ML20023D955

Proposed Findings of Fact & Conclusions of Law Re Seismology,Water Hammer & ALARA

ML20023D955
Person / Time
Site:	Byron
Issue date:	05/31/1983
From:	Bielawski A, Copeland V, Mark Miller COMMONWEALTH EDISON CO., ISHAM, LINCOLN & BEALE
To:
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ML20023D952	List: ML20023D951 ML20023D955
References
NUDOCS 8306060196
Download: ML20023D955 (6)

Text

apET,ED UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION O BEFORE THE ATOMIC SAFETY AND LICENSING BOARD.g3 gj 3 f0():0()

In the Matter of )

)

COMMONWEALTH EDISON COMPANY ) Docket Nos. 50-454 OL

) 50-455 OL (Byron Nuclear Power Station, )

Units 1 & 2) )

COMMONWEALTH EDISON COMPANY'S PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW REGARDING SEISMOLOGY, WATERHAMMER, AND ALARA May 31, 1983 O

8306060196 830531 PDR ADOCK 05000454 0 PDR

TABLE OF CONTENTS O Or1N10N................................................ 1 I. PROCEDURAL HISTORY................................ 1 II. CONTENTIONS....................................... 14 A. League Contention 106 - Seismology........... 14 Applicable Law.......................... 14 Capable Faulting........................ 17 Strain Guage Testing.................... 19 SSE and OBE............................. 20 Conclusion.............................. 22 B. DAARE/ SAFE Contention 9(a) - Waterhammer..... 23 Applicable Law.......................... 23 Byron and Krsko Steam Generator Design.. 25 The Krsko Waterhammer Event............. 26 The Westinghouse Recommendations........ 27 Conclusion.............................. 29 C. League Contentions 42,111, and 112 - ALARA... 30 Applicable Law.......................... 31 Permissible Occupational Exposures...... 38 Cancer Risk Estimate............... 38 ALARA Considerations.................... 39 Dosimetry.......................... 40 In-Plant Monitoring................ 40 Record Maintenance................. 41 Health Physics Staff............... 41 Training........................... 42 Declared Pregnant Women............ 42 Temporary Employees................ 43 Industrial Sabotage................ 44 Design of Byron Station............ 44 Conclusion.............................. 4S O .

)

i

1 1

FINDINGS OF FACT....................................... 46 ,

1 46 III. CONTENTIONS.......................................

A. Seismology................................... 46 62 B. Waterhammer..................................

C. ALARA........................................ 68 APPENDIX Comprehensive Index of Exhibits........................ A-1 Comprehensive Index of Witnesses....................... A-6 i i

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I l

trtXQED '

l UNITED STATES OF AMERICA l

NUCLEAR REGULATORY COMMISSION

, , , , mA R O '.01 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD IN THE MATTER OF )

)

COMMONWEALTH EDISON COMPANY ) Docket Nos. 50-454

) 50-455 (Byron Station, Units 1 )

and 2) )

COMMONWEALTH EDISON COMPANY'S PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW ,

REGARDING SEISMOLOGY, WATERHAMMER, AND ALARA OPINION I

I. PROCEDURAL HISTORY On November 30, 1978, Commonwealth Edison Company

(" Commonwealth Edison" or " Applicant") filed with the Nuclear Regulatory Commission (" Commission") an application for facil-ity operating licenses authorizing it to possess, use and operate the Byron Station, Units 1 and 2, and the Braidwood Station, Units 1 and 2, four pressurized water nuclear reactors each having a core power level of 3411 megawatts thermal and an electrical output of 1120 megawatts electric. The Byron Sta-tion is located in Rockvale Township., Ogle County, Illinois and e

These proposed findings are presented in the form of a partial initial decision which addresses three of the eight litigated issues, specifically, the issues concerning seismol-ogy, waterhammer, and ALARA. In accordance with the procedural discussion herein, the proposed findings on the remaining five issues, which concern emergency planning, quality assurance /

quality control, class 9 accidents, liquid pathways / hydrology, and steam generator tube integrity will be presented in subse-quent submissions.

the Braidwood Station is located in Reed Township, Will County, Illinois. Also on November 30, 1978, the Applicant filed, pursuant to the regulations of the Commission in 10 CFR Part 51, an Environmental Report for each facility, discussing environmental considerations related to the proposed operation of that facility.

On December 15, 1978, the Commission published in the Federal Register a Notice stating that it had received the application; that it would consider the issuance of the facil-ity operating licenses applied for; that the environmental reports filed by the Applicant would be made available to the public; and that there would be an opportunity for a hearing on the issuance of the licenses. 43 Fed. Reg. 58659-60 (Decem-ber 15, 1978).

On January 23, 1979, an Atomic Safety and Licensing Board (" Licensing Board" or "Ec,ard") was established to rule on i

petitions for leave to intervene and/or requests for hearing I

and to preside over the proceeding in the event a hearing was ordered. Timely petitions to intervene were filed by the League of Women Voters of Rockford, Illinois and (jointly) by the DeKalb Area Alliance for Responsible Energy and the Sinnis-sippi Alliance for the Environment (the " League" and "DAARE/

SAFE", respectively; collectively, "Intervenors"). The League submitted 13 proposed contentions, and DAARE/ SAFE submitted 10.

On August 21 and 22, 1979, the Licensing Board held a special prehearing conference in Rockford, Illinois. The Board-ruled that the Intervenors had demonstrated the requisite in-

l terest in the subject matter of the proceeding to establish ,

O standin, to intervene, and they were admitted as intervenin9 I

parties (Tr. 103). The Board further ruled that the Inter- l 1

venors had each stated one or more valid contentions. The Board, however, did not rule on the admissibility of specific proposed contentions. Rather, it directed the parties to meet and conduct negotiations in an effort to refine and rephrase proper contentions for the further conduct of the proceeding (Tr. 104-10). Such negotiations were conducted.

On March 10, 1980, the League filed its " Revised Contentions", consisting of 146 numbered contentions. The Applicant filed its Answer to these contentions on April 18, 1980, and the NRC Staff filed its Answer on April 25, 1980. In an order entered on December 19, 1980, the Licensing Board admitted some 114 of these contentions and ordered that dis-covery commence forthwith on all issues included in the ad-LBP-80-30, 12 NRC 683 (1980). On May 9, mitted contentions.

1980, DAARE and SAFE filed a " Supplemental Statement of Conten-tions." The Applicant filed an answer on May 27, 1980 and the NRC Staff answered on May 29, 1980. In an order entered on December 19, 1980, the Licensing Board admitted 9 of DAARE/

SAFE's contentions as issues in this proceeding.

On February 19, 1981, the Applicant petitioned the Licensing Board for reconsideration of its order of Decem-ber 19, 1981 regarding the admissibility of the League's con-tentions. Applicant argued that the revised contentions ad--

mitted by the Board were untimely and failed to comply with the O

Commission's regulations and applicable decisional law. This petition was denied by the Licensing Board on August 18, 1981.

LBP-81-30A, 14 NRC 364 (1981).

On March 12, 1980, the League served its first set of on the Applicant. On March 19, 1980, the interrogatories Applicant objected to these interrogatories on the ground that the Licensing Board had not yet ruled on the admissibility of any of the proposed contentions to which they referred. Also on March 12, the League served interrogatories on the NRC Staff; on March 26, the NRC Staff similarly objected that the interrogatories were premature. In its order dated Decem-ber 19, 1980, the Board opened discovery on all admitted con-tentions of the League. On July 8, 1981, the Applicant served its first round of interrogatories on the League. On July 30, 1981, after the League and DAARE/ SAFE had failed to respond within the required period, the Applicant moved to compel 1

discovery. On August 18, 1981, the Licensing Board issued a memorandum and order granting the Applicant's motion to compel the League and DAARE/ SAFE to respond to pending discovery LBP-81-30A, 14 NRC 364 (1981). After the League requests.

failed to provide such discovery, the Licensing Board issued an order on October 27, 1981 dismissing the League as a party to this proceeding for "the League's total failure to provide responsive answers to interrogatories." LBP-81-52, 14 NRC 901, 906 (1981).

On June 17, 1982, the League was reinstated by order of the Atomic Safety and Licensing Appeal Board (" Appeal Board").

O

ALAB-678, 15 NRC 1400 (1982). The Appeal Board agreed that the O League's failure to answer Applicant's interrogatories consti-tuted a patent violation of the Licensing Board's discovery 15 NRC at 1416. The Appeal Board, however, while order.

concluding that the League's conduct warranted a serious sanc- I i

tion, determined that the sanction of dismissal was too severe.

15 NRC at 1417. Instead, the Appeal Board limited the number of contentions that the League would be allowed to litigate "to that number the Licensing Board concludes it can comfortably decide on the merits without unjustifiably delaying operation 15 NRC at 1420. The Appeal Board of the Byron facility."

noted that because the League would have to " revise its brcad-side approach so as to concentrate on those few contentions it l

is best prepared to advance," this approach would be "most likely to lead to a useful examination of important safety or environmental issues." M.

On April 15, 1982, the Licensing Ecall granted a l

l notion by the Staff to dismiss DAARE/ SAFE Contention 1(i),

on dealing with the financial qualifications of the Applicant, the basis of a recent change in NRC regulations eliminating the financial qualifcations review of an electric utility applicant for an operating license. 47 Fed. Reg. 13750 (March 31, 1982).

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' On July 6, 1982, the League filed petitions under 10 CFR 2.758 for waiver or exception to the new Commission regula-for tion regarding review of financial qualifications and waiver or exception to a Commission regulation eliminating consideration of need for power or alternative energy sources O

from operating license proceedings. 47 Fed. Reg. 12940 (March O 26, 1982). On July 30, 1982, DAARE/ SAFE filed similar petitions.

on August 2, 1982, the Board issued orders denying the League's petitions for failure to demonstrate the requisite special circumstances under 10 CFR 2.758 for waiver or exception to the Commission's regulations. On August 26, 1982, the Board issued l

an order denying DAARE/ SAFE's petitions on the same grounds.

On June 4, 1982, the NRC Staff moved, pursuant to 10 l

CFR Section 2.749, for summary disposition of all of the ad-mitted contentions of DAARE/ SAFE except Contention 1.

I On June 7, 1982, the Applicant filed a similar motion requesting summary disposition of all admitted DAARE/ SAFE contentions except for 9(c), concerning issues . involving steam generator tube integrity. On July 14, 1982, the NRC Staff filed a re-sponse supporting Applicant's motion as to Contention 1 and identifying possible as yet unresolved questions as to Conten-tions 3 and 9(a). On July 19, 1982, the Applicant filed an answer supporting the NRC Staff's motion as to Contention 9(c)

I except as to one issue. On July 15, 1982, DAARE/ SAFE filed answers to Applicant's and NRC Staff's motions for summary ,

disposition.

On July 15, 1982, the Applicant moved the Licensing Board to enter an order striking certain of the League's con-tentions on the ground that the League had failed to respond to interrogatories with respect to the contentions or had respond-ed inadequately.

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On August 18 and 19, 1982, a prehearing conference O was he1d at nocxford, 111inois for the purpose of determining which contentions of DAARE/ SAFE would be the subject of an evidentiary hearing which contentions the League would be l

permitted to 1itigate, as well as establishing a procedural schedule. On August 30, 1982, the Licensing Board issued an order dealing with the latter two subjects. The Board granted Applicant's motion to strike certain League contentions because i

the League had failed to answer fully interrogatories pertain-ing to them. The Board granted Applicant's motion with regard to 78 contentions and parts of another contention and denied it with respect to 2 contentions. Two other contentions were l

stricken as dealing with financial qualifications and need for l

power, issues not litigable in operating license proceedings.

These rulings left 23 contentions remaining among the League's previously admitted contentions. 'ihe League grouped these l remaining contentions into 13 groups. In regard to the proce-dural schedule, the Licensing Board used a stipulation entered into by the parties as the basis for setting a schedule. The schedule provided that discovery be initiated by November 15, 1982 and be concluded within 30 days; that motions for summary disposition of League contentions be filed by December 6, 1982; j that prefiled testimony for evidentiary hearings be filed by February 15, 1983; that the hearing commence on March 1, 1983; l

and that the parties, except the NRC Staff, file proposed findings of fact and conclusions of law within 30 days after the record of hearing is closed.

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On September 10, 1982, the Licensing Board issued an O order rutin 9 on Avv11 cant's mnd the =RC Staff's motions for

summary disposition of DAARE/ SAFE contentions. The motions were granted as to 9 contentions or parts of contentions and denied as to 4 contentions or parts of contentions.

On September 28, 1982, Commonwealth Edison moved for clarification of the September 10 Board raling on DAARE/ SAFE Contention 9(a) regarding the waterhammer issue. On January 7, 1982, the Board issued an order limiting the 1 -tigation of DAARE/ SAFE Contention 9(a) to consideration of whe'.i r a water-hammer event similar to the one that occurre:d at the Krsko plant in Yugoslavia could occur at Byron.

On December 6, 1982, all parties entered into a stipulation, which was approved by order of the Licensing Board on December 16, 1982. The stipulation represented a comprehen-sive agreement by the parties as to issues to be litigated at

the evidentiary hearings. The League agreed to withdraw 11 of its remaining contentions and portions of a twelfth. The parties agreed to litigate League contentions IA as revised, 8, l

19, 22, 39, 42, 62, 106, 108, 109, 112 and those portions of 111 which pertain to Edison's in-plant monitoring of radio-activity from Byron; and DAARE/ SAFE Contentions 2a, 3, 9a and 9c. The League and DAARE/ SAFE recognized that certain of their respective contentions to be litigated raised similar issues and agreed to consolidate their presentation of evidence, i

briefs and proposed findings of fact by designating one or the' I

other of them as the lead intervenor on certain groups of con-

tentions. The stipulation also contained agreed discovery O provisions. in approving the stipulation on December 16, 1982, the Licensing Board reworded the League's Contention 1A.

On January 24, 1983, the Licensing Board published in the Federal Register a Notice that an evidentiary hearing on  ;

the issues relative to the application for operating licenses for the Byron Units 1 and 2 would be held in Rockford, Illi-nois, commencing at 9:30 a.m. on March 1, 1983 and continuing for several weeks thereafter.

On February 21, 1983 Intervenors moved to substitute a revised emergency planning contention for DAARE/ SAFE Conten-tion 3 and League Contentions 19 and 108, each of which raised emergency planning issues. Intervenors' revised contention j raised 13 issues pertaining to the adequacy of offsite emer-gency plans and the Applicant's evacuation time study. The parties entered into a stipulation which provided that Inter-venors would withdraw DAARE/ SAFE Contention 3 and League Con-tentions 19 and 108; that the parties would litigate certain aspects of the revised contention, described below; and that they would attempt to informally resolve the remaicing aspects of the revised contention. The procedure agreed upon for  !

I informal resolution requires that Applicant satisfy specific commitments set forth in the stipulation before the Byron Station Unit 1 exceeds 5% power operation. The stipulation provides that Intervenors may petition the Board, within cer-tain time constraints, for a hearing on the question whether Applicant has satisfied any of the commitments. As part of the O

stipulation, Intervenors agreed that in the event the Board i finds in Applicant's favor with respect to the nonemergency planning issues adjudicated in this proceeding, they would not object to the full power operation of the Byron Station based upon a claim that a commitment has not been satisfied, unless and until it was determined that the public health and safety  ;

! requires restriction of operation.

Hearings commenced on March 1, 1983, in Rockford for the presentation of evidence on the issues admitted for litiga-tion. At the commencement of the evidentiary hearings, the Board informed the parties that a failure to file proposed findings of fact and conclusions of law on a particular issue could be deemed a default as to the issue. (Tr. 382-383).

Hearings continued, with some adjournments, through the months ,

l of March, April and May, 1983. The following is a brief de-scription the issues litigated during the course of the hear-ings.

Seismology -- League Contention 106 asserts that serious problems regarding the seismic characteristics of the Byron site have come to light since issuance of the construc-tion permit and that the Applicant's seismic analysis applies i

ground acceleration values for a safe-shutdown earthquake and an operating basis earthquake that are not sufficiently conser-vative.

Waterhammer -- DAARE/ SAFE Contention 9(a) concerned waterhammer events caused by rapid condensation of steam in' steam generator feedwater systems. As limited by the Board l

order of January 7, 1983, the issue is whether a waterhammer O event could occur in the steam generator feedwater bypass V

system at Byron similar to the type of event which is believed to have occured at the Krsko plant in Yugoslavia.

ALARA -- League Contentions 42, 111 and 112 assert in general that the Byron Station will not be operated so as to

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keep radiation exposure levels for plant workers as low as reasonably achievable ("ALARA") and that radiation levels at the plant will not be adequately monitored. They also assert t

that the Applicant and the Commission have underestimated the health effects of exposure to low levels of ionizing radiation.

Class 9 Accidents -- League Contentions 8 and 62 and DAARE/ SAFE Contention 2a assert generally that neither the Applicant nor the Staff has presented an adequate environmental and health and safety assessment of severe accidents in connec-tion with the proposed operation of the Byron Station.

Quality Assurance / Quality Centrol (QA/QC) -- League Contention 1A, as reworded by the Licensing Board, asserts that the Applicant lacks both the ability and the willingness to maintain adequate quality assurance and quality control pro-l l grams, as evidenced by the past performance of both the Appli-cant and its architect-engineers. It also asserts that quality assurance functions under the Applicant's program are not suf-ficiently independent from other functions within the company.

As part of the QA/QC contention, on March 21, 1983, during the course of the hearing, the Board directed Commonwealth Edison to make an evidentiary presentation as to the implications for

y the Byron plant of the February, 1983 failures of the automatic O reactor scram system at the Salem plant. ,

Steam Generator Tube Integrity -- League Contention 22 and DAARE/ SAFE Contention 9(c) allege generally that the steam generator tubes at Byron Station are likely to become significantly degraded, due to corrosion, cracking, denting and vibration-induced wear. Such degradation could lead to loss of tube integrity, which might result in radioactive primary system water leaking out of containment, either during normal operation or under accident conditions.

Liquid Pathways / Hydrology -- League Contentions 39 and 109 allege that the Applicant has not adequately assessed the potential for radioactive contaminants to infiltrate local groundwater and wells because of various deficiencies in the studies that have been performed.

