering from the Unive rsity of Arizona (1960) and an M.B. A. fro m the University of Santa Clara 0969). M r.

Hubbard has sixteen years' experience in nuclear power plant ele c tro nic s, instru mentation, and controls, including elaven years' expe:1ence in responsfbla managerial positions in the N uclear Instru m entation D e part m ent (196 5-19 71),

Ato mic Power Equip n ent D epart m ent (1971-1975), and Nuclaar Energy Control

and Instrumentation Department (1975-76) of General Electric Company. Mr.

Hubbard is a member of the IEEE Nuclear Power Engineering standards subcommittee responsible for the preparation of Quality Assurance standards for safety-related aspects of nuelear power facilities. He has testified on safety-related aspects of nuclear power facilities as an expert witness before Nuclear Regulatory Commission Atomic Safety and Licensing Boards; before (and at the request of) the NRC's Advisory Committee on Reactor Safeguards; before the Joint Committee on Atomic Energy of the United States Congress; and before various State legislative and administrative bodies. He has also provided technical consultation to the Swedish and West German governments concerning safety-related aspects of nuclear power plants. Exhibit A hereto further details Mr. Hubbard's experience and qualifications.

1.1. 2. Affiant Gregory C. Minor is an electrical enginaer, a technical consultant, and a founder (in 1976) of MHB Technical Associates. Mr. Minor, who among other things is a co-holder of United States Patent No. 3,565,760 for a nuclear reactor power monitoring system, holds a B.S. In Electrical Engineering from the University of California at Berkeley (1960) and an M.S. In Electrical Engineering from Stanford University (1966); he also completed the three-year General Electric Company Advanced Course in Engineering. Mr.

Minor has nineteen years' experience in the design, development, research, start-up, and management of nuclear reactor systems, including thirteen years' experience as a Design Engineer (1963-1970), Manager for Reactor Control Systems D esign (19 7 0-197 2), and Managec for Advanced Control and Instrumentation Engineering (1972-1976) in the Nuclear Energy Division of General Electric Company. Mr. Minor is a member of the Nuclear Power Plant I Standards Committee of the Instrument Society of America. He has testified l

0 on safety-related aspects of nuclear power facilities as an expert witness before Nuclear Regulatory Commission Atomic Safety and Licensing Boards; before the Joint Committee on Atomic Energy of the ilnited States Congress; before various State legislative and administrative bodies; and in official proceedings in Canada and West Germany. He has also provided technical consultation to the Swedish and West German governments concerning safety-related aspects of nuclear power plants, and recently served (at the request of that Group) as a consultant for the Nuclear Regulatory Commission's Three 3111e Island Special Inquiry Group headed by Stitchell Rogovin. Exhibit B hereto further details Afr.

311nor's experience and qualifications.

1.1. 3. In addition to their training, experience, and qualifications summarized above, for the past four years 31r. Hubbard, Str. 311nor, and 31HB Technical Associates have devoted nearly all of their professional attention to analyzing, evaluating, and consulting with regard to the technical, economic, and environmental aspects of unresolved safety-related issues concerning nuclear power plants, including (a) the more than 100 such issues which had been identified by the Nuclear Regulatory Commission even before the Afarch 28, 1979 accident at Three Stile Island Unit 2 ("T3tI-2") and (b) the additional unresolved safety issues which have been identified as a result of T31I-2 and the various inquiries undertaken into that accident.

1.2 The Issues 1.2.1. The purpose of this Affidavit is to provide six examples of major safety issues, with specific applicability to the Byron Nuclear Station,

-which according to the NRC itself are (a) known and (b) unresolved, and which have a direct impact on the safety of the Byron nuclear plant. The NRC has identified each of these issues as a high-priority safety-related issue in either l

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Its January 1979 pre-TMI Report to Congress concerning such issuesl or its May 1980 TMI Action Plan 2-and, for several of the issues, in both. Four of the issues are among the top 20 high-priority issues identified in the NRC's Report to Congress; five are within the top 22 such issues; and the remaining issue is identified in the NRC's TMI Action Plan as a priority issue having high safety significance. These ax sets of issues are:

(i) Systems interaction and multiple and common-cause failures (par. 3.1 below);

(ii) Steam generator tube integrity and degradation (par. 3.2 below); ,

(iii) Equipment environmental qualification, deterioration, and aging (par. 3.3 below);

(iv) Evaluation of potential accidents, including Class 9 (TMI was a Class 9 accident), from the standpoint of risk, consequences, release of radoactivity, and corrective measures (;ar. 3.4 below);

(v) Conformance to current and anticipated regulatory practices, including . quality assurance and quality control measures during plant construction (par. 3.5 below); and (vi) Consideration of open and unresolved generic safety issues, such as Tasks A-43 and A-44 identified in the NRC's Report to Congress (par. 3.6 below).

1.2.2. The six sets of unresolved safety issues discussed in this Affidavit have been selected from NRC publications NUREG-0410, NRC Program For The Resolution Of Generic Issues Related To Nuclear Power Plants (NRC, Jan.1978); NUREG-0510, Identification Of Unresolved Safety Issues Relating To Nuclear Power Plants-Recort To Congress (NRC, Jan.1979); and NUREG-0660, l NRC Action Plan Develooed As A Result Of The T M I-2 Accident L  :

(NRC, May 1980). They are by no means all of the unresolved safety issues applicable to the Byron plant; NUREG-0410, NUREG-0510, and NUREG-0660 identify more than 150 additional such issues. However, the six sets of issues we will discuss are particularly significant because:

Each of the issues is a high-priority issue matter having a direct and important safety impact on the Byron plant; None of the issues has been resolved or satisfactorily dealt with to date in the Byron design and construction process (in fact, several have not been considered at all); and There is no assurance that any of the issues will be adequately dealt with as Byron construction continues, or in the Byron operating license proceedings.

I 1.2.3. If Byron construction is permitted to continue on its present schedule while these sets of unresolved safety issues remain outstanding, there is strong reason to conclude that the implementation of adequate responses to them at Byron will be severely compromised or will not occur at all. The specific reasons for this in the context of each issue are discussed in Parts II and III below; there are more general reasons as well. To begin with, both the NRC Atomic Safety and Licensing Appeal Board 3 and members of NRC Licensing Boards 4 have recognized-and our combined 35 years' experience with commercial nuclear power plants in the United States amply attests-that as nuclear plant construction proceeds, it becomes increasingly difficult and expensive (and may become impossible as a practical matter) to make design changes or "back-fit" equipment in order to keep abreast of safety developments. This is particularly true of the Byron facility. The Byron

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T construction permits were issued in 1975;5 even before then limited construction had begun under limited work authorizations.6 At present construction of the

Byron plant is more than 50% complete,7 and despite a slowdown resulting from

" cash constraints that have limited the rate of work," Commonwealth Edison expects to bring the. Byron units "on line" in October 1983 and October 1984.8 Thus, the Byron construction permits were issued, and construction was underway, even before the NRC's original listing of unresolved safety issues in the now-superseded Technical Safety Activities Reoort, ("TS AR")9-let alone j NUREG-0410, or the NRC's January 1979 Report to Congress, or the NRC's May .

.. 1980 TM1 Action Plan.

1.2.4. In addition to the fact that continuing construction itself increasingly forecloses design modifications to accommodate new nuclear safety developments, there is evidence that Commonwealth Edison's financial position ana construction schedule for the Byron facility leave little room for factoring ,

in adequate responses to presently outstanding safety issues. In July 1980, Commonwealth Edison officials submitted testimony to 1 'llinois Commerce Commission stating that Edison began 1980 with its " credit ;tanding at an all-time low;n10 that the " cash constraints" which had impeded work on the

< Byron plant during 1979 and 1980 " appear certain to continue for several years;"Il that Edison cannot increase its present construction budget for 1980 and 1981 without " unjustifiable risks;"l2 that the budget already exceeds by $100 million the " amount that can safely be raised;n13 and that simply to " accommodate its l'

present construction program," Edison's " financing capability will be strained to the utmost during 1980 and 1981."14 While we cannot comment on this financial p testimony by 'dison officials, we can and do observe that the financial and -

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l substantial. The Atomic Industrial Forum (an industry group) has estirrated that the short- and long-term nuclear plant modifications proposed i,y the NRC's T31I Action Plan might require as much as 100 additional man-years of engineering work for each nuclear unit (Byron has two) over the next three to five years.15 In testimony before the Illinois Commerce C >mmission, Commonwealth Edison has estimated the cost of responding to only a few of the short-term recommendations of the NRC's T311 " Lessons Learned" Task Force 16 at more than $35 million for its Zion, Dresden, and Quad Cities nuclear plants.17 In the same testimony, Edison estimated that it will have to spend 570,000 engineering man-hours responding to NRC Bulletins concerning safety matters issued in 1979, and 060,000 man-hours responding to NRC Bulletins issued during the first half of 1980.13 The Edison testimony notes that "[t]he number of these Bulletins has grown significantly since T31I, and responses to them have seriously impacted I (Edison's] internal and contracted engineering manpower, to the point that Sargent & Lundy--the architect-engineer for the Byron' facility (among others)--plainly does not have the required man-hours available to respond to those Bulletins.19 1.3 This Affidavit

1. 3.1. In the following portions of this Affidavit, we will begin with an overview of the specific Byron plant design and site characteristics, drawn from the Byron Station Final Safety Analysis Reoort ("FS A R")20 and i

Environmental Reoort-Ooerating License Stag ("ER/OL")21 prepared by Commonwealth Edison and the Final Environmental Statement-Bvron Station f

("F ES")22 prepared by the NRC. Paragraph 2.1 below. We will then describe the evolution and present status of the NRC's lists of unresolved safety issues  ;

(par. 2.2 below) and the NRC's T311 Action Plan (par. 2.3 below), from which

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are drawn the six sets of safety issues we shall discuss in detail. We will then take up each issue, showing for each its nature, its specific importance to Byron, and why there is no assurance that it will be adequately resolved for Byron (pars. 3.1-3.6 below). Finally, in Part IV of this Affidavit we set forth the conclusions we have reached, and which we believe follow inescapably as a result of our analysis. In brief: Continued construction of the Byron facility should not be permitted to proceed unless and until specific solutions to the high-priority unresolved safety issues discussed below are designed for, and implemented at, the Byron facility. The public health and safety require no less.

H. BACKGROUND AND CONTEXT 2.1. The Byron Facility 2.1.1. The Byron Nuclear Station censists of two nearly identical generating units. Westinghouse Electric Corporation, Sargent & Lundy, and Commonwealth Edison jointly participated in the design and construction of each unit; Sargent & Lundy is the architect-engineer for the Byron facility; and Commonwealth Edison is to operate it.23 Byron is similar in overall design concept to Ec" son's Zion Nuclear Station, and several of Byron's major elements are similar in design to other nuclear projects; for example, the Byron reactor, reactor coolant system, and reactor vessel are similar to those of the Comanche Peak nuclear facility in Texas, and the Byron emergency core cooling system is similar to that of the Trojan nuclear facility in Oregon.24 2.1. 2. Each Byron generating unit includes a pressurized water reactor

("PWR") nuclear steam supply system ("NSSS") and turbine-generator designed and furnished by- Westinghouse. Each NSSS is designed for a thermal power output

of 3425 megawatts ("31Wt"),- which is the license application rating. The equivalent warranted gross electrical output and approximate net electrical output of each Byron unit are 1175 and 1120 megawatts ("MWe"), respectively.

The NSSS is evaluated for safety analyses at 3565 MWt.25 The Byron units will be the largest nuclear units installed by Edison, and the largest completed by Sargent & Lundy as architect-engineer, to date.26 2.1. 3. The Byron facility is located on approximately 1,330 acres of land (the mala site area occupying 1,000 acres, and the balance being the transmission and pipeline corridor to the Rock River) in north central Illinois, about two miles east of the Rock River, three miles southwest (downstream) of the city of Byron and five miles northeast (upstream) of the city of Oregon.27 On 1980 estimates, some 21,500 persons live within ten miles of the Byron site (one-third of them within five miles of the site), and more than 280,000 persons live within 20 miles of the site. The site is approximately 17 miles southwest of the city of Rockford, with an approximate populatien of 166,100 persons.28 The site is divided into a flat upland portion containing the plant buildings and a small portion two miles to the west, in the Rock River Valley, containing the intake works througit which Rock River water will be pumped to the plant (from 61 to 107 cubic feet per second).29 2.2 The NRC's Generic Safety Issue Lists

2. 2.1. The term " unresolved safety issues" refers to deficiencies in i nuclear plant equipment, operating procedures, or licensing which may--by causing nuclear accidents, making them worse when they occur, or impeding a proper response to them by plant operators--contribute to increasing the public health and safety risk inherent in the operation of nuclear plants. The term

" generic," as applied to unresolved safety issues, refers to deficiencies which affect, or may affect, many commercial nuclear plants--for example, l

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all pressurized water reactors ("PWR's"), or all boiling water reactors ("BWR's"),

or all plants using a given type of equipment (e.g., Westinghouse steam generator tubes). For nearly a decade, evidence of potential inadequaeles of

, lightwater reactors such as Byron's has been accumulating in the form of 4

generic unresolved safety issues.

2.2.2. In December 1972 the NRC's Advisory Committee on Reactor Safeguards ("ACRS') began publishing a series of lists, periodically updated, of generic problems which are of concern for lightwater reactors. A recent list identifies a total of over 75 such problems, of which some 25 are considered unresolved, including such important and fundamental safety aspects as reactor containments, reactor pressure vessels, emergency core cooling system ("ECCS")

components, piping, and electrical equipment.30 Ten of the ACRS' problems have been listed since 1972, but are still officially unresolved. While some 30 items have been declared " resolved," it is far from clear just what that means.

According to the ACRS:31

' Resolved as used in the Generic Items reports refers to the following: In some cases an item has been resolved in an administrative sense, recognizing that technical evaluation and satisf ac tory imolementation are vet to be comoteted. Anticipated Transients Without Scram represents an example of this category. In other instances, the resolutien has been accomplished in a narrow or specific sense, recognizing that further steos are desirable, as practical, or that different asoects7f the oroblem require further investigation. Examples are the possibility of improved methods of locating leaks in the primary system, and of improved methods or augmented scope of in-service inspection of reactor pressure vessels." (Emphasis added.]

Accordingly, even when an i*am has been declared " resolved," there is no assurance that a solution has in fact been implemented at any plant, let alone specifically at Byren.

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l 2.2.3. The ACRS lists are by no mear.s a complet6 accounting of outstanding safety issues. In 1974 the NRC compiled its own list of unresolved safety issues, in the form of an internal Technical Safety Activities Recort

(" TSAR"), which when released in 1975 listed 223 items the NRC .5und to be of concern and categorized 173 of them as having "an important impact on the licensing review process.,32 The TSAR was "apparently" (the word is the NRC Appeal Board's) discontinued,33 and replaced by the unresolved-issues lists described below.

2.2.4. In the fall of 1976, several members of the NRC Staff identified 27 unresolved issues as " problems whose priority, progress, or resolution was, in their or. inion, unsatisfactory,"34 and the NRC Commissioners directed the Staff to dcvelop a program plan for the timely resolution of outstanding generic tssues.35 This process began in April 1977, when eacn of tne four NRC Divisions reporting to the Office of Nuclear Reactor Regulation submitted a list of those generic issues the Division considered to warrant the highest priority. Proposals for 365 issues or " tasks" were received; af ter consolidation and elimination, a list of 133 issues was eventually developed and published in January 1978 as NUREG-0410.36 These 133 issues were classified, u:ing a set of uniform criteria, into Categories A (warranting the highest priority attention), B, C, and D. Tabh 2.2-? on page 12 below sets out the 31 highest-priority (Category A) issues which the NRC determined are applicable to pressurized water reactors such as those at the Byron facility; we have marked with an asterisk those Category A issues which are discussed in detail in Part III of this Affidavit. NUREG-0410 also listed 86 other unresolved issues which the NRC determined were relevant to nuclear facilities such as Byron.

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TABLE 2.2-1 CATEGORY A HIGHEST-PRIORITY UNRESOLVED ISSUES (Original NUREG-0410 Listing, January 1978)I TASK NO: TITLl!:

A-1 WATER HAMMER A-2 ASYMMETRIC BLOWDOWN LOADS ON THE REACTOR VESSEL

• AJ WESTINGHOUSE STEAM GENERATOR TUBE INTEGRITY A-9 ANTICIPATED TRANSIENTS WITHOUT SCRAM A-11 REACTOR VESSEL MATERIALS TOUGHNESS A-12 FRACTURE TOUGHNESS OF STEAM GENERATOR AND REACTOR COOLANT PUMP SUPPORTS A-13 SNUBBERS e l4 FLAW DETECTION A-15 DECONTAMINATION

• A-17 SYSTEMS INTERACTION IN NUCLEAR POWER PLANTS A-18 PIPE RUPTURE DESIGN CRITERIA A-21 MAIN STEAM LINE BREAK INSIDE CONTAINMENT A-22 PWR MAIN STEAM LINE BREAK - CORE AND PRIMARY COOLANT BOUNDARY RESPONSE (MSLB OUTSIDE CONTAINMENT)

A-23 CONTA'NMENT LEAK TESTING

• A-24 QUALIFICATION OF CLASS 1E SAFETY-RELATED EQUIPMENT A-25 NONSAFETY LOADS ON CLASS 1E POWER SOURCES A-26 REACTOR VESSEL PRESSURE TRANSIENT PROTECTION (OVERPRESSURE)

A-27 RELOAD APPLICATION GUIDE A-28 LNCREASE IN SPENT FUEL STORAGE CAPACITY A-29 DESIGN FEATURES TO CONTROL SABOTAGE A-30 ADEQUACY OF SAFETY-RELATED DC POWER SUPPLIES A-31 RHR SHUTDOWN REQUIREMENTS A-32 EVALUATION OF OVERALL EFFECTS OF MESILES

• A-33 NEPA REVIEWS OF ACCIDENT RSKS A-34 INSTRUMZNTS FOR MONITORING RADIATION AND PROCESS VARIABLES DURING ACCIDENTS A-35 ADEQUACY OF OFF-SITE POWER SYSTEMS A-36 CONTROL OF HEAVY LOADS NEAR SPENT FUEL A-37 TURBINE M5SILES ,

A-38 TORNADO MISSILES A-40 ]

SEISMIC DESIGN CRITERIA - SHORT-TERM PROGRAM 1 A41 SEISMIC DESIGN CRITERIA - LONG-TERM PROGRAM l l

i L TASKS *A-43 AND *A-44 WERE ADDED TO THE LET IN NUREG-0510.

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2.2.5. Af ter the issuance of NUREG-0410, the NRC also conducted a preliminary evaluation of the unresolved issues on a relative risk basis, in order to Mentify those issues having the greatest safety significance.37 The purpose of this evaluation was not to make an " absolute" risk determination, nor to decide that any issue presented an " acceptable" level of risk; rather, the object was to sort out the 133 NUREG-0410 issues on the basis of which ones had the greatest potential impact on safety. Four risk categories were developed, ranging from "high" to "none." Two of the Category A high-oriority issues discussed in Part III of this Affidavit (Nos. A-17, Systems Interaction, and A-24, Qualification of Class IE Safety-Related Equipment) were also placed in the

" potential high-risk" category as a result of this evaluation.38 2.2.6. In January 1979, the NRC issued NUREG-0510 (its Report to Congress on unresolved safety issues), in which the 133 unresol';?d issues were once again reclassified.39 NUREG-0510 classified seventeen Issue,$, consisting of 22 " tasks" to be worked on, as most important, and labeled them as the

" unresolved safety issues" to be reported to Congress. Four of the issues discussed in Part III of this Affidavit wee included: Nos. A-3, Steam Generator Tube Integrity; A-17, Systems Interaction; A-24, Qualification of Class IE Safety-Related Equipment; and A-43 and A-44, respectively Containment Emergency Sump Reliability and Station Blackout.40 See paragraphs 3.1, 3.2, 3.3, and 3.S below.

2.2.7. Also in January 1979, the NRC's Director of Nuclear Reactor Regulation (Harold Denton) and a Staff Steering Committee further reassessed the unresolved issues as part of an attempt to redirect the NRC's manpower toward the highest-priority issues.41 They used a system of assigning " point values" to each issue, based on a set of standardi::ed criteria. Each of the

issues was assigned a point value (ranging from 230 to 0). As a result, the top twenty issues or " tasks" (all those having from ISO to 230 points) were selected for the priority assignment of manpower, and each NRC Division was ordered to

" commit resources to (them] as necessary to assure that these tasks are completed in a timely fashion."42 These top 20 tasks included all of those noted in paragraph 2.2.6 above, which we discuss in Part III of this Affidavit:

Tasks A-3 (230 points), A-17 (230 points), A-24 (220 points), A-43 160 points),

and A-44 (190 points). In addition, Task A-33 (Evaluation of Accident Risks),

discussed in paragraph 3.4 below, was assigned 150 points and barely missed the

" top 20;" it was number 22.43 2.2.8. Between them, NUREG-0410 and NUREG-0510 collated some thirty-three unresolved issues deemed "high priority" by the NRC--several of which had even then been outstanding for yea.x-relevant to pressurized water nuclear reactor plants like Byron. Twelve of those unresolved issues were also in the " top 20" mandatory highest-priority issues identified by the Denton/ Steering Committee review; a thirteenth issue was number 22 (out of a total of 123 issues)-just outside the " top 20"-in that review. It is now nearly three years since NUREG-0410 was issued (par. 2.2.4 above) and nearly two years since the N RC's Report to Congress, NUREG-0510, and the Denton/ Steering Committee manpower allocation directives (pars. 2.2.5 and 2.2.7 ,

above). Yet none of those highest-priority isIsues has yet been resolved; for at least two (Nos. A-43 and A-44) there is lacking even a plan for resolution.44 Only two of the highest-priority issues (Nos. A-12 and A-24) have progressed to the point of an NRC Staff recommendation.45 2.2.9. Given these circumstances, it is difficult to avoid the conclusion that the NRC's activities with regard to the unresolved issues have i

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I consisted largely of re-sorting and re-listing them without making any real progress. As the Kemeny Commission found, the NRC "has a history of leaving generic safety problems unresolved for periods of many years.'46 And the NRC's own post-TM1 Special Inquiry Group found that "there are many institutional disincentives to safety, and safety issues that are Identified at some point in the system of ten fall through the cracks."47 There is no assurance that the issues which the NRC has identified as relevant to Byron will be meaningfully resolved in a timely fnshion, so that solutions can be implemented at Byron before ongoing construction renders appropriate design modifications unfeasible or impracticably expensive. Waiting until Bvnn has commenced operation before implementing solutions is undesirable at best; at that point, making changes is much more complex and expensive (a cost which ultimately will be borne by the ratepayers), and also more hazardous to workers because of l

radiation exposure. In this regard, we note that occupational exposure to radiation at operating nuclear facilities in the United States has been increasing over time. At present, the reported occupational exposure of workers and station personnel during normal reactor operation averages nearly 500 man-rem per reactor year-or twice the predicted value for annual public eccident exposure calculated in the NRC's Reactor Safety Study (WASH-1400) only five years ago.48 (This too has been an unresolved safety issue-No. B-34-since NUREG-0410 and NUREG-0471 were issued in 1978.)

