ML20040C047
| ML20040C047 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 01/13/1982 |
| From: | Hartley F ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | |
| Shared Package | |
| ML20040C043 | List: |
| References | |
| NUDOCS 8201270203 | |
| Download: ML20040C047 (22) | |
Text
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w UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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ARIZONA PUBLIC SERVICE
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COMPANY, et al.
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Docket Nos. STN 50-528
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STN 50-529 (Palo Verde Nuclear Generating )
STN 50-530 Station, Units 1, 2 and 3)
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AFFIDAVIT OF F. W. HARTLEY ON CONTFSTION NO. 6B STATE OF ARIZONA
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) ss.
County of Maricopa )
I, F. W. Hartley, being duly sworn, upon my oath state as follows.
1.
I am employed by Arizona Public Service Company as Manager of Nuclear Operations.
2.
In such capacity I am responsible for the day-to-day operation-and maintenance of the Palo Verde Nuclear Generating Station:("PVNGS"). My resume is. set forth in Attachment FWH-1.
3.
This affidavit is made with reference to Intervenor Patricia Lee Hourihan's Contention'No. 6B concerning the subject of ATWS.
4.
ATWS is an acronym for " anticipated transients without scram."
5.
Anticipated transients are deviations from normal operating conditions which can be foreseen as probable occurrences during the service life of a nuclear power plant.
8201270203 820115 PDR ADOCK 05000528 o
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6.
An ATWS event refers to the failure of the reactor pro-tection system to shut down the reactor following the occurrence of an l
anticipated transient requiring reactor shutdown.
7.
ATWS is an unresolved generic safety issue which has been included by the NRC Staff in its " Task Action Plans for Generic Activities,"
l NUREG-0371 (November 1978), as Task No. A-9.
8.
The NRC staff has issued its Safety _ Evaluation Report related to the operation of the Palo Verde Nuclear Generating Station Units 1, 2, and 3, NUREG-0857 (November 1981) aad its Safety Evaluation Report related to the final design of the Standard Nuclear Stean Supply Reference System, CESSAR System 80, NUREG-0852 (November 1981).
9.
The Staff's review of AIWS for PVNGS is set-forth at pages 15-1 to 15-2 of the Safety Evaluation Report for FVNGS.
10.
In its Safety Evaluation Report for PVNGS at page 15-2, the NRC Staff has identified two procedural requirements which in the Staff's I
view serve as an acceptable basis for operation of PVNGS pending completion j
of any plant modifications ultimately required by the Commission in its final resolution of ATWS as a generic safety issue.
j 11.
As set forth at page 15A-29 of the PVNGS Final Safety Analysis-Report, Joint Applicants have committed to meet the NRC Staff's ATWS pro-cedural requirements set forth at page 15-2 of the Safety Evaluation Report for PVNGS.
12.
As set forth at page 15A-29 of the PVNGS Final Safety Analysis Report, Joint Applicants have committed.to have the required procedures im-plementing the Staff's requirements available for NRC review at least 60 days prior to fuel loading.
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F. W. Hartley, Yb Subscribed and sworn to before ce this /3 -
day of
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, r.,y o,1, 1982.
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Notary Pubiic W
My cocnission expires:
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Attachment.EWd-l NAME:
F. W. Hartley ADDRESS:
7820 N. 107 Dr., Glendale, Az. 85307 EDUCATION & MILITARY SERVICE:
B.S. Degree La Management - Arizona State University Retired USN Master Chief Steam Propulsion PRIOR EMPLOYERS:
United States Navy Connecticut Yankee Atomic Power Company Northeast Utilities Arizona Public Service
SUMMARY
Thirty plus years experience in operation, mairtenance and management of fossil and nuclear power plancs. The past twenty-two years have been in the nuclear field -
six in the Navy and sixte =a in the commercial nuclear power field.
Certified as a Navy Reactor operator in 1960 and an NRC Senior Reactor Operator License holder from 1967 to 1976.
PROFESSIONAL HISTORY:
5/81 - Present:
Manager of Nuclear Operations, Arizona Public Service Co.
10/76 - 5/81:
Manager of Palo Verde Nuclear Generating Station 1/76 - 10/76:
Superintendent of the Millstone Nuclear Power Station, Northeast Utilities, Organization size 250 personnel j
12/1/69 - 1/76:
Superintendent of Connecticut Yankee Atomic Power Station, Haddam, Conn., Organization sizc 95 personnel 9/8/68 - 12/1/69:
Assistant Superintendent, Connecticut Yankee 10/1/67 - 9/8/68:
l Operations Supervisor, Connecticut Yankee l
l l
Page Two..........F. W. Hartley 3/21/66 - 10/1/67:
Shift Supervisor, Connecticut Yankee 12/20/62 - 3/15/66:
Nuclear Chief Operator and Engineering Watch Officer on the U.S.S. Long Beach (CGN-9), USN 5/1/60 - 12/1/62:
Chief Operator, Engineering Watch Officer and Shift Training Coordinator at AIW, Idaho Falls, Idaho (National Reactor Testing Station). USN 6/1/59 - 5/1/60:
Nuclear Power academic and prototype schools, Vallejo, California and Idaho Falls, Idaho. USN PROFESSIONAL AFFILIATIONS:
American Nuclear Society - Chairman of ANS 55.4, Member Executive Ceanittee ROD Division EEI - Member Nuclear Power Committee since 1968, Past Chairman, Nuclear Operating Experience Task Group under the Nuclear Power Subcommittee.
Founder and past Chairman, Western States Plant Managers Association PERSONAL:
Height 5' 10", Weight 175 lbs.
Heath - Excellent Marital Status - Married Children - Four L
NUREG-0857 Safety Evaluation Report related to the operation of Palo Verde Nuclear Generating Station, Units 1, 2, and 3 Docket Nos. STN 50-528, STN 50-529, and STN 50-530 Arizona Public Service Company, et al.
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation November 1981
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15 ACCIDENT ANALYSES
15.1 INTRODUCTION
The analyses of normal operation, anticipated transients and generic accidents are provided in the CESSAR System 80 FSAR.
Staff evaluations for those transients and accidents within CESSAR Scope are provided in the CESSAR SER.
15.2 NORMAL OPERATION AND ANTICIPATED TRANSIENTS The staff evaluation is presented in the CESSAR SER.
15.3 LIMITING ACCIDENTS Staff evaluations for the following accidents 15.3.1 through 15.3.8 are presented in the CESSAR SER.
15.3.1 Steam Line Breaks 15.3.2 Feedwater System Pipe Breaks 15.3.3 Reactor Coolant Pump Shaft Seizure 15.3.4 Reactor Coolant Pump Shaft Break 15.3.5 Inadvertent Opening of a Pressurizer Safety Value 15.3.6 Double-Ended Break of a Letdown Line Outside Containment 15.3.7 Steam Generator Tube Rupture 15.3.8 Loss-of-Coolant Accident 15.3.9 Anticipated Transients Without Scram A number of plant transients can be affected by a failure of the scram system to function.
For a pressurized water reactor, the most important transients affected include loss of normal feedwater, loss of electrical load, inadvertent control rod withdrawal, and loss of normal electrical power.
In September 1973, the staff issued WASH-1270, " Technical Report on Anticipated Transients Without Scram tor Water-Cooled Power Reactors," establishing acceptance criteria for anticipated transients without scram.
In conformance with the requirements of Appendix A to WASH-1270, and as discussed in the CESSAR SER, Section 15.3.9, Combuntion Engineering submitted an evaluation of anticipated transients without scram in Topical Report CENPD-158, " Topical Report Anticipated Transients Without Scram."
On December 9, 1975, tne staff issued a report, " Status Report on Anticipated Transients Witnout Scram for Combustion Enegineering Reactors."
In response, Combustion Engineering issued Revision 1 to CENPD-158 in May 1976.
A reeva~luation of the potential risks from anticipated transients without scram (ATWS) has been publishea in NUREG-0460, Volume 1 through 4.
The status of this NUREG is described below:
)
(1)
In March 1980 the 4th Volume of NUREG-0460 was issued by the NRC staf f.
The recommendations included design criteria for plants such as PVNGS and recommended rulemaking to establish such criteria.
(2) Ine NRC staff presented its recommendations on ATWS to the Commission, inc'.uding the recommendation for rulemaking, in September 1980.
15-1 b
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(3) After deliberation, the Commission will act on the matter.
Whether it will agree tc rulemaking is s)eculative at this time.
If rulemaking is i ni ti ated by the Com...i ssion, the staff would expect that any rule adopted would include an implementation plan for all classes of plants.
As discussed in the CESSAR SER, all reference plants, including PVNGS 1-3, would be required to provide plant modifications in conformance with ATW5 criteria and schedular requirements provided in the rule or as adopted by the Commission.
The following discussion presents the bases for vperation of PVNG5 1-3, prior to the adoption of a rule.
In NUREG-0460, Volume 3, the staff states:
"The staff has maintained since 1973 (for example, see pages 69 and 70 of WASH-1270) and reaf firms today that the present likelihood of severe consequences arising from an ATWS event is acceptably small and presently there is no undue risk to the public from ATWS.
This conclusion is bascd on engineering judgment in view of:
(a) the estimated arrival rote of anticipated transients with potentially severe consequences in the esent of scram failure, (b) the favorable operating experience with current scram systems; (c) the limited number of operating reactors."
In view Gi these considerations and the staf f expectation that the recessary plant mod 5 fica-tions will be implemented in one to four years f ollowing a commission descision on a"Licipated tran2ients without scram, the staff has generally concluded that pressurized water plants can continue to operate because the risk from anticipated transient without scram events in this time period is acceptably smali.
As a prudent course, in order to further reduce the risk from anticipated i
transient w:thout scram events during the interim period before completing the plant modification determined Dy the Commission to be necessary, the staff, as discussed in the CESSAR SER, required that the following steps be taken:
(1) Develop emergency procedures to train operators to recognize anticipated transient without scram event, including consideration of scram indicators, l
rod position indicators, flux monitors, pressurizer level and pressure j
indicator, and any other alarms annunciated in the control room with emphasic on alarms not processed through the electrical portion of the reactor scram system.
(2) iroin operators to take actions in the event of an anticipated transients nithout scram, including consideration of manually scramming the reactor ty u;'ng the manual scram button, prompt actuation of the auxiliary Teedwater system to assure delivery to the full capacity of this system, and initiation of turbine trip.
ihe operator shoald also be trained to
'nitiate boration by actuation of the high pressure safety injection system to bring the facility to a safe shutdown condition.
The statt considers these procedural requirements an acceptable basis for interim operation of the facility based on our understanding of the plant response to postulated anticipated trannsients without scram events.
The applicant has committed to develop emergency procedures for and train operators to respond to anticipated transients without scram per requirements (1) ard (2) above.
The staff finds this acceptable, 15-2
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57521
- proposed Rules Vol. A No 22ti Tinia). Nurmber 24.19u1 l
s s
This secbon of the FEDERAL REGISTER action has been determined to be "not 23.1982. will be considered if prat.In al contains notices to tne pubhc of the maior."
to do so. but only those comments proposed est uance of rules and The Regulatory flexibihty Act (Pub. L received on or before this date can hr regulations. Tr>e purpose of these noticeh 90-354)is not applicable to this action; assured of consideration.
- s to give interested persons an therefore. a Regulatory Flexibility ah,' r5cr to infaYopton of ADDRESSES: Comments should he I
Analysis will n t be prepared.
final This proposed action is intended t submitted in writing to the Secretary n!
- chminate an unnecessary bulletin.
the Commission. U.S. Nuclear thereby saving the Covernment the cost Regulatory Commission. Washington.
DEPARTMENT OF AGRICULTURE of periodic revisions.
D.C. 20555. Attention: Docketing and This program is listed in the Catalog Service Branch. All comments rercisnl Rural'lectrification Administration of Federal Domestic Assistance us and all referenced and other NRC lotso-Rural Electrification Loans and documents relevant to the ATWS mue 7 CFR Part 1701 Loan Guarantees.10.851-Rural will be available for public inspection in Proposed Rescission of REA Bulletin Telephone Loans and Loan Guarantees the Commission's Public Document 81-7:381-11 nd 10.852-Rural Telephone Bank Room at In7 H Suut,ggL to Washm, ston. D.C. Copics of referented CGENCY: Rural Electrincation I ritWi submissions made NRC reports may be purchased from the Ad ninistration. USDA.
pursuant to th s action will be made 85("[C n T'
l.
I r Action: Proposed rule.
available for pub,hc mspection during regular business nours at the above Regulatory Commission. Washington.
SUMMARY
- The Rural Electrification address.
0555.
Administration (REA) proposes to amend Appendix A-REA Bulletins to Dated: November 18,1981.
FOR FURTHER INFORMAT!oN CONTACT:
provide for the rescission of REA II*Id V. Hunter.
David W. Pyatt. Office of Nuclear Bulletin 81-7:381-11. " Changes or Adsu.rustroror.
Regulatory Research U.S. Nuclear Regulatory Commission. Washington.
Corrections in Line Construction."
p14 ou es-w:rma.x nw D.C. 20555. (301) 443-5960.
which has become obsolete. The euc coot mo-is-u pnmary purpose of REA Bulictin 81-regarding protection against anticipated SUPPLEMENTARY INFORMATION: Concern 7.381-11 is to provide REA Form 2;6.
~ Construction Choage Order. Sinc NUCLEAR REOULATORY transients without scram (ATWS1 REA f orm 210 was rescinded in un CONMISSION events has long been a subject of e!Iort to climmate unnccessary REA extensive and continuing study by the forms. REA Bulletin 81-7:381-11 ts 10 CFR Part 50 NRC staff.The significance of ATWS for considared to be unnecessary.
Standards for the Reduction of Risk events could' result in melting of the reactor safety is that some ATWS oatt:1 ublic comments must be received From Anticipated Transients Without by REA no later than January 25,1982.
Scram (ATWS) Events for Light Water
- reactor fuel and the release of a large AconEss: Submit written comments to cooled Nuclear Power Plants amount of radioactive fission products.
the Duector. Engineering Standards The principal benchmark for decidme AGENCY: Nuclear Regulatory whether and to what extent nuclear ~
Dnision. Rural Electrification Commission.
power plants should be modified Admmistra tion. Room 1270. South Badding. U.S. Department of ACTION: Proposed rule.
because of ATWS.related safety concerns is set forth in subsection l
Agriculture. Washington. D.C. 20250.