Emergency Plans -- Paragraphs 2(c), 2fe) and 2(k) of I

Intervenors' revised emergency planning contention allege that Applicant's evacuation time study does not address the relative significance of alternative assumptions, does not considar the impact of peak populations, including behavioral aspects, and does not use site weather characteristics as presented in the FSAR. Paragraph 3 alleges that there are inadequate facilities to provide treatment and transportation of injured persons disaster. Paragraph 8 alleges that during a radiological emergency plans are incapable of offering sufficient guidance for the choice of actions to protect the public. Paragraph 10 alleges that emergency plans rely too heavily on volunteer l 0_

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personnel to effect an evacuation. Finally, paragraph 13 O a11 ees that there has been insufficient communication detween l emergency planners and response organizations.  ;

Pursuant to a stipulation dated August 18, 1982, all parties except the NRC Staff agreed to file proposed findings of fact and conclusions of law no later than 30 days after the close of the record and to file rebuttal to any other party, except the NRC Staff, within 10 days thereafter. The NRC Staff agreed by that stipulation to file its proposed findings and conclusions within 40 days after the close of the hearing record.

As of April 28, 1983, the record was closed on the following issues: seismology, waterhammer, ALARA and steam generator tube integrity. At that time, the Board ordered Applicant and Intervenor to file proposed findings and conclu-sions on those issues by May 31, 1983. NRC Staff proposed findings and conclusions on those issues are due June 10, 1983.

Pursuant to the Bo nd's granting of a motion by the League, the filing date for the proposed findings and conclusions on the ALARA and steam generator tube integrity issues has been ex-tended for all parties until June 14, 1983. (Applicant none-theless submits its proposed ALARA findings herein.) By the Board's order of May 19, 1983, proposed findings and conclu-sions on the emergency planning issues are due from all parties by June 20, 1983. The record on liquid pathways / hydrology and Class 9 accidents closed as of May 25, 1983 and, in accordance^

with the August 18, 1982 stipulation, proposed findings and O

conclusions on those issues are due on June 24, 1983. At the time of this writing, the record on quality assurance / quality O

control issue is subject to a pending motion to reopen the record for additional evidence; thus no schedule has yet been established for submission of proposed findings and conclusions on that issue.

II. CONTENTIONS A. League Contention 106 - Seismology i

' Contention 106 filed by the League asserts (a) that it is not known if the Plum River Fault, which comes within 5.3 miles of the Byron site, is a capable fault, (b) that strain gauge testing should be performed by Applicant on faults cut-ting basement rock located in the Northern Illinois region where earthquakes of Modified Mercalli VII intensity are ex-pected to occur, and (c) that neither Byron's designated safe shutdown earthquake (SSE) peak ground acceleration value of

.20g nor its operating basis earthquake (OBE) pe.ak ground acceleration value of .09g is sufficiently conservative.

Applicable Law Under the Commission'a regulations, nuclear power plants must be designed to protect the public from radioactive releases that might otherwise result from an earthquake. In general terms, the regulations require an investigation of the geology of the power plant. environs and an analysis of historic O 1 l

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1neismic activity, among other factors, to assess the maximum O eartauu *e that couta re on dir de vo= tut ted to trect the site of the pcwer plant during its lifetime. The plant must be designed such that its critical safety systems remain func-tional when subjected to the maximum vibratory ground motion produced by the postulated earthquake.

Specifically with respect to assertion (a), 10 CFR Part 100, Appendix A, Section III(g) defines a capable fault as one that has exhibited one or more of the following charac-teristics:

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1 (1) Movement at or near the ground I surface at least once within the past 1

35,000 years or movement of a recurring nature within the past 500,000 years.

(2) Macro-seismicity instrumentally determined with records of sufficient pre-l cision to demonstrate a direct relationship with the fault.

f (3) A structural relationship to a

' capable fault according to characteristics (1) or (2) of this paragraph such that movement on one could be reasonably expect-ed to be accompanied by movement on the l

other.

With respect to assertion (b), 10 CFR Part 100, Appendix A states that seismic, geologic and engineering in-vestigations "shall be carried out by a review of the pertinent literature and field investigations and shall include the steps L outlined in paragraphs (a) through (c) of this section."

Paragraph (a) discusses the required investigations for obtain-ing information regarding the vibratory ground motion produced by the safe Shutdown and operating Basis Earthquake. Paragraph

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(b) outlines the required investigations for surface faulting.

These include a determination of the lithologic, stratigraphic, O

hydrologic and structural geologic conditions of the site, a determination of geologic evidence of fault offset at or near the ground surface at or near the site, and a determination of whether faults greater than 1000 feet long within 5 miles of tae site are capable faults. Paragraph (c) pertains to the investigations required for seismically induced floods and water waves. There is no requirement in 10 CFR Part 100, Appendix A, nor in 10.CFR 50.57, that strain gauge testing be performed on faults cutting basement rock as part of a seismic investigation.

With respect to assertion (c),Section III(c) and (d) of Appendix A defines Safe Shutdown Earthquake and Operating Basis Earthquake as follows:

3 (c) The " Safe Shutdown Earthquake" is

, that earthquake which is based upon an e .-

evaluation of the maximum earthquake poten-tial considering the regional and local geology and seismology and specific charac-teristics of local subsurface material.

produces It the is that earthquake which maximum vibratory ground motion for which

'certain structures, systems, and components are designed to remain functional. These structures, systems, and components are those necessary to assure:

(1) The integrity of the reactor 4 / coolant pressure boundary, (2) The capability to shut down

, the reactor and' maintain it in a safe

'" shutdown condition, or (3) The capability to prevent or mitigate the consequences of accidents O',,  !

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which could result in potential off-site exposures comparable to the O guideline exposures of this part.

The " Operating Basis Earthquake" is that(d)earthquake which, considering the regional and local geology and seismology and specific characteristics of could reasonably be local subsurface material, expected to affect the plant site during the operating life of the plant; vibratory it is that earthquake which produces the growth [ sic] motion for which those fea-tures of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public are designed to remain functional.

Section V(a)(2) of Appendix A provides that the maximum vi-bratory ground acceleration corresponding to the Operating Basis Earthquake be at least one-half that of the Safe Shutdown Earthquake.

Capable Faulting There are three sets of fault zones on or in the vicinity of the Byron site: (1) small displacement faults discovered onsite during construction excavations; (2) the Sandwich Fault Zone; and (3) the Plum River Fault Zone. All of the witnesses, i.e., Intervenor's, NRC Staff's, and Appli-cant's, agreed that the displacement faults and Sandwich Fault Zone were noncapable. (Findings 8, 14.)

The Illinois State Geological Survey (ISGS) performed a study on the Plum River Fault Zone that included detailed field mapping, well records, drill cores, seismic refraction and the development of computer-constructed base maps. (Find' O

Staff's witness, Dr. Ina B. Alterman, a ings 17 and 18.)

geologist with the Nuclear Regulatory Commission, testified that the ISGS study demonstrated (1) the Illinoian glacial till, which is at least 125,000 years old, overlying the Plum River Fault is undisturbed, (2) there is no seismicity asso-ciated with the fault zone, (3) there is no fault escarpment and (4) the regional tetonic history indicates that faulting in Illinois is not younger than 65 million years. (Finding 19.)

As a result, Dr. Alterman concluded that the Plum River Fault is clearly noncapable. (Finding 19.)

Intervenor's witness, Dr. Henry Woodard, Chairman of the Geology Department at Beloit College, testified that be-cause the ISGS study does not specifically state that the ISGS actually observed that the till overlying the Plum River Fault Zone is undisplaced, he is not convinced that the Plum River Fault is noncapable. However, Dr. Woodard stated that he is not aware of any evidence which indicates the Plum River Fault Zone is capable according to the definition in 10 CFR Part 100 nor is he aware of any evidence of a fault in Northern Illinois that displaces overlying Illinoian-age soil deposits. (Find-ings 20 and 21.) Moreover, Dr. Woodard testified that the ISGS, NRC Staff and the Applicant were justified in arriving at their conclusion regarding the age and noncapability of the i

l fault. (Finding 22.)

Dr. Alterman pointed to several aspects of the ISGS study, including lack of surface displacement and seismic-refraction work, that demonstrate that the Illinoian till over-O

lying the Plum River Fault is undisturbed. (Finding 23.)

O irg11 cant's itness, nr. ^1 n x. vonx, a Senior oeo1osist at Sargent & Lundy, the Byron station architect-engineer, also concluded from the ISGS study that there is no evidence of displacement in the overlying Illinoian till, that the last movement in the Plum River Fault Zone occurred prior to the deposition of Illinoian-age till and that the Plum River Fault Zone is noncapable. (Finding 24.) No evidence with respect to the Plum River Fault Zone was presented to demonstrate (1) that there has been movement at or near the ground surface at least once within the past 35,000 years or movement of a recurring nature within the past 500,000 years, (2) instrumentally deter-mined macroseismicity demonstrating a direct relationship with the fault or (3) a structural relationship to a capable fault.

(Findings 25-28.) ,

Strain Gauge Testing Strain gauge testing is designed to measure the strain rate, or the change in strain over a given period of l

time. (Findings 32 and 33.) Intervenor's witness testified I

that strain gauge testing could measure differential strain in the rocks on opposite sides of faults thereby providing in-formation with respect to future fault movement. (Finding 35.)

Dr. Woodard stated that because earthquakes in northern Illinois are believed to be caused by movement along faults at depth, this testing should be performed in basement rocks at depths greater than 3,500 feet. (Finding 35.) However, Dr. Woodard O

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has never used strain gauges in the field nor has he used the O type of strain eauses in the 1aboratory that cou1d be uti11 zed for down-hole measurements. (Finding 36.) Dr. Woodard is not certain that strain gauge testing has ever been performed at j depths greater than 3,500 feet nor does he know if strain gauges can even be used to measure strain at those depths.

(Finding 37.) Indeed, there is no evidence to indicate that any strain gauge testing as suggested by Dr. Woodard could be used to measure strain at a depth of 3,500 feet or greater.

(Finding 39.) Moreover, the undisturbed residual soil at least 125,000 years old overlying the Plum River and Sandwich Faults indicates that no measurable strain has occurred over this period. (Finding 41.) Therefore, even if strain gauge testing could be performed at depths greater than 3,500 feet, it would not lead to any meaningful results in a reasonable period of time. (Finding 43.) Further, in addition to the above fac-tors, a technique has not been developed for utilizing strain gauge testing to accurately predict fault capability. (Finding 44.) Finally, there is no requirement in 10 CFR Part 100, Appendix A, nor in 10 CFR 50.57, that strain gauge testing be performed.

SSE and OBE l

Dr. Woodard also questioned the sufficiency of Appli-cant's SSE peak ground acceleration value of .20g and its OBE peak ground acceleration value of .099 However, he admitted he is not a seismologist. He does not know how earthquake O

l intensity, magnitude or around acceleration parameters are used O in deriving a nuc1 ear go er 91 ant seismic design. (Finding 49.) The SSE for the Byron plant is an earthquake with a Modified Mercalli intensity of VIII and a magnitude of 5.8, which is greater than any earthquake ever recorded in the Byron area. (Findings 55-57.) The .20g peak ground acceleration value was derived from a site-specific spectrum generated from actual accelograms of earthquakes recorded at rock sites with ,

magnitudes of 5.8. This site specific spectrum was first developed for the Sequoyah Nuclear Power Plant, which is found-ed on rock, as is Byron. (Findings 58-60.) Because the Sequoyah site specific spectrum corresponds closely to the Byron seismic design basis spectrum, and because the Sequoyah spectrum was developed based on data obtained from earthquakes with magnitudes similar to the magnitude of the Byron SSE, the

.20g value selected for Byron is appropriate.

The .09g ground acceleration value for the OBE was based upon an earthquake recurrence interval calculated to be at least 200 years and as great as 2,150 years. (Finding 68.)

By any calculation, the return period for an earthquake with a

.099 peak ground acceleration is far greater than the 40-year operating life of the Byron plant. (Finding 70.) No evidence was presented refuting the testimony of Applicant's and Staff's witnesses that the SSE and OBE peak ground acceleration values for Byron are sufficiently conservative.

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i conclusion O The Board finds that with respect to the faults on or near the Byron site (1) there is no evidence of movement at or near the ground surface at least once in the past 35,000 years or movement of a recurring nature within the past 500,000 years, (2) there is no evidence of macroseismicity demonstrat-ing a direct relationship with the fault and (3) there is no evidence of a structural relationship to a capable fault.

Inasmuch as none of the characteristics of a capable fault as defined by 10 CFR Part 100, Appendix A Section III(g) are present, the Board finds these faults to be noncapable. The Board finds that there is no valid premise on which to conclude that strain gauge testing could be carried out at depths sug-gested by Intervenor and, even if it could, there is no evi-dence that strain gauge testing could measure in a reasonable time the small changes in strain that might occur around the Plum River Fault Zone or that it could accurately predict future fault movement. Accordingly, the Board finds that strain gauge testing, which is not required by the regulations, need not be performed. Finally, the Board finds that the unre-futed testimony of Applicant's and Staff's witnesses demon-strates that the SSE peak ground acceleration value of .20g and the OBE peak ground acceleration value of .09g are sufficiently conservative to fall within the ambit of 10 CFR Part 100 Appen-dix A, Section III(c) and (d).

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B. DAARE/ SAFE Contention 9(a) - Waterhammer O DAARE/ SAFE Contention 9(a) raises a concern over the possibility of a waterhammer event occurring in the feedwater bypass line at the Byron Station. Specifically, the contention addresses the question of whether or not the steam generator feedwater bypass line installed in Byron will be susceptible to bubble collapse waterhammer such as the one that occurred in the feedwater bypass line at the Krsko plant in Yugoslavia in mid-1981. Bubble collapse waterhammer is a phenomenon that can occur when a volume of steam is trapped in an enclosed area, such as a pipe, by slugs of water. Cold water in the slugs would cause the steam to condense rapidly, thereby triggering a sequence of events that could cause a waterhammer event and potentially damage the pipe and/or its supports.

Applicable Law l Before an operating license may issue, NRC must find that reasonable assurance exists such that the activities authorized by the operating license can be conducted without endangering the health and safety of the public and that such activities will be conducted in compliance with the Commis-sion's regulations. 10 C.F.R. i 50.57(a)(3). One of the means for satisfying the requirements imposed by this regulation is to comply with the General Design Criteria for light water nuclear power reactors set forth in Appendix A to 10 C.F.R. Part 50. These criteria establish principal design require-O

i ments for determining reactor safety. With regard to the

, potential for the occurrence of waterhammer in the feedwater i

bypass line, General Design Criterion 4 " Environmental and missile design bases" states in pertinent part:

Structures, systems, and components impor-tant to safety shall be . . . appropriately j

protected against dynamic effects . . .

i that may result from equipment failures and

! from events and conditions outside the nuclear power unit. (emphasis added)

Waterhammer events such as occurred at Krsko, create a " dynamic effect" against which systems importa.nt to safety are to be

" appropriately protected." The feedwater bypass line is "im-portant to safety" because this line serves as part of the Auxiliary Feedwater System which is designed to facilitate safe plant shutdown in the event of a loss of heat sink accident.

(Finding 75). The issue before us then is whether the feed-water bypass lines to each of the eight steam generators in the Byron plant (four in each of the two units) are appropriately protected against the dynamic effects of a Krsko-type water-hammer event.

DAARE/ SAFE Contention 9(a) suggests that there should be a demonstration that a Krsko-type waterhammer event will 1

i never occur at Byron. (Finding 73.) In view of the standard ,

set forth in the regulations, the Contention sets forth an overly strict acceptance criterion. The Commission regulations I

do not require a guarantee that a Krsko-type waterhammer event

will not occur at Byron but rather that there is reasonable.

assurance of no danger to the public health and safety. Such

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assurance can be found by compliance with General Design Cri-O terien 4 which requires thee the systems imgortant to sefety that would be affected by a Krsko-type waterhammer event are appropriately protected against its effects.

Byron and Krsko Steam Generator Design Byron and Krsko share the same model of steam genera-tors, the Westinghouse Model D counterflow preheater type steam generator. They also have a common design of feedwater systems which serve these steam generators. There are basically three feedwater systems: the Main Feedwater System, the Feedwater Bypass System, and the Auxiliary Feedwater System. The last two systems introduce feedwater into the steam generator through an auxiliary nozzle located in the upper shell of the vessel. The Feedwater Bypass System was specifically designed I

to avoid the possibility of occurrence of waterhammer events in the preheater section of the steam generator which is located inside the lower shell of the vessel adjacent to the main feedwater nozzle. The Feedwater Bypass System accomplishes this purpose by diverting cold feedwater to the auxiliary nozzle through the feedwater bypass line. (Finding 75.) The waterhammer event considered here, however, does not concern the preheater section. Rather, the issue before us concerns the potential for waterhammer events in the Feedwater Bypass System, specifically, the feedwater bypass line.

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The Krsko Waterhammer Event O The Applicant's expert witness on the waterhammer phenomenon reconstructed the Krsko waterhammer event based on data obtained during the ensuing engineering investigations of the event. The event was a " bubble collapse" type waterhammer which occurred during testing of the Auxiliary Feedwater System pumps which took place in July of 1981, as part of Hot Func-tional Testing. The first indications of the waterhammer event l

were the discoveries in early August of 1981, of blistering of the paint on the pipes of the Auxiliary Feedwater System, slight damage to several pipe hangers, and a quarter-inch bulge on a six to eight inch section of the bypass piping near the secondary shield wall. (Findings 77, 78.)

The evidence gathered during the investigation of the l damage at Krsko provided a basis for a reconstruction of the event as related by Applicant's expert witness. Blistered paint on the piping of the A ?xiliary Feedwater System evidenced that the check valves in that system were leaking thereby permitting a reverse flow of hot water and/or steam from the steam generator vessel. Concurrently, the water level in the steam generator fell below the internal discharge of the auxil-iary nozzle, thus allowing steam to flow out through the auxil-iary nozzle and into the Feedwater Bypass System piping. With steam present in the bypass piping, the Auxiliary Feedwater System motor driven pumps were started as part of Hot Func-tional Testing, thereby introducing cold water into the bypass O

piping. The contact with the cold water caused the steam to rapidly condense thereby creating a local reduction in pressure.