2.3 The NRC's Post-TMI Safety Issues List

2. 3.1. The listing and re-listing of unresolved nuclear safety issues described above all occurred before the March 28, 1979 accident at Three Mile Island, Unit 2 ("TMi-2"). A perennial problem with such lists, which has arisen with some frequency over the history of the United States nuclear power

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program, is that in many cases the accidents which occur are not the ones which have been analyzed in the licensing process (though after the fact, the accident . initiators or major contributors may well be found to have been somewhere on someone's list of safety concerns). This was true, for example, i

of the. accident at Commonwealth Edison's Dresden Unit 2 In 1970 and of the l similar accident at Edison's Dresden Unit 3 in 1971;49 it was true of the Browns Ferry Nuclear Plant fire in ~ 1975;50 and it was true in multiple respects of the T31I-2 accident. For instance, a major element in the T3fI-2 accident was a pressurizer relief valve or '~PORV" which stuck open (a block valve behind it was also open), . causing pressure to fall, while pressurizer level increased.51 This had happened oefore at ocnte nuclear plants; both the NRC and T311-2's designer knew of an ' earlier occurrence; but nothing had been done about it.52 The PORV itself had been effectively excluded from scrutiny during the licensing process, on the theory that it was not " safety-related."53 Nonetheless, the NRC's post-T31I Special Inquiry Group concluded that because of the open PORV and block valve, "on the morning of Starch 28, before anyone appreciated the seriousness of the situation, Three Stile -Island came close to being the accident we had been told by many in the industry could not happen: a

- (nuclear reactor] core meltdown."54 2.3.2. The T3tI-2 accident spawned at least nine different inquiries, including those of the NRC Special Inquiry Group 55 and the President's

! Commission on the Accident at Three Stile' Island (the "Kemeny Commission"),56 and led the Kemeny Commission to present what the NRC Special Inquiry Group termed "some.of the most severe criticism-of a Government agency ever leveled 4

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by a Presidential Commission."57 It is generally accepted that the TMI-2 accident identified numerous safety-related areas of serious weakness and deficiency in the design, construction, operation, licensing, and regulation of nuclear plants in the United States. The accident also led to still more re-evaluation of unresolved safety issues. After the a.cnident, Harold Denton, Director of the PRC's Office of Nuclear Reactor Regulation, briefed the NRC Commissioners (in document SECY-79-344) on the Staff's plans to continue work on the pre-TMI unresolved safety issues discussed in paragraph 2.2 above. M r.

Denton al% anticipated (correctly) that TMI would result in expanding the scope of some of those existing issues, as well as in identifying new unresolve'd safety issues.

i 2.3.3. In May 1980 the NRC issued a TMI Action Plan (N UREG-0660).38 This was divided into five general categories: Operational safety; siting and design; emergency preparedness and radiation effects; practices and procedures; and NRC policy, organization and management. It included 176 different " tasks"-all safety-related-of which 58 fell in the category of siting and design. Of the six sets of safety issues discussed in Part III of this Affidavit, three were identified in the siting-and-design portion of NUREG-0680 (see pars. 3.1, 3.3 and 3.4 below), one in its emergency preparedness section (see par. 3.4 below), and two in its practices-and-procedures section (see pars. 3.5 and 3.6 below).

2.3.4. As had been done before the TMI-2 accident with regard to the NRC's list of unresolved safety issues, the 176 " tasks" identif!ed in the TMI Action Plan were given priority rankings on a " point value" basis. This priority ranking is shown in Tables I and B.2 of NUREG-0660; Table B.1 of NUREG-0660 shows the " point" system which aas used. These Tables are attached to this

Affidavit as Exhibit C. From Table B.1 one can determine that, though others may also fall in this group, any Action Plan " task" with a point value greater than 160 is necessarily considered by the NRC to have high safety significance.

Three of the six sets of safety issues discussed in Part III of this Affidavit clearly fall in this category (see pars. 3.1, 3.4, and 3.5 below). A fourth, having been assigned 130 points, probably falls in the NRC's high-safety-significance category for purposes of this ranking (see par. 3.3 below). Five of the safety issue sets discussed in Part III fall in priority groups 1 or 2 of the Action Plan.

I 2.3.5. A substantial amount of Congressional and oti;er governmental attention has been given to the NRC's response to the T311-2 accident, and it appears likely that pressure will be exerted to implement the NRC's Action Plan programs relating to accident probability in a relatively timely mannner.

liowever, ther- is no assurance that the Byron nuclear facility will benefit from these effort--particularly as to the unresolved safety issues discussed in Part III of this Affidavit, and particularly if Byron construction continues on its present schedule. This is so for three reasons, discussed in pars. 2.3.6, 2.3.7, and 2.3.8 below. .

2.3.6. First, many of the T5!I Action Plan " tasks" are geared to future nuclear plants, not plants which (like Byron) are already largely designed and constructed; and it is at best extremely doubtful that these " tasks" will be adequately factored into the Byron facility on present schedules. For example, the 58 " tasks" listed in the siting-and-design category of the T31I Action Plan are further subdivided into ten groups: Siting; consideration of reactor core degradation or melting; reliability engin.cring and risk assessment; reactor coolant system relief and safety valves; system design; instrumentat.5n and I

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controls; electrical power; TMI-2 cleanup and examination; general implications of TMI-2 for design and construction activities; and measures to mitigate small-break loss-of-coolant accidents ("LOCAs") and loss-of-feedwater accidents.59 2.3.6.1.' Most of these subcategories are significant to the

, safety of the Byron facility. But their timing, or the NRC's present plans for their applicability, will preclude an adequate consideration of them as to the Byron nuclear plant. The major thrust of the " siting" tasks will be to apply to applicants f r new construction permits; Byron's permits were granted five years ago. The " reactor core degradation" tasks involve matters which were not considered when Byron was designed and licensed, and may not be resolved until its construction is complete on present schedules. The implementatio.. of the

" reliability engineering and risk assessment" tasks for plants (like yron) under construction is officially " undecided." The " system design" tasks are in the main generfe-type studies which on past history (see pars. 2.2.3 and 2.2.9 above) will require substantial time for completion. The " general implice'.fons of TMI-2" tasks will relate primarily to the design and construction of new plants, not those now being built. And the " instrumentation and controls" tasks, though of great importance and relevance to Byron, may result in little or no impact on its design and construction. For example, a top priority task under the

" instrumentation and control" heading concerns accident monitoring instrumentation. This bs been discussed for years. The serious instrumentation inadequacies disclosed by the TMI-2 accident 60 emphasized the problem. But the NRC's newly revised (and more stringent) Regulatory Guide has not been fully implemented,61 and the TMI Action Plan is vague as t' what portions of the revised Guide plants such as Byror. will be required to meet.

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1 2.3.7. Second, even if the TMI Action Plan were better geared to designed-but-not-completed nuclear plants such as Byron, financial and other constraints may impede the implementing of new safety modifications and corrective programs at Byron. From the NRC's standpoint, agency management has recently indicated that budgetary shortfalls of $40-60 million for 1980 and .

1981 may further delay the schedule-relaxed to begin with, in a number of areas-for completing the TMI Action Plan " tasks."62 From Commonwealth Edison's standpoint, the statements of its officers regarding its extreme financial

" squeeze" (par.1.2.4 above) and the ineluctable increase in the cost of safety modifications which occurs as construction progresses (par.1.2.3 above) cast doubt on its ability or willingness to implement Action Plan developments and corrective measures at any point short of formal adoption of a specific requiremer t by the NRC, which may not occur until after Byron constructica is complete.

2. 3. 7 .1. In this regard, it is pertinent to observe that both the Kemeny ComrcisJion and the NRC's Special Inquiry Group found NRC operating license or "OL" proceedings (those in which the Byron plant is now engaged) generally inadaqcitte in terms of safaty issues. While "the full safety review" of a nuclear generating plant like Byron "does not occur until the OL stage," the Kemeny Commission pointed out that:63 "By then, hundreds of millions of dollars have been spent or committed in the construction process.

Therefore, the ultimate safety review may be influenced by economic considerations that can lead to a reli.ctance to order major changes at the GL stage."

The NRC's Special Inquiry Group agreed,64 adding that "most Licensing Board members who responded to a questionnaire" it sent out felt that "the formal

hearing process does little to enhance the quallt a reactor safetyn65 and

't emphasizing several " financial disincentives to safety" which become increasingly strong as nuclear plant ecnstruction nears completion 66-as is now happening at j the Byron facility. Of course, the longer construction continues, the worse these safety "disincentives" become.

2.3.8. Third, while one might have thought (and many hoped) that the af termath of the T31I-2 accident would lead to an amelioration of these difficulties, the indications are that it has not. If anything, the contrary has happened. Seven months after T31I-2, the Kemeny Commission found that the NRC (like its predecessor, the AEC) was still engaged in promoting nuclear power rather than in assuring nuclear safety, that "there is no well-thought-out, integrated system for the assurance of nuclear safety within the current NRC,"

and that "[t]he old AEC attitude is also evident in reluctance to apply new safety standards to previously licensed plants.n67 Nine months after T31I-2 (and two months after the Kemeny Commission " stung" the NRC with "some of the a

most severe criticism of a Government agency ever leveled by a Presidential

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Commissionn68), the NRC's Special Inquiry Group found a disturbing

" business-as-usual approach" at the NRC, "suggest[ing] that the Commission has not fully appreciated the importance of the problem" of identifying and adequately resolving safety issues, "or the kinds of changes that will be required to meet it."69 Proposing the same complete restructuring of the NRC as had the Kemeny Commission, the Special Inquiry Group dismissed almost sarcastically the NRC's post-Kemeny assurances of reform: "The (NRC] management structure which hadn't worked well before T3tl would now police itself."70 2.3.8.1. The criticisms of the Kemeny Commission and the NRC's Special Inquiry Group are, unfertunately, borne out by the NRC's handling

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of the TMI Action Plan. During the development of the Action Plan, the nuclear industry-the same group whose " financial disincentives to safety" were stressed by the NRC's Special Inquiry Group 7L-was consulted on an ongoing basis; but the public was never consulted or invited to comment.72 "(I] n almost all cases" in which Action Plan proposals were changed as a result of industry comments, the result was "a reduction of the r3quirements initially proposed by the (NRC] S taf f."73 As it emerged, the Action Plan addressed some of the most crucial safety areas by identifying serious problems and uncertainties, but suggesting no solutions other than still more studies which may offer some prospect of solutions at some unspecified future date; the section on core degradation and fuel melting ITMI Action Plan, section II.B.5) is a prime example. For a large number of Action Plan issues, the schedule for Implementing solutions is exceedingly long, h important areas, the Action Plan mandates measres the effleacy of which is seriously disputable.

2.3.8.2. Despite the manner in which the TMI Action Plan was developed md its deficiencies, in June 1980 the NRC decreed that while the nuclear industry could continue to attack new Action Plan safety requirements as unnecessary in licensing proceedings, no one would be permitted to question those requirements as inadequate or insufficient.74 Thus the public and Intervenors, excluded from input or comment while the Action Plan was being developed, are now forbidden to ask whether it is in fact adequate. One of the NRC Commissioners described this as " manifestly unfair and unwise."75 Another agreed, and doubted whether the June 1980 policy statement was even legally permissible.76 Certainly the policy is directly contrary to the strong views expressed by both the Kemeny Commission and the NRC's Special Inquiry Group, and (as one of the dissenting NRC Commissioners wrote) " embodies precisely the

complacency" about serious unresolved saf ety issues "that the Kemeny Commission, among others, suggested as a strong contributing factor to the accident at Three Mlle Island."77 2.4 Summary

2. 4.1. The history of the NRC's handling of unresolved nuclear safety issues, briefly summarized in paragraphs 2.2 and 2.3 above, is not calculated to inspire confidence that those issues will be resolved in an adequate and timely fashion, or that solutions to the many issues which directly affect the safety of the Byron nuclear facility will be adequately implemented at Byron. Neither the NRC's unresolved-safety-issue lists (par. 2.2 above), nor its post-TMI Action Plan (par. 2.3 above), nor the nature and timing of the operating license proceedings in which the Byron facility is now engaged (set. par. 2.3.7 above),

provides any assurance at all on these points--particularly if Byron construction continues on its present schedule. See pars.1.2.3 and 1.2.4 above. In Part III of this Affidavit we discuss these problems in greater detail, with regard to six sets of safety issues which the NRC itself has identified as (a) known, (b) unresolved, and (c) of high priority.

III. SIX UNRESOLVED SAFETY ISSUES 3.1. Systems Interaction 3.1.1. Traditionally, the NRC has approached accident and safety analysis on a system-by-system basis, using the " single failure criterion." See 10 C.F.R. Part 50, Appendix A, Criterion 21, for example. The effect of this is to require "that a system designed to carry out a specific safety function must be l able to fulfill its mission in spite of the failure of any single component within m - -e&. '-

the system, or failure in an associated system that supports its operation."78 As a result, emphasis tends to be placed on major failures within a single system. There is inadequate consideration of smaller failures (as both the Kemeny Commission 79 and the NRC's Special Inquiry Group 80 have pointed out, the tendency is to assume that if the large failures can be controlled "we need not worry about the analysis of 'less important' accidents"), and there is inadequate consideration of what happens if m_ultiple pieces of equipment fall from a common cause.81 There is also inadequate consideration of systems i n t e r a c t i o n-- t h a t is, the potential for adverse, accident-causing or accident-contributing interactions between or among different nuclear plant systems.32 This lack of consideration is aggravated by the tendency to assign

. the design and safety analysis of each system (e.g., mechanical, electrical, or nuclear) to a different team of engineering specialists, without ensuring that the I

work of those teams is sufficiently integrated to enable them to identify or i assess adverse interactions between systems.33 3.1.2. We will refer to these related problems of accid'ent and safety analysis--the lack of systems interaction analysis, the lack of multiple or

" common-cause" failure analysis, and the tendency of the " single-failure criterion" to exclude a large number of potential accident-causing events-as the

" systems interaction issue." This issue became extremely significant af ter the T311-2 accident, which itself involved not a single failure but rather a series of failures, or domino effect, which included both dependent and independent multiple failures. The Kemeny Commission found that "[t] he accident at T3tI-2 was a multiple-failure accident,"84 as did the NRC's Special Inquiry Group.85 But "[t] n the licensing process, applications are oeily required to analyze

' single-failure' accidents. They are not required to analyze what happens when 1

J two systems fall independently of each other,"86 nor to assess possible adverse interactions among systems. As a result, the Kemeny Commission called upon the NRC to emphasize:87 "a systems engineering examination of overall plant design and performance, including interaction among major systems and increased attention to the sossibility of multiple failure."

The NRC's Special ' Inquiry Group also criticized the NRC's safety cnd accident analysis as inadequate,88 noting that "one of the obvious lessons" of TMI-2 "is the critical need for overall plant and systems analysis," and (with particular regard to the concentration of engineering design and analysis teams on single specialized systems) that "[t] here is as much or more of a chance that safety rnatters will ' fall in the cracks' between two or more highly proficient technical groups as there is for a safety error to be made in any of the specifle groups." 39 3.1. 3. The systems interaction issue, while extremely important for nuclear plants generally, is of particular significance.for the Byron plant. The Dyron plant was designed by Westinghouse,90 and is quite similar in design concept to Commonwealth Edison's Zion facility.91 In 1977 (almost two years af ter Byron's construction permits were issued) Dr. Stephen Hanauer-described by the NRC's Special Inquiry Group as "one of the NRC staff's leading safety experts"92-wrote as follows in a letter regarding Edison's Zion plant, also designed by Westinghouse:93

" Westinghouse designs are characterized by the large number and types of interactions between control systems and related safety systems. They think this is 4

great. I think it is unsafe. This feud has been going on for years." { Emphasis sdded.]

Similarly, in late 1979 the NRC's Advisory Committee on Reactor Safeguards recommended an investigation of the systems interaction issue at another

Westinghouse-designed nuclear plant (Indian Point Units 2 and 3, in New York),

the design of which has substantial similarities to Byron's, as follows:94 "Thus, uncovering the potential for interaction of nonconnected systems will usually require careful, In-situ examination of the physical plant. This examination must consIifer all features having the potential 3 damage safety systems, including the safety systems themselves. The physical inspection of the plant. could oe approached by dividing the plant into ' compartments' following discernable structures-such as walls, ceilings, and floors with appraisable strengths and weaknesses. Loors, stairs, ventilation ducts, piping, and other penetrations would be evaulated for potential influence transport (fire, steam, hot air, etc.). Structures, which act as barriers to the flow of a damaging influence, would be assessed for the adequacy of their resistance to such influences.

"In each compartment the elements of the safety systems, including such extensions as instrument lines and power of control wiring should be identified on a ' train' basis. The physical vulnerability of the safety system elements to nonstandard conditions (temperature, pressure, water, spray, etc.) should be identified. The character!stics of such svstems as influence generators under faulted

conditions would have to be assessed if such system elements exist as redundant elements within the identified ' compartment' boundaries.

"The influence potential of all non-safety elementsWH including such items as sewer and drain lines, combustible gas transport and storage, compressors, and heavy-power circuits and transformers, within the given compartment should be

, assessed with respect to potential for damaging or disrupting (as with induced electrical noise) critical system (s) within the ' compartment' and the

' compartment' boundary itself.

"The invasion of damaging influences through the barriers or boundaries into the identified compartment would also have to be assessed. This would include consideration of entry of personnel carrying influence generators such as welding equipment.

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"Jecial S consideration would have to be given to

,t_he h identITfeaIlon of convergence of safTty functions into single compartments and the degree of ~

convergency within the gen space. The study M interactions between nonconnected systems would also have to include the possibility of non-visible interactions, such as the possibly adverse effect of

failure of one buried i e on a neighbor due to scouring. A study o p ant drawings would be

. required in cor.nection with this aspect." [ Emphasis added.)

F 3.1.4. Despite its direct applicability to the Byron nuclear plant and the fact that the NRC's Office of Nuclear Resetor Regulation has " clearly recognized" the 'need for this form of analysis,"96 there is no assurance that i

the systems interaction issue will receive adequate treatment in connection with the Byron facility. The issue has been a known, unresolved safety issue since at least 1974, when the ACRS requested that the NRC Staff give attention to it.97 It was a Category A high-priority issue in NUREG-0410, published in January 1978. See par. 2.2.4 above. It was classed as a " potential high risk" issue in the NRC's 1978 risk-based evaluation of unresolved safety issues.- See par. 2.2.5 above. It was one of the top 20 unresolved safety issues listed in the NRC's January 1979 Report to Congress (par. 2.2.6 above), and in the Denton/ Steering Committee " point value" manpower allocation directives (par. 2.2.7 above); in fact, no issue had a higher point value. It is part of the TMI Action Plan (as

" task" II.C.3), with a priority 1 classification and a point value indicating that it I was again determined to be of high safety significance fcr the Action Plan as well: pars. 2.3.3 and 2.3.4 above. Yet the Byron Final Safety Analysis Report PFSAR"), as amended to date, does not include, mention, or even refer to any systems interaction study or analysis. i l

3.1.5. Nor can it be expected that the- systems intersetion issue can or will be _ adequately dealt with for the Byron facility simply through a 1  :!

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generalized process. As with other " generic" issues (see pars. 2.2.8, 2.2.9, 2.3.8 above), the history of the systems interaction issue has been one of repeated j recognition, classification, reclassification, and relisting without much progress.

The ACRS felt that at lesst some types of systems interaction which might lead to a significant degradation of safety could be identified and then dealt with .

through a study of Licensee Event Reports (LERs), which each reactor owner '

i must make when abnormal incidents occur. However, tne ACRS itself concluded in NUREG-057298 that a detailed review of LERs cannot be expected to identify all systems interactions. By far, the bulk of the LERs deal with failure of individual components and equipment, with relatively few cascades of f ailures-such as the T311-2 accident-resulting from an initiating event.99 Furthermore, both the Kemeny Commission 10 0 and the NRC's Special Inquiry Group 101 concluded that (again, as with the known but igncred reports of similar occurrences which preceded the TS1I-2 accident: see par. 2.3.1 above) the LER

system itself was seriously inadequate as a means of ensuring reactor safety; -

1 and the Special Inquiry Group further concluded that while the NRC had taken some post-TSII steps to deal more effectively with LERs, no real progress had been made and a " business-as-usual approach" was dominant.102 3.1. 6. Similarly, in 1975 the NRC-commissioned "Rasmussen Report" (WASH-1400) recognized that the greatest accident risk is posed not by the large single (or. so-called " design basis") accidents, but by small loss-of-coolant accidents compounded by multiple failures or human error,103 such as the 1970 and 1971 accidents at Commonwealth Edison's Dresden nuclear facilityl04 and the T311-2 accident. While WASH-1400 was severely criticized as over-optimistic by the NRC's Lewis Committee in 1978,105 and the NRC then formally disavowed WASH-1400's ot erall conclusions about the low risk of nuclear accidents,N6 l '

_. . _ _ - . _ . . ~ _ . ,

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its recognition of where the major accident risks lie has not been faulted. Yet the NRC's Special Inquiry Group found that "[t] hese types of potential accident sources have...been all but ignored by the NRC in the regulatory review process."107 3.1. 7. Af ter the systems interaction issue became a high-priority issue in NUREG-0410 (par. 2.2.4 above), NRC contracted with Sandia Laboratories to study it. In December 1979, Sandia issued its Phase I report, clarifying some of the potential undesirable interaction areas not adequately taken into account in the NRC review process.108 In parallel, the NRC's post-TMI " Lessons Learned" Task Force concluded nearly a year ago (in NUREG-0585) that the systems interaction issue required prompt and thorough attention during the licensing process:109 "The interactions between non-safety-grade and safetvg equipment are numerous, varied, and complex and have not been systematically evaluated. Even though there is a general requirement that failure of non-safety grade equipment or structures should not initiate or aggravate an accident, there g n_o comprehensive and systematic demonstration that this has been iccomolished.

F u rther m ore, the term 'f ailure' w hen 9pplie d to non-safety-grade _ equipment has generally been defined as

' failure to operate upon demand.' There is evidence from Three Mile Island and other operating and licensing experience that the failure modes should also include unintended operation or unusual operation that might result from process or environmental conditions accomoanying an event. For example, the high humidity or temperature following a loss-of-coolant acaldent might cause a relay, control circuit, or other component in a non-safety-grade system to operate or to function in a manner that unacceptably exacerbates the event.

"The Task Force concludes that comorehensive studies of the interaction of non-safety-grade comoonents, equipment, systems and structures with safety systems and the effects g these interactions during normal ooeration, transients, gd accidents need to be made Dv all licensees and license apolicants (see Recommendation 9).

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I l This would constitute a significant alteration of the current- -

! unresolved safety issue concerning systems interactTon. The '

Office of Standards Development has previously been requested to develop a Regulatory Guide that would specify generic requirements for some . safety-related systems that do not presently fall within the safety-grade classification.

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' This . effort would have to be closely coordinated with the study by licensees that.we are now recommending. In the interim, the effects of the abnormal conditions that

accompany transients and accidents on the operation and failure of non-safety-grade items would be reviewed by all '

licensees to determine if there are any probable adverse interactions. The extent of simultaneous interactions j-considered in this review should reflect the number of non-safety grade. items simultaneously exposed to conditions j

for which they were not designed. Eculoment identified as; the cause of unacceptable interactions should be appropriately modified to reduce the probability of that s

interaction, or the safety system that is adversely aTfected I should be mod 71ed to cope with the interaction. In either i event, op,erating orocedures and operator training must be expanded to include consideration of the possible permutations and combinations of non-saletv-grade system interactions with safety systems.' TEmphasis added.]