SUMMARY
- The Commission is i
FOR FURTHER INFORMATION CONTACT:
considering three alternatives for 1611[3] of the Atomic Energy Act. That section grants to the Commission the ilt. Edwin N. Limberger. telephone (202) amending its regulations to require authenty to " prescribe such regulations 447-7040. A~ Draft impact Analysis has improvements in the design und or order's as it may deem necessary been prepared and is available from the operation oflight. water-cooled nuclear D: rector. Engmeeting Standards power plants to reduce the likelihood of
- in order to protect health and to minimize danger to life or property /
Daision, at the above address.
failure of the reactorprotection system Throughout the history of regulating SUPPLEMENTARY INFORM ATION: Pursuant to shut down the reactor (scram) nuclear reactors. the dual concept of to the Rural Electnfication Act, as following anticipated transients and to presenting accidents and mitigatmg amended (7 U.S C. 901 et seq.) REA miti; ate the consequences of anticipated their consequences should they occur.
preposes to amend Appendix A-REA transients without scram (ATWS) i.e.. defense in depth. has been used to Belictins to provide for the rescission of events. This will reduce the overall risk achieve this ojective. Thus. conservata r REA Bulletm 81-7:361-11. " Changes or of nuc! car pewer plant operation. The design. construction. testing.
C<rachons in Line Construction." Since consequen. es of this regulation wdl be nu significant effect on the economy will to require electric utilitics to install maintenance and operation of plams.o.
required so that accidents will not cccur smce no significant increase in certain equipment in nuclear power hapt.en (i.e.. have a low probab: hts ot' est for consumers. subscnbers.
plants and. possibly. to imp!cment a occurrence).Then. to proude defenw m I.
.istnes or Government will result, rehability assurance program.
depth. the capabihty to mitigate th.dr cd smte no sigmficant impact on DATES: Comment period expires Apnl consequences is required for acudems uanemic conditions will be caused. this 23.1982. Comments receis ed after April that are postulated to occur eten thoud
i 24, 1981 / Proposed Rules Sf522
' Fed:ril R gist:r / Vol. 40. No. 220 / Tu:sday Novemb:r neither of these attemative proposed t
submitted to the Commission for rules will. if promulgated, have a i
ihe ih. sign is required to ine.lude consideration in an early version in measures,to preent them.
SECY B0409. September 4.1980. and in significant economic impact on a ATWS accidents are a cause for final form in SECY BG409C. November substantial number of small entities. The concern because a mismatch can 7.19a0.The second NRC. proposed rule alternative proposed rules affect only devnlop hetween the power generated in is a recent proposal by former NRC the licensing and operation of nuclear the reactor und the power dissipated in Chairman Joseph M. IIendrie.' Dr.
power plants. The companies that own controlled ways if the scram system ifendric's aim in starting afresh was to these plants do not fall within the scope
! ails to shut down the reuctor following try an approach.that would make forth in the Regulatory Flexibility Act or of the definition of "small entities" set a fault in the normal heat dissipation licensees look carefully at their plants functions (transient events). The power for ATWS-related vulnerabilitia and the SmallBusiness Size Standards set i
mismatch can threaten the integrity of the Lurriers that confine the fission then fix these vulnerable erecs.
out in regulat ons issued by the S.rudi i
employing systems analysis or Bus l ness Administretion at 13 CFR Part products. A core meltdown accident. in
- 21. Since these companies are dominant some cases accompanied by a failure of reliability techniq;es containment and a wrylarge r6 ease of The Commission beneves that the in their service areas. Gis proposed rule does not fall within the purview of the radioactivity is a possible outcome of hkchbood of severe consequences mmn ATWS uccident sequences.Thus, arising from an ADVS crent during the Act.
the consequences of some postulated two to four year period required to First NRC-Proposed Rule (the Staff Implement a rule is acceptably small.
Rule)
ATWS accidents are unacceptable.
This judgment is based on (a) the There have been roughly one favorable experienco with the operating The review and evaluation by the thousand reactor years of experience reactors. [b) the limited number of NRC staff of the information that has accumulated in fo.eign and domestic operating nuclear power reactors. (c) the been developed ow the past ten years commercial light. water-co9 fed reactors inherent capability of some of the on ATWS events aad of the mannerin without an ATWS accident.This experience suggests that the frequency operating VNRs to partially or fully which they should be considered in the of ATWS accidents is less than or of the mitigate the consequences of ADVs design and safety evaluation of nuclear events. (d) the partial capability of the power plants is contained in the report order of once in a thousand reactor recirculation pump trip feature to
" Anticipated Transients Without Scram 3 ears. There have been several mitigate ADVS events that has been for Light Water Reactors." NUREC-precursor events, i.e., faults detected implemented on all BWRs of high power 0460. Vohnaes 1 through 4. There are that could have given rise to ATWS esents. This suggnsts that Le buency level, and (e) the interim steps taken to two primary factors in the staff's of ATWS accidents, though less i m develop procedures and train operators evaluation.The first is the degree of Ii to further reduce the nsk from some assurance that ATWS cvents can be once in n thousand rt.Se yer s say ATWS events.On the basis of these prevented which depends on the I
not be very much less. Such trequen ;ies considerations. the Commission believes reliability of current reactor protection i
are too high for accidents of the severity that there is reasonable assurance of systems.The second is the capability of described above. Thus the NRC has safety for contmued operation until existing reactor designs to miti;; ate the determined that reductions must be implementation of a rule is complete.
consequeras of ATWS events, made in the frequency. severity, or both The implementation schedule contained The reliability of current reactor the frequency and severity of ADVS in this rule balances the need for careful protection systems has been estimated accidents.
The Nuclear Regulatory Comm,ission analysis er 1 plant modifications with l>ased on the operating experience to the des,re to carry out the ob)cctives of date and reliability analyses. Ilowever.
has under consideration three proposed i
alternative rules, each intended to the rule as soon as possible.
the very high level of reliabdity required is difficutt to demonstrate wn, h reduce the risk po:ed by ATWS Paperwork Reduct. ion Act c nfidence because it depends on i
accidents.Two of these originated A request for cicarare of any acorately determmmg the r ite of
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within the NPC, and are described application and reporting requirements common cause failures. Common cause below.The thbd is set out in a petition for rulemaking filed by twenty utilities of the alternative finally selected will be failures, volve failures of multiple
(
m submitted to the Office of Management components res@ing from a sm, gle
(" Electric Utilities Petition." PRM 50-29.
and budget under the Paperwork cause or event. Reactor protection o rn 73030. November 4.1980 and the Reduction Act (Pub. L 96-511). At the systems are carefully reviewed to supplement to the petition published on time, the SF-83 " Request for C!carance.- identify and climinate all but the most February 3.1981. 46 FR 10501). The Supporting Statement. and related unlikely commm cause failures.
utilitlen* petition will not be reproduced documentation submitted to OMB will llowever, one connon cause failure in here. however, the current period for the be available for insp(ction and copying the react r trip portion of the protection utihty petition is hereby re?pened to rua for a fee in the NRC Public Document system of a commercial r%1 car power wncurrently with that of tac two NRC l
Room at 1717 ii Street NW reactor has occurred during proposed rules for the purpose of comparing and contrasting the utility Wa shington. D.C.
approximately 1000 reactor-years of petition with the two proposed rules Regulatory Flexibility Certification operating experience.The failure was detected d.mng normal surveillance and pubbshed he:cm.Both of the NRC-in accordance with the Regulatory correcte d before any event requiring a prope ed ruten niandate improvements Flexibility Act of 1980. 5 U.S.C. 605(b).
reactor scram occurred. There has also in ATWS prevention and mitigation.
the Commission hereby certifies that been one partial failure to scram in n They differ in scope, approach, and commercir I power reactor. which I
rriteria.
% ihe memarandam of chorman losesA M.
UCCurred at low power and resulted in The first NRC-proposed rule is known HmJre to Conamiones Cd.ns4 Dmdford. and no core dama;;c or radiation rea.se.
l as the staff rule and is a direct j $ ^,
[ *d))"*[.,' h ^,'g M'g, Common cause failures have also outgrowth of NUREG-atco " Anticipated l
Transients Without Scram for Light comm mn, potmc oocum,ni goom., grir in occurred in other systems in nuclear Water Reactors." Volumes 14 It was smt Nw w. Nnenn.u c 1
~
Fcdcral R: gist:r / Vol. 40. No. 220 / Tuesday. November 24, 1981 / Proposed Rules 57523 power plants and other potential slightly revised form in Volume 4.The Alternative 2 for the ten older plants common cause failures in reactor intent of the proposed rule is to adopt a that began operation before late 1969.
protection systems have been identified. combination of the alternatives Decause of their unique characteristics.
Ilecanse of the kiw rate of occurrence of recommended in Volume 4 (except for the staff believed that more extensive common cause failures, operating one change for reactors designed by modifications would not he appropriate experience is not, und cannot be.
Westinghouse and licensed to operate for these plants.The proposed rule does j
suf ficient to conclusively determine on a before 1984).The proposed rule would not explicitly address these plants I
statistical basis whether reactor implement the requirements in a (except in the implementation schedule).
l proicction systems are reliable enough different manner from that described in but the intent is to consider any to m.ke the probability of unacceptable Volume 4 of NUREG-4160. The forrr. of exempnons from the acceptance criteria
{
conscquences from ATWS events the requirements in the proposed rule is of the proposed rule for these older I
acceptably small. The prediction of also different from that recommende lin plants based on analyses by the
?
I common cause failures is as much art as NUREG-Gl60 in that the picposed rule licensees and eveluations similar to I
it is science. System reliability analyses specifies acceptance criteria for ATWS those conducted under the that attempt to predict the nature and mitigating systems while the required Commission's systematic evaluation frequency of common cause failures mitigating systems are specified in program (SECY-77-561 October 1977) in suffer from problems of completeness Volume 4.
context with the overal; safety of these and ar. curacy, particularly when the Alternative 1 is to make no facilities.
desired failure rate is extremely small.
modifications at all. As discussed, the Alternative 3. as modified in the While quantitative estimates of NRC has concluded that the reliability proposed role, would increase the protection system reliability provide of current reactor protection systems is reliability of the reactor trip portion of important information, the conclusion as insufficient with respect to ATWS and the reactor protection system for some to the adequacy of protection system that the probability of ATWS events is plants and provide for the mitigation of reliability must be based on engineering sufficiently great to warrant most ATWS events.The reliability of judgment. The NRC has concluded that improvements. Therefore, this the protection system would be the reliability of current reactor alternative is not represented in the ir creased in the same manner as in protection systems hus not been proposed rule.
Alternative 2. llowever, this increased demonstrated to be adequate and most Alternative 2. as modified in the reliability of'he reactor protection likely is not adequate.
proposed rule, would increase the system wei d not be required in plants The probability of severe reliability of the reactor trip portion of that have a greater capability tu mitigate consequences resulting from ATWS reactor protection systeme and improve ATWS events.The mitigation of most events is also affected by the capability the capability of existing systems to ATWS events in PWRs was expected to of nuclear power plants to mitigate mitigate some ATWS events. Reliability be accomplished as in Alternative 2.
ATWS events.This capability varies of the reac:or trip systems would be except that means would be required to depending on the design of the reactor increased by the addition of isolate the containment early in an system and the status of systems and suprieme.itcry protection systems that ATWS event upon detection of radiation j
the values of system process variables would be independent and diverse from released from failed fuel. The mitigation at the time the event occurs.The the reactor trip portion of the current capability of BWRs was expected to be capability of a plant to mitigate ATWS reactor protection systems. Diversity increased by providing automatic events can be assessed by analysis.
would be achieved by the use of initiation of the Standby Liquid Control liowever, uncertainties in the design components from different System and increase its flow capacity.
I tharacteristics of the reactor, the manufacturers. by the use of Considering the state of design and l
probability of failure of the mitigating components having different principles construction. and a balancing of public i
systems and the probability that the of operation or power sources, and by safety benefits against economic cost values of system process variables will the use of components in differer t the Commission proposes in this first be different from those assumed in the operating modes (normally energ; zed vs.
rule that plants receiving an operating analysis all combine to produce normally deenergized). This alternative license before 1984 should be required to uncertainty in the results. Therefore, the would not provide increased reliability implement Alternative 3 as modified in difficulty of demonstrating a capability of the reactivity control portion of the the proposed rule.
to adequately r.dtigate ATWS events is protection system, i.e., the control rods Alternative 4. as rzdified in the similar to the difficulty of demonstrating and control rod drives. However. in the proposed rule. would increase the that ATWS events can be prevented.
case of reactors designed by General reliability of the reactor trip portion of Based on analyses performed to date.