Water slugs driven by the pressure differential into the void left by the condensed steam collided and consequently pro-pogated forceful pressure waves throughout the feedwater bypass piping and support hangers. (Finding 79.)

Despite the damage done to the piping and hangers at Krsko from the waterhammer event, the capability of the Auxil-iary Feedwater System and Feedwater Bypass System to perform their intended functions was not impaired. (Finding 78.)

The Westinghouse Recommendations Westinghouse has made four recommendations to Appli-cant to avoid a Krsko-type waterhammer event in the Feedwater Bypass System at the Byron Station. First, temperature sensors should be installed on the bypass piping close to the auxiliary nozzle to detect backleakage of hot water or steam. Second, in the event backleakage is detected, the bypass piping should be slowly refilled or the plant brought to a cold shutdown condi-tion. An analytical study performed by the Westinghouse R & D Center shows that the safe refilling flow rate is in the range of 15-123 gpm. (To be conservative, Westinghouse recommended the rate of 15 gpm or as close to it as can be provided.)

Third, Westinghouse recommended that the steam generator water level be maintained above the auxiliary nozzle discharge pipe as much as possible. Fourth, the check valves of the Auxiliary O

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Feedwater System should be maintained to minimize backleakage.

O (Finding 80.)

Applicant's Assistant Superintendent of Operations at Byron testified that the Westinghouse recommendations are being implemented. Temperature sensors will be installed on the feedwater bypass p.cing adjacent to the auxiliary feedwater nozzle on each of the steam generators at the Byron Station.

These sensors will feed information to the plant process com-puter and will initiate an alarm when an abnormally high tem-perature is detected in the bypass piping. Applicant is devel-oping procedures which will instruct the reactor operator how to properly purge the bypass piping of steam or hot water when it is detected by introducing feedwater into the bypass piping at a flow rate at or near 15 gpm. The water level in the steam generator will be maintained above the auxiliary nozzle dis-charge pipe under normal operating conditions. The Maintenance Department at Byron Station will establish a regular schedule for testing the check valves for backleakage. (Finding 85. )

Prior to power operation, Applicant will test the ability of the feedwater systems to prevent backleakage and to introduce a flow at or near 15 gpm. (Finding 87.)

The NRC Staff witness, a Senior Task Manager in the Generic Issues Branch of the NRC Office of Nuclear Reactor Regulation who is the task manager for the Unresolved Safety Issue (USI) A-1 "Waterhammer", conducted an evaluation of the Krsko event, the Krsko and Byron plant designs, and the Westing-O

He concluded that the Westinghouse house recommendations.

O recommendetions constitute good engineering design practices ,

and prudent operating procedures. The NRC Staff witness stated that if the recommended corrective measures had been in place at Krsko, the waterhammer event likely would have not occurred '

and that if the recommendations are followed, a Krsko-type waterhammer event should not occur at Byron. (Finding 83. )

In addition to the soundness of the Westinghouse recommendations, it was noted by Applicant's witnesses that the occurrence of a Krsko-type waterhammer event during power operation is virtually impossible since a constant flow is maintained through the auxiliary nozzle during all phases of power operation, thus precluding the possibility of backleak-age. (Finding 81.) Further, there are two check valves on each flow path by which backleakage into the Auxiliary Feed-water System could occur, thus providing redundant protection against backleakage. (Finding 82.)

As noted previously, the Krsko waterhammer event did not compromise the ability of the affected systems to function.

The NRC Staff witness testified that similarly, there would be no consequences if a Krsko-type waterhammer event of the same magnitude occurred at Byron. (Finding 88.)

Conclusion Based on the uncontroverted evidence in the record, l

the Board finds that the Applicant's implementation of the l

O

recommendations at Byron provide appropriate Westinghouse a-protection for the steam generator feedwater bypass line .

gainst the dynamic effects of a Krsko-type waterhammer event.

The unlikelihood of such an occurrence plus the limited con-sequences of such an event lead us to conclude that there is no I significant health and safety concern. (Findings 78, 83, 84, 88, 89.)

C. League Contentions 42, 111, and 112 - ALARA In general, Contentions 42, 111 and 112 assert that Byron Station cannot be operated so as to maintain radiation ex-posures to werkers as low as is reasonably achievable ("ALARA").

Moreover, the contentions assert that even if the Applicant maintains occupational exposures within the limits set by 10 C.F.R. Part 20, new information on the effects of low-level radiation exposures shows that such limits will not ensure worker safety. Because of this, we are told that we cannot make the findings required by NEPA and the Commission's regula-tions.

Although the specific wording of the three conten-tions encompasses numerous issues, the League presented evi-dence on only a few topics. Its primary area of concern ap-peared to be whether Applicant has accurately assessed the potential risks from occupational exposure to radiation. The remaining issues the League raised relate to the general abil-l ity of Applicant's ALARA program to keep occupational radiation doses ALARA and to specific concerns about radiation exposures O

and protection. These issues include: (1) whether Applicant's O ao i try proer ror =oattorias r ai tioa no ure to worker-l is sufficient to maintain doses ALARA; (2) whether Applicant's program for monitoring radiation levels inside the plant will provide accurate results; (3) whether Applicant's procedures for maintaining occupational exposure records are sufficient to maintain doses ALARA; (4) whether the size and training of Applicant's health-physics staff is sufficient; (5) whether workers at the station, including contract or temporary workers who are not Applicant employees, are adequately trained in how to keep radiation doses ALARA; (6) whether Applicant's policies on radiation exposure to declared pregnant women will adequate-ly protect the fetus; (7) whether the risk of possible indus-trial sabotage by anyone, especially a contract worker, is suf-ficiently small so as to maintain doses ALARA; and (8) whether the design bases of Byron Station and, more specifically, its steam generators, include features for reducing occupational radiation exposure. The Applicant presented testimony on all of these issues.

Applicable Law Part 20 of the Commission's regulations, 10 C.F.R. Part 20, provides two basic mechanisms for assuring that the 1

I health of workers in nuclear plants is not endangered by expo-sure to excessive doses of radiation. First, Part 20 contains i specific standards limiting permissible radiation doses and dose rates for individual workers. Compliance with these O

limitations is insured by specific requirements (contained in O Parts 19 and 20) that workers be equipped with personnel radia-tion monitoring devices, that individual records on worker ex-posures be maintained, that sources of radiation in the plant be evaluated periodically, and that workers be trained in radia-tion protection techniques. Second, Part 20 provides that in addition to these specific requirements, licensees "should ...

make every reasonable effort to maintain radiation exposures ...

as low as is reasonably achievable." 10 C.F.R. $ 20.1(c).

As we read the contentions, they do not challenge Applicant's ability to maintain exposures of individual workers within the permissible limits set forth in the regulations.

Rather, the contentions assert that even if Applicant maintains occupational exposures at permissible levels, such exposures are unsafe, as demonstrated by new scientific information. We may not consider this allegation here because Commission regu-lations are not subject to attack in licensing proceedings except under special circumstances, which no party has alleged exist here. 10 C.F.R. 5 2.758. Two of the contentions also refer to NEPA, however, and we take this as implying the argu-ment that if low doses of radiation are more hazardous than generally believed, the cost-benefit balance struck in the NRC Staff's Final Environmental Statement ("FES") may be incorrect.

Accordingly, we will consider the League's assertions as chal-lenges to the adequacy of the FES. See 10 C.F.R. 5 51.52(a)-

(b).

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The contentions do contain general challenges to O c-guance with severa1 of the other specific Aggucant's requirements of Parts 20, 19, and 73, such as radiation moni-l toring, record keeping, personnel training and plant security.

I These requirements are set out below. In addition, the conten-tions challenge the Applicant's ability to comply with the general principle of maintaining occupational radiation expo-sures ALARA. Because 10 C.F.R. 5 20.1(c) states only that licensees "should" attempt to keep radiation doses ALARA, we find it necessary to comment on the possibility that absent a condition imposed on the operating license Applicant need not establish and follow a program to do so. Despite the regula-tion's use of the word "should" in place of the mandatory "shall", we conclude that under well established Commission practice Applicant would not be permitted to operate a nuclear power plant without an established program for maintaining radiation exposures ALARA. In arriving at this conclusion, the Board examined Regulatory Guide 8.8, Revision 3, dated June 1978, entitled "Information Relevant to Ensuring that Occupa-tional Radiation Exposures at Nuclear Power Stations will be As Low As is Reasonably Achievable" (NRC Staff Testimony, Attach-ment C, ff. Tr. 1883), and Regulatory Guide 8.10, Revision 1-R, dated September 1975, entitled " Operating Philosophy for Main-taining Occupational Radiation Exposures As Low As is Reason-ably Achievable" (NRC Staff Testimony, Attachment D, ff. Tr.

1883). Both documents militate against a permissive inter-pretation of the word "should" in the regulation.

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Regulatory Guide 8.8 states that its purpose is to provide information to licensees to enable them "to meet the criterion that exposures of station personnel to radiation" will be ALARA. Reg. Guide 8.8, p. 8.8-1 (emphasis added, footnote omitted). Regulatory Guide 8.10 describes "a general operating philosophy acceptable to the NRC staff as a necessary basis for a program of maintaining occupational exposures to radiation" ALARA. Reg. Guide 8.10, p. 8.10-1 (emphasis added).

The regulatory guides thus presume the necessity of an ALARA program. Although regulatory guides are not regulations, they set out methods of implementing NRC Regulations which are acceptable to the NRC Staff. Because of these regulatory guides, and especially because every licensed nuclear power plant has an ALARA program, the Board finds it inconceivable that the Commission would issue a operating license for a plant without an acceptable ALARA program.

In addition, the scope and magnitude of Applicant's company-wide ALARA program provides further assurance to the Board that a program for maintaining occupational radiation doses ALARA will be carried out at Byron Station. During the hearing on these contentions, Applicant presented testimony by its corporate lead health physics - technical services engineer and the head of Byron Station's health physics department.

l These men testified under oath that both Applicant and Byron Station have ALARA programs. Indeed, both witnesses provided detailed descriptions of the programs and explained how they. 1 would operate. (Finding 111.) For these reasons, the Board j

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does not find it necessary to impose a license condition that O Applicant carry out its ALARA pregram at Byron Station. Well settled Commission practice, Applicant's commitment to its ALARA program and the oath of Applicant's witnesses overwhelm-ingly convince the Board that Applicant cannot and will not operate Byron Station without implementing and abiding by its ALARA program.

The specific requirements of Part 20, Part 19 and Part 73 applicable to issues raised by these contentions are as follows:

Permissible Occupational Doses. No individual may receive an occupational dose of more than 1.25 rems per calen-dar quarter, or more than 3 rems per quarter if the licensee has determined that his total accumulated occupational dose will not exceed a defined limit. 10 C.F.R. $ 20.101.

Dosimetry Program. To determine individual occupa-tional doses, a licensee must equip every person who enters a defined high radiation area with personnel monitoring equip-l ment, such as film badges, pocket ionization chambers, pocket l

dosimeters, or film rings. 10 C.F.R. I 20.202.

1 In-Plant Monitoring. A licensee must evaluate the radioactive material in a plant, including levels of radiation, to insure compliance with the requirements of Part 20 and to l

l determine the extent of any radiation hazard that may exist.

10 C.F.R. 5 20.201.

Record Maintenance. A licensee must maintain indivi-dual exposure records on Form NRC-5 (NRC Staff Prepared Testi-O

1 l

1 mony, Attachment G, ff. Tr. 1883); the information includes the O dose to the individua1 for a given period by type of radiation, .

the method of monitoring and the individual's accumulated I In addition, Regulatory lifetime dose. 10 C.F.R. $ 20.401(a). J Guide 8.7, "Occupr. tion Radiation Exposure Records System",

l I ,

describes an acceptable records system and in particular the recording of doses by task so the licensee can obtain feedback for ALARA reviews. While not binding on the Board, this gives us guidance on a program's adequacy.

Health Physics Staff. The regulations contain no requirements regarding a health physics staff. The NRC Staff has developed guidelines regarding the appropriate size and 1.8, training of health physics staffs in Regulatory Guide

" Personnel Selection and Training." ANSI 18.1, " Selection and Training of Nuclear Power Plant Personnel," contains additional guidelines.

Under 10 C.F.R. 5 19.12, every person Training.

including contract working in a radiation controlled area, workers or temporary employees, must be instructed about the potential health risks associated with exposure to radiation, procedures to minimize exposure, and the use of protective devices.

Declared Pregnant Women. Regulatory Guide 8.13, Rev.

1, dated November 1975, provides guidelines on instruction to including the be given to women on prenatal radir. tion exposure, recommendation of the National Council on Radiation Protection and Measurements that the maximum permissible dose to the fetus O

during the gestation period not exceed 0.5 rem. Reg. Guide O 8.13, p. e.13-1.

Temporary Employees. When any contract worker could receive, in one calendar quarter, an occupational dose in excess of 25 percent of the applicable standards, the licensee must require him, before entering a restricted area, to dis-clos in a signed statement his prior occupational dose, if any, during the current calendar quarter. 10 C.F.R. 5 20.102.

Under 10 C.F.R. 5 20.101, a licensee must obtain a certifica-tion on Form NRC-4 (NRC Staff Prepared Testimony, Attachment B, ff. Tr. 1883) or an equivalent signed statement before permit-ting an individual to enter a restricted area where he could receive doses defined in the regulation.

Industrial sabotage. To protect against the possi-bility of industrial sabotage by either temporary workers or employees, a licensee must have a security plan that meets the requirements of 10 C.F.R. 5 73.55(b)-(h). These regulations include provisions on restricting access, screening all workers to be given unescorted access into the plant, searching vehi-cles, individuals and deliveries, and protecting vital equip-ment.

It is against the above regulations, statutes, and guidelines that the Board must review the Applicant's evidence of its ability to maintain occupational doses as low as is reasonably achievable.

O I

Permissible Occupational Exposures O Cancer Risk Estimate The League claims that Applicant has inaccurately assessed the effects of occupational radiation exposure at Byron Station. As noted above, we are considering this as a challenge to the assessment of such effects in the NRC Staff's FES. Specifically, the League's expert witness claims that the risk of cancer and the risk of genetic effects from exposure to low levels of radiation, such as those a nuclear power plant worker might be exposed to, is much higher than accepted by national and international standard-setting committees. Conse-quently, the witness contends the collective radiation dose from Byron Station will be too high because individual occupa-tional exposure limits do not insure worker safety. (Find-ing 96.) To support his claim, the witness relied on the results of several recent studies and on his cwn manual ploting of the data from three studies. (Findings 99, 103.)

The scientific evidence against this claim, however, is overwhelming. The Applicant's witness pointed out that all of the recent studies the League's witness relied upon have serious methodological or statistical deficiencies. (Findings 100, 101, 102.) In additon, the League's witness admitted that his own determination of the cancer risk estimate would be significantly lower if he had excluded the data from one of the three studies. Both the Applicant's and the NRC Staff's wit '

nesses criticized that study and do not consider its results

I reliable. (Finding 104.) Neither of those witnesses, nor any O standard-setting committee, accept the League's witnees's estimation of the risk of cancer from radiation exposure. On j the contrary, both those witnesses and the committees conclude that the effects of radiation exposure are approximately the same as those set forth in the FES. (Findings 105, 106, 108, 109.) Because the vast amount of scientific evidence and the opinions of the scientific community do not support the Lea-gue's assertions, the Board finds that the FES has adequately assessed the effects of radiation exposure to workers.

ALARA Considerations In addition to his testimony on radiation protection standards, the League's witness expressed concern about numer-ous aspects of Applicant's ALARA program. Although the Board admitted this testimony, we find ourselves unable to give much weight to it. The witness' testimony on cross-exmnination gen-erally revealed that he had little or no knowledge of Appli-cant's practices in the areas in which he raised questions, and we do not think that any of his testimony poses a serious challenge to the adequacy of Applicant's ALARA program. (Find-ing 95.) Of course, the burden of proof under the contentions is the Applicant's. The Applicant introduced extremely compre-hensive evidence detailing its committment to pursue the ALARA principle at Byron Station. The Applicant's witnesses testi-fied as to the involvement of top corporate and station manage-ment in the ALARA programs and the priority given to ALARA O

l considerations. Both witnesses play important roles themselves I O in overseeins these pro 9 tams and the Board not on1r found them to be highly intelligent and credible witnesses, but also believes that they demonstrated genuine dedication to pursuing ALARA goals. (Finding 111.) We review below the specific ALARA issues raised.

l Dosimetry The League contends that Applicant's dosimetry pro-gram is inadequate to monitor a worker's exposure to radiation.

In compliance with 10 C.F.R. I 20.202, however, Byron Station currently plans for each worker to carry a film badge, a pocket ionization chamber, and various other dosimeters as needed. In addition, CR-39 will be used to determine neutron dosimetry.

(Findings 113, 114.) Through these devices, Applicant can monitor a worker's exposure to neutrons and gamma and beta radiation. (Findings 112, 113, 114.) Applicant also has procedures designed to decrease radiation exposure and to determine the amount of a worker's internal body burden.

(Findings 118, 119.) Based on the evidence, the Board finds that Applicant's dosimetry program is sufficient to provide an accurate assessment of the dose equivalent received by a worker at Byron Station and to maintain doses ALARA.

In-Plant Monitoring Byron Station will have over 200 in-plant area and air monitoring instruments. The NRC Staff has reviewed Appli- l l

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_ _ _ _ __ __ l

cant's in-plant monitoring program and found it to meet the O requirements of 10 C.F.R. Part 20. (Finding 122.) The League P asserted that this program is inadequate, but presented no evidence to support that claim. Indeed, the League's witness admitted he did not know how many monitors the station had. I (Finding 121.) We find that Applicant's in-plant monitoring program is sufficient to maintain occupational doses ALARA. l 1

Record Flaintenance Applicant maintains records which contain all the information required by 10 C.F.R. 5 20.401(a). Applicant keeps records of all radiation exposures to individuals and makes them readily available to any worker who wishes to know the (Finding 123.) In addition, Applicant amount of his exposure.

has developed a records program that will provide feedback for reviews of the ALARA program, and has imposed administrative daily dose limits on workers. (Findings 124, 125.) The Board finds that Applicant's system of maintaining records is suffi-cient to keep radiation exposures to workers ALARA and satis-fies the regulatory requirements, l

l Health Physics Staff Byron Station's health physics department will con-sist of 15 full-time employees and 28 part-time employees who will spend approximately 70 percent of their time on health physics responsibilities. These employees will recieve appro-(Finding 127.) The League's witness stated priate training.