3.1.8. Thus the systems interaction issue has been recognized again

• 4 and again as a high prio.av, high-risk-potential, unresolved safety issue. It is directly applicable to the Byron facility, as Dr. Hanauer pointed out; yet the j Byron FSAR fails even to mention it, and is generally silent concerning the

potential safety implications of "non-safety" systems. At least three different

, possible approaches to the issue have been suggested: a " fault-tree" approach,

, as in WASH-1400; a physical plant inspection of interaction possibilles, as the t

ACRS suggested for the Westinghouse-designed Indian Point nuclese plant; and a '

site-specific ' analysis . based on a single initiating event (e.g., an earthquake), as the ACRS requested be performed for the Diablo Canyon plant in California.110 Yet it is officially ' undecided" whether any approach to .the issue will be implemented for " plants under construction," such as Byron.lll That is not -

J. acceptable. Byron-particularly Byron, given Dr. Hanauer's and the ACRS' j concern over Westinghouse plants in this regard-needs a thorough, systematic  !

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review from the standpoint of the systems interaction issue to assess the potential for -design modifications and corrective programs to minimize the safety risks inherent in potential adverse interactions and multiple common-cause failures.

3.2. Steam Generator Tube Integrity

3. 2.1. In a pressurized water nuclear reactor ("PWR") such as those at the Byron facility, water in the reactor coolant system (a closed system of pipes running between the reactor and the steam generators and sometimes a

called the " primary loop") is kept at high pressure-approximately 2250 pounds per square inch-in order to keep the water from boiling at the high temperature (600 oagrees FrNenheit) at the outlet of the reactor pressure v essel. The water in the primary loop is heated in the reactor core, and then carried by the primary loop piping from the outlet of the reactor vessel to a steam generator. In the steam generator, a transfer of heat takes place. The primary loop water travels down through the steam generator tarough thousands of thin tubes; the large overall surface area of these tubes facilitates the use of the primary loop water to heat another, completely separate stream of water (the feedwater system or " secondary loop"); and this separate, secondary loop water is then converted to steam. The secondary loop steam is then piped through the main steam line to the steam turbine, which runs the electricity-producing generator.

3.2.2. The thousands of thin ster', get'erator tubes through which the extremely hot, pressurized, and radie-hize primary loop" water travels are notoriously subject to deterioration, h ; #1 ems with PWR steam generators, some or all of which have been experienced at the majority of PWRs after only a comparatively few years' service, include contamination, vibration, fretting,

-water hammer, cracking, wastage, pitting, denting, high-cycle fatigue, and

erosion / corrosion.ll2 The general problem of steam generator tube integrity has been an unresolved nuclear safety issue for years. It was listed as a high-priarity unresolved issue in NUREG-0410 (par. 2.2.4 above); it was classified as one of the seventeen most important unresolved safety issues in NUREG-0510, the NRC's January 1979 Report to Congress (par. 2.2.6 above); and 'It was one of the top 20 unresolved issues in the Denton/ Steering Committee " point value" ranking and manpower allocation directives (par. 2.2.7 above).ll3 3.2.3. . The steam generator tube integrity issue applies specifically to the Byron nuclear p; ant. The specific problem of steam generator tube integrity at Westinghouse nuclear plants such as Byron is Task A-3 in NUREG-0410 and is also listed as a high-priority unresolved safety issue in N UREG-0510.

Westinghouse-design . ' nuclear power plants such as Turkey Point in Florida, Surry in Virginia, r4nd Prairie Island in Minnesota have experienced serious tube degradation problems, resulting in substantial amounts of outage time and worker radiation exposure well above " normal" levels (themselves unexpectedly high:

.ee par. 2.2.9 above), after only five or six years af operation. In at least two cases-(Surry and Turkey Point) complete replacement of the steam generator tube bundles-at a very large total cost, estimated by Florida Power & Light Co. to approximate as much as $380 million for Turkey Point--has been required.U4 Anc this is projected to be a future necessity at a number of other Westinghouse-designed plants, including plants significantly similar to Byron.115 The Surry and Turkey Point problems are particularly disturbing in light of the fact that both plants had previously been touted by Westinghouse as examples of the absence of steam generator tube problems,116 and the recent Prairie Island problem adds a further dimension of concern, both because it appears to be the result of a new, as yet not understood phenomenon l li and because both Prairie l

Island Unit 2 (where the problem occurred) and Byron utilize the all volatile treatment ("AVT") water control program as the means-or so it was hoped-to minimize steam generator tube wall thinning and degradation.118 3.2.4. The great economic cost of dealing with steam generator tube degradation, in terms 'of lengthy plant outa' esg (from six months to a year or more if tube replacement is required) and in terms of direct tube replacement costs, is not the- only element of the steam generator tube integrity issue.

Safety is also importantly at stake. On reported occasions, " transients" (i.e.,

small accidents) occurring at nuclear plants operating with degraded steam generator tubes have resulted in the release to the environment of undesirable and uncontrolled quantities of radioactive material leaking from the " primary loop" reactor coolant system. The October 19 7 9 event at the Westinghouse-designed Prairie Island nuclear plant is a prime example.Il9 In addition, the potential for a major accident exists, and has been identified by the NRC, if a critical number (not by any means large) of the steam generator tubes fail during a loss-of-coolant accident ("LOCA") or main steam line break accident ("MSLB").120 If the LOCA-imposed shock load causes a critical number of tubes to fail, the in-flow from the secondary side can retard reflooding of the core by the emergency core cooling system ("ECCS"), preventing adequate cooling of the core and thus leading to core damage or even meltdown: another TMI-2, in effect. Or in the event of an MSLB, steam generator tube failure provides the potential for releases of the " primary loop" coolant and asso'clated  !

radioactivity to the environment, through the opening of the main steam ~ safety

. 1 valves. Opening of those valves occurs frequently during transients; for example, it was experienced during the October 1979 Prairie Island event.

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l 3.2.o. Despite its longstanding recognition as a major unresolved proble , the' steam generator tube integrity issue has not adequately been dealt with to date. There is no assurance that present practices will control deterioration to the point that a normal steam generator lifecycle (40 years) without tube replacement can be expected; the evidence-including Surry, Turkey Point, and Prairie Island-is to the contrary. Indeed, the NRC's review of the Prairie Island 1979 tube degradation indicates that the issue is if anything increasing, and presenting new problems. In January 1980 the NRC Staff noted that "because of the need to review recent occurrences of steam generator tube problems at plants, manpower problems are de reloping" in terms of Staff efforts to pursue the issue.121 And following an inspection of the Prairi2 Island Unit 2 steam generator tubes, the NRC concluded in February 1980 that its review of the possible causes of the tube defects yielded no evidence of cold work; the cause was " believed" to be corrosien due to local concentration of resin fines at the tube support phte. But "similar cold leg indications have been found in the Takahama Unit 1 plant [a Westinghouse-designed nuclear facility in Japan]

although not due to the same corrodent." The thinned tube surfaces were clean; the cause was therefore " presumed" to be a dissolved corrodent (we do not know what), rather than particulate matter.122 3.2.6. Clearly not enough is known. The steam generator tube integrity issue remains open and unresolved. Measures to correct this significant safety deficiency must be developed and implemented for the Byron facility, before construction is completed, operation begun, and the cost and radiation hazard of implementing adequate solutions reach or pass acceptable limits. See pars.1.2.4, 3.2.3 above. Yet the NRC's handling of the steam generator tube integrity issue on a generic basis, and the Byron FS AR, give no

indication that adequate corrective measures will be either considered or implemented at Byron.123 3.3. Equipment Qualification and Deterioration 3.3.1. What is called " environmental qualification" of nuclear power plant equipment relates to the ability of the equipment, as designed, to withstand the stresses placed upud=it b[Aormal operation. These stresses

! _ include not only 'the physical wear and tear to which a piece of equipment is i

suajected by its own operation (e.g., the constant compression and release of a spring) or by what .it is used for (e.g, the corrosivc effect of primary coolant on the steam generator tubing which carries it), but also the physical wear and tear to which equipment is subjected by the particular environment in which it must op'erate-for example, at the Byron plant Edison personnel have asserted that the " environment" of a splice box and conduit for electrical control cable is water, since they expect the box and conduit will be filled with water during normal operation.124 Equipment is " qualified" if it is demonstrably dedgned to withstand these stresses arising from environment, operation, and use.

Otherwise it is not qualified, and may fall unexpectedly, perhaps dangerously, during operation. (This was why the NRC's Office of Inspection and Enforcement, not surprisingly, complained of the splice box mentioned above:

No one knew whether the "eavironmental extreme" of an electrical splice box being filled with water had been " considered in [its] design."125) 3.3.2. Several years ago an NRC review of eleven nuclear plants show ed that environmental qualification of safety-related equipment was inadequate on older nuclear plants.126 The issue of environmental qualification of Class IE safety-related equipment was listed as a high-priority unresolved safety issue in NUREG-0410 (par. 2.2.4 above); it was also listed as a potential ,

high-risk safety issue in the NRC's risk-based evaluation (par. 2.2.5 above), was

M one of the seventeen most important unresolved safety issues listed in

' NUREG-0510, the NRC's January 1979 Report to Congress (par. 2.2.6 above), and was one of the top 20 issues identified in the Denton/ Steering Committee " point value" rankings and manpower allocation directives (par. 2.2.7 above). Only six issues (out of 12^ had a higher point value.127 It, was recognized that the environmental qualification issue applied to plants under construction, such as Byron, as well as to currently operating plants.128 In fact, in 1975 the ACRS specifically raised the issue with respect to Byron, urging that it be " resolved by [Com.nonwealth Edison] and the NRC Staff" for the Byrou plant.129 3.3.3. There are at least three major problems with environmental qualification of nuclear plant equipment. The first is that the present NRC qualification requirements apply only to " safety-related" equipment--a determination initially made by the nuclear plant license applicant.130 The applicant has obvious incentives to minimize the amount of " safety-related" equipment, since " items not labeled ' safety-related' need not be reviewed in the licensing process, are not required to meet NRC design criteria, need not be testable, do not require redundancy, and are ordinarily not subject to NRC inspection."131 Both the Kemeny Commission and the NRC's Special Inquiry Group found that this " safety-related" limitation is artificial, unworkable, and

! " arbitrary...as a boundary of' NRC's attention.n132 Consider, for example, the PORV and block valve which were major factors in the TMI-2 accident (par.

2.3.1 above):133

"[ A] t T M I-2, the PORV was not a

' safety-related' item because it had a block valve behind it. On the other hand, the block valve was not ' safety-related' because it had a PORV in front of it."

3.3.3.1. S cond, the effects of aging on equipment qualification ire a serious unresolved problem. Aging is a very important consideration. It may be meaningless to prove that a new piece of equipment, just off the production line, will operate properly during accident conditions if the same piece of equipment degrades after a few years of normal use (like, for instance, steam generator tubing: see par. 3.2 above) to the point where an accident environment causes it to fail or misoperate. Manifestly this is untenable for safety equipment. Yet although the American standard for qualification of safety-related electrical equipment (IEEE 3'!3-1974, released in 1971 and revised in 1974)l34 identifies aging as an important consideration, it was

greeted with almost unanimous lethargy in the nuclear industry, has been given only lip service at best, and is not complied with today.

3.3.3.2. Third, there are several national standardsl35 and other guidesl36 which relate to equipment qualification. However, these documents are subject to diverse interpretations, and both their meaning and application in the context of nuclear facilities such as Byron are unclear in important respects.

3.3.4. The issue of nuclear plant equipment environmental qualification is, as previously noted, of substantial safety significance (according to the ACRS as well as a number of recently issued NRC Regulatory Guidesl37) and directly pertinent to the Byron nuclear facility. But there is no assurance that the issue has been or will be adequately considered in the context of Byron. The Byror FSAR138 provides a review of the status of Byron's compliance with NRC Regulatory Guides L1 through 1.143. But (as is more fully discussed in paragraph 3.5 below) as to these Regulatory Guides, some of which deal with equip m ent q uali fica ti on, more often than not the

FSAR announces only a " commitment" to comply with the " intent" of the Guide, providing few (if any) details of how this will be achieved or of whether t.%

specific requirements of applicable equipment qualification are being met.

3.3.5. Thus the Byron FSAR is inadequate even to indicate the status

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of Byron's compliance with existing equipment qualification requirements-let alone to address the serious safety-related unresolved problems in this area.

The NRC Staff has recently circulated for comment a draft of a recommended

" resolution" of some of the unresolved problems, in the form of NUREG-0588.139 But NUREG-0588, like the Byron FSAR in this area, raises more questions than it answers. For example, NUREG-0588 attempts to. provide guidance and interpretation of the existing national standards and guides related to equipment qualification (par. 3.3.3.2 above). But it fails completely to address the crucial, unresolved issue of the unworkable, arbitrary, and potentially hazardous limitation of equipment qualification requirements to so-called " safety-related" equipment (par. 3.3.3 above). And NUREG-0588 deals with the equally crucial unresolve; issue of aging (par. 3.3.3.1 above) by stating that aging should be considered, without clarifying the problem of how to include aging effects. This is an important gap. In the past, accelerated life tests were used in an effort to quickly simulate- the impact of the numerous operating cycles and limiting values of environmental parameters a given piece of equipment is expected to experience over its lifetime. Ilowever, it is now well known that this approach is frequently inadequate and misleading, because it cannot reproduce all the effects of aging over what is in reality a long, slow process. NUREG-0588 fails j to address this issue.

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3. 3. 5.1 In fact, NUREG-0588 is uncertain and incomplete.

on its face. This is apparent from the following excerpt:140 "To promote more orderly and systematic implementation of equipment qualification programs in industry and to provide guidance to be used by the NRC staff for use in the ongoing license reviews, the staff has develooed a number of positions on selected areas of the qualification issue. These positions, which are presented in this report, provide guidance on the establishment of service conditions, methods for qualifying equipment, and other related matters. They do not address in detail all areas of qualifications, since certain areas are not yet well understood and are the subjects of research studies conducted a the NRC and g the induTtry. For example, the effects oT_ aging, sequential versus simultaneous testing, including synergistic effects, and the potential combustible gas and chloride formation in the equipment containing organic materials are being evaluated. It is expected that these studies will lead to the develooment of more detailed guidance in the future and may reouire changes M these positions.

"These positions were develooed prior y the staff comoletion of the TMI-2 event evaluation, and any additional requirements or, modifications to these positions as a result of, this evaluation will O_e identified later. In addition, seismic qualification is being pursued on a case-by-case basis by the Seismic Qualification Review Team (SQRT) and is outside the scope of this document."

[ Emphasis added.)

3.3.6. Given the manif est incompleteness and uncertainty of NUREG-0588, the seriously inadequate treatment even of existing equipment qualification requirements and standards in the Byron FS AR, and the l long-recognized safety importance of the environmental qualification issue, it is apparent that this issue is one of safety significance which must be pursued and resolved at the Byron nuclear plant-as the ACRS said five years ago. Yet neither the Byron FSAR nor NUREG-0588 provides any assurance that this will be done. If anything, both indicate the contrary, and suggest that here as with

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other safety issues, in the words of the NRC's Special Inquiry Group, although

" serious safety problems" have been identified and " underscored by ringing statements," nevertheless they remain outstanding.I41 As with steam generator tube degradation (par. 3.2 above) or the malfunctioning PORV which was a major factor in the TMI-2 accident (par. 2.3.1 above), a failure to deal adequately with equipment qualification can have extremely serious consequences. Ti.e issue must be confronted and resolved for Byron before additional equipment with potentially inadequate qualification is installed in the plant.

3.4. Evaluation of Potential Accidents and Corrective Measures 3.4.1. Nine years ago the NRC's predecessor (the AEC) published (36 Fed. Reg. 22851, Dec.1,1971) a proposed Annex to 10 C.F.R. Part 50, Appendix D, classifying potential nuclear plant accidents into nine categories. Class 9 accidents were rather loosely defined as those involving " sequences of postulated-successive failures more severe than those postulated for the design basis for protective systems and engineered safety features"-that is, accidents beyond the designed capacity of the plant's safety systems to control. In shorthand, these Class 9 accidents, which include significant fuel damage or core melt events, are sometimes termed " accidents beyond the design basis." The proposed Annex conceded that "[t] heir consequences could be severe," but took the position that they need not be considered in environmental analyses under the National Environmental Policy Act ("NEPA") because "the probability of their occurrence l

is so small that their environmental risk is extremely low."142 Since the I issuance of the proposed Annex, NRC environmental statements, including the FES143 and ER/OLl44 for the Byron facility, and the Byron FSAR,145 have not

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discussed the consequences of a Class 9 accident at Byron from a safety standpoint. In fact, because of the supposed improbability of Class 9 accidents, nuclear plants-including Byron--are not designed to guard against their .

occurrence.146 Hence there are two major unresolved issues concerning Class 9 accidents shich have -not been considered for the Byron nuclear facility from -

either a safety or an environmental (NEPA) standpoint:

a

-How probable is a Class 9 accident at Byron, and what can be done (including design modifications) to reduce that possibility?

--What' would be th4 consequences of a Class 9 accident at Byron, and what can be done (including design l

modifications) L reduce or mitigate those consequences?

It is important to note that the environmental aspect of these questions is by no means unrelated to the safety aspect. The NRC has read NEPA to require not only the analysis of environmental consequences (e.g., potential radiation release from an' accident) to determine what

• hose consequences are, but also the " tax [ing] (of] appropriate measures to mitigate or eliminate" those consequences.147 As to Class 9 accidents, however, this inquiry has never been undertaken for Byron.

3.4.2. The exclusion of Class 9 accident considerations has in the past few years been based on two premises: first, the " low probability" I

statement in the proposed Annex to 10 C.F.R. Part 50, and second, the

, numerical estimate of the overall risk of reactor accidents in the "Rasmussen i

Report" (W ASH-1400).148 Like most such documents, the Byron FES (in Section

7) and ER/OL (at page 7.1-21) deal with Class 9 accidents only by restating the l

conclusion announced in 1971 in the proposed Annex and briefly referring to the t

l quantitative analysis in WASH-1400. However, the proposed Annex has now been i

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officially withdrawn,I49 and in January 1979 the NRC formally concluded that NRC "does not regard as rel%ble the (WASH-1400] numerical estimate of the overall risk of reactor accidents.nl50 -Consequently, the theoretical basis on which the exclusion of Class 9 accident considerations rested no longer exists.

Nor is there a practical basis for that exclusion. We now know that Class 9 accidents can and do happen. The NRC 5 taff has formally concluded that the TMI-2 accident was a Class 9 accident.151 In somewhat more detail:

3.4.2.1. First, the source of the generalized conclusions stated in the proposed Annex to 10 C.F.R. Part 50 has never been altogether clear. The assumptions which it directed should be used in analyzing the environmental consequences of acccidents "[do] not contribute to objective consideration;n152 its exclusion 1 Class 9 accidents (for which it provided no supporting analysis) ap arently did not rest on any then-existing accident risk assessment studies;153 r.1d the scientific basis for its accident assessment policy has been described as "at best unidentifiable, at worst non-existent."l54 In fact, the NRC stated at one point that it was "not aware of any" documents prepared by or for the Staff describing the reasons or facts supporting the Class 9 accident exclusion.155 After the proposed Annex >ias published, a number of severe criticisms were received during the public comment period, raising (among others) those deficiencies. The NRC, however, neither responded to those critical comments nor ever formally adopted the Annex as official policy--though for the past nine years it has been generally treated as binding.156 Among the other Federal agencies whien criticized the Annex and j expressed the need for a better treatment of accident risks (and in particular of Class 9 accidents) were the Environmental Protection Agency (" EPA") and the Department of the Interior.157 l

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l l i 3.4.2.2. Second, in 1972 the AEC commissioned a Reactor Safety Study to analyze systematically the probabilities and consequences of various nuclear accidents, including serious accidents. The results, made public in 1975 as WASH-1400 (the "Rasmussen Report"), though obviously not forming the basis for the Annex's conclusions four years es: lier,158 were thought for a time to lend support to those conclusions because they assigned extremely low probabilities to the occurrence of a serious nuclear accident. However, the Rasmussen Report was itself criticized, and in 1977 the NRC organized a Risk Assessment Review Group headed by H. W. Lewis to re-evaluate the Rasmussen Report. The Lewis Group's report to the NRC, published in 1978 as N UR EG/C R-0400,159 concluded that the accident probability estimates contained in the Rasmussen Report were highly suspect. In January 1979, the NRC issued a forrnal S tatenent of Poliev disavowing the Rasmussen Report accident probability estimates as "not... reliable.n160 Since then, the President's Council on Environmental Quality ("CEQ") has concluded that the exclusion of Class 9

' accidents from consideration is "without credible scientific support.,161 3.4.2.3. Third, the theoretical basis (such as it was) for excluding Class 9 accidents from consideration having thus disappeared, on March 28,1979 the TMI-2 accident removed any practical justification for that exclusion as well. In August 1979 the NRC Staff, "[a]pplying [its] information" about TMI-2 "to the description of a Class 9 accident contained in the Annex,

! formally " concluded that the occurrence at Three Mile Island was a Class 9 accident."16 2 And TMI-2 was not a freak, or the " exception that proves the rule." Occurrences similar to TMI-2, in which serious consequences were perhaps averted only by luck, have taken place at other pressurized water reactor nuclear plants.163 In addition, both the Kemeny Commission

l and the NRC's Special Inquiry Group concluded that TMI-2 could not be treated as an isolated event. The Kemeny Commission noted that even on the probability analysis of WASH-1400, the likelihood of a TMI-2 was "high enough" that "such an accident should have been expected," and cautioned that "we must not assume that an accident of this or greater seriousness cannot happen again, even if the changes (the Commission] recommend [ed] are made.n164 The NRC's Special Inquiry Group noted that TMI-2 "could have happened in a lot of places" and that absent "ftmdamental changes," "similar accidents...are likely to recur,"

and specifically called for a change in the NRC's Class 9 exclusion policy.165 3.4.2.4. Fourth, af ter the TMI-2 ace! dent several governmental groups, including the NRC's own Advisey Committee on Reactor Safeguards ("ACRS"),166 have recognized the need for a plant-by-plant assessment of nuclear accident probabilities and mitigating measures in

, connection with plants due to begin operation in the "near term." As the NRC's Siting Task Force said in NUREG-0625,167 "the risk to the public from a range of accidents including accidents beyond that for which the plant is designed (Class 9, or serious nuclear accidents) is sufficiently high to be a consideration in siting." In NUREG-0642,168 the ACRS bluntly stated that Class 9 accidents "should be considered in deciding on the future appececn to siting,3 reactor design, and to o emergency measures." CEQ (which has statutory oversight duties with regard to whether other federal agencies are fulfilling their NEPA obligations, and whose readings of NEPA are " entitled to great weightn169) concluded that the NRC has no " legal justification" for failing to consider Class 9 accidents in environmental impact statements, that "the need for a policy Z i

1 revision [is] compelling," and that the NRC should not only. require consideration of Class 9 accidents in future environmental statements but should also supplement existing statements in this regard.170 A March 20, 1980 letter from CEQ Chairman Speth to NRC Chairman Ahearne was quite forthright:

"The results of our review of imoact statements orecared by the NRC for nuclear powTr reactors are very

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disturbing. The discussion in these statements of potential accidents and their environmental impacts was found to be largely perf unctory, remarkably standardized, and uninformative to the public...[V)irtually every EIS contains essentially identical, ' boiler-plate' language written in an unvarying format. The typical EIS does not consider or analyze the possibility of a major accident even though it is tiiese ' Class 9' accidents which have the potential for greatest environmental harm and wh:3h have led to the greatest public concern. Moreover, for o those accidents which are tyoically discussed in an EIS the pot ential imoacts on human health and the enTirEmE, a.m on sented in a cursory and inadecuate manner with little attention to public understanding."