Electric it was proposed to increase the the ractor protection system of all however, it is clear that,in most cases, reliability of a portion of the control rod plantc and provide for the mitigation of present reactor designs have inadequate drive system. i.e., the control rod drive almest all ATWS events.The reliability capability a miti ate the consequences scram discharge volume. The capability of the protection systems would be of many petulaied ATWS evnts to mitigate ATWS events would be increased in the same manner us in should they occur.
improved by providing actuation Alternative 2.Tlic mitigation of virtually llaving concluded that improvements circuitry that is separate from the all ATWS events was expected to be are needed to reduce the probabihty of reactor protection system for some
. substantially increased by additional sescre consequences from ATWS existing systems such as primary system pressure relief capacity in the reactor events. the staff developed four relief valves, turbine trip, and auxiliary coolant system. The mitigation alternatives, three of which would feedwater in pWRs and the recirculation capability of DWRs was expected to be reduce this probability by increasing pump trip in BWRs. This alternative is increased by the addition of high increments and would require increasing very similar to the proposed rule offered capacity neutron poison injection amounts of modifications.The by the utility group.
systems. In balancing pubhc safety alternatn es were first described in
'Jhe staff proposed in Volume 4 of benefits against economic cost. the Volume 3 of NtJREG-0160 and again in NUREG-ot60 to implement only Commission proposes in this first rule I
L
57524 Fed:r:1 Regist:r / Vol. 40 No. 226 / Tuesdty, November 24, 1981 / Proposed Rul:s f
e tlI.it the e extensive design changes criteria for ecceptable evaluition where the level of safety is already high.
unihl only be practically inuirpurated in models. Since the parameters in the the Advisory Committee on Reactor t.mts not near completion und not to be evaluation model are uncertain to some Safeguards (ACRS) recommended d
lii.cnsed before 1904.
degree und some may vary over the omitting the requirement for The proposed requirements in Volume lifetime of the plant, the level of safety is improvements in the protection system L
i of NUREG-0400 were in the form of determined to a large extent by the reliability.Thus, the proposed rule specihc design changes.The proposed degree of conservatism in the Slows the protection system rule also spec;fies the design changes paran eters used in the evaluation improvements to be omitted if more required to improve the rehability of the models, which affect the conservatism conservative values of the parameters.
protection systern and the rec.nse for of the calculated consequences of such as moderator temperature omtainment isolation, but the changm.
postulated ADVS events.The proposed coefficient are used in the evaluation in nutigation capability ure required rule specifies that realistic values of models and the capability to comply i%gh the spec &ation of acceptance parameters may be used when the value with the acceptance criteria is
.riteria, criteria for evaluation models.
is known with reasonable accuracy, bLt demonstrated. In plants licensed after and mitigating system design criteria.
that parameters with large uncertainties January 1,1984 or later the time The specification of criteria requires must be conservatively treatsd.The available to design and install the hu nsees and applicants to demonstrate intent is to obtain realistic analyses of tw$lications to the protection system is that the designs of their plants are in the course of ATWS events. yet predict sufficent to ensure that the design I
compliunce and thus provides more the consequences conservatively. In process would not be compromised and assurance that the safety objective is order to ensure that the consequences of improvements in the protection systems being attained. This form also allows the most ADVS events will be within the of all of these plants is required by the designer more flexibility in design and a acceptance criteria, the proposed rule proposed rule.
greater potential for minimizing costs.
specifies that the value used for One plant modification that would be Abhough the ultimate safe'y objective parameters that vary over the lifetime of required by the proposed rule is a!reWy is ta iimit the release of radit. activity to the plant (the most significa# of these in being imphmented on boiling wate.r the environment, the acceptance criteria
!be moderator temperature wefficient) reactors. In an order dated February 21.
in the proposed rule are directed toward mod be a value that is not ext. ceded 1980. licensees of DWR plants were ensuring We integrity of the reactor over most or virtual;y all of the plant directed not to operate after December coolant system and the reactor core lifetime. In the case of the moderator 31,1980. without a recirculation pump following ADVS events.The staff temperature coefficient, the value used trip installed. I.fcensees have also been recognizes that failure to satisfy these in the evaluation model that was less directed (IE Bulletin No. 80-17 dated au.eptance criteria does not necessarily negative than the value expected to be July 3,1980, and NUREG-0737, rtault in severe radiological experienced during 90 or 99 percent of
" Clarification of TMI Action Plan unsequences and has considered the the design lifetime of the plant would Requirements") to ensure that operating additional safety margin in developing ensure that the ccnsequences of most or procedures and operator training the proposed rule. In formulating the virtually all ATWS events would not address the actions to be taken in the proposed rule. the Commission has violate the acceptance criteria.
plants as now designed if an ADVS did tunsidered the need to compare for each Although improvements in the occur.These requirements are prudent plant the offsite doses that might result capability to mitigate ADVS events measures that will reduce the risks from from ATWS events with 10 CFR Part 100 provide a significant increase in the ATWS events during the interim period guidelines. Based on conservative level of safety, there I, some uncertainty before the plant modifications generic calculations performed by the associated with this cc usion.This determined by the Commission to be stalf. there is reasonable assurance that uncertainty derives from the uncertainty necessary can be insa!!ed.
calculated offsite doses from ATWS will in the reliability of mitigating systems in particular cases, additional be within the Part 100 dose guidelines if and in the evaluation models used to requirements or earlier implementation the acceptance criteria of the proposed define them. Because of this uncertainty may be apptc;f ate. For example, rule are met. Accordingly, the the staff believes that i;aprovements in cand; dates would be those existing Commission has decided that applicants reactor protection system reliability malear power plants that are and licensees will not be required to should also be required. These considered to be at Eqh risk sites owing calculate the potential offsite modifications to present reactor to e combinWon of population density, r.nliological doses resulting from an protection systems, as with any me aerologl cal conditions and other ATWS event under i 100.11. !f only modifications to a nuclear plant, base factors.
these guidelines for calculated offsite the potential for introoucing "Ite proposed rule would provide for doses were specified, the flexibility for unrecognized failure modes that could implementation of the requirements in De designer would be increased, but the result in a decre ase in the level of stages in order to gain the greatest attainment of the safety objective would safety. A careful design process in increase in safety in the shortest time be rnore difficult to demonstrate. If conjunction with the quality assurance, and at the least cost. The modifications systems designs were specified. the verification, and test programs is to improve the reliability of the flexibility of the designer would be necessary to ensure that this will not protecton system and the mitigating reduced, and the demonstration that the occur. Ilowever, the implementation of s" stem actumn circuitry would be safety objective had been attained these improvements in reliability in required within two years of the would be generic rather than for some plants is to be accomplished effective date of the rule. In order to specified plants. Prior attempts at such a within two years. and such a short accomplish this, desenptions of the generic demonstration have been design and installation schedule might modtfications are to be submitted for unsuuessful. as discussed above.
compromise the design program. In review by the NRC within one year of
'I he level of safety, that is, whether plants such as those designed by the effective date of the rule.
most or virtually all ADVS events can Westinghouse, which have a capability Pursuant to the Atomic Energy Act of be mitigated. is specified through the to mitigate nearly all ATWS events and 1954. as amended the Energy i
m
Fid:r:I Regist:r / Vol. 46. No. 226 / Tuesday November 24, 1981 / Proposed Rules 57525 e
Rc ramixation Act of 1974. as amended. the RCS pressure boundry does not models must represent the effect of the and see.fion 553 of titic 5 of the United exceed that permitted by the " Level C failures in mitigating systems that are a States Code, notice is herchy given that Service Limit" as defined in Article NS-direct consequence of the ATWS ment adoption of the following amendments 3000 of Section ill of the ASME flailer being modeled. For facdities issued to in CFR Pari 50 is contemplated.
and Pressure Vessel Code 'and the operating licenses on or after i.maary I.
calculated deformation of RCS 19% and not standardiwd to a facibiy
!L PART 50-DOMESTIC LICENSING OF components is limited so that the at the same site that was issued an PRODUCTION AND UTILIZATION operability of components neccssary to operating hcense before Januaiy 1. WM.
)
FACILITIES safely bring the reactor to und traintain evaluation models must also represent F
- 1. The authoiity citation for 10 CFR it at a cold shutdown condition is not the effed of the likely random smgle Part 50 reads as follows:
impaired, or (D) the integrity or f...h res of acte.o romponents in i
Authorily; Sev.s.103. If>t.101.182.183. 00 D
St at. fou. 937,948,953.954, as amended (42 t.ymonstrated based on conservative D.i Tne value of parameters that vary t
ttS C.~.a 2134. 2201. 2232. 2233); secs. 202.
ussessments of tests conducted to ever h hfmme of the facility or 20r G. Wt.1244,1246 (42 U.S.C. 5842. 5846),
determine the integrity or operability of reyesent tim characterntics of N n otherwne noted. Section 5o.78 ntno components under the conditions mitigating systems that are permitied by w.en nn.ici s,s 122. wt Stat. m9. 42 U.S.C.
accompanying postulated ATWS events procedure to be moperde for any 2.N) Sci.t..ms r.o nn-rnal also issued under and based on the likely condition of the period during operation raust he S, t n.t. f an Sta t. r.4. as u rnende<l. Secn.
50 tua-T.n 102 issued under sec.100, c8 Stat.
components over their design hic.
selected so that values that would result 955; (42 U.S C. 2236). For the purposes cf sec.
(ii) Fue/ integrity. The calculated in violation of the 4.cceptance criteria 223 ra Slat. nsa. as urneneled;(42 U.Sf.
dam ge to the reactor core as a would not ba expMted to occur during 2273). I So.54fi) issued under sec.1811. (4 consequence of postulated ATWS (A) Most of the design lifetime of S:at. 949; (42 U.S.C. 22o1(ij). and il 50.70.-
events. includmg osediations of power facilities issued operating licenses 50 71 and i 50.78 issued under sec.161o,68 and flow, must be limited to ensure that Stat. 950. as amended; (42 U.S.C. 2201(o)) and the core geometry is not distorted to an before January 1.1984 or of facilities the Laws referred to in Appendices.
extent that would impair core cooling or standardized to a facility at the same
- 2. A new i SOLO is added to read as safe shutdown.
site that was !ssued an operating licxnse follows:
(iii) Radioactivity release. The before January 1.1984.
calculated release of radioactivity from (B) Almost all of the design hfetime of i 50.60 Acceptance criteria for protection the fuel rods to the reactor coolant facilities issued operating licenses on or against anticipated tranWnt without scram system during postulated ATWS events after Januay1.m except facmtks Gwents for light-water-cooled nuclear power plants.
must not exceed one percent of the radioactivity within the fuel rods of a site that was issued an operating license (a) Definitions. (1) " Anticipated pressurized water reactor or ten percent before Januan 1. m.
Transient Without Scram"(ATWS) of the radioactivity within the fuel rods (3) Mitigating System Criteria. ATM,S means an anticipated operational of a boiling water reactor.
mitigating systems must be independent.
occurwce as defined in Appendix A of (iv) Conto /nment. The calculated separate and diverse from the reactor this pe followed by the failure of the containment pressure, temperature, and protection system. ATWS mitigating reactor protection system specified in humidity resulting from postulated systems must be designed. quahfied.
General Design Criterion 20 of Appendix ATWS events must not exceed the m nitored and periodically tested to A of this part.
design values of the containment ensure continuing functional capability (2)"ATWS evaluation model" means stru;ture and components or the under the conditions accompanying the calculational framework for l
evaluating the behavior of the nuclear contained mitigating sysMms. equipment postulated ATWS events. including and components. For boiling water natural phenomena such as power plant during a postulated ATWS i
reactor pressure suppress;on earthquakes, storms, tornadoes.
}
es en t.
(3)"ATWS mitigating systems" means containments. the relief or rafety v:.lve hurricanes, and floods expected to occur discharge line flow rates and during the design life of the plant.
those systems including associated suppression pool water temperatures ATWS mitigatmg systems must be controls, instruments, power supplies
.nast be limited so that steam quenchmg automatically initiating when the and other systen.s assumed to function instabihty will not result in destructive conditions momtored rear.
when evaluating the behavior of the vibrations.
predetermm, ed levels and continue to l
nur.lcar power plant following on ATWS (v) Long.tcen shutdown onvcooling.
perform their function without operatur t
cent The reactor design must permit the action unless it can be demonstrated (b)(1) Acceptance Criteria. Each light-reactor to be safely brought to and that an operator would have adequate i
water-cooled nuclear power plant mu" maintained at a cold shutdown information and would reasonably be be designed, constructs.d. and operated condition following postulated ATWS expected within the tirne available to so that the consequences of postulated events without insertion of control rods.
take the proper corrective action.
anta mated transient without scram (2) Ero/uation Model Criteria. (i)
(4) Evcluaten models. Each applicant (ATWS) events calculated in ATWS evaluatior. rnodels must, with or licensee shall submit evaluation accordance with an AlWS evaluation reasonable accuracy or acknowledged models as defined in paragraph (b)(2) of r odel approved pursuant to paragraph conservatism. represent the actual this section. together with the (b)(4) of this section conform to the characteristics of the facility modeled description and results of the anal ses 3
following criteria:
and each significant physical and test necessary to verify the validity (i) Primary system pressure. Tne phenomenon that would occur in the of the assumptions made in preparing calculated roctor coolant system (RCS) i pressure and tmerature resulting from reactor and related systerns during the such evaluation models to the Nuc! car course of the modeled event. Evaluation Regulatory Commission for appros al hv resiulaird ATW5 events mW be (within s.x months of the effective date hmited su M tener(A) the calculated 9, g 3o s3, for approut omi incorpormon M *e rule) or prior to issuance of an t
r m ins n, amary stress anywhere in to reference.
operating license. whichever is later.
I A
]-
~
- 24. 1981 / Proposed Rules Iid:ral Register / Vol 46. No. 22G / Tuesday. November I
5752G..
rehabibly deficiences in those funttions (iii) Those modifications necessary to (T.) l'/ons for romp /innre. E.u.h reduce lhe common mode failure and systems that prevent or mitigate ATWS accidents.To cover the apphe.am or licensee shall mdimit a potennal of the control rod scram possibihty thet the reliability assorance elesuiption of all measures to be taken discharge volume in plants designed by to ensure comphance with the criteria the General Electric Company imhdin; programs mig 5! fail to correct an set Im th in paragraph (b)(11. (hll21 and diverse scram discharge volume Icvel obscure reliability defect, some th)!.1) of this section together nith stui additional requirements for ATWS sensing devices: and proposed changes in techniad (iv)Those modifications necessary to mitigation would be selectively
. peufmatmns and license amendments provide a supplementary reactor trip ATWS tolerance of reactor p' ants h.ne mandatnt. Timse improvements in as may be necessary to ensme system that is diverse from the reactor been chosen to afford an opportunity to i ompliance with these criteria to the trip portion of the current reactor Icarn imm experience without incurring Na le.o itegulatory Commission as protection system.
1.diows:
(2) Lemption. Pressurized light-a substantiallikelihood of an bl For all hght-water cooled nuclear water-cooled nuclear power plants unacceptable radiological release.
pouer plants for which operating inued operating licenses before January The NRC is exploring the possibility Imenses have been issued on or before 1.19M or standardivd to a facility at that the regulation of reactor safety may Ac;:nst 22. Itm. no later than (righteen the same site that was issued an evolve ioward regulating the process by munihs after the effective date of the operating license before lanuary 1.1tuW which licensees ensure pubhc heath and need not comply with the requirements safety and away from licensing the inh.).
(id l'or alllight wuter-cooled nuclear pun er plants for which operating of paragraph (c)(1)(iv)if the facility d tails of plant design and operation.
l conforms to the requirements of programs like the reliability assurance In enses have been isst.ed after August paragraph (b) of this section except that progtum in this proposed rule offer
- . ttu. no later than (one year after the the fraction of the design lifetime used promise of growing into a formal, effective date of the rule) or prior to to determine the ulue of parameters auditable way the NRC can determine issuance of an operating license.
must be greater than that specified in that licensees are doing a satisfactory w hichever is later, pd Imp /cmentation. Each applicant or paragraph (b)(2)(i) of this section.
job of en.uring public health and safety.