1 l

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um.

that the Byron health physics staff should consist of approxi-l O mate 1r s0 peorie. hut he offered no hasis for this conc 1usion. '

(Finding 128.) The NRC Staff has reviewed the size of Byron's health physics department and found it to be acceptable and approximately the same as those at other nuclear power sta-(Finding 128.) The Board finds that Byron Station's I

tions.

health physics staff will be sufficiently large and well train- ,

ed t,o carry out all the radiation protection services needed at -

the plant.

Training t

All workers at Dyron, including contract workers,- -

must take a standardized. training course and an annual retrain-ing course on the importance of maintaining doses ALARA and how- x to achieve that goal. Workers with radiation protection, oper-ational, or maintenance responsibilities will receive addi- .

tional training. (Finding 130.) The NRC Staff has evaluated Applicant's training program and determined that it will ade- s (Finding 131.) Based on the quately maintain exposures ALARA.

evidence, the Board finds that Applicant's training program is adequate.

Declared Pregnant Women Because of the potential risks to a fetus from radia-l tion exposure, the League asserts that Applicant should not_use women of childbearing age for jobs that may expose them' to

,j radiation. Although the Board questions the legality of such a l O j l

l l

l

restriction, we find that Applicant has taken sufficient mea-sures to protect both the woman and the fetus. Applicant educates women on the risks of radiation exposure to a fetus and limits a declared pregnant woman's exposure to 500 'nillirem over her gestation period. (Findings 132, 133.) We note that h

s this dose restriction is even more conservative than the recom-s.-

mendation cited in Regulatory Guide 8.13, Revision 1, dated November 1975, that the radiation exposure to the fetus, rather

+,

,; than the woman, not exceed 500 millirem. Indeed, the League's x 4 4 ' witness was only concerned with the possibility that a declared pregnant woman could exceed her 500 millirem limit through a

pre-existing internal body burden caused by exposure to a x

long-lived radionuclide. (Finding 134.) The probability that any worker could receive even a measurable uptake of that type of radionuclide, however, is too remote to change our conclu-sion that adequate protections exist. (Finding 135.)

s Temporary Employees

s The League contends that temporary workers are more likely than station employees to exceed their exposure dose

, limits. Each contract worker, however, is required to complete Form NRC-4, which details his occupational dose history, and s

the worker's previous employers are contacted to verify his Although the Board occupational dose. (Findings 137, 138.)

realizes that a worker could purposely fail to list prior employers, no other method is available to obtain the worker's N employment history. (Finding 138.) The Board t.herefore finds O

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, /'

..' t j that Applicant's procedures regardin'g temporary . workers are sufficient to maintain' doses ALARA.

Industrial Sabotage 1 ,-

>f Tbe League also asserted that the use of temporary workers increases the likelihoo,d of industrial sabotage. As part of its ibdustrial security plan, Applicant has established screeningprodeduresforallemployeesandalltemporaryworkers whoneedunescortedaccesstoBhronStation. Contractors must have their screening procedures approved by Applicant and must submit to periodic audits of their phogram. (Findings 139, 140.) In addition, Applicant has established safeguards that restrict a person's access to the plant and to various loca-tions within the plant. (Finding 141.) We find that these

,

procedures meet the requirements of 10 C.F.R. 5 73.55(b)-(h) and providb the necessary measures to protect against indus-trial sabotage by eitheir an employee or a contract worker.

Design of Byron Station The League asserts that high radiation doses at Byron Station are unavoidable because of the design of the station and its steam generators, but the League failed to present any The design credible evidence of this claim. (Finding 142.)

basis of Byron makes use of shielding and distance from radio-active material, removal of such material and reduction of time spent in a radiation field. The Byron FSAR details the speci-l l

fic features of the station which were designed to implement l0

(

l l ALARA principles. (Findings 143, 144.) The steam generators also have extensive design features that reduce radiation exposure. (Findings 147, 148.) We find that the radiation protection measures incorporated into the design of the plant and of its steam generators provide reasonable assurance that they will operate in conformity with the guideline that occupa-tional doses be kept ALARA.

Conclusion The Board finds the testimony of the Applicant's witnesses and the NRC Staff's witnesses on Applicant's ALARA program consistent and unrefuted. The League's sole witness presented competent evidence only on the cancer risk estimate issue. We have reviewed that testimony in depth and have l

concluded that the weight of the scientific evidence estab-lishes that the FES has adequately assessed the risks of occu-pational radiation exposure.

The Board is impressed by Applicant's committment to l its ALARA program, both at the corporate level and at Byron Station itself. We have reviewed the detailed descriptions of every aspect of this program and find it more than adequate to maintain occupational radiation exposures ALARA.

1 0

1 FINDINGS OF FACT O

III. Contentions A. League Contention 106 - Seismology

1. Pursuant to stipulation, original contention 106 was altered. Alternate contention 106, as litigated, reads as follows:

There exist serious seismic related

' site problems discovered subsequent to the construction permit herein which indicates that the seismic design for Byron is not such that there exists assurance that these problems are adequately resolved in accord-ance with applicable regulations, including but not limited to 10 CFR 50.57 (a)(3)(i),

50.57 (a)(6) and 10 CFR Part 100, Appendix A. Specifically, the Rockford League of Women Voters contends that due to the lack of reliable information regarding the causes of earthquakes which have been experienced in northern Illinois, Edison should be required to perform strain gauge tests on faults cutting basement rock located in the northern Illinois region where earthquakes of modified Mercalli VII or greater intensity are expected to occur.

Further, recent evidence from the central portion of the United States shows that neither the Byron designated safe shut down earthquake peak ground acceleration value of 0.20(g) nor the operating basis earth-quake peak ground acceleration value of 0.09(g) are sufficiently conservative.

Ground acceleration sigr.ificantly greater than both of these values are possible at the Byron site. In addition, it is not known if the recently discovered Plum River Fault is a capable fault. This fault is known to approach the Byron site within 5.3 miles and may even be closer if the fault extends further to the east.

2. Applicant presented the testimony of two wit-nesses, Mr. Alan K. Yonk, a Senior Geologist at Sargent & Lundy O

(the Byron station architect-engineer), and Dr. Anand K. Singh, a structural engine'er and Assistant Division Head of the Dyna-mic Analysis Section of Sargent & Lundy. Witnesses for the NRC were Dr. Ina B. Alterman, a staff geoligist in the geosciences branch in the Office of Nuclear Reactor Regulation of the NRC, and Dr. Robert L. Rothman, a seismologist in the geosciences branch in the Office of Nuclear Reactor Regulation of the NRC.

The League presented the testimony of one witness, Dr. Henry H.

Woodard, Chairman of the Geology Department at Beloit College in Wisconsin.

3. The Byron plant foundations extend into the upper bedrock units which are part of che Ordovician-age Galena Group dolomites. These dolomites are jointed and fractured in the upper formations of the Galena Group. Some solution acti-vity has taken place among the joints. Some of the joints have minor offsets, from one to six inches, which technically quali-fy them as faults. (Yonk, Applicant Prepared Testimony, at 3, l ff. Tr. 478.)

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4. On behalf of Applicant, Sargent & Lundy and Dames & Moore, a geotechnical and environmental consulting firm, performed excavation mapping and a geotechnical investi-gation of these minor displacement faults. (Yonk, Applicant Prepared Testimony, at 5, ff. Tr. 478.)

5. The Illinois State Geological Survey (ISGS) also investigated the small displacement faults. (Yonk, Applicant Prepared Testimony, at 5, ff. Tr. 478.) The 'SGS is a well recognized body of experts in the geology of illinois, and it

is customary for geologists to rely upon the work and geologic O investigations conducted by the ISGS. (Yonk, Tr. 436; Alterman, Tr. 835-836.) Indeed, Intervenor's witness characterized the ISGS as the foremost state geology group in the United States.

(Woodard, Tr. 582.)

6. The investigations revealed that these faults are overlain by undisplaced interglacial residual soil and glacial till. (Alterman, NRC Staff Prepared Testimony, at 3, ff. Tr. 753.) This overlying material was deposited at least 125,000 years ago. (Alterman, NRC Staff Prepared Testimony, at 8, ff. Tr. 753; Yonk, Tr. 461.)

7. There is no evidence of any movement on these faults within the past 35,000 years, or of any movement of a recurring nature on these faults within the past 500,000 years.

(Yonk, Applicant Prepared Testimony, at 6-7, ff. Tr. 478.)

There is no instrumentally determined evidence of macro-seismi-city associated with these faults. (Yonk, Applicant Prepared Testimony, at 7, ff. Tr. 478.)

8. The NRC Staff and Intervenor witnesses did not dispute the noncapability of the minor displacement faults.

(Alterman, NRC Staff Prepared Testimony, at 4, ff. Tr. 753; Woodard, Tr. 567.)

9. The Board finds the minor displacement faults at the site to be noncapable.

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10. The western limit of the Sandwich Fault Zone is approximately 6 miles southwest of the Byron site. (Yonk, I

Applicant Prepared Testimony, at 5, ff. Tr. 478.)

11. Since the construction permit stage, the ISGS has performed a detailed investigation of the Sandwich Fault Zone to determine its extent, amount of offset, age, and nature of faulting. (Alterman, NRC Staff Prepared Testimony, at 2, ff. Tr. 753.)

12. The investigation confirmed that no glacial material nor subjacent residual soil was offset anywhere along the entire length of the fault wherever the young material was 1

observed. (Alterman, NRC Staff Prepared Testimony, at 2, ff.

Tr. 753; Yonk, Applicant Prepared Testimony, at 6, ff. Tr.

478.)

13. As with the small displacement faults, there is no evidence of movement on the Sandwich fault within the past 35,000 years; no evidence of movement of a recurring nature within the past 500,000 years; no evidence of instrumentally determined macro-seismicity associated with this fault, and no evidence that the fault is structurally related to a capable fault. (Yonk, Applicant Prepared Testimony, at 6-8, ff. Tr.

478.)

14. The NRC Staff and Intervenor witnesses did not dispute the noncapability of the Sandwich Fault Zone. (Alter-man, NRC Staff Prepared Testimony, at 4, ff. Tr. 755; Woodard, Tr. 567.)

15. The Board finds the Sandwich Fault Zone to be noncapable.

16. The eastern end of the Plum River Fault Zone is approximately 5.3 miles northwest of the Byron plant site.

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,

l (Alterman, NRC Staff Prepared Testimony, at 3, ff. Tr. 753; i Alterman, Tr. 812; Yonk, Applicant Required Testimony, at 5, ff. Tr. 478.)

17. The ISGS performed a detailed investigation documented in Circular 491 that indicated that the Plum River I

Fault Zone contains several hundred feet of offset. (Alterman, NRC Staff Prepared Testimony, at 3, ff. Tr. 753.)

18. The ISGS investigation included detailed field mapping, well records, drill cores, seismic refraction work, i and the development of computer-constructed base maps from the subsurface information. (Alterman, NRC Staff Prepared Testi-mony, at 3, ff. Tr. 753. )

19. Dr. Alterman concluded, based upon the ISGS study, that because (1) the Illinoian glacial till over the Plum River Fault Zone is undisturbed, (2) there is no seismi-city associated with the fault zone, (3) there is no fault escarpment and (4) the regional tectonic history indicates that faulting in Illinois is not younger than 65 million years, the Plum River Fault Zone is noncapable. (Alterman, Tr. 788, 794, 818.)

20. Intervenor's witness, Dr. Woodard, stated that although no evidence indicates that the Plum River Fault Zone is capable according to the definition in 10 CFR Part 100 (Woodard, Tr. 564), the ISGS study does not convince him that the Plum River Fault is noncapable. (Woodard, Tr. 565, 568.)

21. Dr. Woodard's conclusion is based on the absence of a stated observation in Circular 491 that the Illinoian till l

_ _ , . _ _ .n. , n . - - -- , ~ , - - - - - - - - , ~ ~ - - ~ - - - - ' - ^ - ~

. over the Plum River Fault Zone is not displaced. (Woodard, Tr.

565-568.) Dr. Woodard, however, is unaware of any evidence of a fault in northern Illinois that displaces overlying Illinoian-age soil deposits. (Woodard, Tr. 560.)

22. Dr. Woodard does not believe that the Illinois State Geological Survey conclusions regarding the age of the fault are unsupportable. According to Dr. Woodard, the authors of the study used wise professional judgment in interpreting the data. (Woodard, Tr. 710.) However, Dr. Woodard believes that additional evidence regarding the lack of displacement of overlying glacial deposits is necessary before he could con-

] clude that the Plum River Fault Zone is noncapable. (Woodard,

Tr. 565, 567.)

23. Dr. Alterman concluded that observations of undisplaced Illinoian till are implicit in the ISGS study for several reasons. First, the failure of the surface material to give the ISGS any clue as to the presence of a fault, which was originally thought to be an anticlincial structure, indicates that the surface material has not been disturbed. (Alterman, Tr. 762.) Second, the fact that the ISGS had to conduct drill-ing operations in two locations through the glacial deposits down into the bedrock in crder to locate the fault indicates that the surface material is undisturbed. (Alterman, Tr. 763-765, 831-832.) Third, Dr. Alterman has been told by members of the ISGS, including Mr. Dennis Kolata, one of the authors of the study, that they have never observed any disturbance in the till overlying faults in northern Illinois. (Alterman, Tr.

J

764, 830.) Finally, Dr. Alterman has received letters from the ISGS reaffirming the position that displaced till has never been found overlying any of the faults in northern Illinois.

(Alterman, Tr. 764, 787.)

24. Mr. Yonk also concluded from the ISGS study that there is no evidence of displacement in the overlying Illinoian ,

till, that the last movement in the Plum River Fault Zone occurred prior to the deposition of Illinoian age till, and that the Plum River Fault Zone is noncapable. (Yonk, Applicant Prepared Testimony, 416-8, ff. Tr. 478.)

25. The Board finds the investigations performed by the ISGS developed sufficient data from which to conclude that there has not been any movement at or near the ground surface of the Plum River Fault Zone within the past 35,000 years.

26. There is no evidence of any movement of a recur-ring nature on the Plum River Fault Zone within the past 500,000 years. (Yonk, Applicant Prepared Testimony, at 7, ff. Tr.

478.)

27. There is no evidence of instrumentally deter-mined macroseismicity to demonstrate a direct relationship with the Plum River Fault. (Woodard, Tr. 561; Yonk, Applicant Prepared Testimony, at 7, ff. Tr. 478; Rothman, NRC Staff Prepared Testimony, at 2, ff. Tr. 760.)

28. There is no evidence demonstrating that the Plum River Fault Zone is structurally related to a capable fault.

(Woodard, Tr. 562; Yonk, Applicant Prepared Testimony, at 7-8, ff. Tr. 478.)

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29. Accordingly, the Board finds the Plum River Fault Zone to be noncapable.

30. It is believed that earthquakes that have been experienced in northern Illinois may have resulted from strain released in basement rocks. (Woodard, Tr. 703.)

31. Strain is the response of a body to the applica-tion of stress. Stress is the force applied to an object per unit area. (Alterman, Tr. 773, Woodard, Tr. 530.)

32. Strain rate is the percent change in a given period of time. (Woodard, Tr. 611-612; Alterman, NRC Staff Prepared Testimony, at 6, ff. Tr. 753.)

33. Strain gauges are designed to measure the strain rate along faults. (Alterman, NRC Staff Prepared Testimony, at 6, ff. Tr. 753.)

34. Although there are several sophisticated instru-ments, the basic idea of the strain gauge is bolting a wire of known length across a zone that is suspected of being strained measurably, and measuring the change in the length or the straightness of the wire over a given period of time. (Alter-man, NRC Staff Prepared Testimony, at 6, ff. Tr. 153.) i

35. Dr. Woodard testified that strain gauge testing of basement rock in northern Illinois would yield information on the differential strains in the rocks on opposite sides of faults, the release of which causes fault motion. (Woodard, Tr. 621-622.) According to Dr. Woodard, this testing needs to be performed on rock at depths greater than 3500 feet. (Woodard, Tr. 584.)

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36. However, Dr. Woodard has no definite information O that strain gauge testing was ever performed at this depth (Woodard, Tr. 584, 611), nor at what depths it could be per-formed. (Woodard, Tr. 717, 729.) Dr. Woodard has never used strain gauges in the field (Woodard, Tr. 534) nor does he know specifically how to install a strain-gauge in a rock. (Woodard, Tr. 627.) The specific type of strain gauges that Dr. Woodard has used in the laboratory could not be applied to down-hole strain measurements. (Woodard, Tr. 618.)

37. Although Dr. Woodard stated that triaxial strain

gauges and fulcrum strain gauges are usable in the field (Woodard, Tr. 617), he later testified that he doesn't know if strain gauges can be used to measure strain at depths of 3500 feet or lower. (Woodard, Tr. 717, 729, 734; Applicant's Exhi-bit No. 1.)

38. Dr. Woodard also stated that in his opinion the best place to obtain good measurements would be at the foci of earthquakes, at least three miles below the earth's surface,

but acknowledged that this depth would be too great to perform the strain gauge measurements. (Woodard, Tr. 732-733.)

39. The evidence indicates that given the present state of technology, strain gauge testing such as recommended by Dr. Woodard at a depth of 3500 feet or greater is not feasi-ble. (Woodard, Tr. 717, 729-732, 734, 742-743; Applicant's l Exhibit No. 1.)

40. Strai n must be significant to be measurable.

(Woodard, Tr. 619.)

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41. The undisturbed residual soil at least 125,000 (V) years old overlying the Plum River and Sandwich Faults indi-cates that no measurable strain has occurred over this period.

(Singh, Tr. 509-510; Alterman, Tr. 804, Alterman, NRC Staff Prepared Testimony, at 6-7, ff. Tr. 753.)