Hence the CEQ 1etter recommended (among other things):

"We believe that the new policy should be based on the sensible approach of discussing the environmental andd other consequences of the full range of accidents that might occur al nuclear reactors, including accidents now classified as Class 9. This should include core melt events.

In addition, EIS's should present the best estimates of the likelihood off such events. In order to comply with the disclosure requirements of NEPA, the NRC should include in the analyses the likely range of environmental and other consequences from severe and other accidents....

"We also urge the Commission to broaden its range of variables (e.g., radiation oathways) in determining accident imoacts, aM expand its discussions in EIS's of the impacts of nuclear accidents on human health, the natural environment and local economies. Site specific treatment of data should be substituted for 'boilerolate' assessment of

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Ecident initiating events and oTtential impacts, and EIS3 should be comprehensible to non-technical members of the public.

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d 3.4.2.5. Fifth, as a result of these developments, in June 1980 the NRC formally withdrew the proposed Annex, from which the Class 9 accident exclusion policy had grown, and directed that the Annex "shall not hereafter be used by applicants nor by the staff."i71 The NRC concluded that, among other defici ncies, the assumptions prescribed in the Annex "[d]o not contribute to objective consideration" even of accidents which are analyzed in environmental statements, and that the Annex prohibited consideration of precisely those accider.ts (Class 9) which " dominate the accident risk."172 3.4.3. Two of the potential environmental and safety consemences of a nuclear accident are the releases of radiation by the " air pathway" (i.e., into the atmosphere, where wind can carry it) and the " liquid pathway" Q.e., into groundwater which can contaminate rivers and streams). Either " pathway" can lead to public exposure to uncontrolled releases of radiation from an accident.

In addition to the complete exclusion of the " air pathway" consequences of Class 9 accidents, another major deficiency in the Byron FES17 3 and ER/OLl74 and FSAR17 5 is the lack of any specific discussion of the impacts of Class 9 accidents on the " liquid pathway" or potential corrective measures for such an accident. As with the now-rejected Class 9 accident exclusion policy, the failure adequately to consider liquid pathway accident impacts and corrective measures has been severely criticized for some time, and has been an unresolved issue sh.. e NUREG-0410 (par. 2.2.4 above:. This has particular significance for Byron in view of the hydrogeological characteristics of the Byron site, such as permeable rock and high groundwater.176 3.4.4. The failure to consider liquid pathway accident impacts and corrective measures stems in large pect from the accident analysis in the "Rasmussen Report " WASH-1400, whien stated that the " effects of contamination

I on water supplies have not been con,idered in detail" because of a broad assumption that streams and rivers would be contaminated for "only a short tim e."177 No detailed analysis was offered to support this conclusion. The potential economic and safety effects of major liquid pathiray contamination from strontium-90 and other isotopes are very large,173 and the cursory I treatment of water contamination in WASH-1400 is a major flaw. This is particularly true of potential liquid pathway contamination from core melt releases, which (perhaps because of a lack of direct experience and the ill-defined parameters of interaction between a molten core and the surrounding soil and water table) WASH-1400 did not evaluate with the same care given to the more readily observable effects of air pathway accident radiation releases.179 3.4.5. The Department of the Interior disagreed with 'he WASH-1400 asstenptions and conclusions concerning liquid pathway accident impacts, and in 1977 recommended additional study of the problem, including the effects of variations in hydrogeological conditions betweea different nuclear plant sites.180 Both the general inadequacy of NRC accident assessments and the specific liquid pathway deficiency appear from the NRC Staff's 1978 " Description of Problem" in NRC Task Action Plan A-33:

"In 1971, the AEC determined that, consistent with NEPA, the environmental assessments of requests for construction permits and operating licenses should include i consideration of the possible impacts from accidents...

"The approach in these assessments, tyoically is limited ,too p_ reparation of a two-oa_g narrative summary that qualitatively descrioes accident probabilities and the rationale for concluding that accident risks are low and a one-og table that provides numerical estimates o]

consecuences of various categories g accidents (excluding Class 9 events). The aoproach 3 develooing these consequence estimates also involves a largely simolistic analysis; minor adjustments are made from case to case

(basically to account for variations in power level, exclusion boundary distance and population density). These numerical estimates g_e e also limited _to al pathway consequences.

"The Environmental Proter., .on Agency (EPA) and the Department of the Intefior (DOI) expressed the need for an improved treatment of accident risks and an expansion 2

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the Staff assessments Io include quantitative estimate of Class 9 accidents.

1 ... Af ter extended discussions, the NRC Staff reiterated its 1973 commitment to update the standard assumptions in the proposed Annex A. As a precursor to this update, the Staff committed to an extension of the W ASH-1400 study to include a more In-depth evaluation of 1

Class 3-8 accidents and to further explore the significance of variations in site and plant design characteristics. The Department the Interior has routinely suggested that more attenti[on licuid pathway.

be h to the site risks associated with In mid-1977, DOI and NRC Staff met to discuss the DOI's generic concerns. DOI was informed of the Staff's programs to augment the generic studies in WASH-1400, but no commitments were made to revise the current approachTwhich, as,noted above, includes no, discussion on the impacts of accidental releases to the liquid pathwaW."7 Emphasis added.)

Subsequently, the NRC accepted the Department of Interior suggestions and 4

instituted a research program at Sandia Laboratories. The Sandia study results were released in draft form to the NRC in January 1980.181 3.4.6. The differences between the radiation effects of air pathway f

releases and liquid pathway releases are significant. While liquid pathway releases may have less immediately obvious effects, their long-term effects can be serious, and both the liquid pathway dispersal mechanisms and the dominant liquid pathways themselves are more complex than their air-pathway counterparts. Importantly, Interdiction or prevention of liquid pat.%ay releases at the source is of ten possible if adequate design and cor. trol measures are taken; because of the overall failure to' consider liquid pathway releases, l ..

however, and also because this kind of interdiction is usually not possible with air pathway releases, appropriate liquid pathway interdiction design and control measures have been examined inadequately or not at all for the Byron facility.

Table 3.4-1 on page 50 below, drawn from the b80 draft Sandia study,182 summarizes some of these significant air / liquid pathway differences.

3.4.7. Af ter considering the air / liquid pathway differences, the Sandia authors drew the following three conclusions pertinent to the Byron facility:

-"The most probable [ WASH-1400] meltdown categories result in the largest releases to the hydrosphere," i.e., liquid pathway releases.

This is generally not true of air pathway releases.

"Significant amounts of radioactivity are generally expected to be released to the hydrosphere during any meltdown accident."

In the case of plants with the hydrogeologic features of the Byron site Q.e.,

high groundwater and permeable rock), classified as " medium to high-risk" by the Sandia Study, the Sandia authors calculate that the potential radiation dose is approximately 2 to 5 x 107 person-rem (with uncertainties of an order of magnitude), which current studies 183 translate into several thousand probable deaths from a major accident.

-The Sandia calculations indicate that "If interdictive

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measures are not taken, then the liquid pathways can perhaps contribute significantly to the risk of a core meltdown accident."

Thus a nuclear accident at the Byron facility presents significant potential for liquid pathway radioactive contamination and lethal radiation dosages to members of the public. In particular, in any core melt accident (Class 9),'

massive quantities of radioactive materials may leach into the Byron site j groundwater and eventually migrate into the Rock River.

(

TABLE 3.4-1 DIFFERENCES BETWEEN ATMOSPHERIC AND HYDROSPHERIC PATHWAYS (From Draf t Sandia Study, Table 1.1)

ATMOSPHERIC RELEASES HYDROSPHERIC RELEASES SOURCE Atmosphere: Primarily Melt debris: Primarily more volatile radio- less volatile radio-nuclides (I, Cs, nuclides (Ru, Se, Te,...). L a,. .. . ).

Sump water and depressurization: Primarily more volatile radionuclides.

DISPERSAL Population reached rapidly Population usualPJ reached (hours). slowly (months to centuries longer).

All radionuclides move Each radionuelide moves essentially together; through the ground at its deposition mechanisms differ own rate; each radionuclide only for the noble gases. moves through the surface waterbodies with its own set of interactions.

PATHWAYS Primarily inhalation and Primarily ingestion external (ground). (drinking water, aquatic food) and external (shorelines).

Dominant pathways are Some dominant pathways relatively simple. are very complex.

Populations are Populations are not straightforward. obvious.

HEALTH EFFECTS Acute, latent and chronic. Primarily chronic.

INTERDICTION Source: not possible. Source: often possible.

Pathway: possible. Pathway: possible.

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3.4.8. Despite the obvious importance of evaluating Class 9 accidents and liquid pathway accident impacts and corrective measures at Byron, no such evaluations have been undertaken and tnere is no assurance that any such evaluations will occur. The NRC h.as now abandoned its Class 9 accident

, exclusion policy; but Byron may be exempted from the NRC's new interim policy (par. 3.4.2.5 above), since the Byron FES was submitted prior to July 1,1980 and the interim policy is mandatory only for post-July 1st environmental reports.184 Similarly, as part of the TMI Action Plan (par. 2.3.3 above) the NRC instituted a program for plant-specific assessment of accident probabilities, the Integrated Reliability Evaluation Program ("IREP").185 IREP studies will use techniques developed in WASH-1400 (e.g., event-tree and fault-tree analysis and accident radioactive release categorization) on individual plants in order to assess Class 9 accident sequence probabilities and (to a limited extent) consequences.

However, no IREP is scheduled for the Byron nuclear plant.186 And the liquid pathway accident evaluation issue remains--like other priority safety issues-unresolved, with no assurance that any resolution will be adequately implemented at Byron. Models to assess the consequences of possible liquid pathway contamination and appropriate interdiction techniques have not been adequately addressed, either generically in WASH-1400 (par. 3.4.5 above) or specifically for Byron (par. 3.4.3 above). Although the 1980 draft Sandia study emphasized that liquid pathway interdiction measures are possible, few or no preparations have been made at Byron to interdict the flow of contaminated groundwater in the event of a serious accident.

3.4.9. Byron can, should, and must be given the benefit of a thorough evaluation of Class 9 and liquid pathway accidents, impacts, and mitigating measures-both preventive and corrective. As the NRC has stated

(par. 3.4.2.5 above) and as the Kemeny Commission and the NRC's Special Inqu!ry Group found (par. 3.4.2.2 above), Class 9 accidents are precisely those which will recur unless proper steps are taken and which the WASH-1400 analysis fouad " dominate the accident risk." As the 1980 draft Sandia study found (par.

3.4.7 above), it is precisely the most orobable WASH-14Ci meltdown categories ,

which result in the highest liquid pathway radiation releases. And these evaluations should be undertaken for the Byron facility-a potentially risky site, on the Sandia evaluation-before continued construction forecloses, or renders financially or otherwise impracticable, appropriate design modifications and corrective measures. Changes to the Byron site and plant should be made to recognize the facts that Class 9 accidents do happen, that liquid pathway releases are a potentially serious risk at the Byron site, and that emergency response may -be ineffective in highly populated areas like Illinois.187 Safety devices which may prevent or delay the impacts of serious accidents should be implemented (these includ'e, for instance, reactor containment venting systems which reduce explosive pressures while filtering out radioactivity; additional systems _ to flood runaway reactor cores with cooling viater; and " core catchers" to contain a melting core for several_ days). Preventive measures, including systems interaction evaluation and modification (par. 3.1 above) and an IREP study (par. 3.4.3 above), should be implemented. An evaluation of the effects l of liquid pathway interdiction (both close to the source and farther along the pathways to human population exposure) for the Byron site should be conducted, using the models employed in the 1980 draft Sandia study. A design of liquid pathway interdiction systems and the resulting safety improvements should be developed and implemented for Byron. We know that all of these things can be done. To deprive Byron of them is needlessly and unjustifiably to risk the

public health and safety.

3.5. Conformance To Current Regulatory Practices 3.5.1. The NRC has for years placed heavy reliance on conformance with regulations by nuclear power plants as a primary (if not virtually the only) means of assuring safety. . The Kemeny Commission no'ed this " preoccupation with regulations," observing that "[t]he satisfaction of regulatory requirements is equated with safety" by the NRC--though, as the Kemeny Commission emphasized, while compliance with regulations is necessarv, it is by no means sufficient to assure safety.188 The utilities which operate nuclear power plants have similarly-in part, one suspects, due to the " financial disincentives to safety" identified by the NRC's Special Inquiry Groupl89- " regarded bare compliance with NRC minimum regulations as more than adequate for safety,"

to use the Special Inquiry Group's words.190 For example, the utility operating T31I-2 did not see fit to go beyond minimum NRC regulatory requirements in a number of ways, the lack of which " impaired" the " safe operation of the T31I-2 plant."191 3.5.2. Given the heavy reliance on regulatory compliance, it is obviously essential at a bare minimum (though, as we have said, not of itself safety-sufficient) to ensure that existing regulatory requirements are in fact being met. Two sets of important regulatory requirements are the NRC's quality assurance and quality control or "QA/QC" requirements (which the NRC has described as a primary line of defense against safety problems 192) and the NRC's Regulatory ruides, which deal with a large number and variety of design safety matters. Following the T311-2 accident, it was found that the utility operating T311-2 was in violation of applicable NRC requirements-notably including the QA/QC requirements-in multiple respects (though as frequently happens its Q A/QC program was adequate o_n paper to meet the l

requirements),I93 and both the Kemeny Commis::fon194 and the NRC's Special Inquiry Groupl95 expressed serious concern over the general level of compliance with r,1ulatory requirements throughout the nuclear industry. Both the Kemeny Commission 196 and the NRC's Special Inquiry Groupl97 also found that the NRC's policing of regulatory compliance-carried out by the .NRC's Office of Inspection and Enforcement ("I & E")-was sericusly inadequate and ineffective.

The Kemeny Commission noted that in 1978 two separate reports had found that I & E inspectors "did little independent testing of construction work, relied heavily on the utility's self-evaluation," and " felt their procedures were unclear and lacking in sufficient technical guidance."198 The NRC's Special Inquiry Group urged that I & E be given "new technical resources" and "substantially more manpower," and indicated grave concern over the adequacy of I & E's own post-T311 self-evaluation.199 3.5.3. One cannot safely assume, in light of these points, that I & E will succeed in policing regulatory compliance; as the NRC's Special Inquiry Group observed,200 even the post-T3tI placing of a resident I & E inspector at each nuclear plant site is of dubious adequacy given "the inability of a single inspector to assess all of the different systems in a large nuclear plant, and the danger that he may become ' captive' to the utility staff." Nor can one assume that even those regulatory noncompliances which are reported by I & E personnel will be adequately acted upon, given the NRC administrative shortcomings and institutional disincentives to safety identified by the NRC's Special Inquiry Group 201 and the history of inadequate I & E enforcement 1

(including " difficult [y] [in] having safety issues that (I & E inspectors] have l

raised seriously considered within the office") identified by the Kemeny l Com mission.202 One must to a large degree rely, for any nuclear plant, on the

a l

1 utility's ability and willingness to comply with regulatory safety requirements l

I without NRC prodding; cnd as it should be, an affirmative finding in this regard is a prerequisite to the issuance of an operating IIcense.203 Unfortunately, the Byron nuclear facility ex'drits at least two serious ' deficiencies in this regard.

3.5.4. First, following the reports of the Remeny Commission and the' NRC's Special Inquiry Group, in June 1980 Congress added to the NRC 4

authorization bill a requirement (the Bingham Amendment) that the NRC conduct a safety reassessment of all operating nuclear plants.204 This i

assessment will consist of reviewing the plant design against those current NRC Regulatory Guides and standards applicable to safety, from the standpoint of both compliance with previously existing requirements and "backfitting" (which the NRC has historically been slow to req uir e 2 0 5) to meet newer require m ents.206 - Because the legislation is in terms directed to " operating" nuclear plants, there is no assurance that the Byron plant will have the benefit of such an assessment. However, such an assessment should be require'd for Byron. The status of Byron's compliance with current Regulatory Guides is dubious in important respects.

3. 5. 4.1. As part of the Pyren Final Safety Analysis Reoort

'"FSAR"), Commonwealth Edison presented, as Appendix A to the FSAR, a report on the status of Byron's degree of compliance with the NRC Regulatory Guides. There are at present 141 Division 1 Regulatory Guides "in place," five (Nos.1.42,1.51,1.66, L104, and 1.119) having been withdrawn. Of these Division 1 i

Regulatory Guides,128 are pertinent to the Byron nuclear facility (the balance involve reactors of a different type, e.g., No. L5, or duties to be undertaken by the NRC Staff rather than the license applicant, e.g. No.1.36). As amended to date, Edison's status report in Appendix A to the Byron FSAR covers 125 of these pertinent Division 1 Regulatory Guides, Nos. 1.144,1.145, and 1.146 not being

mentioned. A summary of Edison's Appendix A status report appears in Table LS-1 on pages 57-60 below. Table 3.5-1 shows that, based on Edison's own assessment of its degree of compliance with the Division 1 Regulatory Guides at the Byron facility:

~ Byron does not comply or complies only in part (or, as to three Guides, the degree of its compliance is unknown) with 27, or over 20%, of the pertinent Regulatory Guides.

-As to 29 additional pertinent Regulatory Guides, or 23%, the most Edison is able (or willing) to provide is a

" commitment" that Byron will comply at some unspecified point with the "!ntent" of the Guide.

-As to 42 additional pertinent Regulatory Guides, or 33%, Edison asserts that Byron complies with the " intent" of the Guide, in three instances hedging even this statement with " clarifications" or a statement that the status of compliance is "under review."

-Byron is in actual compliance with only 30 (or 23%) -

of the pertinent Regulatory Guides. Even here, the precise extent of compliance is doubtful; for ~ of these 30 RegulaMau Guides, Edison's assertion of compliance is sutyect to " qualifications" or " clarifications" or statements that Byron " generally" complies with the particular Guide.

Of the remaining 23 " actual compliance" Guides, one (No.

1.70) relates to the format of the FSAR and ten concern Q A/QC issues-as to which, while Edison's programs may conform to the Guides on paper, its performance in practice has not been. encouraging. See par. 3.5.5 below.

3.5.4.2. As a result, it is difficult-even though Byron is now more than half complete 207-to gauge accurately the extent of Byron's compliance with over 55% of the current Division 1 Regulatory Guides. Byron does not comply, wholly or partially, with over 20% of those Guides; and for a

further.23%, we have only a promise to comply at some point with their

" int en t." This is inadequate. Though not binding, many of the Regulatory Guides deal with serious safety issues. For example, Byron at least partially fails to comply' with each of five Guides (Nos. 1.25,1.52,1.67,1.77, and 1.97) l -

i l

TABLE 3.5-1 STATUS OF BYRON COMPLIANCE WITH DIVISION 1 REGULATORY GUIDES (From Byron FSAR, Appendix A)

Guide No: Status:

1.1 " Commitment" to comply with " intent" L2 Compues L3 Not pertinent 14 Compues 1.5 Not pertinent L6 " Commitment" to comply with " intent" 1.7 " Commitment" to comply with " intent" L8 " Commitment" to comply with " Intent" L9 " Commitment" to comply with " Intent" 1.10 Complies with " intent" 1.11 " Commitment" to comply with " intent" 1.12 " Commitment" to comply with " Intent" 1.13 " Commitment" to comply with " intent" 1.14 Partial compliance with " intent" L15 " Commitment" to comply with " intent" 1.16 " Commitment" to comply with " intent" 1.17 - " Commitment" to comply with " intent" 1.18 " Commitment" to comply with " intent" 1.19 " Commitment" to comply with " intent" L20 May comply with " intent," will " justify" deviation L21 Complies with " intent" (with " clarification")

L22 " Commitment" to comply with " intent" 1.23 " Commitment" to comply with " intent" L24 Compues 1.25 Partially complies l L26 " Commitment" to comply with " intent" L27 " Commitment" to comply with " intent" L28 Complies 1.29 " Commitment" to comply with " intent" L30 Complies 1.31 Partially aamplies, " essentially" L32 " Commit wnt" to comply with " intent"

L33 Complies L34 Partially complies 1.35 " Commitment" to comply with " intent" l

TABLE 3.5-1 (Continued)

Guide No: Status:

1.36 Complies L37 Compues L38 Complies L39 Complies L40 " Commitment" to comply with " intent" L41 " Commitment" to comply with " intent" 1.42 Withdrawn L43 Complies with intent" L44 Partially complies with " intent" (with

" clarification")

L45 Complies with " intent" (with " clarification")

L46 Partially complies with " intent" 1.47 Complies with " intent" 1.48 Does not comply L49 Complies 1.50 Partially complies with " intent" (with

" clarification")

L51 Withdrawn 1.52 Partially complies 1.53 Complies 1.54 Complies with " intent" L55 Complies with " intent" L56 Not pertinent 1.57 Complies (with " clarification")

1.58 Complies L59 " Commitment" to comply with " Intent" 1.60 Complies (with " clarification")

1.61 Complies with " intent" L62 " Commitment" to comply with " intent" 1.63 Complies 1.64 Complies 1.65 Partially complies 1.66 Withdrawn 1.67 Partially complies 1.68.1 Not pertinent 1.68.2 Complies with " intent" L69 Complies with " intent" L70 Compues

l TABLE 3.5-1  !

(Continued)

Guide No: Status:

1.71 Does not comply 1.72 Not pertinent 1.73 " Commitment" to comply with " intent" 1.74 Compues L75 " Commitment" to comply with " intent" L76 Partially complies L77 Partially complies L78 Complies with " intent" 1.79 " Commitment" to comply with " intent" 1.30 Not pertinent 1.31 " Commitment" to comply with " intent" 1.82 Complies (with " clarification")

L83 Complies 1.84 Complies with " intent" 1.85 Complies with " intent" L86 Not pertinent 1.87 Not pertinent 1.88 Complies L89 Complies with " intent" 1.90 Not pertinent 1.91 Complies with " intent" L92 Complies with " intent" L93 Complies with " intent" 1.94 Complies 1.95 Complies with " intent" 1.96 Not pertinent L97 Partially complies L98 Not pertinent 1.99 Complies with " Intent" 1.100 Complies with " intent" ("under review" by NRC)

L101 Complies with " Inter.'."

1.10 2 Compues with " intent" 1.103 Complies with " intent" 1.10 4 Withdrawn 1.10 5 Partially complies

l l

TABLE 3.5-1 (Continued)

Guide No: Status:

1.10 6 Complies with " intent" L107 Not pertinent L108 Complies with " Intent" 1.10 9 Not pertinent 1.11 0 Not pertinent 1.111 Complies with " intent" 1.11 2 Complies with " intent" 1.11 3 Complies with " intent" 1.11 4 Complies with " intent" 1.11 5 Complies with " Intent" 1.11 6 Complies 1.11 7 Complies with " intent" Lil8 Complies with " intent"

1. 11 9 Withdrawn 1.12 0 Complies with " intent" 1.121 Partially complies L122 Complies with " intent" L123 Complies 1.12 4 Complies (with " qualification")

1.12 5 Complies with " intent" L126 Complies (with " qualification")

1.127 Complies with " Intent" L128 Partially complies 1.12 9 " Generally" complies L130 Complies (with " qualification")

1.131 Partially complies L132 Does not comply 1.133 Partially complies L134 Complies with " intent" 1.135 Complies with ' intent" L136 Complies with " intent" 1.137 Complies with " intent" 1.138 Complies with " intent" 1.13 9 Complies with " intent" Ll40 Partially complies with " intent" 1.141 Partially complies 1.142 Partially complies with " intent" 1.143 Complies 1.144 Unknown (not mentioned) 1.14 5 Unknown (not mentioned) 1.14 6 Unknown (not mentioned)

dealing with different aspects of nuclear accident prevention and mitigation.