[a)Submittul. A description of the A number of diverse regulatory to ensee shallimplement those measures measures together with such proposed initiatives are supportive of this tn:ad.
necessary to ensure compliance with the
< riteria set forth in paragraph (b)(1) of changes in technical speci'ications or Among them are the requirements on license amendments as may be I censcc staffing and organization. the this u;ction on the following schedule:
necessary to ensure comph,ance w.th the proposal thut licensees employ i
UI l or ull light-water nuclear reactor criteria set forth m paragraph (c)(1) of pmbabilistic risk assesament methods l
power plants for which operating this section must be submitted to the us design and operations management I
bcenses have been issued on or before Nuclear Regulatory Commission no later tools, and the pilot studies of August 22.1969. by dates agreed upon with the NRC. These dates must be than (nine months after the effective independent design reviews.'
submitted for approval not later than date of the rule) or prior to issuance of The necessity for and content of the Iduce years after the effectwe dale of an operating license, whichever is later.
proposed rule is based on (1) operating (4) /mplementation.Those measures experience to date with power reactor the rule).
(ii) For alllight water-cooled nuclear required under paragraphs (c)(1) of th.
scrum systems. (2) system reliability is reactor power plants for wh ch section must be completed:
analysis. (3) the qualitative findings of (ill-or alllight-water cooled nuclear reactor nsk assgssment. and (4) ATWS operating licenses have been or may be reactor power plants for which
""h""I ""*b 8
issord after August 22,1!w. but befre Ohree vcars after.cffective date of the operating licenses have been or mey be
'lhere has been one partial fajh.re of ruleJ. Ell modifications shall be issued ufter August 22,1969 but before the scram system in a commerem., sower completed prior to startup following the (two years after effective date of the mactot h occurred at Browns Ferry tinit hrst n fueling that begins (three vents rule). all modifications shall be 3 on [une 28,1980. Although the uher effective date of the rule). '
completed prior to startup following the particular scram system failure mode On! For alllight. water-cooled nuclear first refueling that beg:ns (two years that caused the event is,very unlikely to reattor power plants licensed on or after after effective date of the rule).
cause a severe radiological release Dhren vears nher effective date of the (ii) For all light. water cooled nuclear accident, the event and the reviews enh L all mod,firmtions Aall be reactor power plants license d on or after
"'5"Ih"N I""" '! "*.eayd a numlnt of "Weted prior to issuauce of an Dwo scars after effective date of the whabaty defu:n nms in du M ru!c). all modifications shall be w: ram systems. These are now I,cing opewting license.
completed prior to issuance of an rectified by the mdustry subject to the
{c) Ariditionalrequirements--(1)
.b /untion:In addition to those operating hcense.
review and appr va! f the NRC Staff.
tr<porements set forth ia paragraph (b) idl Dose calculations. Applicants or me oMective I the proposed reliabihty of this serimn. each huht. water-cooled licensees are not required to cadculate
"".surance program is to insututionatiu cer leur power plant cuept as provided the potential offsite radiological doses widun the in.ensed industry the m paragruph {c)(2) of this section. must resulting from an anticipoed transient thomugh evaluadon and implementat,on i
without scram event under { 100.11 of he provided with:
O) Actuation circuitry for ATWS this chapter.
A* A sm a...r ne nu : A.non r"i..n reu'm N km mek N ^d=
Plan D m:
+ mgalina., systems that is independent Second NitC-Proposed Rule (the
" P' " "* "2W"h Pra'hns
, nd da erse from the reactor protection llendrie Rule)
"n'".,y' Mo "r%,'.,r.d Minut uunng ta.. -
n,rnoe 9 stem; The essence of the second NRC.
Aen and saa-et-:co. and -U.c c,r Ud piompt automatic containment
'W* "h D, v Heurws (IDF siin ebe md. tion mitrated by a signihr. uni suun e proposed rule is that power reactor
['f,",' ""',[,hyl '[y'i"' j 7'[,;
l of 1.uhation in the containterni resulting brensees would be required to Oom fadure of the fuel rods following implement a reliabilay assurance v..
- mo..a c w m u e nent m,.m t u n s.....
sw u.wmim program to seek out and rectify postulated ATWS events:
Federal Register / Vol. G. No. 220 / Tuesday. November 24. 1981 / Proposed Rules 57527 ni n.c h ssons of experience with compromise the availability of one of studies hke tho-e now being m.nic m b tmns important to Anys the systems required io mitigate nn response to the lhowns Feir3 un.ith ni prnention or mitigation.
ATWS cvent, or both.
These can be i.ounted o, to make a iht.aliilitv defic.icncies in saIct.s Thus, a third objectise of the recurrence of that f..ilure muile much
~
spiems ibffer substantially in the kmd reliability assurance program is to in s hkely in the future. -
mt Ircquency of opportunities to th tect scarch out and evaluate the potenti.d Calculations of the expecied nd icpair them. Some faults are self-common cause failures that might i nr.,cquences of s cry seu se rewtoi
.cmoum.ing and thus clicit prompt contribute to failure in two or mcre act; dents bas e been made in tin-mpair. Others show up in each systems whose reliability is important to /teur I,,e Sarcty Study (W ASi l-14ool '
wrteillance test. Some faults may not ATWS accident sequences. This search and other studies.The results indicate hr sn ealed by routine surveillance should embrace not only auxiliary that the accidents that could uts. l'or instance, the reliability defect systems but also human factors vin test.
reabstically be espected to result m responsil le for the partial scram failure maintenance and operations; technical lethal radiation doses outside the plant
.a Drowns Ferry could not have been specifications dealing with equipment site are those denoted as release detected in routine surveillance tests I availability: and environmental categnry 1. 2 or 3 accidents in the L. sr.rarn system.
cor 'itions in the plant.
notation of WASII-1400.These are 4dso system reliability cah.ulations by the
/. fourth objectise of the reliability the accidents that nre espected to c.one llectric Power Researuh Institute and
.rance program is to search out and substanti4J offsite propertv damage.
ethers base shown tht, component evaluate the susceptibihty of the Studies M ATWS accidents in i.nlures of reactor scram systems that redundant divisions of each safety pressurized water reactors (PWRs) are detected and corrected in each system important to ATWS prevention suggest that only a small pen.cntage of wnci!!ance test are very unlikely to or mitigation to common cause failure.
reactor scrams are limiting transients.
i.ause ATWS events. Other system Concern with common cause failure That is, only a small fraction of the 61ure modes can only be detected in modes of the scram system has been opportunities for ATWS accidents occur some but not all surveillar.cc tests. Still central to the history of the ATWS under circumstances that most severely mhers show up only in some or all cont 2 0s ersy.
challenge the ATWS tolerance of the
. naine demands upon !be system.
A ommon cause failure of an plant. In addition, the qualitative p
Some reliabihty defects cannot be electrical nature has already occurred in findings of PWR risk assessment stud;e*
detected even in genuine system a reactor scram system in a commercial stiggest that even the most limiting
&mands unless triggered by other nuclear power plant (Kahl reactor) that classes of ATWS accidents in PMs are
'..!ures. Examples of the latter category could have resulted in its failure to unlikely to produce a release categors 1.
are the hydraulic design deficiencies in operate on demand. That failure was
- or 3 radiological outcome.
de DWR scram discharge system detected during normal surveillance and In boiling water reactors (BWRs) a resca!ed by the incident ut Browns rectified. A similar common cause substantial fraction of scrams take place ieny.Such blind spots in the experience failure was detected and corrected in under circumstances that can lead to a base for safety systems can conceal the startup testing of the Monticello limiting innsient. BWRs are least serious flaws in reliability.Thus a reactor. Estimates of the upper limits of forgiving of those ATWs events in second objective of the reliability the frequency of ATWS events for the which the reactor is isolated. lsven if assurance program is to conduct a commercial power reactor industry are reactor isolation does not cause the norough analysis of the startup test of the order of 10-' per reactor year. The transient in the first place. the effects of l
program the survedlance test program.
NRC staff has concluded that operating a failure to scram are likely to trigger ed the record of system functional experience is not sufficient to determine reactor isolation. l'urthermore. UWR sisk
.ppenence to identify and-where conclusively on a statistical basis assessment studies suggest that ATWS
- casible-close loopholes through which whether reactor scram systems are accidents may give rise to release des.gn deficiencies. construction reliable enough to make the probability category L : or 3 (as described in
.!cf.ciencies, s ulnerability to test or of unacceptable consequences from WASil-1400) outcomes.
maintenance error, or component ATWS events sufficiently small.
These arptments s 7est that PWRs w!ures might escape detection and thus The improvements emanating from may ulready achieve t: i minimum
<orrection for considerable periods of the proposed reliability assurance ATWS tolerance neces,ary to t 'ne.
program will make ATWS accidents less supplement the reli..bility assurance S'uthes a tiated in response to the likely and the systems that mitigate i
m rogram, whereas imp. vements shouhl auw ns 1 erry partial scram failure A'l WS events more reliable.
be mandated for BWRs'to strengthen mhcated that two auxiliary sy stems,
hevertheless. it is necessary tornsure their provis;ons for ATWS mnigation.
t serve the scram sytem us well as that mitigating systems wdl render the However, a more careful analysis of
. er systems. could have caused partial outcome of rnost ATWS events ATWS. tolerance is required m the co nplete scram failures. 't his acceptable.The principle of defens.c m.
,r proposed rule to provide the basis for mery is suggestive of a class of depth calls for reactor plants to be and form of actions to be taken by i
amman cause failures that might designed ur d aperated in such a way
.coromise the safety of a reactor-that a rare ATn S accident can be 5 Srcs in auxiliary systems might tole ra ted.
In pWRs. the limitm; transient with l
use the initiatmg transient as well ah The requirements for ATWS tolerance gesputt ATWS is a comp!cte code the rehability of the scram m bght water cooled commercial power intenupu n in the delivery of feedwater
.s cm or they might contribute to the reactors are intended to afford an to the steam <tenerators at full power.
Should the scram fail to shut the reartor un failure and also could opportunity to learn from experience l
without placing the public health and
- "'"b"'""""'
"" J '"' F "
n a.r. n *m. or n a...o r u.' "*
I","n H+ D4 W'on of T.,"h,"a al inform in..
safety n jeopardy.The first occurrence i
m o ri er the mi ram do.c harge wlen end n
-w d me s p +rt. scrug the oir +presh.el o an AID,S precursor due to any p,nu,,e d Comnil. U S Mr.er ke; M.iq J.,,.
particular failure mode will result in c..n--on w hen. O c. m b
1 i
1 Feder:1 R: gist:r / Vol. 46 No. 220 / Tuesday, Novemb';r 24, 1981 / Proposed Rules 57528 f
dem o. il.e continued power generation gross above-ground failure of pressure injection system and (5) ti.,
and the dethning heat removal, as the containment.This is not among the more integrity of reactor coolant pressure scumdary cool.m boils awey, causes n probable outcomes of even the most boundary valves through which a I.OCA i
r smge in pneuve of the reactor coolant.
severe und damaging pressure would bypass containment and could The severity of this pressure excursion excursions associated with ATWS in not be isolated.
is a sensitive function of the moderator PWRs.
In some PWRs, the very rapid i
temperniure uncfficient the capacity of Analyr.is of ATW3 trunsients by the autostart of the auxiliary feedwater the rehef vues uttached to the reactor NRC staff and the reactor suppliers system following a feedwater transient E
unnlant system.und the speed with suggest that Westinghouse reactors can overcool the reactor if the scram is which the auxiliary feedwater system have sufficient relief capacity so that successful. In such plants, the rapid start starts. The pressure surge will subside pressure excursions expected of limiting logic may be interlocked to take place as the power decreuses due to the ATWS transients will not be damaging, only if the scram fails. llowever, such increasing moderator temperature.
provided that the auxiliary feedwater interlocks must not degrade the Subsequent reactor coolant syste:n starts promptly. Combustion reliability of the auxiliary feedwater replenishment and reactivity control is Engineering and Babcock and Wilcox system for the more frequent loss-of.
provided by the high pressure injection reactors muy be subject to severe feedwater transients in which the scram Illpl) vslem. which pumps cooling pressut o excursions even with prompt is successful and in which a delayed water containing a reactivity poison into start of the auxiliary feedwater system, autostart of nuxiliary Icedwater is the seactor coolant system.
should the ATWS accident take place appropriate.The identification of the The most severe test of the ATWS when the moderator temperature required instrumentation and the tolerance of a PWR lies in its survival of coefficient is unfavorable.The NRC staff training of operators may be made a the pressure excursion and in the has argued in NUREG-0400 that these part of the rehabilit/ assurance successful start of the auxiliary plants should install additional relief program, and the verite % that the feedwater and high pressure injection capacity to improve their ATWS instruments and the critical pressure systems.The possible outcomes of the tolerance. The industry has argued that boundary valves on the reactor coolant pressure excursion are (1) the reactor such modifications are very expensive, system have the required tolerance for coolant system and interfu m will produce substantial occupatior:a!
the limiting pressure excursions would equiptrens are undamaged. (2) the exposures to radiation to those be part of the ATWS tolerance reactor coolant system remains intact installing them, and are unnecessary requirements.
but instruments on the pressure because the plants already have The moderater temperature boundary fail or the valves for the HPI sufficient tolerance of the pressure coefficient, which strongly influences system are damaged. (3) the reactor excursion. according to their analyses in the severity of the reactor coolant coolant system is ruptured producing a proprietary reports.
pressure excarsion forlimiting ATWS loss-of-coolant accident (LOCA) to The NRC, in reassessing its position, transients, in at its least favorable value t
containment (4) steam generator tubes has concluded that the minimum ATWS during the early months of operation rupture causing a primary-to-secondary tolerance necessary to complement the with the first fuelload. The early months LOCA or a LOCA to other interfacing reliability assurance program does not of plant operation are also characterized systems, or (5) combinations of (2). (3) or dictate additional pressure relief by a higher.than-average frequency of (4). The first outcome is clearly capacity in CE and B&W plants in light transients and safety system failures as preferred. The second outcome makes it of the several mitigating factors noted the plant is shaken down and the plant clear that care must be taken to ensure above. Ilowever, there are a number of personnel gain experience with the that the operutzs have sufficient other safety.related incentives to alter equipment.Therefore, much of the risk information about the status of the the provisions for reactor coolant associated with ATWS accidents is reactor to manage the recovery. Should pressure reduction or relief in PWRs.