42. Strain gauges have not been devised that can measure the very small strain rates that may exist along the Plum River Fault Zone. (Alterman, NRC Staff Prepared Testimony at 7, ff. Tr. 753.)

43. Even if a strain gauge could be inserted to depths greater than 3500 feet, the strain rate would not lead to any meaningful results in a reasonable period of time.

(Singh, Tr. 509, 512-513; Alterman, Tr. 804, Alterman, NRC Staff Prepared Testimony, at 6-7, ff. Tr. 753. )

44. Further, even if a strain gauge could be insert-ed to depths below 3500 feet and even if strain rate could be measured in a reasonable period of time, a technique for trans-lating strain gauge data to predict fault movement has not been

developed due to the numerous variables affecting predictabil-l l

ity. (Alterman, Tr. 783.)

l 45. The Board finds that strain gauge testing of the l

faults cutting basement rock in northern Illinois would serve no useful purpose in connection with the Bryon plant.

46. The Byron plant is designed for a safe shutdown earthquake (SSE) peak ground acceleration value of .20g and an operating basis earthquake (OBE) peak ground acceleration value g of .09g. (Singh, Applicant Prepared Testimony, at 3, ff. Tr.

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431; Rothman, NRC Staff Prepared Testimony, at 3-4, ff. Tr.

760.)

47. As set forth in the NRC's regulations, 10 CFR Part 100, Appendix A, the safe shutdown earthquake, also com-monly referred to as the design basis earthquake, is that earthquake which is based upon an evaluation of the maximum earthquake potential considering the regioncl and local geology and seismology and specific characteristics of local subsurface material. It is the earthquake that produces the maximum i vibratory ground motion which the structures, systems, and components that are necessary to enable a reactor to shut down and avoid major offsite exposures are designed to withstand.

(Singh, Applicant Prepared Testimony, at 3, ff. Tr. 431.)

48. The operating basis earthquake is that earth- a quake which, considering the regional and local geology and seismology and specific characteristics of local subsurface material, could reasonably be expected to affect the plant site during the operating life of the plant. It is the earthquake j that produces the vibratory ground motion for which those features of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the i public are designed to remain functional. If the vibratory I

ground motion exceeding that of the operating basis earthquake i occurs during the life of the plant, the Commission't regula-tions require that the plant be shut down and that, prior to resuming operations, it must be demonstrated that no functional damage has occurred to those features of the plant necessary

for the continued operation without undue risk to the health and safety of the public. (Singh, Applicant Prepared Testi-mony, at 3-4, ff. Tr. 431.)

49. On voir dire examination, Intervenor's witness, Dr. Woodard, admitted he is not a seismologist and does not consider himself an expert with respect to determining the appropriate ground acceleration for which a structure should be designed based upon the geology and seismology of the site.

(Woodard, Tr. 521-522.) Dr. Woodard candidly disavowed any knowledge on how to calculate the seismic design basis for a nuclear power plant. (Woodard, Tr. 528-29.) Finally, Dr.

Woodard testified that he did not know how earthquake inten-sity, magnitude or peak ground acceleration parameters are utilized in developing the seismic design for a nuclear power plant. (Woodard, Tr. 589-590.)

50. On cross examination, no information was eli-cited to refute the testimony of Applicant's witness, Dr.

Singh, or that of the NRC Staff's witness, Dr. Rothman, with respect to the appropriateness of the SSE and OBE peak ground acceleration values.

51. The selection of ground acceleration values is connected to the determination of the intensity of the earth-quake for which a facility is designed. (Singh, Applicant Prepared Testimony, at 4, ff. Tr. 431.)

52. Seven earthquakes hav; occurred in northern Illinois between 1804 and 1972 with, according to P.C. Heigold, Modified Mercalli intensities ranging from IV to VI. (Yonk, Tr. 446; Woodard, Tr. 551, 553-556, 558.)

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53. Although the seven earthquakes depicted by Heigold have been interpreted as having Modified Mercalli intensities as high as VII, (Yonk, Applicant Prepared Testi-mony, at 10, ff. Tr. 478; Yonk, Tr. 446-447) there is no evi-dence that any earthquake has occurred in northern Illinois with a Modified Mercalli intensity greater than VII. (Woodard, Tr. 558-559; Rothman, Tr. 849; Singh, Applicant Prepared Testi-mony, at 5, ff. Tr. 431; Yonk, Applicant Prepared Testimony, at 10, ff. Tr. 478.)

54. The controlling earthquake for the Byron plant is the 1937 Anna, Ohio (which is also located in the Central Stable region) Modified Mercalli intensity VII-VIII earthquake.

(Singh, Applicant Prepared Testimony, at 5, ff. Tr. 431.)

55. The SSE for Byron is based upon an earthquake with a Modified Mercalli intensity of VIII, which is greater than any earthquake ever recorded in either northern Illinois or the entire Central Stable Region. (Singh, Applicant Pre-pared Testimony, at 5, ff. Tr. 431, Rothman, Tr. 849.)

56. The earthquake magnitude selected for the Byron SSE which corresponds to a Modified Mercalli intensity VIII earthquake is 5.8. (Singh, Applicant Prepared Testimony, at 5, ff. Tr. 431; Rothman, NRC Staff Prepared Testimony at 2, ff.

Tr. 760.)

, 57. The magnitude of the 1937 Anna, Ohio earthquake I

is estimated to range from 5.0 to 5.3 and the magnitude of the i

largest historical earthquake in the Byron area, the May, 1909 northern Illinois earthquake, is estimated to be 5.1. (Singh, Applicant Prepared Testimony, at 5,'ff. Tr. 431.)

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58. A site-specific response spectrum was generated for the Tennessee Valley Authority's Sequoyah Nuclear Power Plant based upon actual accelerograms of earthquakes with body wave magnitudes of 5.8 + 0.5 (5.3 to 6.3) recorded at rock sites, at epicentral distance of approximately 25 kilometers (15.5 miles) or less. (Singh, Applicant Prepared Testimony at 6, ff. Tr. 431; Rothman, NRC Staff Prepared Testimony, at 3, ff. Tr. 760.)

59. Like Byron, the Sequoyah plant has a Modified Mercalli intensity VIII design basis earthquake. (Singh, Tr.

480 )

60. Like Byron, the Sequoyah Plant is a rock site.

(Singh, Tr. 480.) In applying the Sequoyah response spectrum to the Byron site the exact nature of the rock is not crucial so long as the site has a shear wave velocity above 3,000 to 3,500 feet per second. Both the Byron and Sequoyah plant sites have the requisite shear wave velocities. (Singh, Tr. 481.)

61. Grouting of foundation rock has little effect, if any, upon the shear wave velocity of the rock. If it has any effect, it would be a small increase in the shear wave velocity. (Singh, Tr. 482, 485, 502.)

62. The grouting material in the Byron site has no significant effect on seismic input. (Singh, 483-484.)

63. A Byron site specific spectrum would have uti-lized parameters identical to those used in generating the Sequoyah site specific spectrum and would have yielded identi-cal results. (Singh, Applicant Prepared Testimony, at 6, ff.

Tr. 431; Singh, Tr. 497.)

64. In the frequency range of interest, the Sequoyah site specific spectrum corresponds closely to the Byron design basis response spectrum. (Singh, Applicant Prepared Testimony, I at 6, ff. Tr. 431.)

65. On July 5, 1982 at 04:13:49.81 GMT (4 July 1982, at about 11:14 p.m. CDT) there was a magnitude 3.8 earthquake with an epicentral location of 35 11.l' North latitude, 92*

13.72' West longitude, near the town of Enola, Arkansas. An SMA-1 strong motion seismograph was located about 200 meters from the epicenter and recorded a peak acceleration of 0.59g on its east-west component. Another strong motion seismograph, a DR-100 which was co-sited with the SMA-1, recorded a peak hori-zontal acceleration of 0.19g. The discrepancy in accelerations between the co-sited SMA-1 and DR-100 instruments is currently unexplained. The Tennessee Earthquake Information Center (TEIC), the agency that is monitoring the earthquake, has stated: "A distinct possibility is that the high SMA-1 accel-

eration is an installation effect and does not represent a true 1

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free-field acceleration. " The entire earthquake recording had a duration of about 3 seconds and the high acceleration had a frequency of about 17 hertz. (Rothman, NRC Staff Prepared l Testimony, at 6, ff. Tr. 760.)

66. If the 0.59g acceleration is not due to instal-lation effects, then it would represent a very close (near field) high frequency, short duration record of an earthquake with little energy. There was no damage reported to the shed in which the SMA-1 instrument was located or to any other build-l

(

ing from this earthquake. Inasmuch as there was no damage to these buildings that were not designed to withstand earthquake motion, there is no reason to believe that earthquake motion of this type cculd cause damage to a nuclear power plant which is designed using a broad band response spectrum encompassing the wider frequency range and higher energies of larger earthquakes.

(Rothman, NRC Staff Prepared Testimony, at 7, ff. Tr. 760; Rothman, Tr. 758, 806-811; Woodard, Tr. 588.)

67. The Board finds that the Byron Station safe shutdown earthquake peak ground acceleration value of .20g is sufficiently conservative and complies with 10 CFR Part 100, Appendix A.

68. The expected recurrence of an earthquake with a Modified Mercalli intensity VI and corresponding peak ground acceleration value of .09g has been calculated to be 2150 years by Applicant and from 200 to 1000 years by the NRC Staff's consultant, Lawrence Livermore Laboratories. (Singh, Applicant l

Prepared Testimony, at 6-7, ff. Tr. 431; Singh, Tr. 437; Roth-man, NRC Staff Prepared Testimony, at 4, ff. Tr. 760; Rothman, Tr. 757.) A third study conducted by Dr. Robert B. Herrmann l

predicts a return period of approximately 1000 years for peak i ground accelerations of about .09g in the site area. (NRC Staff Prepared Testimony, at 5, ff. Tr. 760; Rothman, Tr. 758.)

69. Applicant's calculation of a return period of 2150 years was based upon a Modified Mercalli intensity VI or greater earthquake occurring at this site. (Singh, Tr. 505).

The site is defined as an 18-mile radius around the plant.

v (Singh, Tr. 493.)

70. By any calculation, the return period for a .09g peak ground acceleration value is far greater than the 40-year operating life of the Byron plant. (Rothman, NRC Staff Prepared Testimony, at 5, ff. Tr. 760; Rothman, Tr. 758; Singh, Applicant Prepared Testimony, at 5, ff. Tr. 431.)

71. The Board finds that an operating basis earth-quake with a peak ground acceleration value of .09g complies with 10 CFR Part 100, Appendix A, in that it is a conservative assessment of the earthquake which could reasonably be expected to affect the plant site during the quality life of the plant.

B. DAARE/ SAFE Contention 9(a) - Waterhammer j

72. DAARE/ SAFE Contention 9(a) concerned waterhammer events caused by rapid condensation of steam in feedwater systems. The scope of this contention was narrowed by the Board on September 10, 1982 in its Memorandum and Order Ruling on Motions for Summary Disposition of DAARE/ SAFE Contentions.

That order was subsequently clarified by the Board on January 7, 1983 in its Memorandum and Order Ruling on Applicant's Motion for Clarification. By those two orders, DAARE/ SAFE Contention 9(a) was limited to a consideration of whether a waterhammer event could occur in the Feedwater Bypass System at Byron similar to the type of event that occurred at the Krsko plant ,

in Yugoslavia.

73. Applicant, Staff, DAARE/ SAFE, and the League stipulated by an agreement dated February 15, 1983, that, inter alia, the final language of DAARE/ SAFE Contention 9(a) shall read as follows for litigation:

[ _._ _ _ _ _ _ _ _ -_ - - -

1 1

During recent start-up tests at the Krsko plant in Yugoslavia, which has steam generators which are similar in design to those at Byron, the plant experienced a bubble collapse waterhammer event in the feedwater bypass line. Applicant should be required to demonstrate that a similar event will net occur at Byron.

74. Applicants presented the testimony of Robert W.

Carlson, a Principal Engineer in the Balance of Plant Systems Design Group of the Nuclear Technology Division of Westinghouse Corporation, who testified as to the cause and prevention of a Krsko-type waterhammer event, and Richard Pleniewicz, Assistant Superintendent of Operations at Byron, who testified as to Applicant's modifications in design and operation of the steam generator feedwater systems to minimize the occurrence of a Krsko-type waterhammer event. NRC Staff presented the testi-

mony of Aleck W. Serkiz, a Senior Task Manager in the Generic I

Issues Branch, in the NRC Office of Nuclear Reactor Regulation, who reviewed the documented information surrounding the Krsko event and leviewed the design and operational controls at Byron intended to minimize the occurrence of the Krsko event at Byron. Intervenors presented no testimony on Contention 9(a) .

75. Byron and Krsko both have steam generators referred to by Westinghouse as the Model D counterflow pre-heater type steam generator and similar Feedwater Bypass Sys-tems and Auxiliary Feedwater Systems. The Feedwater Bypass System is designed specifically to prevent waterhammer in the preheater section by diverting cold feedwater to the auxiliary nozzle. The Auxiliary Feedwater System is designed to provide O

feedwater to the steam generator through the auxiliary nozzle O via a section of the feedwater bypass line in the event of a loss of heat sink accident. (Carlson, Applicant Prepared l

Testimony, at S-8, ff. Tr. 930.)

76. The Krsko waterhammer event was a " bubble col-lapse" type. (Carlson, Applicant Prepared Testimony, at 8, ff.

Tr. 930.)

77. The Krsko waterhammer event was believed to have occurred in the Feedwater Bypass System during Hot Functional Testing of the Auxiliary Feedwater System pumps in July,1981.

(Carlson, Applicant Prepared Testimony, at 8, ff. Tr. 930; Carlson, Tr. 1086-1090.)

78. The resultant damage consisted of loosening and movement of some pipe hanger components and some change in the locaticn of the bypass feedwater piping. Also a bulge was discovered on the upper surface of the bypass piping near the secondary shield wall. Outside of containment there was negli-gible pipe movement and blistering of the paint on the auxil-iary feedwater piping indicating unusually high temperatures as far back as the motor driven Auxiliary Feedwater System pumps.

Despite the damage, the design functions of the Auxiliary Feedwater System and the Feedwater Bypass System were not adversely affected. (Carlson, Applicant Prepared Testimony, at 8, 11-12, also Figures 3, 4, ff. Tr. 930; Carlson, Tr. 1091, 1117-1118.)

79. Two concurrent conditions were primarily respon-sible for the Krsko waterhammer event. First, the check valves

which are provided to prevent reverse flow in the Auxiliary l O Feedwater System were leaking. Second, the water level in the steam generator fell below the discharge end of the internal extension of the auxiliary nozzle. The leaking of the check valves allowed hot water and/or steam to leak back into the bypass piping and subsequently into the Auxiliary Feedwater System piping. The abnormally low water level in the steam generator, allowed steam to enter the discharge end of the auxiliary nozzle internal extension and thus backleak into the bypass piping. With steam present in the bypass line, the motor driven pumps of the Auxiliary Feedwater System were started as part of Hot Functional Testing thereby introducing cold water into the bypass piping. The cold water rapidly condensed the steam and thus caused the waterhammer. (Carlson, Applicant Prepared Testimony, at 9-10, ff. Tr. 930; Carlson, Tr. 1086-1090.)

80. Westinghouse has made several recommendations to Applicant to avoid a Krsko-type waterhammer event in the Feed-water Bypass System at the Byron Station. First, temperature l sensors should be installed on the bypass piping close to the auxiliary nozzle to detect backleakage of hot water or steam.

Second, in the event backleakage is detected, the bypass piping should be slowly refilled or the plant brought to a cold shut-down condition. The recommended refill rate is in a range which was shown to be safe by an analytical study performed by the Westinghouse R & D Center. Third, Westinghouse recommended that the steam generator water level be maintained above the l

auxiliary nozzle discharge pipe as much as possible. Fourth, the check valves of the Auxiliary Feedwater System should be maintained to minimize backloakage. (Carlson, Applicant Pre-pared Testimony, at 12-13, 16, ff. Tr. 930.)

81. Assuming a failure of the check valves and a low water level in the steam generator, backleakage during power operation is very unlikely since, at between 0 and 100 percent power, a continuous flow is provided through the steam genera-tor auxiliary nozzle which effectively prevents the back flow of steam from the steam generator. However, during the normal non-power operations of heat-up, cool-down and hot standby, the flow which continues through the auxiliary nozzle is relatively small and the opportunity for backleakage is greater than during power operation. (Carlson, Applicant Prepared Testi-mony, at 10-11, ff. Tr. 930; Pleniewicz, Applicant Prepared Testimony, at 5-6, ff. Tr. 896.)

82. The proposed arrangement of check valves for the Auxiliary Feedwater System at Byron Station provides that there will be two check valves in each flow path by which backleakage into that system could occur. This check valve arrangement is consistent with Westinghouse recommendations in that it pro-vides redundant check valves along each flow path. (Carlson, Applicant Prepared Testimony, at 14-15, ff. Tr. 930.)

83. The NRC Staff conducted an evaluation of the Krsko waterhammer event and the Westinghouse recommendations.

The NRC Staff Evaluation relied in part on NUREG/CR-3090 "Eva-p luation of Waterhammer Potential in Preheat Steam Generators",

d

which was introduced into evidence as Board Exhibit 2. The NRC Staff witness concluded that the Westinghouse recommendations constitute good engineering design practices and prudent oper-ating procedures. The implementation of the recommendations should prevent the occurrence of a Krsko-type waterhammer event at Byron. (Serkiz, Tr. 962, 963, 965, 968, 980-981.)

84. If the Westinghouse recommendations are adopted and implemented and applicant installs the check valves in the Auxiliary Feedwater System as proposed, the likelihood of occurrence of a Krsko-type waterhammer event is reduced to an acceptable level meaning that such an event should not occur.

i

) (Carlson, Applicant Prepared Testimony, at 16-17, ff. Tr. 930; l

Carlson, Tr. 968, 1112, 1130.)

85. Applicant confirmed that all of the Westinghouse recommendations have been or will be implemented at the Byron Station. (Pleniewicz, Applicant Prepared Testimony, at 4-8, ff. Tr. 896.)