Similarly, its spproach to two others, Nos. 8.8 and 8.10 (Division 8 Guides dealing with occupational exposure to radiation) is that Edison " believes" it complies with the former, and is willing to comply with the latter only "to the degree considered reasonable." (Edison similarly expresses disagreement or reservations, or has " qualifications" or interpretations of its own, as to some 27 of the pertinent Division 1 Regulatory Guides.)

3.5.5. A second majoe area of doubt concerning Ryron's compliance with current regulatory requirements relates to its Quality Assurance / Quality Control ("QA/QC") programs. These programs are "self-policing" mechanisms, intended to assure that the design, construction, and operation of a nuclear facility comply with applicable requirements in practice as well as on paper.

But as the Kemeny Commission found with regard to the utility operating TMI-2, the dichotomy between "in oractice" and 'on paper" which QA/QC programs are intended to bridge can apply to the QA/QC programs themselves:

However reassuring they may be on paper, they are not worth much unless they are rigorously carried out in practice.208 The record concerning Byron is not reassuring in this regard.

3. 5. 5.1. Periodically (e.g.,11 times in 1978, 17 times In' 1979, and at least 14 times is 1980), representatives of the NRC's Office of Inspection and Enforcement ("I & E") Inspect various aspects of the ongoing work at the Byron nuclear facility. These inspections vary in scope and intensity, and may be limited to only a few areas. Each inspection results in an I & E report, one purpose of which is to identify problem areas (" unresolved" matters) and violations (" infractions," " deviations," or " deficiencies") observed by the I & E inspector. A summary of the I i E inspection reports on the Byron

facility for 1978 and 1919 is provided in Table 3.5-2 on page 63 below. Table 3.5-2 shows that out of 28 I & E inspection reports in 1973 and 1979, 21--or 75%-noted at least one violation or problem area, the great majority of which concerned QA/QC matters. During this two-year period, a total of 30 violations and 33 other problem areas were formally noted in the I & E reports, as well as other doubtful matters which were resolved during inspections. Many of these matters remained outstanding for lengthy periods after they were ffrst noted; for example, in 1978 there still remained outstanding iteme arising from a 1974 QA audit of Sargent & Lundy, architect-engineer for the Byron facility.209 3!any items were recurrent; for example, in May 1979 the I & E inspectors noted as a recurrent item a " failure to provide adequate storage, cleaning, and preservation" for important equipment (e.g., the reactor vessel and coolant piping),210 which has caused damage to equipment,211 and also complained of the

" number of outstanding unresolved and noncompliance items in the electrical area."212 Inadequate welding has been a continuing problem.213 The qualifications of QA/QC personnel and the sufficiency of QA/QC inspection programs at the Byron site have repeatedly been criticized as inadequate by I &

E inspectors.214 3.5.5.2. In May 19791 & E representatives held a special meeting with Commonwealth Edison to indicate their belief that " improvement was warranted" in several areas relating to Byron, including the " direction and overview" provided by electrical QA/QC personnel, the training of QA personnel,

" housekeeping as it concerns proteeMon of equipment and attitude of workers,"

and " supplemental (QA] inspection...where intial inspection was not adequate.n215 But it does not appear that much was achieved. During the remainder of 1979, l

1 j

-1

~I

TABLE 3.5-2

SUMMARY

OF 1978 AND 1979 I & E REPORTS PERTAINING TO BYRONI Report No.: Pindin_gs 50-454 & 455/78-01 None 50-45M455/78-02 2 unresolved matters, one concerning QA inspector qualifications 50-454&455/78-03 2 unresolved matters 50-454&455/78-04 3 unresolved matters, one concerning QA inspector qualifiestions 50-454&455/78-05 1 QA violation 50-454&455/78-06 2 unresolved matters 50-454&455/78-07 1 QA violation, 5 unresolved matters (3 concerning QA problems) 50-454&455/78-08 1 unresolved matter 50-454&455/78-09 1 QA violation, 2 deviations, 3 unresolved matters (1 concerning QA problems) 50-454& 455/78-10 None 50-454&455/78-H 1 "open" matter concerning lack of response to I & E safety Bulletin issued ten months previously 50-454& 455/79-01 1 QA violation, 2 unresolved matters (1 concerning QA problems) 50-454&455/79-02 2 QA violations, I unresolved QA problem 50-454&455/79-03 2 unresolved matters, one unlisted instance of poor equipment stor:.ge practices 50-454&455/79-04 None -

50-454&455/79-05 1 unresolved QA problem, one unlisted instance of lack of QA inspector training 50-454&455/79-06 1 QA violation, I unresolved matter 50-454&455/79-07 1 unresolved matter 50-454&455/79-08 6 QA violations, 2 unresolved matters 50-454&455/79-09 Report of a special meeting 50-454 & 455/79-10 None 50-454 & 455/79-11 1 unresolved QA problem 50-454&455/79-12 2 QA violations, I unresolved QA problem 50-454 & 455/79-13 None 50-454 & 455/79-14 8 QA violations, I unresolved matter concerning a safety design change 50-454&455/79-15 1 QA violation, I unresolved matter 50-454 & 455/79-16 1 QA violation, I unresolved matter 50-454 & 455/79-17 None l 50-454 &455/79-18 3 QA violations

1. Unresolved matters listed include only new ones identified in the referenced report, not previously identified problems which the report describes as still outstanding.

i

- . --- .- ~ _- .. - - _. - -

c 9

i 15 violations were noted in I & E reports, including violations in the same areas i

raised at the May meeting. While we have not been able to review all of the I

& E reports -for 1980 to date, those we have seen216 note 3 violations and 7 I

problem areas, again concerning (among others) the same subjects covered at the May 1979 meeting. In July 1980, I & E inspectors noted "what appears to be excessive rework" at Byron and expressed concern over "its impact on Construction and Quality Systems management.n217 Thereafter, I & E called upon Commonwealth Edison-not for the first time 2 18-to " perform an in-depth examination and evaluation of [its] design / engineering organizations and

l. function."219 4

3.5.6. Byron's significant QA/QC problems and the dubious status of I

its compliance with NRC Regulatory Guides are both matters of high safety significance. Both are also directly relevant to the ongoing Byron constructicn process, during which lax QA/QC enforcement may lead to poor work and to a failure to identify and correct serious deficiencies, and which as it proceeds tends increasingly to foreclose or render unfeasible a subsequent compliance with Regulatory Guides impacting on safety design and construction. These outstanding deficiencies should be thoroughly explored and resolved promptly, and before Byron construction proceeds further.

3.6. Open Generic Issues 3.6.1. In paragraphs 2.2 and 2.3 above, we pointed out that the issues we have discussed in paragraphs 3.1-3.5 are not the only unresolved safety issues applicable to the Byron nuclear plant. Rather, they are examples of a large number of such issues, all of which the NRC has identified (sometimes repeatedly) and a sizable portion of which the NRC has repeatedly termed high-priority matters. Many of these issues have been unresolved for years.

l The prospects ~ for their resolution are not encouraging. The NRC's Special

Inquiry Group found that owing to " institutional disincentives to safety," even

" issues that are identified at some point in the system often fall through the cracks"220-and in an extremely disturbing finding, also pointed out that just as financial considerations deter nuclear plant designers and licensees from pushing for the resolution of safety issues,221 so als.o the NRC Staff may not aggressively pursue them. The NRC's Special Inquiry Group found:222 "It appears well understood by the Staff that assertien of safety concerns, particularly those that may be controversial, is most unlikely to advance one's career and i.s far more likely to result in stigmatization. In short, at the NRC

' whistle blowing' and ' rocking the boat' are likely to lead to ' career paralysis."'

3.6.2. In a November 1977 decision, Gulf States Utilities Co. (River Bend Station, Units 1 & 2), ALAB-444, 6 N.R.C. 760, 774-75 (1977), the NRC's Atomic Safety and Licensing Appeal Board imposed an affirmative duty on the 4

NRC-as part of licensing proceedings-to identify, and to evaluate the impact t 'on plant safety of, unresolved safety issues. At least three of the " tasks" contained in the NRC's TMI Action Plan 223 (Nos. IV.E.2, IV.E.3, and IV.E.4) are addressed to the need to resolve outstanding safety issues and apply the results to nuclear plants. Yet notwithstanding these recognitions of the problems posed by outstanding safety issues, the NRC schedule for resolving most of those issues continues to erode;224 the NRC has not yet developed a firm plan for resolving two issues (NUREG-0510, Nos. A-43 and A-44) which were amcng the 17 top unresolved safety issues submitted to Congress in 197S and also among the top 20 issues in the Denton/ Steering Committee " point value" rankings. See pars. 2.2.6 and 2.2.7 above. In. whole or in part, antecedents to these issues (which themselves are an " upgrading" of three other issues) were also placed in the " potential high risk" category in the NRC's risk-based evaluation.225

See par. 2.2.5 above. These issues, concerning respectively containment emergency sump reliability and station blackout, are directly relevant to the Byron nuclear plant and its safe operation. Both relate to the ability to cool the reactor core in the event of an accident.

3.6.3. No definitive finding of safety such as the River Bend decision requires, assessing the singular and cumulative impact of unresolved safety issues, has been prepared for the Byron facility-nor can it be, given the points we have discussed in paragraphs 3.1-3.5 above, under present circumstances.

Given the slippage in scheduling which afflicts the resolution of these unresolved issues, there can be no assurance that safety improvements to which they may ultimately lead will be implemented at Byron if construction continues on its present schedule. In particular, the emergency sump reliability and station blackout issues require attention. As to the former, the status of Byron's compliance with even the existing Regulatory Guides on the subject (Nos.1.79 and 1.82) is questionable (see Table 3.5-1 above); as to the latter, no present regulatory' guidance exists.226 In connection with both issues, the necessary design modifications and corrective measures should be implemented at Byron.

IV. CONCLUSIONS 4.1 The issues Are Important 4.1.1. From the foregoing discussion and background information, we have no hesitation in concluding that the safety issues analyzed in this Affidavit are important to nuclear plant safety. Indeed, the NRC itself has--on multiple occasions, as to most of the issues--reached the sam e conclusion.

l

1 i I This is true of each of the issues we have' analyzed. It is doubly true when the cumulative impact of these issues on plant safety is considered, particularly as to those (e.g,~ systems interaction and equipment qualification, pars. 3.1 and 3.3 above) which may tend to aggravate each other with regard to the same risks.

4.2. The Issues Are Unresolved For Byron 4.2.1. It is equally clear from the foregoing discussion and information that each of the important safety issues we have analyzed is unresolved, both generically and with specif2 reference to the Byron nuclear facility. Each of the issues has a direct bearmg on the ultimate safety of the Byron plant. Aporopriate evaluation, design modifications, and corrective measures for each issue should be implemented at the Byron plant, in order to avoid undue and needless c'sks to the public health and safety. Yet even though in several instances action ,Is being taken at other nuclear facilities, there is no assurance that adequate measures will be put into effect at Byron. This is exacerbated by the tendency of the NRC to allow unresolved safety problems to linger "for many years,"227 and by the continuing slippage (despite considerable public pressure) in even the post-TMI resolution schedules.228 4.3. Ongoing Construction Worsens Matters

., 4. 3.1. Despite the importance of the unresolved safety issues we have discussed, both the Kemeny Commission 229 and the NRC's Special Inquiry j Group 230 found that neither utilities nor nuclear plant designers can be relled upon to pursue and address those issues on their own. Both the Kemeny Commission 231. and the NRC's Special Inquiry Group 232 also found that NRC operating license hearings-those in which the Byron facility is now engaged-are in general inadequate to assure proper consideration of safety issues, in part

.,-m m- or

because " economic considerations" Q.e., the " hundreds of millions of dollars" already spent on plant construction) "can lead to a reluctance to order major changes at the OL stage." The Kemeny Commission also found that the NRC has historically been reluctant to require "backfitting" of new safety modifications or corrective measures on nuclear plants which have previously been built.233 4.3.2. All of these problems worsen the longer construction continues.

As we pointed out in paragraph 1.2.3 above, the NRC itself has recognized more than once that the longer construction continues, the more difficult and expensive it becomes to " factor in" appropriate. design safety modifications or corrective measures, and that continuing construction may even foreclose appropriate modifications. This is pointedly true of the Byron nuclear facility, given its present status (more than 50c6 complete) and construction schedule (Unit 1 is expected to go "on line" in just two years, and Unit 2 a year later),

and given the recent testimony by -Commonwealth Edison officials concerning Edison's extremely tight construction budget (par.1.2.4 above).

4.4. Construction Should Be Suspended 4.4.1. Based on the discussion and information we have provided in this Affidavit, and for the reasons we have given, we believe that it is imperative to address and resolve, in the specific context of the Byron nuclear plant, the outstanding safety issues we have analyzed. Those issues have been known for years. What is needed now is action by the NRC, and a timely implementation of solutions by Commonwealth Edison at Byron. We also believe that further construction of the Byron facility should be suspended while those issues are addressed and resolved. Otherwise there is no assurance that solutions will be implemented, and corrective measures taken, at Byron. There l

i

-6 8 -

f-l

is, if anything, contrarv assurance. We emphasize that neither the importance of these outstanding safety issues, nor the fact that allo wing construction to c o n tinu e while they re main unaddressed seriously prejudices the taking of ap p ro priate measures to meet the m, is a new ide a. The NRC itself has recognized both points. It is time to act on that recognition.

Jkga wc, 'h RICHARD B. HUBBARD

.w-:ns:+Ncac.>we:: :-;;

Jt t fLAA4 x*L #4 KAREN L. ENGn3 $# -

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'e O- fr!ncip i Office sam cm county gy commission escues 143. 13. 1984 p

TIREG0jlY

/

. HINOR

t. IFICATION Richard B. Hubbard and Gregory C. Minor, being first duly rworn, on oath state that they have read the foregoing Affidavit and are familiar with the contents thereof and that the same is true and accurate to the best of their knowledge and belief. .

SSs s , ht RICHARD B. HUBBARD Subscribed and sworn to before me this /1 'C day l of November, 1980:

[ g[

MINOR GREGORY /C

, ,r y? .. ,, 4

/-t n d-Notarr Public l

REFERENCES L NUREG-0510, Identification of Unresolved Safety Issues Relating to Nuclear Power Plants, Reoort to Congress (USNRC, Washington, D.C.,

January 1979).

2. NUREG-0660, NRC Action Plan Develooed As A Result Of The TMI-2 Accident, Vols. I, I (USNRC, Washington, D.C., May 1980).

3. See Consumers Power Co. (Midland Plant, Units 1 & 2), ALAB-395, 5 N.R.C. 772, 779 (1977). See also the Commission's recognition, in its June 1980 Policy Statement (45 Fed. Reg. 40101, June 13,1980, p.

40103), that " substantive changes in plant design features...may be more easily incorporated in plants when construction has not yet progrested very far."

4. See Tennessee Valley Authority (Hartsville Nuclear Plant), LBP-77-28, 5 N.R.C.1081,1119-20 (1977).

5. Com monwealth Edison Co. (Byron Plant, Units 1 & 2), LBP-75-74, NRCI-75/12 972 (1375).

6. Ibid. note 5, p. 973.

7. See "World List of Nuclear Plants - June 30, 1980," Nuclear News, ,

August 1980, stating that Byron Unit 1 is 68% complete and Unit 2 is 55% complete.

8. Testimony of Byron Lee, Jr., Executive Vice President, Commonwealth Edison Co., in In the Matter of An Investigation of the Plant Construction Program of the Commonwealth Edison Comoany (Illinois Commerce Commission, Docket No. 78-0640), July 1980, at pp. 2, 9.

9. See Investigation of Charges Relating to Nuclear Reactor Safety, Hearings Before the Joint Committee on Atomic Energy (U.S. Gov't Printing Office, Washington, D.C.,1976), Vol. 2, pp.1200-1245.

10. Testimony of Robert J. Schultz, Vice President, Commonwealth Edison Co., in In The Matter of An Investigation of the Plar'c Construction Program of the Common'vealth Edison Comoany (Illinois Commerce Commission, Docket No. 78-0646), July 1980, at p. 4.

11.

Schultz testimony, suora note 10, p. 3; Lee tes'timony, suora note 3, p.

2.

12. . Schultz testimony, suora note 10, p. 2; see Lee testimony, suora note 8, pp. 5-7.

13. Schultz testimeny, suora note 10, p. 2.

l l

14. Schultz testimony, supra note 10, pp. 2-3, 6,10.

15. Lee testimony, supra note 8, p.12.

16. NUREG-0578, TMI-2 Lessons Learned Task Force Status Report And Short-Term Recommendations (USNRC, Washington, D. C., July 1979).

17. Lee testimony, supra note 8, p.12 and Ex. A. The testimony refers to ten of the recommendations in NUREG-0578 (suora note 16), which the NRC Task Force characterized as "short-term actions" of a

" narrow, specific, and urgent nature." NUREG-0585, TMI-2 Lessons Learned Task Force Final Report (USNRC, Washington, D.C., October 1979), p.1-1. - The Lee testimony does not indicate whether these recommendations will be implemented at the Byron facility, nor does it estimate the cost of doing so.

18. Lee testimony, supra note 8, pp.12-13.

19. Lee testimony, supra note 8, pp.12-13.

20. Final Safety Analysis Report - Byron Station (Commonwealth Edison Co., Chicago, 111.,1978), Vols.1-14, and Amendments thereto.

2L Byron Station, Environmental Report - Ooerating License Stage (Commonwealth Edison, Chicago, 111. 197 8), V ols. I and 2, and Amendments thereto.

! 22. Final Environmental Statement - Byron Station (USNRC, Washington, D. C.,1974).

23. FSAR, suora note 20, Sections L1,1.4.

24. FSAR, supra note 20, Section L3.

25. FSAR, supra note 20, p. L1-6.

23. FSAR, supra note 20, tables L4-1, L4-2.

27. FSAR, supra note 20, Sections 2.1.1.1, 2.1.1.2, 2.5.1; ER/OL, suora note 21, Section 2.1.1.2.

28. FSAR, supra note 20, Sections 2.1.3.1, 2.1.3.5, tables 2.1-1, 2.1-2, 2.1-9.

29. FSAR, supra note 20, Section 2.5.1; ER/OL, suora note 21, Section 5.1.3.2.  !

30. Letter from M. Bender, ACRS Chairman, to J. Hendrie, NRC

' Chairman, Nov. 15, 1977, titled " Status of Generic Items Relating to Light Water Reactors: Report No. 6." (An updated report was issued i

March 21,1979; further updating has routinely occurred.) l l

- 3L Bender letter, supra note 30, p. 2.

32. See Joint Committee Hearings, supra note 9, p.1203.

33 . Gulf States Utilities Co. (River Bend Station, Units 1 & 2), ALAB-444, 6 N.R.C. 760, 767-68 (1977).

34. NUREG-0410, NRC Program For The Resolution Of Generic Issues Related To Nuclear Power Plants (USNRC, Washington, D. C., January 1978), p. 7. See also N UREG-0138 and NUREG-0153.

35. NUREG-0410, _suira note 34, p. 6.

J

36. NUREG-0410, supra note 34; see NUREG-0510, supra note 1, pp. 6-3.

NUREG-0371, Approved Task Action Plans for Category A Generic Activities, Vol.1, Rev.1, appears as Appendix F to N UREG-0410; NUREG-0371 was itself published in November 1978.

37. NUREG-0510, supra note 1, Appendix C.

38. Ibid. note 37. See also Taylor et al., Summary Reoort On A Risk

, Based Categorization Of NRC Technical And Generic Issues (preliminary draf t issued by NRC), p.1. In addition, issues Nos. 3-57 and C-3, which were later transmuted into Nos. A-43 and A-44 (discussed in par. 3.6 of this Affidavit), were placed in the high-risk category. See NUREG-0510, supra note 1, p. A-19 and Appendix C.

39. NUREG-0510, suora note L

40. NUREG-0510, supra note 1, pp. 2-3.

41. See NRC Document SECY 79-76 (January 30, 1979), a memorandum from Harold R. Denton, Director, Office of Nuc ear Reactor Regulation, to the NRC Commissioners.

42. Memorandum from Harold R. Denton to Roger S. Boyd, g al., l January 23,1979 (attached to SECY-79-76, supra note 41), pp.1-2 and Enclosure 1.

43. Denton Memorandum, supra note 42, Enclosure 1.

44. NUREG-0606, Unresolved Safety Issues Summary (USNRC, Washington, D.C., January 1980), Vol. 2, No.1, at p. 2.

45. NUREG-0649, Task Action Plans For Unresolved Safety Issues Related To Nuclear Power Plants (USN RC, Washington, D. C., Feoruary 1980).

The proposed resolution of No. A-24 is discussed in par. 3.3.5 of this Affidavit.

p e 1

I

7 46.

The Need For Change: The Legacy Of TMI, Report of the President's Commission on the Accident at Three Mile Island (U. S. Gov't Printing Office, Washington, D. C., October 1979), p. 51. This is hereaf ter referenced "Kemeny Report." See also SIG Report, infra note 47, pp.

139-140.

47.

NUREG/CR-1250, Three Mlle Island. A Report To the Commissioners And D. C.,To The Public, NRC Special Inquiry Group (USNRC, Washington, .

January 1980), Vol. I, p. 90; see Ibid., pp. 139-40, 163-64. This is hereafter referenced "SIG Report." (Volume II of the SIG Report contains technical material from which the conclusions and recommendations Volume I.) in Volume I are drawn. The references herein are to 48.

See letter from M. S. Plesset, ACRS Chairman, to J. F. Ahearne, NRC Chairman, March 11, 1980, titled "ACRS Report on Near Term Operating License Items from Draft 3 of NUREG-0660" (NRC News Release 80-56), at p. I-23. Comparison of the risks from occupational exposure with the risks from accident-caused exposure is difficult.

However, it will be observed that accident exposure risk estimates are largely hypothetical calculations, while the occupational exposures are actual, documented experience.

49.

Kendall, et al., The Risks Of Nuclear Power Reactors (Union of Concerned Scientists, Cambridge, Mass.1977), Appendix B, pp.131-87.

50.

See Browns Ferry Nuclear Plant Fire, Hearings Before the Jo it Committee on Atomic Energy (U.S. Gov't Printing O.ffice, Washingtw, D. C.,1975), Part 1, p.1, where Senator Montoya described the incident: "As I understand it, the fire began by the use of an almost primitive inspection technique of holding a candle close to a hole which contained flammable material The initial attempt to beat out the flames was made by using a flashlight. When that failed, rags were used to attempt to stifle the flames. And when that effort was unsuccessful, one of the men attempted to actuate the carbon dioxide fire control system, only to find that the power had been cut off and a metal plate had been placed over the CO2 controls so they could not be immediately activated. This became the genesis of the l

disabling of the multi-million-dollar plant."

51.

Kemeny Report, supra note 46, pp. 27-28,110-11,119-20; SIG -Report, sup_ra note 47, pp.14-20.

52.

Kemeny pp. 94-95.Report, supra note 46, pp. 28-29; SIG Report, suora note 47,

! $3.

Kemeny pp. 148-49.Report, supra note 46, pp. 44-52; SIG Report, supra note 47,

]

54. SIG Report, supra note 47, p. 91. In fact, the Special Inquiry Group found that if the block valve had remained open for as little as 30 to 60_ minutes longer, "our projections show that...a substantial amount of the reactor fuel would have begun to melt down." Ibid.