expected to be concentrated in the first the HPI pressure boundary valves all These include deliberate months of plant operation. One seite in the closed alignment, the core depressurization to enable low-pressure mitigating factor is the less-than-will melt. This is one of several paths safety injection in small LOCAs and equilibrium inventory of fission products from ATWS to a contained core melt feedwater transients with scram, to accumulated in the fuel at this time.
accident. A LOCA to containment is avoid the melt.through of reactor vessels Nevertheless, PWR reactor licensees likely to be mitigated by the Emergency while at elevated pressure, and to would be required to propose and Core Cooling System (ECCSI even enable the ECCS accumulators to implement particularly stringent limiting though the initial pressure conditions extend the point of no return for the conditions of operation in the technical are outside the design envelope for restoration of AC power in station specifications to constrain operation ECCS analysis. Thus no core reelt is blackout accidents. The NRC expects to when combinations of the unavailability expected (although a contained core take up the case for and against altered of mitigating or preventive equipment.
melt is a remote possibility), and a core pressure relief provisions for PWR the prevailing moderator temperature meh with missile damage to reactor coolant systems in the coefficient, and the power icvel containment is a still more remote forthcomin8 rulemakings on severe encroach upon the tolerance of the plant possibility. Steam generator tube rupture accidents.
for the pressure excursions to be can ruth e leakage path to the The required ATWS tolerance of expected of limiting ATWS transients.
outsa umosphere that bypasses PWRs rests: (1) upon the prompt start of In large, modern boiling water contamment. Ilowever, ECCS is likely to the auxiliary feedwater system, (2) the reactors, a transient with failure to be successful, so the core would not availability of instruments necessary for scram from fu!! power is very likely to melt. All but one steam generator can the operators to diagnose the ATWS cause. or may follow, the isolation of the very probably be isolated, thus accident sequence and successfully reactor, notably a trip of the main steam termmatmg a minor release.
maneuver the plant to m%imize the isolation valvet. lf the reactor coolant The ses ere relem category 1. 2. or 3 release of radiation,(3) the training of recirculation pumps continue to run, the events occur only for a core melt and a operators. (4) the availability of the high power level will remain N3h and a l
I
0 f
Fed:rrl Register / Vol. 40. No. 22G / Tuesday November 24, 1981 / Iboposed Rules 57529 l
'. s me pr essure excursion will take Hear. tor Core Isolation Cooling sy, tem ATWS events. thus threatening p!.ios Es en if the reactor coolant system (HCIC) should be expected to autostart successful miligation. In some seipwne sm t n es ilm pressure surge. the very and run. delivering coolant to the vari.ints. Operators might he templed to A mssurize the reactor to enable low hmh sic.on flow will rapidly be.it the reactor.The flow rate delivered h3 the T
suppression pool and pressurize the itCIC is lower than that of the lipCI. If prc.sure reactor omlant inp ;.Jon but, in mot.iinment. In addition, the high.
the RCIC is the sole operative mc. ns of so doing. disable turbine.drm n coolant pressure coolant injection (!!pCI) may replenishing reactor coolant, the injection systems or otherwise L
not suffice to cool the core: m erheating adequacy of core cooling. rather than compromise possible avenues of 4
and core dumuge may follow. Ultimately the heat deposited in the suppression successful ATWS mitigation. The h
the omtainment is expected to rupture pool. is likely to be the factor limiting reliability assurance program must
<!ne to overpressure while the core the time allowed to shut down the entail a thorough imstigatioc of such H
sustain. Jamage. Continued ~a e react.r without unacceptuble ATWS uccider' sequences, of the coolant replenishment is questionable censequences. The RCIC can instrument indac 6tions available. and of after containment rupture. A large successfully cool the reactor once it is the possible range of operator actions.
radiological release is a plausible shut down and it can slow the boitoff of Operator training should familiariu outcortm. A necessary mitigating feature reactor coolant in the reactor.
operators with the optium str.A.;:es anil i
is thus a prompt automatic trip of the The N'tC has concluded that tht.
alert them to serious errors that could recirr.uhition pumps to avoid the hquial reactivity poison injection system occur in deahng with ATM acculents pressure excursion and diminish the in large, modern BWRs must have a DWR reactor operators nm be power and the consequent steam flow to start time and poison injection rate such subject to a strong disincentwe to the suppression pool. Given a trip of the tt.et either of two redt.ndant trains of actuate the Standby Liquid Control recirculation pumps, the reactor power Mgh. pressure reactor coolant (SLC) system because of the costly wdl stabilize at roughly 30% power until replenishment systems. either of which nature of spurious SLC actuations.They the rcactor coolant boils down and may be expected to be avt /~ble under may also be inclined to override an l
steam bubbles (void formation)in the ATWS conditions, can successfully autostart of the SLCif they doubt that l
wre throttle the chain reaction.
mitigate ATWS transients.The two an ATWS indication is genuine or the 1hercafter, a static or oscillatory trains may be the lipCI and RCIC.
failure of the scram system is eymlibrium will be maintained in which The criteria of successful mitigation irreparable.The NRC recognizeas the the reactor sustains the average power ure: (1) The containment temperature ligitimacy of the concern with the cost of necessarv to boil off however, much and pressure must remain within the spurious SLC actuation reator c'uolant is delivered, up la about design envelope. (2) the core must retain To deal with these conflicting W. power. Analysis shows that !!"Cl or coolable geometry, and (3) neither concerns, the NRC proposes to require raain feedwater can adequately cool the prompt fatalities nor serious offsite the automatic start of the SLC system core to avoid extensive core damage.
property damage are predicted by under circumstances diagnosed to be flou ever, the power delivered to the analyses whose conservatism is ATWS sequences. Licensees are free to r
wppression pool will be greater than the compatible with that employed in employ rehability engineering methode pool toeling system can dissipate.
W ASil-1400.5 to minimize the likehhood of spurious Therefore, containment overpressure Concern has been exp;essed t'r.at the actuation.s under ne>ATWS f.olute remains a distinct possibility RCIC. though capable of meeting these circumstances provided these provisions r
un'ess the reactor is shut down either success criteria. does not prevent the do not compromise the reliability of the 13 control rod insertion or by liqu:d automatic depressurization of the essential SLC safety function in genuine reactaity poison injection. Well before reacta coolant rystem. Operator action ATWS sequences.
- he containment is significantly is necessary in less than ten minuted to
- n light of the analysis and operator pressunzed, the suppression pool will override the automatic depressurization training associated with the reliability l
approach saturation, and the steam or to throttle l>w pressure ECCS should assurance program. it is not deemed l
l condensation will become unstable.
the depressunzation eccur.The NRC necessary to r reclude provisiens for i
Chugging steam condensation may stuff does not wuh to forr.e an alteration manually ourriding the natostart of the mreaten containment intergrity or of the logic go.erning the utomatic SLC. As part o! the reliability assurance pressure 6 ippression and thus shorten depresurizatior. system (ADS) which program. a thorough analysis is to be ne time available to shut down the might compromise the reliability of the made of the circumstances in which an reactor without unacceptable ADS in non-ATWS events. Options to operator might be tempted to werride a f
l consequences. In limiting transients. the resolve Sese competing concerns w:ill genuinely needed SLC actuation.
n
'. Ne of the main feedwater system be evaluated by the NRC staff during Consideration should be giver' to i
my be the initiator ef, or companion of the comment period. We are interested improved instrumentation if the correct F e initiating event. The llPCI is a in receiving comments on the potential diagnosis of such sequences is de traia system. The fault or F urr.an effects of the three proposed rules on ambiguous. Operators must be trained i:ror that precipitates the initial this subsystem (high pressure makeup) to give first priority to safety rather than ansient might also disable the llPCI. In of the DWR design.
to the availability of the plant for power Several factors complicate the analysis of the ATWS-tolerance of BWR ~ generation.The anticipation that
..Mt.on. s> stem reliability unalyses repeated manual scrams or quick fixes u o mdianed that !!PCI may fail or be m.n a:lable in us many as Nm v. to pl nis.The delivery of main feedwater, in the control cabinets may succeed in P ef the cases in which a demand is which may be avanable in some ATWS inserting the control rods would be an mJe of tha system.This may be accident sequences, may dilute liquid unacceptable justification for overriding mfficient reliability for the mitigation poison and increase the power level in SLC actuation.
- a potentially serious accident having in conjunction with this form of the
,,]Nl,llQ'.l'j]/lg7 frequency of occurrence that might be rule the NRC does not deem it
... h gh as once in a thousund reactor necessary that th SLC meet the single.
t nmn,cni conirot tr s socie., ac,oi.ior3 s A second diverse system. the en n mm w uwon. o c. ro.ss.
failure criterion as well as the indir.acd i
im
57530 Fed;ril Regist:r / Vol. 40. No. 226 / Tuesday. Novemb:r 24, 1981 / Proposed Rules r
S soccess* criteria. In the very unlikely licensee and to relieve the NRC staff of (a) hiitialre/labihty assurance event of an ATWS cvent und a failure of much of the detailed involvement in program. The initial reliability i
automatic and manual starts of the SLC cxperience review and the selection of assurance program must include un system. a f.ilitsack strategy is available procedural or hardware backfits in the unalysis und classification of the s"
through manual rod insertion and context of ATWS risks. For this reason, principal deterrainants of the intervention in the reactor protection the proposed rule emphasizes criteria radiological severity of each class of system control cabincts. Nevertheless, for the sound implementation of the ATWS accident sequences in terms of the S!.C must not depend upon a sing!c reliability assurance program and limits the initial plant conditions, the type of daision of an auxiliary system the the staff review to these criteria, initiating transient, the failure mode of failure of which would also compromise together with t.he conventional review the reactor protection system. and the g
the reliability of the scram system or of and approval of the license amendments state of operability or inoperability of ihe recirculation pump trip or precipitate associated with changes in design or other active systems affecting the the initiating transient.
operation.
UWRs rnust also operate under outcome.This analysis must be specified Limiting Conditions of Pursuant to the Atomic Energy Act of employ ( d in each of the following Operation that constrain power 1954, as amended, the Energy elements of the reliability assurance
)
gr ncration under circumstances in Reorganization Act of1974, as amended, program:
li whic.h erguipment unavailability and section 553 of title 5 of the United (1) Training oflicensed reactor compromises the reliabihty of systems Sintes Code, notice is hereby given that operators in the diagnosis and prognosis
(
I important to ATWS prevention or adoption of the following amendmen;s of the several ATWS accident t
rnitiga tion.
to 10 CFR Part 50 is contemplated.
sequences. Operators must be trained to The older lower-power. level reactors PART S0--DOMESTIC LICENSING OF make productive use of their time during li may differ significantly in the levels of PRODUCTON AND UTILIZATION ATWS accidents to effect mitigation..
I ATWS. tolerance provided. These plants FACILITILS Consideration must be given to would be required to submit analyses of improving instrumenta tion, displays, S. tolerance for review by the
. The authority citation fur 10 CFR and emergency procedures to minimize i
I Part 50 reads as follows:
the likelihood that misdiagnosis or
{
The dualapproach of ATWS Authority: Sees.1c3.104.161.182.183. 68 delayed diagnosis of ATWS sequences tolerance and the reliability assurance Stat. 936. 937. 948. 953. 954, as amended (42 may substantially increase the I;
prograrn provides defense in depth. Each U.S.C 2133,2134. 2201. 2232. 2233): secs. 202.
radiologicaI severity of the outcome.
allows the oth:r to be implemented 206,88 Stat.1244.1248(42 U.S.C 5842,584el.
. (2) An analysts of hypothetical errors j
3 without highly conservative mrpins.
unless otherwise noted. Section 50.78 also in or erroneous departures from proper The margin provided by ATWS-issued under sec.122,68 Stat. 939. 42 U.S C.
test and maintenance pre.cedecs fur m!crance allows realistic cost. benefit
- 52. Secuons Sa8%5a81 also issued under systems whose reliabihty n bportant to considerations to govern the selection sec.184. 68 Stat. 954, as amended. (42 U.S.C ATWS prevention or mitigation.
and schedule ofimplementation for 22341. Sees. 50. loo-50.1021ssued under sec.