86. The Westinghouse recommendations do not impose significant changes in the manner of operation of the plant.

(Pleniewicz, Tr. 1119-1120.)

87. Prior to power operation, Applicant will test the implemented recommendations. (Pleniewicz, Applicant Pre- ,

pared Testimony, at 7, ff. Tr. 896.)

88. If an event identical to the Krsko waterhammer event occurred at Byron, there would be no consequences.

(Serkiz, Tr. 1019-1020.)

O

i t

89. The Board finds that the implementation of Westinghouse recommendations by applicant will provide appro-priate protection for the Byron steam generator feedwater systems against the dynamic effects of a Krsko-type waterhammer event. Further, in the unlikely event of its occurrence, a Krsko-type waterhammer event does not endanger the health and i

safety of the public.

C. League Contentions 42, 111, and 112 - ALARA l

)

90. In general, Contentions 42, 111 and 112 assert that Byron Station cannot be operated so as to maintain radia-tion exposures to workers as low as is reasonably achievable

("ALARA"). Moreover, the contentions assert that even if the Applicant maintains occupational exposures within the limits set by 10 C.F.R. Part 20, new information on the effects of low-level radiation exposures shows that such limits will not ensure worker safety.

91. Contention 42 reads as follows:

As the Staff has recognized in NUREG-0410 and in the Black Fox testimony pre-

, viously cited, occupational radiation

! exposure to station and contractor person-nel has generally been increasing in recent years, and violation of the limits of.10 C.F.R. Part 20 has been avoided by C.E., as by other licensees, by obtaining the tem-porary services of transient workmen rather than by devoting adequate effort to reduc-ing exposures. Among other things, this practice results in using larger numbers of people and thereby increasing the risk of sabotage, operator error and similar safety-related hazards. Furthermore, new informa-tion on low-level radiation effects indi-cates that the Byron design basis will not Os provide safe operation. Accordingly, both

because of the lack of assurance that p proper exposure levels will be maintained and because of the practice of using tran-sient workers, as a result of this serious and unresolved problem the findings requir-ed by 10 C.F.R. $$ 50.57(a)(3)(8) and 50.57(a)(b) cannot be made.

92. By Stipulation dated December 6, 1982, the Parties agreed to litigate only the in-plant monitoring aspects of Contention 111. That contention reads as follows:

111. C.E. has not met the requirements of NEPA and the Regs, including but not limited to 10 C.F.R. SS 50.34(a) and 50.36 (a) because C.E. has not adequately moni-tored and provided a design base for the Byron plant which will keep radiation levels as low as achievable as required for operation of the plant to protect the health and safety of the public. To keep radiation levels as low as achievable, C.E.

should provide and utilize:

A. More adequate environmental and discharge monitoring of radioac-tive emissions from the Byron plant, which include:

(1) Monitoring devices at more locations within and without the plant site.

(2) Provisions for more frequent reading of monitors by independent analysts.

(3) Better monitoring de-vices which include:

(a) An automatic sys-tem of monitoring that notifies local authorities by an alarm when discharge emission exceed design limits;

(

(b) Monitoring devices that measure differences in alpha, beta and gamma dose s levels, which presently are J not proposea to be consi-dered and measured;

i t

( 170- .

N 1

(c) Monitoring and re--

O d

cording of emissions tof all

~

dangerous long Jived radio ~ -

nuclides, including espe-s- cially I-129 and Plutonium;.

.. (d) 'Blyccumulative I 1

s testing in' a tiered 'syctem 4 to assess ",the uptake ~ of ,

t iradioactive and polutants from\ bottom aedi-chemical i,.

~

inents or Soil to lower organisms .an~d to contamina-

~ ' ' tion of ' Ithe food chSin of '

' ' man'and other life.

N c B. More accurate Aalculation of design doses which can be, accomplished '

by utilizing information - from . the x improired 4Taonitoring suggested .al.ove '

and also by: ,, _I . [,  ; ,

(1) Providing i for' 'and con- .'

stant update' and replac6 ment of l equipment and analysis to respond '

to new experimental and 'analyti-cal results.

for construction, Byron was' licensed for example,

'('

vhen some .(including C.E.) as-serted improperly that there was i

sa' threshold to radiation effects; (2) Including in calcula-tion of doses the large . transient populations in the low population zones around the plant, including school children when present in schools and other participating in recreational facilities; (3) Including internal ra-diation doses caused by inhaled and/or ingested radionuclides which are deposited in different parts of the body where they give repeated radiation or until they are eliminated from the body;

(4) Including in calcula-tion of radiation doses, cummula-tive doses to the general popula-tion outside the site boundary Q

V caused by overlapping circles of radiation from any nuclear facil-

M ;5 ity (whether on or off the site),

T including Zion, Dresden, LaSalle,

'(V Quad Cities, and Braidwood Sta-tions, as well as any new propos-ed facility and disposal facili-ties such as the Morris Waste a Disposal Site; and (5) Including in the calcu-lation, calculation of doses to people by utilizing actual radio-nuclides for and in food, animals, plants, soil, water, and in other parts of the environment in and around the Byron site.

As a result, the applicable findings required by the Act, NEPA, and the Regs, cannot be made herein.

93. Contention 112 reads as follows:

112. C.E. has not met the require-ments of NEPA and 10 C.F.R. Part 20 because it has not adequately assessed the effect s of radiation on plant workers and provided a design base for the Byron plant which will provide radiation levels as low as achievable. To keep radiation levels as low as achievable there is a need for better use of preventive measures to reduce radiation, including neutron, exposure levels to regular plant personnel and transient workers. These include but are not limited to:

(a) Plant designs for reducing amount of radiation exposure which take into account new evidence on low levels of radiation which were not considered in design of the plant.

(b) Improved record keeping of radiation exposures, including cummu-lative exposures both at the plant site and at other facilities.

(c) Better training of personnel to prevent radiation exposures, in-cluding more use of regular trained personnel rather than transient or -

temporary workers with little exper-ience and training.

=

(d) Limiting exposure to high levels of radiation to volunteers

(] and/or only older workers beyond the l child bearing age or others incapable of biological reproduction.

(e) Better education about ra-diation dangers to ensure cooperation of workers in keeping radiation expo-sures to a minimum.

As a result, the applicable findings required by the Act, NEPA, and the Regs, cannot be made therein.

94. For these contentions, the Applicant presented Frank Rescek, Lead Health Physics - Technical Services Engi-neer, Commonwealth Edison Company, and James R. Van Laere,

Radiation Chemistry Supervisor and Radiation Protection Manager at Byron Station, to testify on the ALARA programs at the corporate level and at Byron Station, respectively. Applicant also presented Jerome L. Roulo, Deputy Nuclear Security Admin-istrator, Commonwealth Edison Company, who testified about Byron Station's security system. Gerald P. Lahti, Assistant Division Head of the Nuclear Safeguards and Licensing Division in charge of Shielding and Radiological Safety at Sargent &

Lundy, testified for Applicant on the design basis of Byron Station and Dr. Lawrence Conway, Advisory Engineer of Westing-house Electric Corporation, testified on the design of the steam generators at the plant. Finally, Applicant presented Dr. Jacob I. Fabrikant, a physician, radiologist, professor, and researcher, who testified on the risks associated with ex-posure to radiation. The NRC Staff's witnesses were: Michael A.

, Lamastra, a Health Physicist in the Radiation Protection Sec-O L

tion of the Radiological Assessment Branch of the NRC, who testified on the Staff's review of Byron Station's ALARA pro-gram; Edward F. Branagan, Jr., a Health Physicist in the Radio-logical Impact Section of the Radiological Assessment Branch of the NRC, who presented testimony on the risks of radiation exposure; and Robert F. Skelton, Plant Protection Analyst in

! the Power Reactor Safeguards Licensing Branch, Division of Safeguards of the NRC who testified on security aspects of Byron Station.

95. The League presented Dr. Karl Z. Morgan, a I

consultant on radiation safety and Adjunct Professor at Appa-i lachian State University, to testify on the effects of exposure to radiation. Dr. Morgan is admittedly an expert health physi-cist, but he also expressed concern about numerous aspects of Applicant's ALARA program. Although the Board admitted this testimony as a relevant expression of concern, we find our-selves unable to give much weight to it. Dr. Morgan's testi-mony on cross-examination generally revealed that he had little or no knowledge of Applicant's practices in the areas in which he raised questions. He complained, for example, that Appli-cant does not use the personal neutron dosimeters which it in fact uses. (Morgan, League Prepared Testimony at 23-24, ff.

Tr. 1515; Rescek, Applicant Prepared Testimony at 10, ff. Tr.

1157; Tr. 1163, 1174.) Although Dr. Morgan claimed that Byron Station needs more area radiation monitors, he admitted that he did not know how many monitors the station has. (Morgan, League Prepared Testimony at 21, ff. Tr. 1515; Morgan, Tr.

1662.) Dr. Morgan stated that contract workers were more likely to improperly use the dosimetry program and improperly wear protective clothing than other employees, yet after voir dire examination, the Board required him to limit his state-ments to his experience at the Oak Ridge National Laboratory, where he has not worked for over ten years. (Morgan, League Prepared Testimony at 17, ff. Tr. 1515; Tr. 1511-1512.) We do not think any of Dr. Morgan's testimony poses a serious chal-lenge to the adequacy of Applicant's ALARA programs.

Cancer Risk Estimate

96. Dr. Morgan expressed concern that the collective radiation dose per year from Byron Station would not be as low as he considered appropriate. (Morgan, League Prepared Testi-mony at 8, ff. Tr. 1515.) He conceded that Commission regula-tions do not limit the collective dose from a nuclear plant.

(Morgan, Tr. 1524.) The basis of Dr. Morgan's concern is his belief that the risk of cancer and genetic effects from expo-sure to low levels of radiation, such as those allowed for workers under Commission regulations, is much higher than is accepted in the scientific reports published by national and international standard-setting committees. (Morgan, League Prepared Testimony at 12-13, ff. Tr. 1515.)

97. The Applicant presented testimony of Dr. Jacob I.

l Fabrikant, an eminent scientist in the field of radiation, and the NRC Staff presented testimony of Dr. Edward F. Branagan, ,

n Jr., a health physicist. Both witnesses testified that the U

i overwhelming weight of the scientific evidence supports the risk estimates set forth in the reports of the standard-setting bodies. (Fabrikant, Applicant Prepared Testimony, passim, ff.

Tr. 1399; Branagan, NRC Staff Prepared Testimony at 7, ff. Tr.

1883.)

98. Dr. Fabrikant explained that epidemiological studies of exposed human populations and laboratory animal studies show that even low dose levels of radiation exposure carry some risk of cancer. These methods do not allow direct quantification of the risk, however, so a mathemetical model is used to exptrapolate this risk. (Fabrikant, Applicant Prepared Testimony at 9-12, ff. Tr. 1399.) For radiation protection purposes, the scientific community usually endorses a linear dose-response model, because it is conservative. (Fabrikant, Applicant Prepared Testimony at 19, ff. Tr. 1399.)

99. Dr. Morgan's belief that generally accepted cancer risk estimates are too low appears to rest on two bases:

several recent studies which have reached this conclusion and a risk estimate derived by Dr. Morgan himself based on the re-sults of three studies. (Morgan, League Prepared Testimony at 4, 8, ff. Tr. 1515; Morgan, Tr. 1545, 1555-1556.)

100. The evidence showed that neither basis for Dr.

Morgan's opinion was reliable. Dr. Fabrikant offered a lengthy and detailed review of each of the recent reports which have challenged the conservatism of the linear hypothesis. All such studies have serious methodological and statistical deficien-cies, and no scientific commission has accepted them. (Fabri-kant, Applicant Prepared Testimony at 16, 27-45, ff. Tr. 1399.)

l 101. Furthermore, there is strong biological evidence that the linear hypothesis overestimates the effects of low-level radiation doses. (Fabrikant, Applicant Prepared Testi-mony at 22-24, ff. Tr. 1399.)

102. Dr. Branagan testified that no convincing evi-dence exists to support Dr. Morgan's own calculation of the appropriate cancer risk. (Branagan, NRC Staff Prepared Testi-mony at 9, ff. Tr. 1883.)

103. Dr. Morgan conceded that he had simply plotted some curves by hand based on the results of three reported studies. (Morgan, Tr. 1555-1556, 1586-1587.) He did not have access to a computer, which is the customary way to analyze the data in order to develop a dose-response curve. (Morgan, Tr.

1587.)

104. Dr. Morgan conceded that his determination of the cancer risk estimate would be significantly lower if he had i

excluded the data from the Mancuso study, one of the three studies on which he relied. (Morgan, Tr. 1592.) Dr. Fabrikant pointed out _that the results of the Mancuso study have been widely criticized by leading epidemiologists and statisticians.

(Fabrikant, Applicant Prepared Testimony at 27-30, ff. Tr.

1399.) Dr. Branagan also critized the Mancuso Study and stated that better sources of data exist for determining the cancer risk estimate. (Branagan, Tr. 1886-1888.)

105. No scientific commission or council charged with evaluating and recommending radiation protection standards has q endorsed a risk estimate, such as Dr. Morgan's, which states

(>

that lower doses of radiation are much more harmful than pre-O dicted 3r the 1ineer no-thresho1d mode 1. (Fenrikent, App 11 cent Prepared Testimony at 16, 18, 24, ff. Tr. 1399.)

106. The Board finds that there is no credible evi-dence that exposure to low doses of radiation carries a greater cancer risk per unit dose than exposure to high doses. The evidence shows that the linear dose-response model is conserva-tive.

107. Given the linear dose-response model, which we have found to be conservative, the spreading of a given quan-tity of dose over a larger number of workers would not increase the overall risk of health effects. (Branagan, NRC Staff Pre-pared Testimony at 8, ff. Tr. 1883.)

108. The Staff's risk estimates of potential health effects to occupationally exposed workers at the station are contained in section 5.9.3.1 of the Byron Final Environmental Statement ("FES") and were adopted by Dr. Branagan as part of his testimony. (Branagan, NRC Staff Prepared Testimony at 4, ff. Tr. 1803.) Attachment F to Dr. Branagan's testimony con-sists of a table which demonstrates that the risk estimates used in the Byron FES are consistent with the estimates of the major standard-setting scientific bodies. (Branagan, NRC Staff Prepared Testimony, Attachment F, ff. Tr. 1883.)

109. Both the Applicant and NRC Staff witnesses and the major standard-setting bodies conclude that the effects of low-level radiation exposure are approximately those set forth in the FES. (Fabrikant, Applicant Prepared Testimony at 61-64, G'

1 ff. Tr. 1399; Branagan, NRC Staff Prepared Testimony at 7 and Attachment F, ff. Tr. 1883.)

110. Based on the overwhelming weight of the evi-dence, the Board finds that the assessment of the health ef-fects of occupational radiation exposure at Byron contained in the Byron FES is adequate.

ALARA Programs 111. Applicant has a detailed corporate ALARA program and a complementary ALARA program at Byron Station. Mr. Rescek and Mr. Van Laere described these programs in detail, both in their prepared testimony and during cross-examination. (Rescek, Applicant Prepared Testimony at 2-8, ff. Tr. 1157; Rescek, Tr.

1268-1273; Van Laere, Applicant Prepared Testimony at 2-10, ff.

Tr. 1707.) Applicant is strongly committed to its ALARA pro-grams and management supports this commitment. The committees that oversee the ALARA programs include top level management, both at the corporate level and at Byron Station. Other mem-

bers have had extensive experience at nuclear stations in controlling radiation exposures. (Rescek, Tr. 1268-1283.) An i

example of management's dedication to its ALARA programs is its l decision to modify the station's steam generators before the start-up of the plant. This decision raeans no radiation expo-sure will occur as a result of the modifications. (Butterfield, Applicant Prepared Testimony at 5, ff. Tr. 5908; Blomgren,

Applicant Prepared Testimony at 17, ff. Tr. 4126; Blomgren, Tr.

t l 4120.) The Board finds that Applicant's ALARA program as a V

1 whole is both comprehensive and detailed and demonstrates an O

V impressive commitment on Applicant's part to the achievement of ALARA goals. Mr. Rescek and Mr. Van Laere were highly intelli-gent and credible witnesses and answered the League's and the Board's questions in great detail. Both witnesses themselves play important roles in Applicant's ALARA programs, and the Board finds that their intelligence and dedication are further grounds for assurance that the programs will be carried out dilligently.

Dosimetry Program 112. Applicant's dosimetry program is designed to provide an accurate assessment of the dose equivalent received by an individual. (Rescek, Applicant Prepared Testimony at 9, ff. Tr. 1157.) Gamma radiation is the primary type of radiation a nuclear power plant worker could be exposed to. (Fabrikant, Applicant Prepared Testimony at 8, ff. Tr. 1399.) Applicant's dosimetry program will monitor gamma radiation, as well as neutron exposure and beta radiation. (Rescek, Tr. 1165-1168, 1180, 1220, 1223.)

113. Personal film badges will monitor a worker's j exposure to gamma and beta radiation. (Rescek, Applicant i

Prepared Testimony at 11, ff. Tr. 1157; Rescek, Tr. 1223.)

Fast and intermediate neutrons will be monitored by badges containing CR-39. (Rescek, Tr. 1165-1168, 1180.) The League contended that Applicant uses NTA fil.1s for determining neutron dosimetry, but the evidence showed that since 1980 Applicant

l

I l

has used a superior detector containing CR-39, which the League's witness agreed was the best available. (Morgan, League Prepared Testimony at 23-24, ff. Tr. 1515; Morgan, Tr. 1650.)

114. All badged workers will use pocket ionization chambers as secondary dosimeters to monitor gamma dose. (Rescek, Applicant Prepared Testimony at 15, ff. Tr. 1707.) Thermolumi-nescent dosimeters ("TLDs") will be used to measure extremity exposures. (Rescek, Applicant Prepared Testimony at 10-11, ff.

Tr. 1157; Rescek, Tr. 1173-1174.) When a worker must perform a job in a high radiation area containing variable dose rates, meters known as beepers, which emit a sound that increases in intensity as the radiation level increases, may be used. (Van Laere, Applicant Prepared Testimony at 15-16, ff. Tr. 1707; Van Laere, Tr. 1751-1152.) In addition, a worker can carry a hand-held instrument which responds to beta dose, and REM meters will monitor neutrons and gamma radiation. (Rescek, Tr.