55. SIG Report, supra note 47.

56. Kemeny Report, supra note 46.

57. SIG Report, supra note 47, p.170. Nevertheless the Special Inquiry Group made many of the came criticisms and recommendations. The Kemeny Commission found that "With its present organization, staff, and attitudes, the NRC is unable to fulfill its responsibility for providing an acceptable level of safety for nuclear power plants."

Kemeny Report, supra note 46, p. 56. The Special Inquiry Group (staffed in large part by NRC employees) found the NRC's response to the Kemeny Report inadequate, found the President's proposed modifications to the NRC also inadequate, concluded that "We are not reassured by what we see so far," and found (twice) that "the [ Nuclear Regulatory] Commission is incapable, in its present configuration, of managing a comprehensive national safety program for existing nuclear powerplants and those scheduled to come online in the next few years adequate to ensure the public health and safety." SIG Report, suora note 47, pp. ix, 89,112,117,170,171.

58. NUREG-0660, supra note 2.

59. NUREG-0660, suora note 2, Chapter II, p. II-1.

60. See Kemeny Report, e note 46, pp.11, 29-30; SIG Report, supra note 47, pp.102,123,125-28.

61. Division 1 Regulatory Guide 1.97. In particular, the Byron plant compiles only partially with even the go-T311 version of the Guide.

FS AR, supra note 20, Vol.14, p. A1.97-1. The Guide was revised following the T31I-2 accident and is still in a " comment mode."

62. Nucleonics Week, Stay 29,1980, pp. 2-3.

63. Kemeny Report, supra note 46, p. 52.

64. SIG Report, ' supra note 47, pp.139-40.

65. SIG Report, supra note 47, p.140.

66. SIG Report, supra note 4I, p.140.

67. Kemeny Report, supra note 46, pp.19-21, 53-54.

68. SIG Report, supra note 47, p.170.

69. SIG Report, svora note 47, p. 98. {

m - - , - - - ,v

i

70. SIG Report, suora note 47, p.170.

1

71. SIG Report, suora note 47, pp.163-64.

72. USNRC Statement of Policy, 4.5 Fed. Reg. 41738 (June 20,1980),

dissenting views of Commissioner Bradford.

73. Ibid. note 72, Commissioner Gilinsky's separate views.

74. USNRC Sthtement of Policy, suora note 72. This is true of "new"

(

requirements, but not of requirements which " interpret, refine, or quantify the general language of existing regulations." Anyone may challenge the latter requirements, even NRC Licensing and Appeal Boards. Ibid, i 75. USNRC Statement of Policy, suora note 72, Commissioner Gilinsky's separate views.

76. USNRC Statement of Policy, suora note 72, dissenting views of Commissioner Bradford. As a result of the dissents of Commissioners

, Gilinsky and Bradford, the "no challenge" part of the Policy Statement ,

was adopted by a bare 3-2 majority of the NRC, and Commissioner

^ Bradford "(gave] notice that I intend to seek its reconsideration and revocation.' lbid.

77. USNRC Statement of Policy, suora note 72, dissenting views of Commissioner Bradford. He described the "no challenge" policy as a

" radical act" whose " justifications...smount to no more than a bored yawn toward the concerned puolic." Ibjo.

j 78. SIG Report, suora note 47, p.147.

79. Kemeny Report, suora note 46, pp. 9,19-20, 52.

80. SIG Report, suora note 47, pp.147%d.

81. Ibid. note 80, pp.147-49; Kemeny Report, suora note 46, pp. 9,19-20, 23, 32, 52.

82. SIG Report, suora note. 47, pp.119,148,150.

83. SIG Report, suora note 47, p.119.

84. . . Kemeny Report, suora note 46, p. 52.

85. SIG Report, suora note 47, p.148.

3 6 .- Kemeny Report, suora note 46, pp.19-20. "

l

37. Kemeny Report, suora note 46, p. 63.

38. SIG Report, suora note 47, pp.148-15L l

.-, . .._ _ _ ~ . _ . - . _ . , _ . _. - - _ ~

89. SIG Report, suora note 47, p.119.

90. FSAR, supra note 20, Section 1.4.3.

91. FSAR, supra note 20, Section 1.3.1.

92. SIG Report, supra note 47, p.122.

93. Letter from Dr. S. Hanauer to E. G. Case (NRC), August 18, 1977.

94. ACRS Letter to NRC, October 12, 1979. Sorae of the similarities between Indian Point Units 2 and 3 and Byron are noted in FSAR, '

suora note 20, Table 1.3-2.

95. It should be noted that these "non-safety" systems were not subjected to scrutiny in the prior Byron license proceedings. See Kemeny Report, supra note 46, pp. 44-45, 52-53, and SIG Report, supra note 47, pp.

148-49. See paragraphs 3.1.7 and 3.3.3 of tais Affidavit.

96. SIG Report, suora note 47, p.119.

97. NUREG-0510, suora note 1, Appendix A, p. A-12.

98. NUREG-0572, Review of Licensee Event Reoorts (1976-1978) (USNRC, Washington, D. C., September 1979).

99. NUREG-0572, supra note 98, pp. 3-3, D-9 to D-12, D-19 to D-22.

10 0. Kemeny Report, supra note 46, pp. 54-55.

101. SIG Report, suora note 47, pp. 96-97.

10 2. SIG Report, supra note 47, pp. 97-98,170-71.

10 3. WASH-1400 (NUREG-75/014), Reactor Safg Study - An Assessment Of Accident Risks in U. S. Commercial NuWear Power Plants (USNRC, Washington, D. C., October 1975). The NRC's Special Inquiry Group said that WASH-1400 "shows that the greatest risk of an accident comes not from the design basis accidents, such as the large loss-of-coolant accident, but from small loss-of-coolant accidents and relatively routine transients con ~panded by multiple failures or human error, having a higher probability of occuring than a large pipe break."

SIG Report, supra note 47, p.148. The Kemeny Commission agreed.

Kemeny Report, supra note 46, p. 32. Both groups agreed that TMI-2 was an accident of this type, and the Kemeny Commission added (Ibid., p. 32) that " based on WASH-1400...such an accident should have been expected."

10 4. Kendall, g al_., suora note 49, loc. cit.

I

_7s_

l

4 i

10 5. NUREG/CR-0400, Risk Assessment Review Group Report To The U. S.

Nuclear Regulatory Commission (USNRC, Washington, D. C., September 1978).

10 6. USNRC Statement of Policy, Fed. Reg. (January 18, 1979).

107. SIG Report, supra note 47, p.148.

10 8. N UREG/CR-1321 (SAN D 80-0384), Final Report - Phase I Systems Interaction Methodology Applications Program (USNRC, Washington, D.

C., April 1980).

10 9. N UREG-058 5, TMI-2 Lessons Learned Task Force Final Report l (USNRC, Washington, D. C., October 1979), p. 3-3.

11 0 . NUREG-0660, supra note 2, Vol. II, p. C-9.

11L N UREG-0660, supra note 2, Vol. II, p. C-9. A recent Letter to All Licensees of Ooerating Plants and Apolicants for Ooerating Licenses e

and Holders of Construction Permits (USN RC, Washington, D. C.,

l September 5,1980) states in Enclosure 1 that a system interaction study will be required for the Indian Point Unit 3 plant, and in Enclosure 2 that such a study will also be required for Diablo Canyon,

with other plants to be decided on a " case by case" basis.

11 2 . "R & D Status Report - Nuclear Power Division," EPRI Journal. Vol.

3, No. 3 (Electric Power Research Institute, Palo Alto, Calif., April 1978), pp. 51-56.

i 11 3 . In 1977 the ACRS also identified the problem as a top-priority issue.

Bender Letter, supra note 30, Table L 11 4 . The New York Times reported on April 19, 1977 that the utility 1 operating Turkey Point had reported to its stockholders that the total direct and indirect cost of replacing the degraded steam generator tubes might approach $380 million, and that the utility operating the Surry plant had estimated in a prospectus that replacing the Surry steam generator tubes would approximate $60 million in direct cost

alone, cotpled with a shutdown of the plant for one year.

11 5 . R & D Status Report, supra note 112. In addition to Surry and Turkey Point, older Westinghouse-designed nuclear plants such as San Onofre and Point Beach have experienced serious tube degradation problems.

i 11 6 . See Westinghouse Steam Generator Symcosium (April 1973), p. 5.

11 7 . NRC Memorandum, " Summary of Meeting Held on February 12,1980 to Discuss the January 1980 Unit 2 Generator Tube Inspecion," NRC l Docket 50-306 (March 20,1980), pp. 2-3.

l i -

_ .~, ,- ,, . , _ _ _ . . . - ~ . . - . _ . - .

11 8 . FSAR, suora note 20, p. 5.4-11; NRC Memorandum, suora note 117, p.1.

11 9 . Licensee Event Report, " Prairie Island Unit 1 Steam Generator Tube Break" (Northern States Power Co., Minneapolis, Minn., October 16, 1979).

120. MUREG-0460, Anticioated Transients Without SCRAM For Light-Water Reactors (USNRC, Wastungton, D. C., Aped 1978), Vol.1, p. 8.

121. N UREG-0606, Unresolved Safety Issues Summarv (USNRC, Washington, D. C., January 1980), Vol. 2, No.1, " Program Overview."

122. NRC Memorandum, suora note 117, pp. 2-3.

123. See Byron FSAR, suora note 20, Section 5.4.2 and particularly the discussion in Section 5.4.2.1.3.

124.  ! & R Inspection Report No. 50-454/78-09; 50-455/78-09 (NRC Docket Nos. 50-454, 50-45 5), December 1979, p. 7. It seems unlikely that anyone would subject a splice box to such conditions. However, the I

& E Report states: "The inspector observed work activities relative to the installation of control cable to the essential service water make up pumps in the river screen house. It was noted that the splice box on the lower elevation of the cable run was full of water. The licensee representatives at the exit meeting stated that they expected the boxes and conduit to always be filled with water. The inspector questioned whether the cable and splices were suitable for an extended period of underwater service. The licensee is requesting information from the architect-engineer whether the environmental extremes expected to exist in the cable run were. considered in the design. This item is considered unresolved." Ibid.

125. Ibid. note 124.

126. N UR EG-0690, 1979 Annual Reoort (USNRC, Washington, D. C.), pp.

8 9-9 0, 114-15. See also NRC Commissioner Bradford's September 24, 1980 speech (before ALI-ABA), " Reasonable Assurance, Regulation, and Reality," NRC Press Release No. S-13-80.

127. Denton Memorandum, suora note 42, Enclosure L 128. NUREG-0410, suora note 34, Task A-24; NUREG-0510, suora note 1, Appendix A, p. A-13.

129. Letter, W. Kerr, ACRS Chairman, to W. A. Anders, NRC Chairman, May 13,1975, p. 3. The environmental qualification issue is also raised in tasks I.F.1 and II.F.5 of the TMI Action Plan (NUREG-0660, suora note 2.

130. NUREG-0510, suora note 1. Appendix A, p. A-13; Kemeny Report, suora note 46, p. 52; SIG Report, suora note 47, pp.148-49.

, 1

~

1 1

131. Kemeny Report, supra note 46, p. 52; SIG Report, supra note 47, p.

14 8.

132. Kemeny Report, supra note 46, p. 53; SIG Report, supra note 47, p.

14 9.

133. Kemeny Report, suora note 46, p. 52.

134. Cualifying Class IE Equipment for Nuclear Power Generating Stations, ,

I.iE E-323-19 7 4. This standard is referenced in NRC Regulatory Guide l 1.89.

135. See, e.g., IEEE 323-1971 and IEEE 323-1974 (supra note 134); IEEE 383, i

IEEE 382, IEEE 334, IEEE 317, related to cablas, valves, motors, and i electrical penetrations, respectively.

13 6. An example Is Branch Technical Position 9.5-1. Appendix A is used as i a guide for proper fire protection design and IEEE 383 for flame test and qualification.

137. See, for example, Regulatory Guides 1.89 and 1.100.

13 8. FSAR, supra note 20, Appendix A (in Vol.14).

139. NUREG-0588, Interim Staff Position On Environmental Qualification of Safety Related Electrical Equipment (USN RC, Washington, D. C.,

December 1979) (issued for comment).

14 0. NUREG-0588, supra note 139, p.1.

4 j 141. SIG Report, supra note 47, p. 93. The Special Inquiry Group suggested i

that "the Congressional oversight committees should hold the NRC l

accountable with respect to such issues." Ibid. However, even though Congress took a step in this regard in 1977 by amending the Energy Reorganization Act of 1974 to require the NRC to " develop and submit to the Congress a plan for the specification and analysis of

't'nresolved Safety Issues' relating to nuclear reactors" and report 4

mually on the subject (NUREG-0510, supra note 1, p.1), as noted elsewhere in this Affidavit there has been Itttle real progress.

14 2. 36 Fed. Reg. at 22852 (Dec.1,1971). See also SEC Y-7 8-137, Memorandum to the NRC Commissioners on Assessments of Relative Differences in Class 9 Accident Risks in Evaluations of Alternatives to Sites with Hign Population Densities (March 7,1978).

14 3. FES, supra note 22.

14 4. ER/OL, supra note 21.

14 5. FSAR, supra note 20.

d.

1

t I.

146. SIG Report, supra note 47, pp.147,151. The Special Inquiry Group concluded that "[m] odification is definitely needed in the current philosophy that there are some accidents ( Class Nine accidents') so unlikely that reactor designs need not provide for mitigating their consequences." Ibid.

147. Offshore Power Systems (Floating Nuclear Power Plants), CLI-79-9,' 10 N.R.C. 257, 261 (1979).

14 8.- W AS H-1400, supra note 103. See the NRC's Statement of Interim Policy,, 45 Fed. Reg. 40101 (June 13,1980), at p. 40102.

14 9. Statement of Interim Policy, supra note.148.

1 15 0. NRC Statement on Risk Assessment and the Reactor Safety Study Report (W ASH-1400) in Light of the Risk Assessment Review Group Report (USNRC, Washington, D. C. January.18,1979), p. 3.

4 151. In Re Public Service Electric & Gas Co. (Salem Nuclear Generating Station, Unit 1), NRC Docket No. 50-272, "NRC Staff Response to

' Board ' Question No. 4 Regarding The Occurrence Of A Class 9 Accident At TMI," by B. H. Smith, August 25, 1979.

I 152. Statement of Interim Policy, supra note 148, at p. 40103.

15 3. " Obvious [1y]" the proposed Annex did not rest upon WASH-1400, completed some years later (Statement of Interim Policy, supra note 148, p. 40103); and after a detailed study the President's Council on Environmental Quality concluded that the Annex "[was} not based on the [other) then-existing accident risk assessment studies prepared by i

the AEC." Report of President's Council on Environmental Quality to the Nuclear Regulatorv Commission (CEQ, Washington, D. C., March 20,1979), p. 4.

154. See CEQ Report, supra note 153, pp. 4,19, nn. 63-65, 155.- Public Service Co. of Oklahoma (Black Fox Station, Units 1 & 2), NRC Docket Nos. STN 50-556, STN 50-557, NRC Response to Document Discovery No.12 (March 18,1980).

156. See Statement of Interim Poliev, supra note 148, p. 40102.

15 7. - CEQ Report, supra note 153; see par. 3.4.5 of this Affidavit.

I 158. Statement of Interim Policy, supra note 148, p. 40103.

15 9. NUREG/CR-0400, Risk Assessment Review Group Report To The U.S.

Nuclear Regulatory Commission (USNRC, Washington, D. C. September 1978).

l l

d i

160. NRC Statement of Poliev, supra note 150, pp. vi-x.

16L CEQ Report, supra note 153, pp. 2, 4.

16 2. "NRC Staff Response to Board Question No. 4," supra note 151.

16 3. See SIG Report, supra note 47, pp. 94-95.

164. Kemeny Report, supra note 46, pp.15, 32.

165. SIG Report, supra note 47, pp. 5, 91,151.

166. Letter from M. S. Plesset, ACRS Chairman, to J. F. Ahearne, NRC '

Chairman, titled "ACRS Report on Near Term Operating License Items from Draf t 3 of N UREG-0660," 3f arch 11,1980, NRC News Release 80-56.

16 7. N UREG-06 25, Report Of The Siting Policy Task Force (USNRC, 3 Washington, D. C., August 1979), p. 42.

168. NUREG-0642, A Review Of NRC Regulatorv Processes And Functions (USNRC, Washington, D. C., January 1980), p. 8-3.

169. Warm Sorings Dam Task Force v. Gribble, 417 U.S.1301, mo. to vacate stay denied, 418 U.S. 910 (1974); Andrus v. Sierra Club, 442 U.S. 347 (1979). See 42 U.S.C. Section 4344.

17 0. CEQ Report, supra note 153, pp. 2, 3, 4-5.

17L Statement of Interim Policy, suora note 148, p. 40102.

17 3. Statement of Interim Policy, supra note 148, p. 40103.

17 3. FES, supra note 22.

17 4. ER/OL, supra note 21.

17 5. FSAR, supra note 20.

17 6. See FSAR, supra note 20, Section 2.4.

17 7 . WASH-1400, supra note 103, 3tain Report, pp. 76,134.

17 3. Kendall, g al,., supra note 49, pp. 95-96.

.i 179. See Risk Assessmment Review Group (" Lewis Committee"), afinutes of Ateeting Five, p. 57.

18 0. Letter, Secretary of the Interior to NRC Advisory Committee on Reactor Safeguards, February 9,1977.

l

,- , . - + . . - - m - _ ,,-

w- g - - -,,-- . e---

181. Sandia Study (Draf t) for USNRC, "Effect Of Liquid Pathways On Consequences Of Core Melt Accidents," January 1980. This study was provided to participants in the Black Fox construction permit proceedings (NRC Docket Nos. STN 50-556, STN 50-557).

18 2. Sandia Study, suora note 131, Table 1.1.

18 3. Sandia Study, suora note 181, p. 6-16; R. Hubbard, telephone discussions with Susan J. Niemezyk in October 1980.

18 4. Statement of interim Policy, suora note 148, p. 40103. For other environmental reports, the NRC Staff may-or may not-recommend

" factoring in" Class 9 accident analyses on a case-by-case basis. Ibid.

A letter from M. Karman (NRC Staff Counsel) to the Byron Licensing Board dated October 8,1980 indicates that the Staff plans to issue further Byron environmental scatements in 1982. Even if these consider Class 9 accidents, however, there is no assurance that liquid pathways impacts and corrective measures will be considered, and a 1982 assessment is unlikely to provide adequate time to factor in changes to the Byron design. The NRC said on this precise point in its Statement of Interim Policy (Ibj.) that " substantive changes in plant design features...may be more easily incorporated in plants when construction has not yet progressed very far."

18 5. NUREG-0660, suora note 2, pr. II-3, II.C-1 to II.C-5.

18 6. N UREG-0660, suora note 2, pp. 21, II.C-3 to II.C-9.

187. For example, the city of Rockford is less than 20 miles from the Byron plant site (par. 2.1.3 of this Affidavit), and over 20,000 persons live within 10 miles of the site (Rid.). Both the Kemeny Commission (Kemeny Report, suora note 46, pp.15-17, 33-40, 76-77) and the NRC's Special Inquiry Group (SIG Report, suora note 47, pp.129-30,133) severely criticized the NRC's " low population zone" concept and the chaotic, disorganized response to the TMI-2 accident which resulted from a lack of adequate emergency planning, and found that even a 10-mile radius for evaluaton planning was too small. The Byron FSAR (suora note 20) does not indicate that any consideration has been given to planning for a 20-mile evacuation zone (which is necessary, given

" ripple effects," even if only a 10-mile zone were to be evacuated:

see Kemeny Report, suora note 46, pp. 39-40); in the case of Byron, this would involve nearly 300,000 persons. See par. 2.1.3 of this Affidavit.

188. Kemeny Report, suora note 46, p. 9.

18 9. SIG Report, suora note 47, pp.161,163-64, 19 0. SIG Report, suora note 47, p. 90.

-8::-

19L Kemeny Report, supra note 46, p. 45.

19 2. See, e.g., AEC Doc. No. WASH-1240 (1973), pp. 2-Iff., 3-19; Consumers Power Co. (311dland Plant, Units 1 & 2), ALAB-106, RAI-73-3 182, 183-84 (1973).

19 3. Kemeny Report, supra note 46, pp. 44-47.

19 4. Kemeny Report, supra note 46, pp. 22-23. ,

19 5. SIG Report, supra note 47, pp. 89-90,103,109-11.

196. Kemeny Report, supra note 46, pp. 54-55.

197. SIG Report, supra note 47, pp. 94-95, 97-98,99-100,170-71.

19 8. Kemeny Report, supra note 46, p. 54.

19 9. SIG Report, supra note 47, pp.99-100,170-71.

200. SIG Report, supra note 47, p.100.

20L SIG Report, suora note 17, pp.161,163-64.

202. Kemeny Report, supra ..ote 46, pp. 54-55.

203. See 10 C.F.R. Sections 50.57(a)(2), (a)(3), (a)(4).

204. NRC 1980 Authorization Bill, Section 110 (entitled " Systematic Safety Evaluation Plan"); Congressional Record - House, June 4,1980, pp.

H4472 - H4482.

205. Kemeny Report, _ supra note 46, pp. 53-54.

206. Ibid. note 204.

207. "World List of Nuclear Plants," supra note 7.

208. Kemeny Report, _suora note 46, pp. 44-48. The NRC's Appeal Board has also emphasized this point. Consumers Power Co. (311dland Plant, Units 1 & 2), ALAB-106, RAI-73-3182,184 (1973).

209. 1 & E Report No. 50-454/78-04; 50-455/78-04, July 7,1978, p. 3.

210. Letter from G. Florelli to B. Lee, Staf 30, 1979, Appendix ^ (Notice of Violation), p. 2.

2 11. See, e.g., I & E Report No. 50-454/S0-05; 50-455/80-05, April 25,1980,

p. 5.

212. I & E Report No. 50-454/79-08; 50-455/79-08, May 30,1979, pp. 3-5.

213. See, e.g., I & E Reports Nos. 50-454&455/78-07, /78-09, /79-01, /79-02,

/79-12, /79-16, /80-12.

214. See, e.g., ! & E Reports Nos. 50-454&455/78-02, /78-04, /78-05,

/78-07, /78-09, /79-01, /79-02, /79-05, /79-08, /79-12, /79-14, /80-05.

215.  ! & E Report No. 50-454/79-09; 50-455/79-09, June 12,1979, pp. 3-4.

216. We have reviewed Nos. 50-454/80-01, /80-05, /80-08, /80-09, /80-10,

/80-12, /80-13, and <30-14.

217. I & E Report No. 50-454/80-12; 50-455/80-11, July 10,1980, p. 9.

218. See, e.g., ! & E Report No. 50-454/79-09, June 12,1979; I & E Report No. 50-454/77-01, March 14,1977. A May 14,1977 letter from R. F.

Heishman to B. Lee emphasized that "[t] he thrust of [I & E's]

concern" was that deviations from requirements of your procedures i

and specifications repeatedly occurred without corrective steps being taken."