Intes suggested by the reliability 186. 68 Stat. 955: (42 U.S.C 22381. For Consideration must be given to 9
purposes of sec. 223. 08 Flat. 958, as improved designs, test equipment auurance program.
amended. [42 U.S.C 22731. I 50.54(i) issued pr.o:cdures, and personnel trainin8 to The very ccstly accident at Three Mile under sec. te1L 68 Stat. 94t. [42 U.S.C mmimize the likelihood that the 1
Island has demonstrated that the 2201(ill. and il 50.70-50.71 and i 50.781: sued reliability of these systems will be protection of a licensee's investment in under sec.161o. 68 Stat. 950, as amended; (42 compromised by errors in test and r
a reactor plant provides a powerful 11S C 2 01[o)) and the Laws referred to in maintenance.
economic incentive to search out and A pendices.
cturect reliability defects in the (3) Ar analysis of the blindspots in
- 2. A new 150.00is added to read as
}he experience base with systems functiors that protect a reactor core follows'.
imp rtant to ATWS prevention or from damage. These economic g
mitigation through which reliability considerations together with a realistic
$ 50.60 Standards for the reduction of risk defects might cscape detection for j.
evaluation of offsite iisks affecting from Anticipated Transients Without Scram considerable periods of time, public health and safety, are sufficient (ATWs) events for tight. water. cooled to determine the scope and schedule of "uCI'8"Po*'rPl8nt*-
Hypothetical reliability deficiencies
{
must be classified by (i) kind (design the more expensive or intrusive Each light. water-cooled commercial deficiency. construction deficiency.
niterations in plant operation or design power reactor licensee shall establish vulnerability to test or maintenance emerging from the reliability assurance and maintain a reliability assurance.
error, active or passive failure). (ti)
[
program.
program for functions associated with affected components or tubsystems. (iii)
The reliability assurance program is the prevention and mitigation of not to be a paper study to demonstrate Anticipatt d Transients Without Scram severity of the reliability o! deficiency, and (iv) the frequency ar.i kind of to the NRC staff that the plant is stready (ATWS) c.aploying state-of the. ort opportunity to detect the deficiency. A safe enough. The role of probabilistic methods and procedures to identify test program covering startup or or c.
culuations is secondary to the vulnerabilities to failure. Each licensee time-only tests, tests associated with r
quahtative search for and evaluation of is responsible for the implementation of peric lc plant overhauls, and inservice g
specific types of reliability defects.The cost. effective improvements to reduce surve. ante tests must be developed reliabik'; assurance program is not ATWS risk. Defense in depth must be and im. lemented so that the mean time j
.ntended primarily to assist NRC staff maintained by operating commercial to detect the deficiencies is reduced to j
review. Rather,it is to be integrated into power reactors only in modes that the extent reasonably achievabic.
the conduct of plant management, afford an opportunity to learn from (4) An analysis of the susceptibility of pe sonnei training, and the conduct of experience with ATWS events without the plant to common cause failures of operatmns. It is intended to strengthen severe radioactivity releases. Specific two kinds: those in which a single root the responsibihty for safe design and acceptance criteria are delineated operation of th plant resting with the
- below, cause degrades the reliabdity of redundant divisions of a safety system
Fed:rtl Rzgistir / Vol. 4G. No. 220 / Tucsday. November 24. 1981 / Proposed Ruiss 57531 important to ATWS prevention or (2) pressurized water reactor licensees and approval. Iloiders of operating mitigation, and those in which a sing!c receiving an operating license after hcenses. applicants for operating tool cause degrades the reliability of August 22.1909, shalb licenses, and those expecting to file an tuo or nmre systems whose concurrent (i) Provide for the prompt. uutomatic application for an operating license failune contributes to a severe ATWS start of the auxiliary feedwater system within one year of [the effective date of
.na.idnnt sequence.The kinds of root under circumstances indicative of a the rulel shall file the reliabihty caum to be considered are those listed Iransient entailing loss of main assurance program plan at a time to im in paragraph (a)(3)(i) of this section.
feedwater nnd a failure to scram, ngreed upon by the NRC staff. The time Considerahon must be given to lii) Ensure that the instruments afforded for plan development will be improved design or operation to reduce necessary for the diagnosis of and not less than one year (from the vulnerability to common cause failures.
recovery from ATWS accident effective date of the rule).Those holders (b) Crmtinumg re/iobility assurance sequences will not be disabled by the pmyrom. Each commercial power effects of such accidents, and of construction permits who file an i
reactor licensee shall maintain a (iii) Ensure that those reactor coolunt application for an operating license on continuing relinbility assurance program system pressure boundary valves or after [one year from the effective date hir hmetions important to ATWS through which high pressure mjection of the rule) shall file the reliability pievention nnd mitigation that includes can reach the reactor remain functional assurance program plan at the time of i
Udiom The plans ihr follmving:
nfter limiting ATWS, transients and a es wh must identify (i) the ways the reliability i
(1) Configura tion control for designs, t oge n eg i Qs essential assurance program will be integrated d
procedures, and techmcul specifications to assure consistency with the amtia uncontained loss of coolant accidents into the engineering and operations reliability assurunce analyses.
retain their integrity throughout limiting management of the plant. (ii) the (2) procedures for updating affected ATWS transients.
reporting and approval requirements portions of the initial reliability (3) Commercial light. water-cooled internal to the licensee's organization, assurance analysis for, and prior to, power reactor licensees not covered in (iii) the plans for information evaluation departures from the controlled design.
paragraphs (c)(1) or (c)(2) of this section and exchange among licensees as part ponrdures, or techrucal specifications.
Adl mM u unalysis of the ATWS of the experience feedback function. (iv)
Applications for license amendments t tolcrunce of their plants.
the criteria for reporting to the NRC. (v) implement these changes must include a (4) Euch commercial power reactor the criteria for the adoption and brief analysis of the impact of the licensee shall prepare, submit for review scheduling of alterations to plant design thange on the reliability of systems and approval. and implement IJmiting or operation emerging from the imimrtant to ATWS risk.
Conditions of Operation that proscribe reliability assurance program, and (vi) p) An experience feedback system to operation in, and mandate upeditious the date at whic s the irntial reliability t
retiew operational and test data on retreat from operation under conditions assurance studies can be completed A relevant systems in the licensed plant that compromise the ATWS tolerance of brief summary of findings and plans for and the relovant experience at plants the plant I.imiting Conditions of the resolution of reliability deficienc.es Operation should also minimize must be filed with the NRC upon haung a simd, ar syttem design. Each operational occurrence must be operation under conditions in which the completion of the initial reliability reviewed for clues to oversight or errors ATWS tolerance of the plant would be assurance studies. Subsequent m ihp reliabihty assurance analyses.
severely tested by a hmiting AT%T, discoveries of reliability deficiencin a t he untial rehability assurance unalys" event. Consideration of the prevalling the plant must be repoin 4 in accord
.nnli.ust benefit analyses based thereon plant parumeters as well as equipment with prevailing practices for reporting arn to bu updated when the experience operability is appropriate in the Limiting licensee events. The reliability feedback system reveals oversights or Conditions of Operation.
assurance program will be sub' ject to f
hmitations in these studies.
(5) For the purposes of this pargraph.
audit by the NRC. It is not expected that the ATWS tolerance of a plant is the NRP will cogage ir. *autine review (c) Desi.i;n and operation for A TIFS to/enne. (1) Boiling water reactor nudequate if any of the more limiunE und apt roval of the program unless a transients. followed by a total failure of I
licensees receiving an operating hcense pattern suggestive of eencompliance is nfier August 22,1969, shall:
the scrum system, result in any one of observed.
i the following-(i) provide equipment to trip (i) Containment pressure or (2) Applicants for or holders of automatically the reactor coolant temperature above the design values.
oporaung Ucenses suhet to paragraph n., irculation pumps under conditions (ii) Loss of coolable geometry in tho (c)(1) or (c)(2) of this section shall file ladiadite of an ATWS event.
with the NRC plans for the core, or (ii) proside equipment to (iii) Releases of radioactivo matertut implementation of the requirements of automatically deliver liquid reactivity that may realistically cause any offsite paragraph (c) of this section (withm one pomon so that either of Iwo independent prompt fatalities or serious offsite year of the effective date of the rulej or reactor coolant replenishment system property damage.
upon licerne cpplication. whichever is trains expected to be available during (0) Applicants or licensees are not later.The fullimplem:ntation rd the ATM cvents can successfully bring the required to calculate the potential offsite re luircments of par'gr phs (c)(1). (c)(2).
reactor to stabte hot shutanwn.The radiological doses resulting from nn and (cli4) of this seAn must be pun.on injection system must not depend ATWS event under i 100.11 of this completed:
for its function on a single division of un chapter.
(i) For all hght. water cooled nuclear mdiary t.ystem whose fail are could (d) Schedule ofimplementation ond reactor power plants for which precipitate the transient, degade the reporting frquirements. (1) plans for the operating licenses have been or may be rehainhn of the scram system or defeat implementation of the reliability issued after August 22,1969. but before the retirtulation pump trip. and ussurance program cidled for in (three years after effective date of thn (iii) proside a rehable scram discharge paragraphs (a) and (b) of this section rule), all modifications shall be s olum,e system.
must be filed with the NRC for review c,mpkted prior Ii startup followir a the 1
E 532 Fedir21 R:giltsr / Vcl. 46. No. 226 / Tutsd y Nrvsmber 24, 1981 / Propos:d Rules Tirst refuelin:: that begins (three years existing finsncix! critzria wruld be quotations, and (3) sufficient issuer after effective date of the rule).
relaxed to more closely resemble disclosures. The National Associatum of (ii) For all light water cooled nuc! car requirements established by major Securities Dealers Automated Quotation lear. tor power plants licensed on or after rxchanges.
System ("NASDAQ") now in operatum Ithree years after effective date of the DATE: Comments should be received by for ten years, has greatlyimproved the rule), all modifications shall be junuary 29,1982.
efficiency of the OTC market and has completed prior to issuance of an AponEss: Comments, which should refer addressed the first two concerns of ihr operating license.
to Docket No. R-0372, muy be mailed to Board. The SEC over the pust few
- 13) Holders of operating licenses the Secretary. Board of Covernors of the yeart has improved and strengthened subject to paragraph (c)(3) of this section shall file with the NRC plans for Federal Reserve System. 20th Street and its disclosre rules so that financial
~
Constitution Avenue. N.W., Washington, infermation on foreign as well as the acc.omplishment of the ATWS D.C. 20551, or delivered to Room B-2223 denestic issues is available to the tolerance assessment called for in between 8:45 a.m. and 5:15 p.m.
nLlic in a comprehensive and time!v paragraph (c) of this section [within one year of the effective ble of the rule].
Comment.+ received may be inspected at fashion. In addition the National Room B-1122 between e.45 a.m. and 5:15 Association of Securities Dealers such licensees shall file the results of p.m except as provided in i 201.6(a) of
("NASD") requires that its domestic ami thc.c studies, together with proposed the Board's Rules Regarding Availability foreign issuers file financial data with H changes, if any, in design, procedures, and ter hnical specifications to assure of Information l12 CFR 261.6(a)).
as a prerequisite for trading on ATWS-tolerance, and a proposed FOR FURTHER INFORMATION CONTACT:
NASDAQ.
implementation schedule shall be filed Robert S. Plotkin. Assistant Director.
None of the approximately one-with the NRC for review and approval Laura llomer, Securities Credit Officer.
hundred eight3 i180) foreign stocks
[within three years of the effective date or Jamie Lenoci. Financial Analyst.
currently in the NASDAQ system can be of the rule.1 Division of Banking Supervision and placed on the OTC List. as they do not Regulation (20M52-2781).
meet the existing criterion which daw at Washington. D C., this 19th day of November.1901.
SUPPLEMENTARY INFORMATION:In July requires all OTC List candidates to be for the Nalcar Regulainry Commission.
1969, the Board adopted criteria for
" organized under the laws of the United including stocks on the OTC Lb t. In States or a c' ate." A growing number of samuel l. chilk.
dHcussions leading to the selecticn of requests haw been received from both scorcru,y of the comminion.
r,wh criteria, the Board indicated investor groups and the general public to p u ru m.uu2 ni.a n-am e o =1 generally that (a) stocks to be included include foreign OTC stocks on the OTC twe coot new on the List should have ma:ket List. When the Board first adopted its
_ _ _ _ _ _ _. - characteristics similar to exchange-criteria forinclusion on the List, there listed securities. (b) manipulation by was insufficient financial disclosure for FEDERAL RESERVE SYSTEM sauers to be included or excluded from foreign issues.This problem has now 12 CFR Parts 207,220, and 221 the OTC List should be made as difficult been remedied. Furthermore, foreign as possible, and (c) fluctuations m the issues can and do list on national (Dicket No. R4M2]
number of stocks on the List should Le exchanges and are therefore minimized.
automatically eligible for margin credit.
Pr: pot.a; To Revise Criteria for initial The changes now proposed in the In this connection. the Board also and Ccr.dnued inclusion on the Ust of LaTC List criteria are the result of a review of the OTC margin stock listing proposes to allow American Depository OTC Margin Stocks AGENCY: Board of Cosetnors of the and continued listing requirements in Receipts ("ADRs") to be eligible for inclusion or. the OTC List. ADRs are Federal Reserve System.
light of recent developments in the ACTION: Proposed amendments, securities markets in general, the OTC receipts issued against securities of Market m particular. and staff foreign issuers deposited in an American depository, and are exempt suMMAny:The Board proposes to amend experience s,d h admimstering the..
t the requirements set forth in Regulations requirements. It is believed that revismg frem registration under Section 12 of the l
(L T and U for inclusion and continued the criteria is especially appropriate at 34 Act.There are approximately sixty nelusion on the List of OTC Margin this time because of a recent decision to (60) ADRs current! yin NASDAQ.The Board would allow ADRs to be Stocks ("OTC List"). Brokers and revise the List three times a year.
i dealers may not extend credit on stocks commencing in 1902. rather than twice a ecusidered for inclusion on the OTC which are traded over.the-counter year as is the current practice. This has List, provided the foreign securities unless such stocks appear on the OTC been a frequent recommendation of the against which the ADRs are issued are 1.ist. Loans by banks and other tenders securities mdustry. The following is a registered pursuant to Section 12 of the l
that are used to purchase stocks that discussion of the specific proposals to 34 Act, which imposes certain reporting appear on the O TC Ust are subject to amend OTC 1.ist critena.
requirements upon the foreigi. issuer.
This approach is consistent with the re s
m nd en s w uld A. Deleting Requiremnnt That Issuer be pchey currently employed by stock modify three areas in the existiq rules Organized Under the Laws of the United exchanges with respect to exchange for initial and continued OTC last States or a State listingr and with the Secunties and cligibility. First. they would permit As early as 19G4, when the SEC first Exchun2e Comm.ssion's current equity recurities of foreign issunrs and accommended a broadening of the pr p salt all w ADRs to be American Depository Recei9ts (* ADRs") l'ederal Reserve's margin authorii3 to gnated as national market system e
to be considered for OTC List inclusion, encompass over the coun'er stocks, the securhics. 8 Second, the proposals would replace Daard indicated that secuities, to bc s celain t.riteria which must currently be chgib!c for credit ut a broker, should y
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met in the alternalise und replace them meet the prerequisites of (1) market s-ectora ow.n.d unde seumn u w hh mandatory requirements. Finally, depth. (2) a reliable system of 1ste rwic.m.w maut.
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ATOMIC SAFETY AND LICENSING BOARD
,d Before Administrative Judges :
d/qg7 James P. Gleason, Chairman Paul W. Purdom Glenn O. Bright SERVED E 1 4 531
')
In the Matter of:
)
)'
Docket Nos. 50-387 OL PENNSYLVANIA POWER & LIGHT CO.
)
30- 388. OL and
)
. ALLEGHENY ELE _CTRIC COOPERATIVE, INC.
)
October 12, 1981
)
(Susquehanna Steam Electric Station,
)
)
Units 1 and 2)
)
m MEMORANDUM AND ORDER ON
SUMMARY
DISPOSITION MOTIONS The Board has previously set forth the law applicable to motions for summary disposition (see Board Memorandum and Order, dated March 16, 1981).