1221; Rescek, Applicant Prepared Testimony at 10, ff. Tr. 1157; Rescek, Tr. 1165, 1179, 1274-1275; Van Laere, Tr. 1734.)

l 115. In addition, Applicant surveys an area before a worker is allowed to enter it to determine the type and '.evel of exposures to which the worker may be subjected. (Rescek, Applicant Prepared Testimony at 12-13, ff. Tr. 1157; Rescek, Tr. 1221.)

116. An independent vendor supplies and processes the j film badges and TLD finger rings. Normally they will be pro-cessed bi-weekly, but results can be obtained in as little as g-) four hours if necessary; results are entered into the station's V

, computer system. (Rescek, Applicant Prepared Testimony at 11, ff. Tr. 1157; Van Laere, Applicant Prepared Testimony at 15, ff. Tr. 1707; Rescek, Tr. 1218-1219.) Pocket ionization cham-bers are read daily at the station and the data cross-checked with the results from the film badges. The computer prints out any discrepancies and the health physics department investi-gates them. (Rescek, Applicant Prepared Testimony at 11, ff.

Tr. 1157; Rescek, Tr. 1259-1262.)

117. " Spiked" badges, exposed to known amounts of radiation, are submitted routinely to the processor as a check of accuracy. Both the Applicant and the film badge processor perform independent quality assurance tests on the film badge program. (Rescek, Applicant Prepared Testimony at 12, ff. Tr.

1157; Rescek at Tr. 1224-1225, 1257.)

118. Byron Station will have extensive in-house facilities for testing workers suspected of receiving con-tamination, including nose swabs, urine and fecal samples, and whole body counts. (Van Laere, Applicant Prepared Testimony at 16-17, ff. Tr. 1707; Rescek, Tr. 1209-1212.) Whole body counts will be performed periodically on all workers according to set procedures, and if a worker has received more than 3% of any

maximum permissible individual organ burden or more than 5% of the maximum permissible total body burden, the station will investigate. (Rescek, Tr. 1212, 1198-1199; Van Laere, Appli-cant Prepared Testimony at 17-18, ff. Tr. 1707; Van Laere, Tr.

1708-1711.)

O

119. Applicant has numerous other procedures designed

) to keep radiation doses ALARA, including: instruction in when and how to wear protective clothing (Rescek, Applicant Prepared Testimony at 23, ff. Tr. 1157; Rescek, Tr. 1204-1208), how to test a face mask (Rescek, Tr. 1210-1211) and how to remove skin contamination (Rescek, Tr. 1208-1209); the prohibition of eating, drinking, smoking, or chewing gum in a controlled area (Rescek, Tr. 1214); and the referral of a worker exposed to a dose exceeding regulatory limits to Applicant's medical depart-ment (Rescek, Tr. 1177; Van Laere, Tr. 1732). In addition, the station superintendent at Byron has the authority to delay a hot operation until after short-lived radionuclides have pri-marily decayed. (Van Laere, Applicant Prepared Testimony at 22, ff. Tr. 1707; Van Laere, Tr. 1737-1739.)

120. The Board finds that Applicant's dosimetry program is sufficient to provide an accurate assessment of the dose equivalent received by a worker at Byron Station and to maintain doses ALARA.

l In-Plant Monitoring 121. Dr. Morgan claimed that Byron Station needs more area monitors. (Morgan, League Prepared Teetimony at 21, ff.

Tr. 1515.) Upon examination, however, Dr. Morgan admitted that he did not know how many monitors currently exist in the station.

(Morgan, Tr. 1662.)

122. Byron Station will have more than 200 in-plant area and air monitoring instruments. (Van Laere, Applicant

Prepared Testimony at 20-21, ff. Tr. 1707.) The NRC Staff has evaluated Applicant's in-plant monitoring program and found the system to be acceptable and in compliance with 10 C.F.R. Part

20. (NRC Staff Prepared Testimony at 19A-20A, ff. Tr. 1883.)

Record Maintenance 123. Applicant maintains records containing all the information on Form NRC-5 in compliance with the requirements of 10 C.F.R. 6 20.401. (Lamastra, NRC Staff Prepared Testimony at 15-16, ff. Tr. 1883; Rescek, Applicant Prepared Testimony at 15, ff. Tr. 1157.) Applicant keeps records for all radiation exposures to individuals, both routine and non-routine, and records data from processed film badges and personal ionization chambers. (Rescek, Applicant Prepared Testimony at 15-17, ff.

Tr. 1157; Rescek, Tr. 1184.) These records are readily avail-able to any worker who wishes to know the amount of his expo-sure. (Rescek, T.'r . 1280-1281.)

124. Applicant will follow the guidance of Regulatory Guide 8.7, " Occupational Radiation Exposure Records System" in developing its occupational records system and will record doses by tasks to provide feedback information for ALARA re-I views. (Lamastra, NRC Staff Prepared Testimony at 15-16, ff.

Tr. 1883.)

125. Applicant has imposed an administrative daily dose limit on exposure to personnel of 50 millirem and an administrative weekly dose limit of 300 millirem as a method of p

%s dose management. Before a worker would be authorized to exceed I

- - - - - , , - , - - - - - - - - . , , . , , , , , ---,-,-,,..---e ~. --, v- -

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an administrative dose limit, a Radiation Work Permit would

! have to be prepared. These permits provide detailed radiologi-cal information, including protective equipment requirements, on the job to be performed. (Rescek, Applicant Prepared Testi-mony, at 13-14, ff. Tr. 1157; Rescek, Tr. 1285-1286.)

126. On the basis of the uncontroverted evidence, the i

Board finds that Applicant's recordkeeping program meets the requirements of 10 C.F.R. Part 20 and is sufficient to keep radiation exposures to workers ALARA.

Health Physics Staff 127. Byron Station's health physics department will consist of 15 full time employees and 28 radiation chemistry technicians who will spend approximately seventy percent of their time on health physics responsibilities. The station's training program may require these employees to attend training courses or seminars, both onsite and offsite, or to be tempor-arily assigned to another position to gain on-the-job exper-ience. (Van Laere, Applicant Prepared Testimony at 13-14, ff.

Tr. 1707.) In addition, all radiation chemistry technicians will receive 18 to 20 weeks of specialized training. Byron Station also currently has three health physicists with a Bachelor of Science degree in health physics. (Van Laere, Tr.

1719, 1721.)

128. Dr. Morgan believes that Byron Station's health physics staff should consist of approximately fifty people and that they should have more training. (Morgan, Tr. 1634, 1665.)

Dr. Morgan offered no basis for this conclusion. The NRC Staff O reviewed the number of health physics staff members at Byron Station and found it acceptable and approximately the same size as other nuclear power stations' health physics staffs.

(Lamastra, Tr. 1895-1898.) The evidence also showed that Byron staff members will generally have more training and education than is suggested. (Van Laere, Tr. 1717-1721.)

129. The Board finds that Byron Station's health physics staff will consist of enough people with sufficient training to enable them to carry out all of the radiation protection services that will be needed.

Training 130. All workers at Applicant's nuclear power sta-tions, including temporary workers, clerical staff, and mainte-nance personnel, are required to take the training course out-lined in the Nuclear General Employee Training ("N-GET") Manual (Applicant Ex. 4). This course instructs the workers on the fundamentals of radiation exposure, the importance of maintain-ing doses ALARA, their role in accomplishing that objective, and the biological effects of radiation exposure. (Rescek, Applicant Prepared Testimony at 20-22, ff. Tr. 1157; Rescek, Tr. 1187-1190, 1243, 1277; Van Laere, Tr. 1726.) Workers will also attend an annual retraining course, and workers performing radiation protection, cperational, or maintenatice duties re-ceive additional training in radiation safety. (Rescek, Tr.

1189-1190, 1297-1298.)

l

{

131. The NRC Staff has reviewed Applicant's training program and found that it will adequately maintain occupational radiation exposures ALARA. (Lamastra, NRC Staff Prepared Testimony at 16-17, ff. Tr. 1883.) On the basis of the evi-dence, the Board finds that the training of all workers at Byron Station, including its health physics staff, will be sufficient to maintain occupational exposures ALARA.

Pregnant Women 132. Recognizing that a pregnant worker needs addi-tional protection against radiation exposure, Applicant edu-cates every female worker on the potential risks of radiation exposure to a fetus and requires each woman to notify her supervisor when she knows, or suspects, she is pregnant.

(Rescek, Applicant Prepared Testimony at 26, ff. Tr. 1157.)

Neither a woman's job or position would be adversely affected if she declares herself pregnant. (Rescek, Tr. 1193-1194.)

133. Applicant limits the radiation exposure to a declared pregnant woman to 500 millirem over her gestation period. (Rescek, Applicant Prepared Testimony at 27, ff. Tr.

1157.) The exposure to the fetus will be significantly less.

(Rescek, Applicant Prepared Testimony at 27, ff. Tr. 1157.)

Applicant reviews a woman's radiation exposure records to determine how much exposure she has received since the onset of her pregnancy, so that her total exposure is kept below 500 l

millirem. (Rescek, Tr. 1193; Van Laere, Tr. 1773.)

O V

i

134. The League's witness, Dr. Morgan, was not con-cerned about the possible external exposure to a pregnant woman (Morgan, Tr. 1644), but pointed out that if a woman is exposed to J rem in a single quarter of a long-lived radionuclide, such as strontium-90, her pre-existing internal body burden would cause her to receive an exposure greater than 500 millirem during the next nine month period (Morgan, Tr. 1640-1643).

135. The probability of a woman receiving such a dose is, however, exceedingly low because Applicant has a respira-

tory protection program that satisfies the requirements of 10 C.F.R. Part 20. This program will ensure that workers will not receive any significant uptake of a long-lived radionuclide.

(Rescek, Applicant Prepared Testimony at 25, ff. Tr. 1157.) In addition, the araount of strontium-90 that will be airborne at Byron Station will be below the level at which it is even considered to be present under 10 C.F.R. Part 20. (Rescek, Tr.

I 1195-1197.) The Board therefore does not share the League's concern that Applicant's ALARA program will not adequately protect a fetus.

Temporary Workers 136. Dr. Morgan stated that when he worked at Oak Ridge National Laboratories more than ten years ago, contract workers were more likely to improperly use the dosimetry pro-gram and to improperly wear protective clothing than permanent employees. (Morgan, League Prepared Testimony at 17, ff. Tr.

(3 1515.) We have noted above that we cannot give much weight to N._)

this testimony as bearing on Byron Station. Under Applicant's ALARA program, temporary employees will receive the same N-GET training as permanent employees (Rescek, Applicant Prepared Testimony at 22, ff. Tr. 1157), and Applicant's dosimetry pro-gram (Rescek, Applicant Prepared Testimony at 24, ff. Tr.

1157), respiratory protection program (Rescek, Applicant Pre-pared Testimony at 25, ff. Tr. 1157), and record maintenance program (Rescek, Tr. 1188) will apply equally to contract workers.

137. To make sure a contract worker's exposure does not exceed allowable limits, each worker must fill out NRC Form-4, which requires him to provide his occupational dose history. Applicant will use the information provided in this form to establish restrictions to prevent each worker from exceeding any applicable standard. (Lamastra, NRC Staff Pre-pared Testimony at 2-3, ff. Tr. 1883; Rescek, Applicant Pre-pared Testimony at 24, ff. Tr. 1157.)

138. Initially, the Board was concerned that a worker might provide inaccurate information when filling out the form, but Applicant's witness stated that when a contract worker lists former employers, Applicant will obtain a confirmatory statement from each employer. (Rescek, Tr. 1230-1232, 1266-1267.) The possibility still exists that a contract worker could purposely fail to list all of his former employers, but no licensee currently has any other means to determine a work-er's occupational dose history. (Rescek, Tr. 1299.)

O U

Industrial Sabotage G

U 139. To protect against the risk of industrial sabo-tage, Applicant screens all of its employees and will not allow any contract worker to work at Byron Station unescorted unless he also has undergone pre-employment screening. (Roulo, Appli-cant Prepared Testimony at 2, ff. Tr. 1356.) All contractors must establish a screening program that meets Applicant's written requirements. (Roulo, Applicant Prepared Testimony at 2-3, ff. Tr. 1356.) This program is an integral part of Appli-cant's Industrial Security Program.

140. Applicant has a full time staff member who conducts periodic audits of all contractors who require unes-corted access to Applicant's nuclear plants. Applicant's quality assurance department personnel also conduct periodic audits of contractor screening procedures and practices (Roulo, Applicant Prepared Testimony at 3-4, ff. Tr. 1356) as does the NRC. (Roulo, Tr. 1380-1381.)

141. During the hearing, an in camera session was held for the purpose of questioning both the Applicant's and the NRC Staff's witnesses on confidential plant security infor-l mation. The witnesses were thoroughly questioned on the manner in which a person can enter or leave the plant site or obtain access to various locations within the site. (Roulo, Skelton, In Camera Tr. 13-52.) Because of the sensitive nature of this information, we will not detail what the witnesses said. The Board is, however, satisfied that Applicant's security system

/~h v

i

l at Byron Station provides the necessary measures to protect O

V against any sabotage or espionage by a contract worker. In view of the limited scope of this issue, the Board sees no need to provide separate findings on this subject which are main-tained confidential.

Design of Byron Station 142. The League contends generally that high radia-tion doses at Byron are unavoidable because of the station's design. The only station components alluded to by the League's witness, however, were the steam generators. (Morgan, League Prepared Testimony at 19-20, ff. Tr. 1515.) During voir dire examination the League's witness admitted that he has never participated in designing either a nuclear power plant or a steam generator (Morgan, Tr. 1476), and it was clear, from the tenor of his testimony on cross-examination, that he had little or no familiarity with the design of the steam generators at Byron Station.

143. The Applicant's and NRC Staff's evidence shows that Byron Station was designed to assure that its operation and expected maintenance does not result in radiation doses in excess of the levels specified in 10 CFR S 20.101. The design of the Station also takes into account other NRC regulations regarding occupational radiation exposure, specifically the ALARA provision of 10 CFR S 20.1. (Lamastra, NRC Staff Pre-pared Testimony at 3-4, ff. Tr. 1883; Lahti, Applicant Prepared Testimony at 2-3, ff. Tr. 1830.)

144. The design of Byron utilizes four basic design d principles which reduce radiation exposure: (1) provide shield-ing, in the form of water, concrete, or steel doors (Lahti, Tr.

1843), between the radioactive source and accessed areas; (2) provide distance between the radioactive source and work-ers; (3) reduce the time spent by the worker in the radiation field; and (4) remove the radioactive source material. (Lahti, Applicant Prepared Testimony at 3-4, ff. Tr. 1830.) Sections 12.3 through 12.3.2.1.8 of the FSAR describe the specific features of Byron Station which have been designed to take into account ALARA principles. (Lamastra, NRC Staff Prepared Testi-many at 4, 14-15, ff. Tr. 1883; Lahti, Applicant Prepared Testimony at 5, ff. Tr. 1830.)

145. The design of Byron Station provides for shield-ing at all three locations where neutrons may be a hazard.

Beta rays, which can be emitted when fission products in the i

! coolant stream decay, are shielded by the pipe containing the j fluid. Shielding also is provided in the plant design to protect against gamma rays. (Lahti, Tr. 1832-1835.)

146. The Byron FSAR states that station decontami-nation probably will be necessary at least once during the l

station's life. Details on how this decontamination will occur l

have not yet been formulated because decontamination probably will not be done within the first twenty years of Byron's life.

The design does, however, have shielding, draining, and back-flushing features which will permit decontamination to occur when needed. (Lahti, Tr. 1857-1864, 1875.)

l

h q

147. The Byron steam generator design by Westinghouse O adopts the same design philosophy expressed above in that the design minimizes personnel radiation exposure by providing distance between the radiation source and workers, by reducing the time spent by workers in the radiation field, and by mini-mizing radiation source levels in the steam generators. Mecha-nical and metallurgical design features have been incorporated in the Byron steam generators to directly address these con-siderations. (Conway, Applicant Prepared Testimony at 5, ff.

Tr. 1309.) They are the elimination of " crud traps" (Conway, Applicant Prepared Testimony at 5-6, ff. Tr. 1309; Conway, Tr.

1311-1312, 1331-1332, 1334-1335), material selection and con-trol (Conway, Applicant Prepared Testimony at 6-7, ff. Tr.

1309; Conway, Tr. 1320-1321, 1339-1340), the placement of the primary channel head external drain (Conway, Applicant Prepared Testimony at 7-8, ff. Tr. 1309), adding primary nozzle closure rings (Conway, Applicant Prepared Testimony at 8, ff. Tr. 1309; Conway, Tr. 1313, 1351-1352), the location, size, and number of manway access openings (Conway, Applicant Prepared Testimony at 9-10, ff. Tr. 1309; Conway, Tr. 1329), and the design and placement of secondary side instrument and access openings.

l (Conway, Applicant Prepared Testimony at 10-11, ff. Tr. 1309; Conway, Tr. 1319-1322, 1329-1330.)

148. Byron Station also has provided for the follow-ing additional measures to keep any radiation exposure from maintenance work on the steam generators ALARA: (1) the in-S stallation of a manway handling device; (2) the use of. remote

%)

_. - - _ . - - = - - .

l l

l control inservice inspection equipment, including handling devices for inspecting steam generator tubes; and (3) the fil-tering of the air from inside the steam generator manways.

(Van Laere, Applicant Prepared Testimony at 22-23, ff. Tr.

1707; Van Laere, Tr. 1711-1713.) Applicant also will modify the steam generator, for the purpose of eliminating the possi-bility of flow-induced vibration, before the start up of the plant. (Butterfield, Applicant Prepared Testimony at 5, ff.

Tr. 5908; Blomgren, Applicant Prepared Testimony at 17, ff. Tr.

4126; Blomgren, Tr. 4120.)

149. The Board finds that the Applicant's ALARA program is sufficient to maintain occupational radiation doses ALARA. Applicant has detailed and comprehensive procedures for personal dosimetry, in-plant radiation monitoring, record maintenance, and protection of declared pregnant women. Byron Station's health physics staff is of adequate size and will be properly trained. All employees will receive thorough training on how to keep doses ALARA. Every effort will be n.ade to keep temporary employees from exceeding their legal dose limits.