219. I & E Report No. 50-454/B0-13; 50-455/80-12, August 15,1980, p. 3.

220. SIG Report, supra note 47, p. 90.

221. SIG Report, supra note 47, pp.161,163-64.

222. SIG Report, suora note 47, p.163. This is particularly distressing in

' light of the Special Inquiry Group's finding that as a forum for safety Issues, license hearings are "a sham"-because the Staff (the same group which avoids " whistle blowing") takes the position at such hearings that any significant safety issues have been resolved. Ibid.,

p. 13 9. It would thus appear that safety issues of the kind discussed in this Affidavit will be considered (if at all) only to the extent that Intervenors insist upon them, as is now being done in the context of the Byron plant. Hence the Special Inquiry Group's proposal for Intervenor funding by the NRC (Ibid., pp.143-44).

223. NUREG-0660, supra note 2.

224. See N UREG-0606, Unresolved Safety Issues Summary (USNRC, Washington, D. C., January 11,1980), pp.1-2.

225. NUREG-0510, supra note 1, p.17; Ibid., Appendix C, p. C-L 226. NUREG-0690, supra note 126, pp. 85-86.

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227. Kemeny Report, supra note 46, p. 51. The NRC's Special Inquiry Group pointed out that while generic safety issues are " theoretically" dealt with in various ways other than in Individual license proceedings, "In practice, it appears that many of these issues do not get meaningful attention anywhere." SIG Report, supra note 47, pp.139-40.

228. NUREG-0606, supra note 224.

229. Kemeny Report, supra note 46, pp. 44-48, 54-55.

230. SIG Report, supra note 47, pp. 90,161,163-64.

23L Kemeny Report, supra note 46, pp. 52, 53-54.

232. S!G Report, supra note 47, pp.139-40.

233. Kemeny Report, supra note 46, pp. 53-54. See also Statement of Interim Policy, supra note 148, at p. 40103 (quoted supra in note 134).

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PROFESSION AL OUALIFICATIONS OF RICHARD 3. HU33ARD RICHARD 3. H UB B A RD MH3 Technical Associates 1723 Hamilton Avenue Suite K San Jose, California 95125 (403) 266-2716 EX?ERIENCE:

9/76 -

PRESENT Vice-President - MH3 Technical Associates, San Jose, California.

Founder, and Vice-President of technical consulting firs. Special-is ts in independent energy assessments for government agencies, particularly technical and economic evaluation of nuclear power facilities. Consultant in th is capaci:y to Oklaho=a and Illinois Actorney Generals, Minneso:a Pollution Control Agency, German Minis t ry f or Research and Technology, Governor of Colorado, Swedish Energy Cossission, Swedish Nuclear Inspec: orate, and the U .S .

Depar:sen: of Energy. Also provided studies and testimony for various public interes: groups including the Cen:er for Law in the Public Interest, Los Angeles; ?ublic Law Utility Group, 3a:on Rouge, Louisiana; Friends of the Earth (F0E), Italy; and the Union of Concerned Scientists, Camb rid g e , Mas s a chus e t ts .

Provided testimony :o the U.S. Senate / House Join: Cossi::ee on A:osic Energy, the U.S. House Cossittee on In te rior and Insular Affairs, the Calif o rnia Assembly, Land Use, and Energy Cossittee, the Advis o ry Cossi: ee on Reactor S afe guards , and :he A:o=1c S af e ty and Licensing Board. Performed comprehensive risk analysis of the accident p rob ab ilit ie s and consequences at .the 3 ars eback Nuclear Plant for the Swedish 2nergy Cossission and edited, as well as contributed to, the Union of Concerned S cientis t's technical review of the NRC's Reacto r S af ety 5 :udy (WASH-14CO).

2/76 - 9/76 Consultant, Project Survival, Palo Alto. California.

Volunteer wo rk on Nuclear Saf eguards Ini:iative caspaigns in Cali-fornia, Ore 6on, Washington, Arizona, and Colorado. Numerous presentations on nuclear power and alterna:ive energy options to c iv ic , government, and college groups. Also resource person for public service presentations on radio and television.

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5/75 -

1/76 Manager - Qualiev Assurance Section Nuclear Energy Control and I ns trumen ta tion Depart =ent, General Electric Company. San Jose, California.

Report to the Department General Manager. Develop and imple=ent quality plans, prograss, methods, and equipment which assure that produe:s produced by the Depar: cent meet quality requirements as defined in NRC regulation 10 CFR 50, Appendix 3, ASME 3 oiler and Pressure %ssel Code , cus tomer contracts, and CE Corporate policies and procedures. Product areas include radiation sensors, reactor.

vessel internals, fuel handling and servicing tools, nuclear plant control and protecttan instrumentation systems, and nuclear s teas supply and Balance sf Plant control room panels. Res pon s ib le for approximately 45 exempt personnel, 22 non-exempt personnel, and 129 hourly personnel with an expense budget of nearly 4 million dollars and equip =ent investment budget of approximately 1.2 million dollars.

11/71 -

5/75 Manager - Quality Assurance Subsection. Manufacturing Section of Atonic Power Ecutonen: Decartment, General Electri: Comoanv, San Jose, Calif o rnia .

Report to the Manager of Manufacturing. Sa=e functional and product responsibilities as in Engagement 11, except at a lower o r ga ni: 2 t io na l report level. Developed a quality system which received NRC certification in 1975. The systes was a ls o success-fully surveyed for AS ME "N" and "N?!" symbol authorization in 1972 and 1975, plus AS ME "U" and "S" sy mb o l authorizations in 1975.

Responsible for from 23 to 39 exempt personnel, 7 to 14 non-exempt personnel, and 53 to 97 hourly personnel.

3/70 - 11/71 Manager - Application Engineering Subsection. Nuclear Instrumen-tation Decartment. General Electric Company, San Jose. California.

Responsible for the post order technical in ter f ace with architect engineers and power plant owners to define and schedule the instru-sentation and cont rol sys tems for the Nuclear S teas Supply and Balance of Plant portion of nuclear power generating stations.

Responsibilities included preparation of the plant instrument list with ap p ro xima t e location, review of interface drawings to define f unctional design requirements, and release of func:1onal require-ments for detailed equipment designs. Personnel supervised included 17 engineers and 5 non-exempt personnel.

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12/69.- 3/70- )

i Chairman - Eouipment Room Task Force, Nuclear Instrumentation Department, General Electric Comoany, San Jose. California.

Re s p ons ib le for a special task force reporting to the Department General Manager to define methods to improve the quality and reduce the installation time and cost of nuclear power plant control rooms. Study resulted in the conception of a f actory-f abr1cated control room consis ting of signal conditioning ~and operator control panels mounted on modular floor sections which are 4ompletely asse= bled'in the factory and thoroughly tested for proper operation of 1steracting devices. P e rs onnel supe rvis ed included 10 exempt personnel.

12/65 - 12/69 Manager - Proposal Engineering Subsection, Nuclear Ins trumenta tion Deoartment, General Electric Comoany, San Jose, California.

Responsible for the application of instrumentation systems for nuclear power reactors during the proposal and pre-order period.

Responsible for technical review of bid specifications, preparation of technical bid clarifications and exceptions, definition of material list for cost es tima tin g , and the "as sold" review of

. contracts prior to turnover to Application Engineering. ?stsoanel supe rvis ed varied from 2 to 9 engineers.

4 8/64 - 12/65 S ales En ginee r, Nuclear Electronics 3 us ine s s Section of Atomic  ;

Power Equipment Decartment, General Electric Comoanv, S an -Jose ,

California.

Responsible for the bid review, contract ne go tia tio n , and sale of

ins trumenta tion sys tems and components for nuclear power plants,

[ cast reactors,.and radiation hot cells. Also r es po n s ib le for

-indus trial- s ales of radiation sensing systems for measurement of chemical properties, level, and density.

10/61 - 8/64' Apolication Engineer, Low 7eltage Switchgear Department, General Electric Comoany, Philadelphia, Pennsylvania.

Responsible for the application and design .o f advanced diode and s il ic o n-c o n t r o lle d rectifier constant voltage DC power sys tems and variabit voltage DC power sys tems fo r indus trial app lications .

Designed, f ollowed manuf acturing and personally tested an advanced SCR power supply for pro duc t introduction at the Iron and S teel Show.

Project Engineer for a DC power sys tem for an aluminum pot line sold to Anaconda beginning at the 161KV switchyard and encompassing all the equipment to convert the power to 700 volts DC at 160,000 amperes.

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9/60 -

10/61 GE Ro tational Trainine P rogram Four 3-month as s ignments on the GE Rotational Training Program for college technical graduates as follows:

a. Installation and Service Eng. - Detroit, Michigan.

Installation and startup testing of the world's largest automated-hot strip steel mill.

b. Tester - Industrv Con trol - Ro an ok e , Virginia.

Factory testing of control panels for control of j

steel, paper, pulp, -and utility mills and power 1 plants.

c. Engineer - Light Militarv Electronics - Johnson City, New York.

Design of ground support equipment f or tes ting the auto pilo ts. on the F-105.

d. Sales Engineer - Morrison, Illinois.

S ale of appliance controls including range timers and refrigerator cold controls.

EDUCATION:

S achelo r o f S cience Electrical Engineering, University of Arizona, 1960.

Mas ter o f 3 us ine ss Adstai. u t:i.- . University of Santa Clara, 1969.

PRO FESSIONAL AFFILI ATION :

Regis te red -Quality Engineer, License No . QUS05, S tate of California.

Member of Subcommittee S of the Nuclear Power Engineering Committee of the IEEE Power Engineering Society responsible for the prepara-tion and revision of the f ollowing 4 national Q . A. Standards:

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a. IEEE 498 (ANSI N4 5. 2.16) : S up p l eme n t a ry Requirements for the Calibration and Control.of Measuring and Test Equipment used in th e Construction and Maintenance of

' Nuclear Power Generating Staticas.

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P RO FE S S I ON AL AFFILIATION: ( Con td)

b. IEEE 336 (ANS I N4 S . 2.41 - Ins:allation, Inspection, and Testing Requirements for Instrumentation and l Elec tric Equipment during the Cons true:1on o f Nucler r Power Generating S tations .

c. IEEE 467 (ANSI 45.2.141: Quality Assurance Program Requiremen:s for the. Design and Manufacture of Class IE Instrumentation and electric Equipment for Nuclear Power Generating S ta tions .

d. IEEE Draft: Requirements fo r Replacement Parts for Class IE Equipment Replacement Parts for Nuclear Power Generating Stations.

PE RS ON AL DATA:

Sirth'Date: 7/08/37-Married; three children H eal th : Excellent PUBLICATIONS AND TES T! MONY :

1. In-Core Svstem ?rovides Continuous Flux Mas of Reactor Cores, R .3 . Hubbard and C.E. Foreman, Power, November, 1967.

2. Quality Assurance: Providing I:. Proving It, R .3 . Hubbard, Power, May, 1972.

3. Tes timony o f R.3. Hubbard, D.G. 3 ridenba u gh , and G.C. Minor before ths Uni:ad States Congress, Joint Commi::ee on Ato=ic Energy, Feb rua ry 13, 1976, Washington, DC. (Published by the Unio n o f Conce rned S cientis ts , Cambridge, Massachusetts.)

Excerpts from testimony published in Ouo te Without Co mmen t ,

Chemtech, May, 1976.

4.

4 Testimony of R.B. Hubbard, D.G. 3ridenbaugh, and G.C. Minor to the California S tate Assembly Committee on Re s o urces , Land Use, and Energy, Sacra = ento, California, March 3, 1976.

5. Testimony of R. 3. Hubbard and G.C. Mino r - b e f o r e California S tate Senate Committee on Public Utilities, Transit, and Energy,

! Sacramen:o, California, March 23, 1976.

6. Testi=ony or R.3. Hubbard and G.C. Minor, Judicial Hearings Regarding Crafenrheinfeld Nuclear Plant, March 16 & 17, 1977, Wur: burg, Germany.

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PU3LICATIONS AND TES TIMONY : ( Con td)

7. Testimony of R.3..Hubbard to United States House of Representatives, Subcommittee on Energy and the Environ-nent, June 30, 1977, Washing:on, DC, entitled, Effectiveness of NRC Regulations - Modifications to Diablo Canven Nuclear Units.

3. Testimony of R.3. Hubbard to the Advisory Co==ittee on Reac:or Safeguards, Augus: 12, 1977, Washington, DC, entitled, Risk 1 Uncertaintv Due to Deficiencies in Diablo Canyon Ouality Assurance Program and Fa ilur e to Isolement Current NRC Fractices.

9. The Risks of Nuclear Power Reactors: A Review of the NRC Reactor Safety Studv WASH-1400, Kendall, et al, edi:ed by R.3.

Hubbard and G.C. Minor for the Union of Concerned Scientis :s ,

August, 1977.

10. Swedish Reactor S af etv Studv: Barsebsck Risk Assessment, MH3 Technical Associates, January 1978 (Published by Swedish Depart-ment of Industry.as Document DSI 1973:1).

11. Testi=ony of R.S. Hubbard bef ore the Energy Facility S iting Council, March 31, l '.* 7 8 , in the mat:er of Febble Springs Nuclear Power Plant, Risk Assessmen:: Febble Serings Nuclear Plant,

?ortland, Oregon.

12. Presentation by R.3. Hubbard before :he Federal Minis try for Research and Technology (3 MFT) , August 31 and September 1, 1973, Meeting on Reacto.: Safety Research, Risk Analysis, 3onn, Germany.

13. Tes tisony by R.3. Hubbard, D.G.-3ridenbaugh, and G.C. Minor before the Atomic S af e:7 and Licens ing Board, Septe=ber 25, 1973, in the satter of'the Black Fox Nuclear Power 5:ation Construction Permit hearings, Tulsa, Oklahoma.

14 Tes timony o f R.3. Hubbard bef o re the Atomic Safety and Licensing Board, November 17, 1978, in the matter of Diablo Canyon Nuclear Power Plant Ooerating License Hearings. Ooerating Basis Earth-ouake and S eis mic Reanalysis of Structures, Systems, and Com-ponents, Avila Beach, California.

15. Testimony of R.3. Hubbard and D.G. 3ridenbaugh before the Louisiana Public Service Commission, November 19, 1973, Nuclear Plant and Power Generation Costs, 3aton Rouge, L ou is ia na .

16. Tes ti=ony o f R.3. Hubbard b ef ore the California Legislature, Subcommittee on Energy, Los Angeles, April 12, 1979.

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i PUBLICATIONS AND TESTIMONY: (Contd)

- 17. Tes ti=o ny o f R.3. Hubbard and G .C. Mino r befo re the Federal' Trade' Commission, on behalf of the Union of Concerned-Scientists, S tanda r ds and Certification'?roposed Rule 16 CFR Part'457, May 13,-1979.

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13. ALO-62, Imoroving the Safetv o f LWR Power Plants, MH3 Technical Associates, prepared for U .S . Department of Ener gy . . S andia National Laboratories, September, 1979, available from NTIS.

19. . Testimony by R.3. Hubb ard b e f o re - the Art:ona S tate Legislature,

.Special Interim House Committee on Atomic Energy, Overview of i

Nuclear S af ety , Phoenix, AZ. September 20, 1979.

20. "The Role of the Technical Consultant," Fractising Law Insti-tute program on " Nuclear Litiga tion ," New York City and Chicago, i

November, 1979. Available from ?LI, New York-City.

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21. Uncertainty in Nuclear Risk Assessment Me tho dolo gy , MH3 Technical ,

Associates, January, 1980, preparedSfor and available from the tockholm, Sweden.

Swedish Nuclear Power Inspectorate, Italian Reac tor Safe ty Studv; Caorso Risk Assessment, MH3 22.

Technical Associates , March, 1980, prepared for and available  ;

from Friends of ths I rth, Rome, Italy.

23. Develooment of Study Plans: Safetv Assessment of Monticello and ?rsirie Island Nuclear S tations , MH3 Technical Associates .

4 August,=1980, prepared for and available from the Minnesota

' Follution Control Agency .

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. -PROFES SIONAL 0UALIyICATIONS O F GREGORY C. MIN O R -

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' GRE GO RY C . MINOR MH3 Technical Associates 1723 Hamilton Avenue Suite K San Jose, California 95125 (403) 266-2716 4

E.G E RIEN CE :

1976 -

P RIS ENT Vice-?residen: - MH3 Te chn ical As s ocia tes . San Jose. California.

Engineer:ng and energy consultant to sta:e, federal, and private organizations and individusa,ls. Maj o r activi:1es include studies of safety and risk involved in energy generation, providing tech-nical_ consulting to legislative, regulatory, public and private groups and expert witness in behalf of state organica: ions and citi: ens' groups. 'J a s co-edi:or of a critique of the Reactor S af ety S tudy ('4 AS H - 14 0 0 ) for the Union o f Conce rned S cientis ts and co-author of a risk analysis of Swedish reactors for the Swedish Energy Commission. Served on the Peer Review Group of the NRC/TM1 Special Inquiry Group (1.;ovin Committee). Actively

. involved in the Nuclear Powe r Plant s:andards ;o:mi::ee work for the Instrumen: Society of America (ISA).

1972 - 1976 1

Manager. Adv an ce d Control and Instrumentation Engineering, General Elactric Comoany, Nuclear Energy Division. San Jose, California.

Managed a design and development group of thirty-four engineers

and support personnel designing systems for use in the =easurement,

contro*. and

peratien of nuclear reactors. Involved coordination v'ith other reactor design organi:aciona, :he nuclaar Regulatory Co mmis s io n , and cus tomers , both overseas and domestic. Responsi-b ili tie s included coordinating and managing the-design and development of con:rol sys:eaa, i c f : :,- ' s y s t a = s , and new control concepts for use on the nex: generation of reactors. The position

, included res ponsib ility . f or s tandards applicable to control and ins:ruman:stion, as well as the design of short-ters solutions to field pechlems. The disciplines involved included electrical and mechanical engineering, seismic design and process computer con trol /

programming.

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_l970-- 1972 Manage r, Reactor Con trol .Sys tems Design, General Elec:ric Company,

-Nuclear Energv Division. San Jose, California.

l Managed a group'of seven engineers and two support personnel in the design and preparation of the detailed systes drawings and control documents rela:ing to safety _and emergency. systems for nuclear reactors. -Re s p o n s ib ill:y required coordination wi:h o ther design organiza: ions and interaction wi:h the customer's engineering personnel, as well as regulatory personnel.

4 1963 - 1970 Des ign Engine er , General Electric Company, Nuclear Energv Division,

- San Jose. California.

Responsible for the design'of specific control and instrumentation systems :or nuclear reactors. Lead design responsibility for various

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subsystems of instrumentation.used to seasure neutron fluxPerformed in the reactor during startup and in termedia:c power operation.

lead systes design function in the design of a majorOther systes r e s for p' on s i-seasuring :he powar generated in nucleares:ing reactors.

of a complete reactor b ilitie s included on-site checkou; and Received control system at an experimental reactor in the Southwest.

pa:en: f or Nuclear ?ower Monitoring Sysees.

4 1960 - 1963 1.-- ~ : Assignments 4

Advanced Engineering P ro gram, General Electri: ? :

in '4 as h in 2 t on . Cali f o rn ia , and Arisona.

4 Rotating assignments in a varia:y :f ' t : :ip line s :

1

- Engineer, reactor maintenance and instrument design,

' KE and D reactors, H an f ord , Washing ton , circuit design and equipment maintenance coo rdination.

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- Design engineer, Microwave Department, Palo Alto, Cali-fornia. Worked on design of ca 71:7 ~ couplers fo r TWT's .

l Design engineer, Computer Department, Phoenix, Arizona.

Design of core driving circuitry.

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- Design engineer, Atomic Power Equipmen: Departmen:, San Jose, California. Circui: design and analysis.

> - Design enginee r , Space Sys tems Department, Santa Barbara, California. P repared control portion o f satellite proposal.

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- Technical S taf f .- Technical Milita ry Planning Operation.

Prepare analysis of

- (TEMPO) , S an ta S arb ara, California.

mis s ile exchanges.

During-this period, completed :hree-year General principles Electric program of extensive education in advanced engineering of high-Also completed courses er mathematics, probability and analysis .

in Kep ner-Tregoe, Effective Presentation, Management Training Pro-gram, and various-technical. seminars.

F EDUCATION Univers ity of California at 3erkeley, SSEE, 1960.

4 . Advanced Course in Engineering - three-year curriculum, l General Electric Company, 1963.

S tanf ord Univers ity , MS E E , 1966.

M O N*0 RS AND AS SO CI ATIONS

- Tau 3 eta Pi Engineering Hono rary S ocie ty.

- Co-holder of U .S . 'P aten: No. 3,565-760, " Nuclear Reactor Power Monitorin g S ys tem," F eb rua ry , 19i1.

- Member: American Association for Advance of Science.

- Member: Nuclear ?over Plant Standards Co=mittee, Instru-ment Society of America.

P E RS ON AL D AT A Born: June 7, 1937

' Y.arried, three children Residence: San Jose, California 1

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PUBLICATIONS AND TESTIMONY G.C. Minor, S.E. Moore, " Control Ro d Signal Multiplexing,"

1.

IEEE Transactions on Nuclear S cience , Vol. NS-19, February, 1972.

2. G.C. Mino r, W .G . Milas, "An integra:ed Control Room Sys tem for a Nuclear Power Plant," NEDO-10653, presented at In-ternational Nuclear ladus tries Fair and Technical Mee tings ,

O c tob e r , 1972, 3asle , Switzerland .

3. The above at:1cle was also p ub lis he d in th e German Technical Magazine, NT, March, 1973.

4 T es timony o f G .C. Minor. D.G. 3ridenbaugh, and R.3. Hubbard before the Joint. Committee on Atomic Energy, Hearings held February 18, 1976, and publisned by the Union of Concerned Sciencists, Cambridge, Massachusetts.

5. Testimony of G.C. Minor, D.C. 3 r ide nb au gh , and R.S . Hubb ard before the California S ta:e Assembly Committee on Resources, Land Use, and Energy, March 3, 1976.

6. Tes timony o f G.C. Minor and R.3. Hubbard bef ore the Cali-f ornia S tate Senate Co mmit:a a en ?ublic Utili:ies, Transit, '

and Energy, March 23, 1976.

7. Testimony of G.C. Minor rega'rding the Grafenrheinfeld Nu-clear Plant, March 16-17, 19 7 7, W u r: burg , Germany.

S. Tes timo ny o f G .C . Mino r b ef ore the' Cluf f Lake 3oard of In-quiry, Regina, Saskatchewan, Can:fa. E ap : mb e r 21, 1977.

9. The Risks o f Nuclear Power Reactors - A Review of the NRC Reac to r Saf e ty Studv W AS H - 14 0 0 - ( N UREG -7 5 / 01401, H . Kendall, et al edited by G.C. Minor and R .3 . Hubbard for the Un ion o f Concerned S cientis ts , August, 1977.

S w e d is h Reac tor S af ety Study: 3 ars eb3ck Risk Assessment, 10 .

MHB Te chnical As s ociates , January, 1978. (Published by Swedish Dep artment o f Indus try as Document SdI 1973:1)

11. Testimony by G.C. Minor before the Wisconsin Public Service Co mmis s ion , February 13, 1978, Loss of Coolant Accidents:

Their Probab ility and Consecuence.

12. Tes :1 mony by G.C. 'tinor before the Calif ornia Legislature Assemb ly Committee on Resources, Land Use, and Energy, A3 3103, April 26, 1973, Sacramen:o, Califo rnia .

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l 13. P res en ta tion : by G.C. Minor'before the Federal Minis try

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. -for Research and Technology (3 MFT) , Meeting on Reac to r S af e ty Research, Man / Machine Interface in Nuclear Re ac to rs ,

August.21, and September 1, 1978, Sonn, Germany.