We see no need to repeat it herein, and thus will proceed with our evaluation of same motions pending.
We have previously communicated our decision on some of the motions discussed herein.
1.
The applicant filed a motion for summary c'isposition of q )
contention 5, and subsequently received a Staff response supporting the Applicants' Motion.
The Board has received no response frsm
'b
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any of the other parties in this proceeding.
v Backcround:
Contention 5, as accepted by the Board reads as follows:
5.
Certain models used by the applicants to calculate individual and population radiation doses are in-accurate and obsolete.
The deficiencies are compounded by the arbitrary selection of data from inappropriate sources for the purpose of formulating these models.
Specifically:
a.
the milk transfer coefficient for iodine has been underestimated (see Health Physics, 35, pp 413-16, 1978);
Jm b.
the modelc use factors which correct alpha
. particle dose in rads to rems which are far too low (see Health Physics, 34, pp. 353-60, 1978);
I[ )
the models use factors which underestimate the c.
radiation effect, on a per rad basis ~, for the very los energy beta and gamma radiations, as from H-3 and C-14 (see Health Physics, 34, pp. 433-38, 1978).
There being no models prescribed by regulation for such calculations, the Board accepted the contention as a challenge to the models specifie'd.
We now consider these.
The general thrust of Intervenor's contention is that the models used by Applicant are inaccurate and obsolete.
Both the Applicant and the Staff dispute this, stating that the models used by the Applicant are set forth in NRC Regulatory Guide 1.109.
These models were developed by the Staf f and by experts at Battelle Pacific Northwest Laboratories and such national laboratories as
oak Ridge and Argonne.
These models, which are used to calculate
-)
V both the maximum hypothetical individual and general population doses fr.om exposure to radioactive liquid and airborne releases
.: ru -- :se from routine operation of a commercial nuclear power reactor, are
- .... -.~.. :
subject to continuing peer review and verification by other federal agencies such as the U.S. Environmental Protection Agency and the Bureau of Radiological Health.
(Sta'f f af fidavit, Branagan at 4).
In September 1977, a group of experts, meeting to evaluate models used for the environmental assess ent of radionuclide releases, w
concluded that the transport models given in Regulatory Guide 1.109 are adequate for dea.onstrating compliance with Appendix 1 of 10 CFR
~
m Part 50.
(Branagan at 4).
As these models are adequate for demon-qj strating compliance with the numerical guides in Appendix 1, they are conservatively adequate for demonstrating complian,ce with the higher standards for protection of the public against radiation hazards found in 10 CFR 20.
The NRC models are therefore not
~
inaccurate or' obsolete.
Data used in these models are taken from reports by scientists at national laboratories, peer-reviewed articles in", s cientific
{
j ournals, and recommendations of nationally-and internationally-known radiation protection organizations.
(Branagan at 4).
The models do not, therefore, use data arbitrarily selected from Gd inappropriate sources.
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The Staff also argues, and the Board must agree, that from
~
v the standpoint of the protection of public health and safety it crs.ctbe degree of conservatism in an overall model that is important
.in: determining whether doses have been overestimated or under-t estimated, rather than the degree of conservatism in individual parameters.
The challenged models contain many parameters, some of which depend, in turn, on a number of subparameters.
- Thus, due to the complexity of the model's, an apparent lack of conservatism in one parameter does; not necessarily constitute a
, prima facie _ showing that a model will underestimate doses.
(Branagan at 2-3).
O
-Subpart a of the contention alleges that the value of the
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" milk transfer coefficient" for iodine has been underestimated, and refers to a note contained in Health Phycies.
The Staf'f points out that the values in Regulatory Guide 1.109 are within the range of values given in the referenced note, both for cow's milk and for goat's milk. (Also see Applicant's Bronson A'ffidavi t at 3).
~
The author of the article referred to, Dr. F. Owen Hoffman, subsequently published a more comprehensive report (NUREG/CR-1004) which included a statistical analysis of the entire NRC model, as well as its individual parameters.
The conclusion of this report
-is that doses to real individuals frem ingestion of milk econtaining radio-iodine are more likely to be overestimated J
ll( )
than underestimated.
(Branagan at 5-6).
3-Subpart b alleges that the factors which are used to
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7.:. : :. _
-f convert alpha radiation doses from rads _.to rems give results
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which are t o lc.
Intervenor cites an article by Rossi, et al.,
,4 entitled, " Leukemia Risk from Neutrons" (Health Physics, 34, pp. 353-60(1978)), presumably as support for their~ argument.
The primary thrust of the article is the effect of neutrons, and deals with alpha radiation and other types of hig ~ linear energy transfer radiation in only ;g speculative way. (Branagan
.at 7).
Furthermore, alpha particles are not a significant source of dose to offsite individuals from exposure to radioactive O
effluents from routine reactor operations.'
(Branagan at 8).
Subsection c alleges that factors used to convert dose to radiation effect underestimate the effect for very low energy
- r: es.
n; : r : ;.3:
b'ta and gamma radiation, as from H-3 and C-14, and cites.an e
article by Bond, et,al., in support of their argument.
(Health l
Physics, 34, pp. 433-38, 1978)).
This " quality f actor" ("Q") allows doses of different biological effectiveness to be added and measured relative to a reference standard.
In 1969, the International Commission I
on Radiation Protection (ICRP) established through research and
~tudy that a Q-value of 1 was the best estimate for low linear r~3 s
V energy transfer beta and camma matters.
Since that time,'all
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6_
major national and international advisory and regulatory groups
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have used a Q of 1 for low-LET beta and gamma, radiation.
The.
~
NRC has es tablished, by regulation in 10 CFR 20,.4.(c) (2),. a.O of
_cz
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- Iri T.e
- :Ir; :- :I Oni
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1 for low-LET beta and gamma radiation.
(Bronson affidavi.t, 1.
12:.-
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paragraph 3, 5, 6).
The cited article asserts that different kinds of radiation in the low-LET range (0.2 to 3.5 kiloelectron-volts per micrometer) may have a biological effect varying by as much as a factor of 4,
and proposes that a reference radiation be chosen in the mid-a:
point of this range.
If this were done, some kinds of low-LET radiation would 'have their Q go up or down by a f actor of
(~
approximately 2.
(3,ro,ns on a,f f idavit,,, _a,ragra,ph_ _4 ).
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v--.
. The Susquehanna dose estimates'for'~ low-LET beta and gamma
~
radiation were arrived at using a model developed for the NRC
~
by Batt'elle' Paci:fi~c-Northwest 'Laboratorie's.- ~ This "pa~r~ticular l
model uses a Q of 1.7.
Thus, the instant dose estimates use essentially the same Q as is proposed in the referenced article.
Thus, even if P.he redefinition of Q were cade as proposed in the article, this would not require a change in the dose estimates f or low-LET beta and gamma emitters released from Susquehanna.
(Bronson af fidavit, paragraph 7.)
l Conclusion lD
'/
.The. Board has closely reviewed the affidavits tendered by the Applicant and Staff in support of Applicants' motion for
1 None of
!'3 summary disposition of Contention 5 in this proceeding.
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kJ as discussed above, were controverted by any of the the facts, parties in the case.
The Board finds that the Applicant and Staff have met the burden of showing the absence of a genuine
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issue of material fact, and are ~ entitled to judgment as a matter of law.
The Applicant's motion is granted.
2.
The Applicant filed a motion for summary disposition of contention 8 which is supported by the Staff.
No response was received from any other partg.'
Background:
The contention reads as follows :
The Applicant has not demonstrated adequately I)
.a. compliance with _the requirements of the Standard
(;4 Review Plan, 5.3.3,
" Reactor Vessel Integr.ity",'
Part 11.6.
As a result the ' reactor vessels may
, not survive the thermal shock of cool emergency water af ter blowdown without cracking.
The. Applic, ant 'j _. motion.makes. the. f ol,1,caing aqqument :
s
- f a)TherecommendedcriterionoftheStandardReviewPlanisthat the vessels remain leaktight enough to support adequate c, ore cooling af ter therma 7 shock; b) that the critical location for cracking to occur in the plants pressure vessels way determined by detailed stress analysis on materials with essentially identical wall thicknesses and basic dimensicns; c) that test data and results, using the ASME Code and other relevant lq t..h.e..an.alvsis nerformed by the reactors aanufacturer demonstrates
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y that the ve:sels not only meet the criterion at its most
8-critical point but indicate that the area would survive with considerable. margin the stress that could cause cracking
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_.The Staff's, supporting respons,e, brings out the information
.... _a that the design, material, fabrication, inspection and operation k
of the Applicant's pressure vessels conform to the Standard Review Plan as well as Appendices G & H of 10 CFR Part 50 respecting fracture t.oughness requir'ements.
In addition, the supporting information argues that independent research efforts support the conclusion that the integrity of the vessel could survive a thermal shock from cooling water during a large break t
loss-of-coolant accident.
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==
Conclusion:==
Based on the foregoing information, which is supported
.by af f.idavits., t.he Boar.d _c_onc,ludes th_e App _licant has fulfilled the criterion of the Standard Review Plan as well as the other regulatory requirements in relation to the plant's.
pressure vessel integrity.
In view of the information supplied by the Applicant and Staff, and in the absence of any contrary or contradictory information from any party, the Board finds that contention 8 presents no genuine issue of a material f act and therefore grants the Applicant's motion for summary S.,
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b
- i
-5
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- 1-rs. 5:
1.
1 I
l
The applicant has filed motions for summary disposition
[,-
3.
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of all of contention 7 an'd these motions have been supported
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- :. t No other party has responded.
s r:.. r-1, by a Staff response.
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Background:
Contention 7(a), as accepted by the Board reads as follows:
~
7.
The Nuclear Steam Supply System of Susquehanna 1 i
and 2 contains numerous generic design deficiencies, some of which may never be resolvable, and which, when reviewed together, render a picture.of an unsafe nuclear installat' ion which may never be safe enough to operate.
Specifically:
(a)
The presure supprie'Esion containment structure may not be constructed with sufficient strength to withstand the dynamic forces realized during blowdown.
! _)
The Applicant has provided a brief description of the containment structure and its function.
Basically, this structure consis.ts of two parts:
the wetwell,. whiIch contains the wat'er used
...L.
to condense steam resulting from a' blowdown of the primary system; and a drywell, which is situated directly above the wetwell and contains the reactor vessel and associated piping, valves and equipment.
The two chamber; are separated by a horizon-tal diaphragm slab.
(Affidavit of George R. Abrahamson in Support of Summary Disposition of Contention 7 (a) para. 4-5).
~
There are two mechanisms by which hydrodynamic loads can be placed upon the containment structure.
These are 1) actuation
/~N (SRV), and 2) a
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of one or more of the 16 steam relief valves los s -of -c oolant (LOCA).
Although the end effect is the same,
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i.e.,
the quenching of steam from the primary coolant system, the mechanisms which govern how the hydrodynamic leads are applied differ, and they must be considered separately.
We consider these below.
SRV Discharce Loads The Applicant has described, succinctly but adequately, the complex phenomena which occur during blowdown through SRVs.
(]Q1., par. 6, 20-22).
To determine the. pressures on the pool boundary during this process, an extensive test program was
~ undertaken by' PP&L at Kraf twerk Union (KNU), a German firm with extensive experience in nuclear reactor s. team discharge phenomena.
Tests were made using an actual SSES SRV system which simulated the simultaneous activation of all sixteen SRVs, which is the case that gives the highest loads on the containment' structure.
Gyi., para. 23).
The test program covered the range of reactor operating f
l conditions, including various parameters such as length of discharge line, pool temperature, steam pressure, etc. (gi.,
para. 26).
Pressure measurements, the principal data obtained I
from the tests, showed peak values of the order of 15 psig, with a main frequency of about 6 Hz.
(11., para. 29).
These I) pressure histories at the pool boundary were used as input to a computer model of the containment.
The resulting calculations t
T show that the SRV blowdown loads on the pool boundary produce V
stresses in the containm5nt floor and walls that are within l
the structures' design values.
(;pd., paras. 31, 48).
1 Further tests to measure loads caused by SRV blowdown on i
submerged structures in the cont'ainment pool were done for PP&L by SRI International (SRI).
Uctng the peak pressure and i
oscillation frequency observed in the EdU tests,. the SRI
~
tests confirmed that the loads on submerged structures in the pool are well below design loads.. (Id,., ' paras. 33-34).
LOCA Loads 1
Applicant describes the physical setup of the downcomer t'
s/
system which discharges steam to the suppression pool following a LO' A.
(j[d., para. 7).
A full description of the phenomena C
observed during LOCA blowdown, which are similar but not isentical to those observed during SRV actuation is also set forth.
(Id., paras. 3 6 - 3 8,).
Loads on the containment during LOCA blowdown were
A number of tests were performed covering a ' range of pool temperatures, steam flows and break sizes, including the 1
break size corresponding to the design basis accident.
(i.e. a break in a 28-inch diameter recirculation line).
(pl., paras.
-s
)
v 43, 44).
Measurements were made of the pressures in the drywell, wetwell air space and wetwell water space (pl., para. 41).
,.12. -
+
The pressure histories obtained for the wetwell airspace 7,
and the drywell indicate.that the wetwell' pressure rises to j21.2= -. p.s ig_ a nd.the. dryw.e11 pres sore..:ise.s..t;.o.. 37 7.psis..
The
, recirculation.line break produces the,;mos t_,r_ api,d flow of steam into the drywell, and the drywell and wetwell. pressures for the recirculation line break are greater than for smaller breaks.
Gg[., para. 47).,
Comparison of Hydrodynamic Loads with Containment Desian Capacity and Tast Level The design pressure for the..SSES containment is 53 psig
- for both the wetwell and drywell.
The SSES containment has already been tested by pressurizing it to 61 psig with air.
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These pressures are greatly in excess of the maximum pr, essure of 37.7 psig produced in a recirculation line break.
GB[.,
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para. 49), and the 15 psig produced during the SRV discharge.
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Gg[., para. 29).
The differential press ~ure across the contain-l ment diaphragm slab produces load stresseh'that are within the l
l allowable range GBl., para. 39).
The experimental test results and the accompanying computer calculations show that the SSES containment can withstand l
the hydrodynamic loads from both SRV discharges and LOCAs with ample safety margin.
Gg[.,
paras. 48, 50).