Byron Station is well protected against industrial sabotage and has been designed to maintain doses ALARA. Finally, over-l whelming scientific evidence establishes that the FES has adequately assessed the risks to workers of exposure to radia-tion.

O

APPENDIX 1

/7 d Comprehensive Index to Exhibits )

I A) LICENSING BOARD EXHIBITS Exhibit No. 1: " Modified Mercalli Intensity Scale of 1931", from " Earthquake History of the United States" (U.S. Coastal & Geodetic Survey).

Exhibit No. 2: NUREG/CR-3090, " Evaluation of Waterhammer Potential in Preheat Steam Generators".

Exhibit No. 3. NUREG-0654, FEMA-REP-1, Rev. 1, " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans & Preparedness in Support of Nuclear Power Plants", Nov. 1980.

Exhibit No. 4: Stipulation regarding testimony of Michael Zeise.

B) COMMONWEALTH EDISON COMPANY EXHIBITS Exhibit No. 1: " Rock Mechanics Research Requirements for Resource Recovery, Construction, and Earthquake-Hazard Reduction", Chapter 4: " Determination c f in situ Stress".

Exhibit No. 2: " Radiation Protection Standards", Feb-ruary 28, 1982.

Exhibit No. 3: " Policy & Procedures for Maintaining occupational Radiation As Low As Reasonably Achievable",

November 1, 1981.

Exhibit No. 4: " Instructor's Guide for Nuclear General Employee Training", January 21, 1983.

Exhibit No. 5: Handwritten copy of HC-QA-23, Hunter Intercompany Correspondence re: use of design tolerances and QCWI acceptance of as-built data in the insulation of component supports.

Exhibit No. 6: Letter dated October, 1974 from Bill Wellborn & Paul Hardenbrook re: impementation of recom-l mendations from audit 059-3.

Exhibit No. 7: "Second Follow-up report for audit 059-3, Hangar Process Control".

U l

A-1

B) COMMONWEALTH EDISON COMPANY EXHIBITS (cont.)

() Exhibit No. 8: Letter dated June 24, 1982 from Norelius to C. Reed, with attached inspection report 82-05.

Exhibit No. 9: Personal resume of Lee Alan sues.

Exhibit No. 10: Reactor Trip Breaker Logic Diagram.

Exhibit No. 11: Pictures & diagrams from Westinghouse vendor manual pertaining to breakers.

Fig. 1: Typical installation of Westinghouse DS Breaker.

Fig. 5: Front of DS Breaker.

Fig. 6A': Breaker with front panel in place.

Fig. 6B: Breaker with panel removed.

Fig. 7: Side view of breaker with levering device.

Fig. 8: Rear of breaker showing main disconnects.

Fig. 23: Operation of shunt trip coil.

Fig. 67, 68: Under-voltage trip devices.

Fig. 39, 40: Front & rear views of moveable contacts.

Exhibit No. 12: Reactor trip breaker. (Not admitted into evidence.)

Exhibit No. 13: BHP 4200-005, " Preventative maintenance Inspection of 480V SubBreakers type DS". Procedure cover-ing type DS breakers in effect prior to Salem occurrence.

Exhibit No. 14: BHP 4200 -015, Revision Zero, "Preventa-tive Maintenance Inspection-Reactor Trip Breakers".

Exhibit No. 15: Example of deviation report.

Exhibit No. 16: " Reactor Trip or Safety Injection, Rev. O, BEP-O".

Exhibit No. 17: EPRI, NP-2704-SR, Special Report, "PWR Secondary Water Chemistry Guidelines", October, 1982.

Exhibit No. 18: " Evacuation Time Estimates Within the

Plume Exposure Pathway", December, 1982.

Exhibit No. 19: " Illinois Plan for Radiological Acci-dents", IPRA Vol. 6-Byron Station.

Exhibit No. 20: IPRA Emergency Response Training Plan Matrix.

Exhibit No. 21: " Illinois Plan for Radiological Acci-dents", IPRA Vol. I.

Exhibit No. 22: " Byron IPRA Project-1983-Bar Graph."

A-2

f C) NRC STAFF EXHIBITS n

U (Unnumbered): Illinois State Geological Survey Circular 491, page 16.

Exhibit No. 1: NRC Staff's Safety Evaluation Report for the Byron Station, February, 1982.

Exhibit No. lA: Supplement No. 1 to the NRC Staff's Safety Evaluation Report for the Byron Station, dated March 1982.

Exhibit No. 1B: Supplement No. 2 to NRC Staff's Safety Evacuation Report for the Byron Station, dated January, 1983.

Exhibit No. 2: Final Environmental Statement related to the operation of Byron Station, NUREG-0848, dated April, 1982.

Exhibit No. 3: (Marked for identification only.) Figure 3.35, Diagram of Zion Reactor Building, from NUREG-0850

" Preliminary Assessment of Core Melt Accidents at the Zion and Indian Point Nuclear Power Plant and Strategies for Mitigating their Effects," November, 1981. (ff. Tr. 6791)

D) ROCKFORD LEAGUE OF WOMEN VOTERS EXHIBIT Exhibit No. 1: Letter dated June 27, 1980 from James G.

Keppler, NRC Region III, to James J. O'Connor, Common-wealth Edison Company, with attachments: Appendix A Significant Appraisal Findings; Appendix B Notice of Violation; and IE Inspection Report No. 50-295/80-05 and No. 50-304/80-04.

E) JOINT INTERVENORS' EXHIBITS - (ROCKFORD LEAGUE OF WOMEN VOTERS AND DAARE/ SAFE)

Exhibit No. 1: NRC inspection Report of Braidwood, No.

50-456/80-12, October, 1980.

Exhibit No. 2: Letter from James G. Keppler, NRC Region III, to Cordell Reed, Commonwealth Edison Company, dated December 22, 1980, with cover sheets from six IE Reports (Byron numbers 50-454/80-22, 50-455/80-21.

Exhibit No. 3: Letter from R.F. Heishman, NRC Region III, to Byron Lee, Comi.Mnwealth Edison Company, dated Octo-ber 15, 1978, with attachments; September 28, 1978 letter from Cordell Reed (rith attachments); August 31, 1981 letter from R.F. Hishman, NRC Region III to Byron Lee, Commonwealth Edison, with attached Notice of Violation and Inspection reports 50-454/78-07, 50-455/78-07.

A-3

E) JOINT INTERVENORS' EXHIBITS (cont.)

O V Exhibit No. 4: Letter dated January 26, 1980, from James Keppler, NRC Region III to Cordell Reed, Commonwealth Edison Company, with attached Notice of Violation & Inspec-tion Reports 50-454/79-18 and 50-455/79-18.

Exhibit No. 5: Letter dated A1.ril 17, 1981, from James G.

Keppler, NRC III, to Cordell Reed, Commonwealth Edison Company, with attached Notice of Violation and Inspection Reports 50-454/80-25 and 50-455/80-23.

Exhibit No. 6: Letter dated February 2, 1983, from James G.

Keppler, NRC Region III, to James J. O'Connor, Commonwealth Edison Company with attached Notice of Violation and Proposed Imposition of Civil Penalties, and Inspection Reports 50-456/82-05 and 50-457/82-05.

Exhibit No. 7: Letter dated February 16, 1983, from James G. Keppler, NRC Region III, to James J. O'Connor, Commonwealth Edison Company, with attached Notice of Violation & Proposed Imposition of Civil Penalties, Dresden

& Quad Cities Nuclear Power Stations.

Exhibit No. 8: Letter dated December 30, 1980, from James G. Keppler, NRC Region III, To Byron Lee, Common-wealth Edison Company, with attached Notice of Violation and Inspection Report 50-454/80-04, 50-455/80-04.

Exhibit No. 9: Draft final SAI Report, " Valve Impact Analysis of Recommendations Concerning Steam Generator Tube Degradations and Rupture Even* ", September 23, 1982.

Exhibit No. 10: (There was no Joint Intervenor's Exhibit No. 10 identified.)

Exhibit No. 11*: Ambulance Service Computer Printout, Region I.

Exhibit No. 12*: Ambulance & Hospital Survey Cover letter.

The Table of Contents for the April 22 transcript (Tr. 5563) indicates that Joint Intervencrs' Exhibits Nos. 11 &

12 were admitted at Tr. 5781. However, from Tr. 5781 it is clear that these exhibits were only identified.

A-4

E) JOINT INTERVENORS' EXHIBITS (cont.)

Exhibit No. 13**: Affidavit of Paul Holmbeck.

Exhibi_t No. 14**: Radiological Emergency Response Survey-Ambul i Medical Services.

Exhibi.f.

e. 15**: Affidavit of Gary Montel.

Exhibit No. 16**: Affidavit of J. Michael Maloney.

Exhibit No. 17**: Affidavit of Charles Lamb.

Exhibit No. 18**: Affidavit of David Turner.

Exhibit No. 19**: Affidavit of David Miller.

Exhibit No. 20**: Testimony of James L. Murphy.

• • Admitted pursuant to stipulation of the parties, as approved by the Licensing Board. (Tr. 6854-6860). These documents have been marked on their face to indicate stricken

() portions.

A-5

1 i

)

l 4

APPENDIX 2 Comprehensive Index of Witnesses i

Prepared Issue Sponsor Witnesses Testimony Seismology Applicant Alan K. Yonk ff. Tr. 478 l Anand K. Singh ff. Tr. 431 i

NRC Staff Ina B. Alterman ff. Tr. 753

Robert L. Rothman ff. Tr. 760 League Henry H. Woodard ff. Tr. 548 l

Waterhammer Applicant Robert Carlson ff. Tr. 930 Richard Pleniewicz ff. Tr. 896 NRC Staff Aleck W. Serkiz ff. Tr. 940 Intervenors (none)

ALARA Applicant Frank Rescek ff. Tr. 1157

Lawrence Conway ff. Tr. 1309 Jerome L. Roulo ff. Tr. 1356 Jacob I. Fabrikant ff. Tr. 1399 James R. Van Laere ff. Tr. 1707 Gerald P. Lahti ff. Tr. 1830 1

NRC Staff Michael A. Lamastra ff. Tr. 1883*

l Edward F. Branagan, Jr. ff. Tr. 1883*

Robert F. Skelton ff. Tr. 1883*

Intervenors Karl Z. Morgan ff. Tr. 1515 a

i Steam Generator Applicant John C. Blomgren ff. Tr. 4126 Tube Integrity Mehendra R. Patel ff. Tr. 4126 Daniel Malinowski ff. Tr. 4126

, Michael J. Wooten ff. Tr. 4126 Lawrence Conway ff. Tr. 4126 i

^

Jointly Prepared Testimony.

4 A-6 i

_ , , . . . . _ _ . , . . - _ , . - ...m,.,-. , , , ,, , , , , _ _ ~__ ,_.,__.,,y_,.,v._,_,y,,_,,,_.,,......m__m.,,...,___,,,.,..,-...-,,.,,__.,.y.._,____ . , _ _

Prepared Issue Sponsor Witnesses Testimony Steam Generator Applicant Thomas F. Timmons ff. Tr. 5908 Tube Integrity Wilson D. Fletcher ff. Tr. 5908 (continued) Michael Hitchler ff. Tr. 5908 Laurence D. Butterfield ff. Tr. 5908 Roldolpho Pai11 aman ff. Tr. 4818 Kenneth J. Green no prepared testimony (Tr. 6209)

NRC Staff Ledyard B. Marsh ff. Tr. 4473*

Jai Raj N. Rajan ff. Tr. 4473**

Louis Frank ff. Tr. 4473**

Conrad McCracken ff. Tr. 4473 League Dale G. Bridenbaugh ff. Tr. 6406 Emergency Applicant Jean L. McCluskey ff. Tr. 4834*

Planning Thomas J. Horst ff. Tr. 4834*

John C. Golden ff. Tr. 5035 David L. Smith ff. Tr. 5170 David D. Ed ff. Tr. 5174 E. Erie Jones ff. Tr. 5444 NRC Staff Thomas Urbanik II ff. Tr. 5391 Monte B. Phillips ff. Tr. 5509

] Gordon L. Wenger ff. Tr. 5511 Intervenors Thomas Bowes ff. Tr. 5622 Paul Holmbeck ***

Gary Montel ***

J. Michael Maloney ***

Charles Lamb ***

David Turner ***

David Miller ***

James L. Murphy ***

Joel Cowen not admitted (Tr. 6106)

Carl Swon withdrawn (Tr. 5652)

I

• Jointly Prepared Testimony.

• • Submitted jointly prepared testimony and individually prepared testimony.

• • • Admitted as Joint Intervenors' Exhibits pursuant to stipulation of parties and as approved by the Board. Tr. 6854-60. See Comprehensive Index to Exhibits.

A-7

i Prepared Issue Sponsor Witnesses Testimony Quality Assurance / Applicant Louis 0. Del George ff. Tr. 2344 Quality Control Cordell Reed ff. Tr. 2594 Walter J. Shewski ff. Tr. 2364 Michael A. Stanish ff. Tr. 2619 Robert E. Querio ff. Tr. 2714 John Mihovilovich ff. Tr. 2750 Richard Barnhart ff. Tr. 2797 Donald Pope ff. Tr. 2823 Malcolm L. Somsag ff. Tr. 2883 NRC Staff John G. Spraul ff. Tr. 3562 William L. Forney ff. Tr. 3586*

D.W. Hayes ff. Tr. 3586*

James E. Konklin ff. Tr. 3586*

Cordell C. Williams ff. Tr. 3586*

Isa T. Yin ff. Tr. 3586*

Intervenors Peter Stomfay-Stitz ff. Tr. 2939 Michael Smith ff. Tr. 3243 Daniel Gallagher ff. Tr. 3459 Michael Zeise ****

Class 9 Accidents Applicant Saul Levine ff. Tr. 1930 Saul Levine (Rebuttal) ff. Tr.

(6/25/83)

NRC Staff L.G. Hulman ff. Tr. 2091*

Millard L. Wohl ff. Tr. 2091*

Scott Newberry ff. Tr. 2091*

Edward F. Branagan, Jr. ff. Tr. 2091*

Intervenors (none)

Liquid Pathways / Applicant George C. Klopp ff. Tr. 6750 Hydrology Lawrence Holish ff. Tr. 6750 Gerald P. Lahti ff. Tr. 6750 NRC Staff Richard Codell ff. Tr. 6649*

Gary Staley ff. Tr. 6649*

Intervenors Bernard J. Wood ff. Tr. 6879 k

Jointly Prepared Testimony.

() Admitted as Board Exhibit 4 pursuant to Stipulation of the par-ties and as approved by the Board. Tr. 7025.

A-8

O The foregoing document, " Commonwealth Edison Company's Proposed Findings of Fact and Conclusions of Law Regarding Seismology, Waterhammer and ALARA" is respectfully submitted by the undersigned attorneys for Commonwealth Edison Company.

L Michael I. Miller s, ll d .' '

/ V M( Bielawski

//um 6. W Victor G. Copeland t ISHAM, LINCOLN & BEALE Three First National Plaza Chicago, Illinois 60602 (312) 558-7500 Dated: May 31, 1983 0

C00',UjD oc i UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION.g3 21-3 00:01 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD

?

In the Matter of )

)

COMMONWEALTH EDISON COMPANY ) Dc,cket Nos. 50-454 OL

) 50-455 OL (Byron Nuclear Power Station, )

Units 1 & 2) )

CERTIFICATE OF SERVICE The undersigned, one of the attorneys for Common-wealth Edison Company, certifies that he filed the original and two copies of the attached " COMMONWEALTH EDISON COMPANY'S PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW REGARDING SEISMOLOGY, WATERHAMMER, AND ALARA" with the Secretary of the Nuclear Regulatory Commission and served a copy of the same on

each of the persons at the addresses shown on the attached l

service list. Service on the Secretary and all parties, unless otherwise indicated, was ruade by deposit in the U.S. Mail, first-class postage prepaid, this 31st day of May, 1983.

YW h*

One of the attorneys for Commonwealth Edison Company I

ISHAM, LINCOLN & BEALE Three First National Plaza Chicago, Illinois 60602 (312) 558-7500 D

SERVICE LIST COMMONWEALTH EDISON COMPANY -- Byron Station ~

Docket Nos. 50-454 and 50,-455

• Mr . Ivan W. Smith Secretary Administrative Judge and Chairman Attn: Chief, Docketing and Atomic Safety and Licensing Board Panel Service Section

. Room 428 U.S. Nuclear Regulatory Commission Washington, D.C. .20555 East West / West Towers Bldg.

4350 East West Highway

• Ms. Betty Johnson Bethesda, MD 20114 1907 Stratford Lane Rockford, Illinois 61107

• Dr . Richard F. Cole Atomic Safety and Licensing

• Ms. Diane Chavez Board Panel SAFE U.S. Nuclear Regulatory Commission Washington, D.C. 20555 326 North Avon Street Rockford, Illinois 61103 Atomic Safety and Licensing-

• Dr . Bruce von Zellen Board Panel U.S. Nuclear Regulatory Commission Department of Biological Sciences Washington, D.C. 20555 Northern Illinois University DeKalb, Illinois 60115 Chief Hearing Counsel Office of the Executive

• Joseph Gallo, Esq.

Legal Director Isham, Lincoln & Beale Suite 840 U.S. Nuclear Regulatory Commission 1120 Connecticut Ave.,

Washington, D.C. 20555 N.W.

Washington, D.C. 20036 Dr. A Dixon Callihan **Douglass W. Cassel, Jr.

Union Carbide Corporation Jane Whicher P.O. Box Y BPI Oak Ridge, Tennessee 37830 Suite 1300

• Mr. Steven C. Goldberg 109 N. Dearborn Chicago, IL 60602 Ms. Mitzi A. Young Office of the Executive Legal

• Ms. Patricia Morrison Director U.S. Nuclear Regulatory Commission 55G8 Thunderidge Drive Washington, D.C. Rockford, Illinois 61107 20555 Atomic Safety and Licensing ** Mr. David Thomas Appeal Board Panel 77 South Wacker Chicago, IL 60621 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Via Express Mail Via Messenger