~14 Testimony by G.C. Minor, D.G. 3ridenbaugh, and R.3. Hubbard, before'the Atomic Safe:y and Licensing Board, S ep tember 25, 1973, in the matter of the Black Fox Nuclear Power S tation Cons truction Permit Hearings, Tulsa, Oklahoma.

15. Testimony of G.C. Minor, ASL3 Hearings Related to TMI-2 Accident, Rancho Seco P ower P lan t, on behalf of Friends of.the Earth, September 13, 1979.

16. Testimony of G.C. Minor before the Michigan S tate Legisla-ture, Special Joint Committee on Nuclear Energy, Imp lica t ions of Three Mil e_,I s_l,an d A c c i d e n t f or Nuclear Power Plants in Michigan, 10 /15 / 7 Y .

. 17 ', A Critical View of Reactor Safety, by G.C. Minor, paper presen:ed.co.the Ame rican - As s o cia t ion for the Advancement o f S cience, Sympos ium on Nuclear Reactor S af e ty, January 7, 1980, San Francisco, California.

13. The Effects of Aging on S af ety of Nuclear ?ower Plan ts ,

pap.: presentad a: Fo rum on Swedish Nuclear Referendu=,

S tockholm, Sweden, March 1, 1980.

19. Minnesota Nuclear Plants Gaseous Emissions Study, MH3

!achnica l As sociates , S ep temb e r, 1980, prepared for the Minnesota Pollution Contrbl Asency, Roseville, MN.

j 20. Testimony of G.C. Minor and D.G. 3ridenbaugh before the New York S tat e Public S ervice Commis sion, Shoreham Nuclear-i Plant Construction Schedule, in the matter-of Long Island Lighting Company Temporary Rate Case, September ~22, 1980.

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TABLE B.1 TMI ACTION PLAN PRIORITY RANKING SYSTEM Rank I. Safety Significance High............................................................'100

' Medium.......................................................... 50 Low...................................................... ...... O II. Type of Improvement Improves the human element................ . . ................ 20 Fixes the' hardware............... .............................. 10 III. Utilization of Resources A. Project is ongoing, and resources would be wasted if stopped.... 20 Project has not yet been initiated. .. . . . ....................... 10 B. Staff resource requirement: Total - $50K = 1 my Small (< 2 my)................ ...... ............ . . ....... 20 Medium (> 2 < 10 my)........................ . ...... ......... 10 Large. (> 10 my)... .............. ....... ................... .. O C. Industry resource requirement: Total per unit over 40 yr life - 1 my = $50K Small (< $1.0M)................................................. 20 Large (> $1.0M)................................................. O IV. Timing of Improvement (i.e. , how quickly will the expected benefit begin to be realized after initiation of task)

Short-term (within one year).................................... 30 Ne a r- te rm (wi thi n two ye a rs ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 l

l Long-term (within three years).................................. 10 l Extended beyond three years..................................... O l

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TABLE B.2 - COMPARAIlvt PH10RillE5 Of TASK ACTIONS Key to Symbols Decision Group: A = Items or criteria already .pproved by.the Commission in the course of business apart from the Action Plan.

8 = ltems for which the scope end criteria are sufficiently well-defined in the plan that additiona? study is not requi tect. Commission appraval of the plan means, for these items, implementation in the manner described in the plas,, consistent with a policy to solicit and consider public comments on these and any other IMI related

. requires.ents developed in accord with the plan. .Thir policy may impact the estimated implementation deadlines presently shown for these Decision Group B items in L.*.e plan and in Table 1.-

C = Items which require furth u definition of scope, need, and criteria. Cosmiission approval tf the plan means.

for these items, approval to commit the necessary staf f resources, consistent with other resource priorities, to develop the information needed to bring the item separately to the Cimmission for a decision on the schedule shown in the pina.

D = Items that are related to. but not direc.tly lerived from,.the IMI-2 accident and are.more properly characterized as part of the agency's aormal operating plan. Some Decision Group D items are ongoing. Decision Group D ,

items are included in th plan for. coa.pleteness but are to be scheduled and assigned resources along with the other normal furctions on the agency in its routine operating plan and budgetary process, ticensee implementation details for Decision Group D items are not included in this Action Plan.

N Mey to NIOL Column FL - action must be complete for near-term operating license facilities before fuel loading.

IP - action must be complete for near-term operatinu license f acilities before full power operation. 3 FL & FP - part of action for near-term operating license facilities must be complete before fuel' loading and part before full power operation.

FL 4.1/1/81 - part of the aciion must be complete for near-term operating license facilities before fuel loading and part by Jaa,.sary 1, 1981.

The other items are not applicable to near-term operating applicants. They are either internal NRC actions or longer range license requirements that have not been issued yet.

i

)

n.

Key to AIF Column -

I - high priority

'll - low priority III - task should be removed from Action Plan .

( ) - scopes of tasks assigned priorities by AIF are not comparable with the scopes of NRC tasks.

Key to ACRS Column date - letter from ACRS forwarding recommendation (s) or connent(s) for which the NRC staf f feels the identified task adequately responds to the ACRS concern.

9 N

G B

i ww- == ,

h - - _ _ _ _ _ _ _ - _ _ _ _ _ ._.

P IABLE 8.2 (continued)

^

NRC Alf '

Implementation-Priority List and litle Points Priority ACR5 N10L Dates Decision Group A 1.A.2.1 Immediate upgrading of Operator and 210 S/16/79 Senior Operator Trainiiig and Qualifications I, II, Ill -

5/1/80 to 12/1/80 (Pca r t s)

II.D.3 Relief and Safety Valve Position Indication -210 -

. f t. 1/1/80 II.F.1 Additional Accident Mo.iltoring Instrumentation 210 -

4/7/79, 5/16/19, -

7/1/80 to.1/1/81 8/13/79 II.F.2 . Identification of and Recovery f rom 210 -

4/7/79, 5/16/79 FL Conditions leading to Inadequate Core Cooling 1/1/80 ar.d 1/1/81 p I.A.I.3 Shift Manning 200 (11) -

ft 8/1/80 to 7/1/82 1.8.1.2 Evaluation of Organization and 200 -

12/13/19, FL -

Hanagement Improvements of NIOL 3/11/80 Applicants I.C.5 Procedures for feedback of Operating 200 1 12/13//9 L aperience FL 1/1/81 I.C.7 H555 Vendor Review of Procedures 200 .8/14/19, II FL & IP -

(Part) 3/11/60 li.B.4 Training for Hitigating Core Damage 200

( -

1 FL & FP 1/1/81 and 4/1/81 II.E.1.2 Auxiliary feedwater System Automatic 200 - -

IL 6/1/80 and 1/1/81 Initiation and flow Ir.dication II.E.3.1 Reliability of Power Supplies for Natural 200 -

5/16/79 fP Circulation 1/1/80 e

I e

_ . _ . _. . . - _ . - . ~ . __ _ _ . _ _ . . . _ ..

8 lABLE 0.2 (continue 1) '

I NRC Alf Implementation Priority List and Title . Points Priority ACRS NIOL Dates II.E.4.2 -Containment Isolation Dependability 200 -

-8/14/79 FP 1/1/80 to 11/1/80 II.K.1 IE Bulletins on Measures to Mitigate 200 1 3/11/60 FL 3/31/80 and Small Break LOCAs and Loss of see lable C.1 feedwater Accidents II . K. 2 Commission Orders on B&W Plants 200 - - -

1/1/81 and

see lable C.1 III.A.3.1 NRC Role in Responding to Nuclear 200 -

5/16/79, 3/11/80 FP 2/80 (complete)

Emergencies Ill.B.1 Iransfer of Responsibilities to FEMA 200 -

5/16/19 IL NA I . A. I.1 Shift Teclinical Advisor 190 -

8/13/79,'12/13/19 FL 1/1/80 vi I.B.2.2 Resident inspector at Operating Reactors 190 - -

FL 10/1/80 1.D.1 Control Room Design Reviews 190 Il 12/13/79 FL -

I.E.1 Office for Analysis.and Evaluation of 190 -

5/16/79 -

7/80 Operation Data .

1.E.2 Program Office Operational Data Actiyfties 190 -

5/16/79 -

6/80 I I . D.1 Testing Requirements 190 - -

FL and 1/1/80 and 7/1/80 7/1/81 II.E.1.1 Auxiliary feedwater System Evaluation 190 1 -

FP 6/1/80 an.1 1/1/82

, ll.E.4.1 Dedicated Penetrations 190 -

3/11/80 fL 1/1/80 and 1/1/81 II.E.4.4 Purging 190 I 5/16/79 -

1/1/80, staged

_ ,_ + = = = ,

• O hI

II.8LE B.2 (continued)

NRC Alf laplementation Priority List and Title Points Priority ACRS HIOL Dates 1

1.E.3 Operational Safety Data Analysis 180 -

5/16/79 -

Ongoing II.E.2.2 Research on Small-tireak LOCAs and 180 - 4/7/79, 4/18/79, -

NA Anomalous Transients 5/16/79, and 8/14/79 II.G.1 Power Supplies for Pressurizer Relief 180 -

5/16/19 FL 1/1/80 Valves,. Block Valves, and level Indications III.A.I.2 Upgrade Licensee leergency Support Facilities 180 -

5/16/79 ft and 1/1/81 1/1/80 to 1/1/81 111.B.2 Implementation of HRC's and FEMA's 180 -

5/16/79 -

NA Responsibilities cn

, II.C.1 Interim Reliability Evaluation 180 (1) 5/16/79, 8/14/79, -

7/80 to 3/81 Program (IREP) 12/13/79, and 3/11/80 II.C.3 Systems interaction 180 -

8/14/79, 12/13/79 -

Plant specific I.C.1 Short-term Accident Analysis and 170 -

5/16/79, B/14/79, FL and FP 1/1/80 Procedures Revision 12/13/19, and 3/11/80 11.B.6 Risk Reduction for Operating Reactors at 170 -

12/13/79, 5/16/79 -

Selected sites -

Sites with liigh Population Densities 10/1/80 II.H.1 Maintain Safety of lHi-2 and Minimize 170 - - -

HA Environmental Impact lil. A. l.1 Upgrade Emergency Preparedness 170 -

5/16/79 fL and 1/1/01 Phased 1/1/80 -

1/1/85 1.A.I.2 Shift Supervisor Adininistrative Duties IE0 - -

FL 1/1/80 i .

j TABLE B.2 (continued) .

NRC AIF Implementation Priority List and Title Points Priority ACR5 NIOL Dates ,

I.C.2 Shift and Relief Iurnover Procedures 160 -

Implicitly FL 1/1/80 l.C 3 shift Supervisor Responsibilities 160 - -

ft 1/1/80 1.C.4 Control Room Access 160 - -

IL 1/1/80 li.H.2 Obtain Technical Data on the Conditions 160 - - -

NA

• Inside the lHI-2 Containment Structure ll.H.3 Evaluate and f eedback Information Obtain.:d 160 -

5/16/79 -

NA from IMI I.D.5 Improved Control Room Instrumentation Research 160 -

12/13/79 -

NA

, 11.8.3 Post-accident $ampling 150 -

12/13/79, 3/11/80 FP and 1/1/81 1/1/8'a to 1/1/81 j

'a IV.A.1 Seek Legislative Authority 150 - - -

NA

!.A.3.1 Revise Scope and Criteria for Licensing Exams 140 I, 111 -

FL 3/2e/80 to 11/1/80 (Parts) 1.C.8 Pilot Monitoring of Selected Emergency 140 til 3/11/80 FP -

Procedures for Hl0lc Applicants ,

l.E.8 Human Error Rate Analysis 140 -

5/16/79 -

NA II.B.1 Reactor Coolant System Vents 140 -

4/1/19 FP and 1/1/81 1/1/80 and 1/1/81 II.B.2 Plant Shielding to Provide Access to Vital 140 -

12/13/19 FP and 1/1/81 1/1/80 and 1/1/81 -

Areas and Protect Safety Equipment for Post-Accident Operation

!!.E.5 Design Sensitivity of B&W Reactors 140 - - -

4/1/81 hI

TABLE 8.2 (continued)

NRC Alf implementation Priority List and Title Points Priority ACRS NIOL Dates ll.E.5.2 B&W Reactor Transient Response Task force 140 - -

NA NA Ill.A.3.3 Communications, Iten (1) 140 111 5/16/19 IL 3/1/80 111.D.1.1 Primary Coolant Sources Outside the Containment 140 - -

FP 1/1/80-1.C.1 ' Training Requirements 130 11 12/11/19 ft and IP -

11.8.5 Research on Phenomena Associated with Core 130 -

3/21/19, 12/13/79 -

NA Degradation and fuel Helting 3/11/79 ou 1.A.2.3 Administration of Training Programs 130 - - -

8/1/80 II.D.2 Research on Relief and safety Valve Test 120 - - -

NA Requirements

!!I.D.3.4 Control Room Habitability 120' 11 3/11/80 FP and 3/1/81 1/1/81 and 1/1/8) 1.D.6 Technology Transfer Conference 110 - - -

Complete II.J.2.2 Increase Emphasis on Independent Measurement 110 - - -

NA-in the Construction inspection Program II.J 2.1 Reorient Construction Inspection Progran . 100 - - -

NA III.D.3.3 Inplant Monitoring, item (1) 100 11 -

FL 1/1/80-1/1/81 IV. F.1 Increased IE Scrutiny of Power Ascension 100 - -

FL - until -

Test Program completion 4

h

i m

IABLE B.2 (continued)

NRC All implementation Priority List and Iltle Points Priority ACRS N10L Dates IV.H NRC Participation in the Radiation Policy 100 - -

NA NA Council 1 1 . 11. 4 Determine Impact of INI on Socioeconomic and Real 90 - - -

NA Property values III.D.2;4 Offitte Dose Measurements, Item (1) 90 - 111 -

FP NA Decision Group B ~

1.s.4.1 initial $1mulator Improvesent 200 -

II -

1/1/82 I.E.4 Coordination of Licensee, Industry, and 200 $/16/79 Regulatory Programs 11 -

6/80

• 1.A.2.5 Plant Drills 190 III - -

7/1/81 I.D.1 Control Room Design views 190 12/13/79 Il -

1/1/82 to 1/1/83 1.D.2 Plant Safety Paramu ter Display Console 180 11 12/13/79 -

1/1/82 1.D.4 Control Room Design Standard 180 -

12/13/79 -

NA II . E. 3. 2 Systems Reliability 180 -

8/14/79, 12/13/79 -

NA II.K.3 final Recommendations of B&O Task force 180 -

3/11/80 See Table C.3 See Table C.3 I A.4.2 tong-term Training simulator Upgrade 170 II - -

NA II.E.2.1 Reliance on ECCS 170 11 - -

Beyond 1/1/82 II.F.3 Instruments for Monitoring Accident Conditions 170 11 4/18/79 -

6/82 (Regulatory Guide 1.97)

d IABLE 8.2 (continued)

NHC AIF Implementation Priority List and Title Po ir,t s Priority ACRS NIOL Dates

1. E.1. foreign Sources 160 -

5/16/79' -

NA II.E.4.3 Integrity Check 150 III - -

NA Ill.A.3.2 Improve Operations Centers 140 -

5/16/79 - -

I.G.2 Scope of Test Program 140 - - -

NA I.A.2.2 Iraining and Qualifications of Operations 130 11 2/13/80 .

1/1/82 Personnel I.A.2.3 Administration of Training Programs 130 - - -

NA 111.D.3.1 Radiation Prot'ection Plans 130 III - -

9/1/81 II.F.5 Classification of Electrical, Instrumentation, 130 -

4/17/80 -

NA and Eontiol Equipment I.C.6 Procedures for Verification of Correct 120 - - -

1/1/81 Performance of Operating Activities

1. F.1 Expanded Quality Assurance List 120 -

8/14/19, 12/17/79 -

NA 111.0.2.1 Radiological Monitoring of Effluents I?O 111 - -

NA 111.D.1.3 Ventilation System and Radiolodine Adsorber 110 III - -

NA Criteria II.C.4 Reliability Engineering 1 10 I 10/12/79, 12/13/79 - Beyond 1982 III.D.3.3 Inplant Radiation Konitoring, item (2) 100 11 -

6/1/82 1.A.2.4 NRR Participation in Inspector Training 90 - - -

NA lt

TABLE B.2 (continued) , .

NHC Alf Implementation Points NIOL Dates Priority List and Title Priority ACRS 111.0.2.5 Offsite Dose Calculation Manual 80 til 9/1/82 10/9/79 - NA Ill.D.I.2 Radioactive Gas Management 70 -

50 - - - NA III.D.2.2 Radiolodine, C-14, and Tritium Pattuay Dose Analysis Decision Group C I,. A. 2. 6 long-tern Upgrading of Training and 190 - - -

2/1/82 Qualifications I.E.6 Reporting Requirements' 190 1 5/16/19 -

NA II,J.3.1 Organisation and Staffing to Oversee 180 Ill g Design and Construction v

IV.E.5 Assess Currently Operating Reactors 180 -

10/11/79 - NA I . 8.1.1 Organization and Management Long-tern 170 III 12/13/79 -

5/1/81 Improvements I.C.9 Long-tern Program Plan for Upgrading of 170 III 8/14/79 -

NA Procedures I . 8.1. 3 loss of safety Function 160 Ill 8/13/19 -

NA II.B.7 Analysis of ifydrogen Control 160 11 3/21/79, 12/13/79 IP NA 8/13/79 II.J.4.1 Revise Deficiency Reporting Requirements 160 -

4/17/80 -

NA 1.D.3 Safety System Status Monitoring 150 11 12/13/79, 5/16/79 -

NA I II.E.2.3 Uncertainties in Performance Predictions 150 11 4/1/79, 8/14/79 -

Beyond 1982 h4

lABLE 8.2 (continued)

NRC Alf laplementation Priority List and Title Points Priority ACRS NIOL

. Dates II.C.2 Continuation of INEP 150 1 8/14/19, 12/13/19 -

1983 4/17/80 til.A.3.3 Cosaunications Backup, Item (2) 140 (III) 5/16/19 -

NA 111.0.1.1 Primary Coolant Sources Outside the Containment 140 - - -

NA Structure ,

NA -

l.A.2.7- Accreditation of Iraining Institutions 130 -

Licensing of Additional Operations Personnel 130 - -

NA I.A 3.4 II.E.3.3 Coordinated Study of Shutdown lleat Removal 130 -

3/11/80. 4/17/80 -

NA Requirements

• t*

" IV.C.1 Extend Lessons learned From IMI to Other 130 - - -

NA HRC Programs .

II.A.2 Site Evaluation of Existing Facilities 120 -

2/14/80 -

NA 11.8.8 Rulemaking Proceeding 120 -

12/17/179 FP NA IV.0.1 NRC Staff Iraining 120 -

5/16/19. 12/13/19 -

NA 111.D.3.4 Control Room Habitability 120 11 3/11/80 -

NA Ill.A.3.6 Interaction of NRC with Other Agencies 110 -

5/16/79 -

NA IV.E.2 Plan for Early Resolution of safety issues 110 -

12/11/19 -

NA II.A.1 Siting Policy Reformulation 110 -

12/17/79, 2/14/80 -

HA issue Regulatory Guide 100 - - -

NA II.J.3.2 Il l . A.1. 3 Maintain Supplies of Thyroid Blocking Agent 100 II 3/1/81 (Potassium lodide) .

se

. ~ - ._. . - -_ . . .. - -

IABLE B.2 (continued)

HRC Alf .

laplementation Points Priority ACRS NI0t Dates Priority List and Title Decision Group D 1.A.1.4 Long-Tern Upgrading Plant Drills

1. A. 2. 5 1.A.3.3 Requirements for Operator Fitness 1.A.3.5 Establish Statement of understanding - -

With thPO and DOE feasibility Study of Procurement of - -

I.A.4.3 NRC training $1mulator e - - - -

'd I.A.4.4 Feasibility Study of HRC Engineering Computer -

1.B.2.1 Revise IE Inspection Program

1. B . 2. 3 Regional Evaluations I.B.2.4 Overview of Licensee Performance Nuclear Plant Reliability Data System - -

5/16/79 - -

1.E.5 Develop More Detailed Criteria

- - 8/14/19, 12/17/79 -

1.F.2 II.E.1.3 Update Standard Review Plan and Develop Regulatory Guide Alternate Concepts Research 4/18/79 - -

ll.E.3.4

- - -~

II.E.3.5 Regulatory Guide II.E.6.1 Test Adequacy Study h

I 1ABLE !!.2 (continued)

Priority List and Title f4HC Alt implementation-Points Priority ACRS NIOL Dates II.F.4 Study of Control and Protection Action - - -

Design Requirements -- -

II.J.l.1 Establish a Priority System for Conducting vendor - - -

Inspections - -

ll.J.l.2 Modify ixisting Vendor Inspection Program - - -

.II.J.l.3 increase Regulatory Control over Present -

Nonlicensees

^4/11/80 - -

ll.J.l.4 Assign Resident Inspectors to Reactor Vendors and - - -

Arc hi t ec t-Engineers l l . J. 2. 3 Assign Resident inspectors to all Construction - - -

Sites - -

V III.A.3.5 Training, Drills, and Tests - -

5/16/79 - -

Ill.D.2.6 Independent Radiological Measurements - - -

111,0.3.2 llealth Physics leprovements -

Ill -

111.D. 3. 3 inplant Radiation Monitoring, Item (3) & (4) - - -

III.D.3.5 Radiation Worker Exposure Data Base -

til -

IV.A.2 Revise Enforcement Policy - - -

IV.B.1 Revise Practices for issuance of Instructions -

and Infoemation to Licensees 5/16/79 - -

IV.E.1 Expand Research on Quantification of 5/16/79, 12/13/79 Safety Decision-Haking -

I r

le

I i

I I

e TA8tE B.2 (continued) f4HC Alf laplementation Priority List and Title Points Priority ACR5 - N10L Dates IV.G.1 Develop a Public Agenda for Rulemaking - -

12/13/79 -

IV.G.2 Periodic and Systematic Reevaluation - -

12/13/19 - -

of Exi, ting Rules ,

.IV.G.3 Improve dulemaking Procedures - -

12/17/19 -

IV.G.4 Study Alternatives for improved - -

12/13/19 - -

Rulemaking Process V. NRC Policy, Organization, and Management vs  ;

'"' Develop NRC Policy Statement on Safety - -

5/16/79 - -

l.

2. 5tudy Elimination of Nonsafety - - -

Hesponsibilities

3. Strengthen Role of ACR5 - - 1/15/80, 3/12/B0 - -

Study Need for Additional Advisory - - - -

4.

Committees

5. Improve Public and Intervenor Participation - - - -

in hearing Process Study Construction-During-Adjudication Rules - - - -

6.

Study Need for THI-Related legislation - - - -

7.

8. Study the Need to Establish an Independent - -

' Nuclear Safety Board Study the Reform of the Licensing Process -

9.

i .

TABtl B.? (continued)

_ NRC AIF Implementation Priority List and Title Points Priority ACR$ NIOL Dates

10. Study HRC Iop i:inagement Structure ' - - - - - -

and Process 11, Reexamine Organization and Functions - - - - -

of NkC Offices

12. Revise Delegations of Authority to Staff - -

1/15/80 - -

13. Clarify and Strengthen the Respective Roles - - - - -

of Chairman, Commission, and EDO

• 14. Authority to Delegate Emergency Response - - - - -

functions to a Single Commissioner

15. Achieve Single location - Long-term - - - - -

16. Achieve Single location - Interim - - - - -

17. Reexamine Comunission Role in Adjudication - - - - -

O w

le

_ . - - _ - _ _ _ _ _ _ .