Therefore, the
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SSES containment can withstand the dynamic forces realized L
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during blowdown with an ample margin of safety.
(pd., para. 2).
f
.13
(~s The Staff supported. the Applicants ' motion by its conclusions id that dynamic loads used by the Applicant to assess the design gagagity.to. withstand.such 1 cads wete conservative when a..;
reviewed against the. Commission'.s generic acceptance criteria, and further that the Staff.had concluded that the dynamic forces' realized during blowdown had been considered and that the containment structures had sufficient strength to withstand thoseIorces
==
Conclusion:==
The Board has reviewad the affidavit supporting Applicant's Motion for summary disposition of contention 7(a) in this r~s -
p.roceeding.
Non.e,o". th.e f a_c,ts have_ been controverted by any s_
a.......
of the parties in the case.
The Board finds that the Applicant has met the burden of showing the absence of a genuine issue of a material fact, and is.. entitled to judgment in its favor as a matter of law.
Contention 7 (a) is dismissed.
Background:
Contention 7 (b) states that:
(b) the cracking of stainless steel piping in BWR coolant water environments due to stress corrosion has yet to be prevented or avoided.
In support of their motion for summary disposition of this part of the contention, applicant submitted affidavits from Joseph C. Lemaire (Lemaire affidavit) and Walter J.
,_s i
Q Rhoads (Rhoades affidavit).
te former affidavit generally a
)
describes the problem and various studies, experiments, etc.,
undertaken to understand the mechanisms of intergranular stress corrosion cracking (IGSCC) and determine means to eliminate er mitigate the problem.
The latter affidavit addresses the specific means taken at Susquehanna to cope with the problem.
~
First, the Board is satisfied that the. mechanisms which produce cracks in 304 stainless steel are understood.
While the incidence of cracks has been low, the discovery of hairline cracks in BWR's in late 1974 and,.early 1975 triggered an intensive
, effort to discover its cause.
From this effort, it has been determined that the cause was IGSCC, and that it occurred in O
the sensitized region of the weld heat affected zone.
(HAZ).
(Lemaire affidavit, paras 8, 11-13).
The investigation also determined that three conditions must be present for cracking.
These are:
- 1) tensila stress in excess of the local yield
- stress,
- 2) suitable environmental conditions (i.e., dissolved oxygen, and 3) use of susceptible material.
Stress corrosion is absent
'will not occur if any one of these three conditions.
or reduced below a critical value.
(Lemaire affidavit, t
paras. 14-28).
m y
15 -
w 1
yf Second,cethodsofeliminatingoneor'moreohthe' required -factors. Tor :IGSCC have been exp~erime'ntany v'esifib'dU (Lemaire affidavit, paras. 29-31).
These are:. -1). solution:. heat treatrent (eliminates residual stress and sensiti,zation),
- 2) corrosion-resistant cladding, 3) residual stress improvecent, (field application of induction heating to relieve stress),
- 4) ferrite control in weld metal, 5) use of limited-carbon type 304 stainless steel), and 6) use of ASME code in design which limits design stresses.
(Lemaire. affidavit, paras. 32-43).
Third, Applicants, af ter being made aware of this potential problem in 1975, have undertaken an extensive program to r
i effestively' elimihate IGSCC in the Susqudhinni sys' ed. "This has t
been accomplished through a number of means.
Oxygen levels in the reactor primary coolant will be controlled by a mechanicdl deaerator., Extensive use has been made.of: solution-heat: treatment and induction heating stress relief.
Critical piping and weldments have been made of carbon-limited stainless steel and weld metal.
Redesign of some elements of the system to eliminate crevices, stagnant reaches and built-in stress point has been made.
(Rhoade affidavit, paras 4-12).
Fourth, austenitic stainless steel has a high ductility behavior, which renders sudden, brittle-type fracture highly w e4 e
- = =
[J'l unlikely.
In other words, for a significant crack, the component would leak before it broke.
This has been verified through experience, analysis and experimentaion.
(Lemaire affidavit, paras. 9, 10).
The Susquehanna plant has a continuous, on-line leak detection system capable of sensing small leaks and small leak changes, such that small leaks can be detected before critical crack length is achieved.
(Lemaire affidavit, para.
s 10, Rhoades affidavit, para. 13).
The Staff supported the applicant's motion on the grounds w
, that its program to reduce and evaluate incidents of inter-granular stress corrosion cracking conforms to the in-service m
sl inspection and leak detection requirements of NUREG-1313,
~
Revi'sion 1, which was developed subsequent to receiving recommendations from NRC and General Electric study groups'on the problem.
Af ter a careful review of the information submitted, the Board finds that a substantial case for summary disposition may be present.
However, the Board has some*
reservations which precludes its finding that no gancine issoe of a material fact exist in this part of contention 7.
The principle issue, we believe that should be ventilated during the hearing without excluding any others covered by this part of the intervener's contention has two aspects :
first, the use s
J of low-carbon stainless steel while apparently investigated
- 17
=
thoroughly on an experimental basis deserves further information
- 7 on the record concerning operating experience, if any, in its a: zs :t.. - :. e :- + ::-- -
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application; and secondly, there should be some further illumination of the ef ficacy of the applicant's leak detection system in the areas of concern here.
Accordingly, the Board denies this part of applicant's summary disposition of contention 7.
Backcround:
Contention 7 (c) reads as follows:
BWR core spray nozzles occasionally crack, c.
a problem which reduces their effectiveness.
No cracking of core spray nozzles has ever been reported j
to General _ Electric.Co., nor is GE aware'of any such cracking.
Cracking of these nczzles would not be expected in view of the relatively lcw cyclic thermal stress in'these nozzles and the successful overall performance.oficore spray nozzles throughout four hundred reactor years of servita.
(Affidavit of Joseph C.
Lemaire, para. 4).
The Staff supported Applicant's motion and confirmed the information that no cracking of SWR core spray nozzles has ever been reported.
Cracking has occurred in other parts of the core spray system, namely in external lines, safe ends, internal core
_s i
spray. piping end core spray. spargers._.
This cracking was determined to result from intergranular
}
stress corrosion in the type 304 high-carbon stainlesc steel which was used in these components.
At Susquehanna either low-carbon l
. 1.- ;
l type 304L stainless steel or Alloy'600, both of which are highly
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.:15
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resistant to intergranular stress corrosion, is used (p5., paras.
5-8).
In summary, cracking ha's not been reported in SWR core spray nozzles, and infrequent cracking in other components of the system occurred in materials substantially different from those used at Susquehanna.
The Board finds no genuine issue of material fact to be heard here and Applicants are entitled
()
to a decision in their favor as a matter of law.
Appliants'
.- ;-~ : _
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Moti'on for Summary Disposition of Contention 7(c) is therefore granted.
Backcround:
Contentian 7 (d) reads as follows:
(d)
The ability of Susquehanna to survive antici.cated transients without scram (ATWS) remains to be demenstrated.
In this regard, reliance on probabilistic numbers, as 10-7 per year, is unwise and unsafe.
Unresolved generic safety issues, of which AhWS is one, are rarely litigated absent a showing of special circumstances involving a specific plant.
No such showing has been presented in this case.
In resolving this issue, tnen, the soard must f.
be guided by the Appeal 3 card's ruling in Gulf State Utilities Co.
(River Bend Station, Units 1 and 2), ALAB-444, 6 NRC 760 (1977), and Vircinia Electric and Pcwer Comoany (North Anna Nuclear Power Station, Units 1 and 2), ALAB-491, 8 NRC 245 (1978).
These rulings' basically hold that the primary consideration is whether the Staff review of an unresolved issue is adequate.
Technical resolution of the ATWS problem has been completed by the Staff, and a proposal for rulemaking has been submitted to the Commission.
(SbCY-80-409, dated September 4,
~ 1980).
In this. situation, the Appeal Board has upheld a Licensing Board's ruling that a facility could operate safely, even though ATWS was an unresoived safety issue, pending final Commission action.
(Northern States Pcwer Co. (Monticello, Unit 1'), ALAB-611, 12 NRC 301, 304 (1980)).
A requisite here
~
is compliance with the Staff's interim requirements.
Af ter a. thorough study od thu ATWS problem in boiling water reactors, the Staff has established the following interim requirements:
1.
Applicants must develop emergency operating procedures to recognize and handle an ATWS event; and 2.
train operators to take such actions to respond to an
/
ATWS event; and 3.
Install an automatic trip of the recirculation pumps.
I
.--20 o
The reason for the first two requirements are self-evident.
Automatic trip' of the recirculation pumps has a two-
_ fold beneficial effect.,Fj;r,st, it minfmizes tpe preysure rise
_in,,the ve,s s el in the first few seconds of the event so that the,
reactor coolant system pressure is maintained within acceptable limits by the relief. valve.
Second, it reduces the reactor thermal power output, which in turn minimizes the peak suppression pool temperature and containment pressure.
==
Conclusion:==
e,4; Applicant has implemented, or is implementing on a continuing basis, requirements 1 and 2.
(Affidavit of William L. Fiock, para, 11).
It is in the process of installing an automatic recirculation pump trip.
(pl., paras 9, 10).
The Board.therefore finds that the Susquehanna plant can be operated with no undue risk to the public from an ATWS event.
Applicants' motion for summary disposition for rontention 7 (d) ic granted.
4.
The Applicant filed a motion for summary disposition of part of contention 11 which is supported by the Staff and has not been responded to by any other party.
Background:
The contention alleges that the Applicants have f ailed to provide adequately f or safe on-site storage, for periods of up to 10 to 12 years, of spent fuel and by
~.
sj violating the standards for protection against radiation in 10 CFR. 20.1 and 2 0. lO5 (a), the project create,s an unreasonable risk of. harm to the health and_ safety of,the petitioners and
. their property.
~_
e The Applicants' motion makes the following points:
la)
With respect to the spent fuel facilities, each unit has its can storage facility, located in the reactor building and consisting of a water-filled reinforced concrete basin lined with stainless steel with racks.for storing the fuel, cranes and material. handling equipment, a heat exchanger for cooling the water purity and pumps to circulate the water; b) the O
pool walls are six foot thick reinforced concrete; a leak detection system is provided, the liner's corrosion is insignifi-cant a'nd repairs can be made, if necessary, when fuel is in the pool; c) the fuel racks are designed to withstand any significant degradation; d) the fuel pool cooling system has several backup systems and f our independent sources of make-up water for evaporative losses, if they became necessary; e) that alarms indicating high pool water temperatures, high or low water levels and high area radiation are provided in the control room; f) the design of the spent fuel racks will asure ggg the spent fuel remains in a sub-critical condition under both
1
- 22 normal and abnormal conditions; g) that the major components
)
of the system.are protected against any credible seismic event, and the possibility of the system being impacted by aircraft, spacecraft or meteors is negligible, and h) the spent storage facilities can store spent fuel safely for at least the l
duration of the operating license period or until the year 2013.
2a)
With respect to the capability of the spent fuel to be safely stored during this period, the Commission Advisory Committee on Reactor Safeguards has sta.ted that safe interim storage of spent fuel can be provided weif beyond a thirty-year period;
'b) that spent' fuel in storage is best characterized by its inactivity and thac decay heat from fission products decreases rapid,1y so that the margin of safety for the storage system increases with time in storage; c) that.any credible failures in the cooling system would only result in slow temperature increases in view of the large volumes of water in the system; d) visual monitoring of the fuel in storage is possible and that monitoring of radiation levels of the pool water and of airborne radioactive materials above the pool is pefformed frequently; e) that Zircaloy cladding surrounding fuel pellets is an important containment barrier and such fuel has been stored successfully for periods of over twenty years; h
f) that the uranium oxide ceramic fuel pellets themselves pros 1de a barrier to the leaching of radioactise material into
l basin water, and g) that encapsulation as a means of isolating defective or failed fuel in storage has been used routinely in I
Canada and could be used here if necessary.
The Staff's supporting response indicates that the spent l
fuel facilities of the Applicant including its operating I
systems and backups meet the recommendations of the appropriate l
regulatory guides and General Design Criterion.
It further substantiates that the design of the fuel was adequate to withstand storage well in excess.of the 10 to 15-year period referred to in contention 11 without a loss of integrity and that any corrosion in fuel rods during the lifetime of the plant would be of little significance.
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Conclusions:==
Based on the information contained in the motions and supporting affidavits and in the absence of any information to the contrary in the pleadings, the Board finds that the Applicant has not violated the Commission's standards for protection against radiation in the facility's on-site storag2 of spent, fuel and that there ie no genuine issue of a material fact presented by this contention.
Accordingly, the Applicant 's motien f or summary dispo-sition of this portion of contention 11 is granted.
FOR THE ATOMIC SAFETY AND
" CENSING ~^;RD Dated at Bethesda, Maryland ames P. Gleasca, Chairman this 12th day of October, 1981 ADMINISTRATIVE JUDGE
UNITED STATES OF AMERICA NUCLEI.R REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
ARIZONA PUBLIC SERVICE
)
COMPANY, et al.
)
Docket Nos. STN 50-528
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STN 50-529 (Palo Verde Nuclear Generating
)
STN 50-530 Station, Units 1, 2 and 3)
)
)
CERTIFICATE OF SERVICE I hereby certify that copies of " Joint Applicants' Motion for Summary Disposition of Intervenor's Contention No. 6B" have been served upon the following listed persons by deposit in the United States mail, properly addressed and with postage pre-paid, this 15th day of January, 1982.
Docketing and Service Section U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 Chairman, Maricopa County Board of Supervisors 111 South Third Avenue Phoenix, Arizona 35004 Dr. Richard F.
Cole Atomic Safety arid Licensing Board U.S.
Nuclear Regulatory Commiscion Washington, D.C.
20555 Atomic Safety and Licensing Appeal Board Panel U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 Ms. Patricia Lee Hourihan 6413 S.
26th Street l'ho enix, Arizona 85040
l l'
Robert M. Lazo, Esq.
Chairman, Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission
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Washington, D.C.
20555 Dr. Dixon Callahan Union Carbide Corporation P.O. Box Y Oak Ridge, Tennessee 37830 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Stephen M.
Schinki, Esq.
Office of the Executive Legal Director U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 Edwin J.
Reis, Esq.
Office of the Executive Legal Director U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Rand L. Greenfield, Esq.
Assistant Attorney General P.O. Drawer 1508 Santa Fe, New Mexico 87504
~M Charles A BTs5ho y
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