ML20008F753
| ML20008F753 | |
| Person / Time | |
|---|---|
| Site: | Summer, Wolf Creek  |
| Issue date: | 04/08/1981 |
| From: | Kniel K Office of Nuclear Reactor Regulation |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20008F754 | List: |
| References | |
| ALAB-444, NUDOCS 8104220889 | |
| Download: ML20008F753 (21) | |
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9 Uf: ITER > STATES
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April 8, 1981
. Docket tio. 50-395 MEMORANCUM FOR:
A. Schwencer, Chief 3
Licensing Branch #2, DL lJ FRCM:
Karl Kniel, Chief
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Generic' Issues 3 ranch, OST.
SUBJECT:
SER ItiPUT: VIRGIL'.C. SUFMER, UtilT NO. I
!?lant Name: Virgil C. Su m er, Unit fio. 1 Cocket Number: 50-395 Licensing Stage: OL
- esponsible Branch and Project Manager, LS*2, W. F. Kane
. CST Brant.h Involved: Generic Issues Branch L
Cescription of Review: ' Unresolved Safety-Issues Requested Cr:pletion Date: March 24, 1981 Review Status: Complete The Generic Issues 3 ranch, OST, input to the Virgil C. Summer Unit No.1 Safety Evaluaticn Report is enclosed. This appendix to the SER addresses the status of Unresolved Safety Issues pertaining to these facilit'es, and is in response to the ALAS-444 decision on this subject. That decision specified that "...each SER should. contain a summary description of those generic problems under continuing study which havc, both relevance-to facilities of the type under review and potentially significant l
l public safety implications."
Page 5 of'this Appendix references NUREG. reports providing proposed Tne Summer generic resolution to five of the Unresolved Safety Issues.
SER/SER Suoplement section discussing the plant specific implementation d the generic programs is not available at this time and should be-supplied by the Sumer Praject Manager when available. The Project Yanager should also assure that plant specific implementation of resolved l
USIs is addressed in the body of the SER.
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Karl Kniel, Chief Generic Issues 3 ranch Division of ':afety Technology
Enclosure:
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. Input to SER i
cc: w/ enclosure See next.page m_
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i 2-April 8, 1981 A. Schwence:r cc: w/ enclosure F. Schroeder N. Anderson P. Norian W. Kane J. Wilson R. Stark
- f. Peltier X. Kniel J. %rtore L. Kintner C. Anderson e
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1 APPENDIX C NUCLEAR REGULATORY COMMISSION (NRC)
UNRESOLVED SAFETY-ISSUES C.1 Unresolved Safety Issues The NRC staff continuously l evaluates the safety recuirements used in its crev'ews against new infonration as-it becomes available.
Infornation
'related to the safety of nuclear pcwer plants comes from a variety of t
l sources:includino excerience from ccerating reactors; research results; NRC staff ar.d Advisory Committee on Reactor Safeguards (ACRS) safety reviews;-and vendor, architect /engir:eer. and utility design reviews.
Each time a new concern or. safety issue is identified frem one or more cf these sources, the need for immediate action to. assure safe Operation l
is assessed. This assessmeat includas consideration of the generic imolications of the issue.
In some cases, immediate action is taken to assure safety, e.g., the derating of boiling water reactors as a result of the channel box wear problems in 1975. 'In other cases, interim measures, such as modifications to ocerating procedures, may be suff:cient to allow further study of the issue criar to making licensing decisions.
In most cases, however, the l
initial assessment indicates that imediate licensing actions or chances t
in licensing criteria are not necessary.
In any event, further study may be deemed approcriate to make judgments as to whether existing NRC l
staff recuirenents should be modified to address the issue for new f
plants or if backfitting is appropriate for the long term operation of l
olints alreadv under construction or in operation.
These issues are sometimes called " generic safety issues" because they are related to a particuiar class or type o' quclear facility rather than a scecific plant. These issues have also been referred to as
" unresolved safety issues." However, as discussed above, such issues are considered on a generic basis only after the staff has made an initial determination that the safety significance of the issue does not crohibit continued operation or require licensing actions while the lancer-term generic review is underway.
C.2 ALAB Ja4 Recuirements These longer-term generic studies were the subject of a Cecisicn by the Atomic Safety and Licensing Acceal Board of the Nuclear Regulator
- Comis sion. The Decision was issued on November 23,1977 (ALAB JaQ in connection with the Apceal Board's consideration of the Gulf States i
Utility Comcany anplication for the River Bend Station, Unit Nos..1 and l
2.
L In the view of the Appeal Board, (pp. 25-29)
L "The responsibilities of a licensinc board in the radiolonical health and safety schere are not confined to the consideraticn and C-1
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v disposition of those issues which may have been cresented to it by a party or an " Interested State" with the required degree of soecificity.
To the contrary, irrespective of what matters may or may not have been properly placed in controversy, prior to authorizing the issuance of a construction permit the board must make the finding, inter alia, that there is ' reasonable assurance" that "the prooosed facility can be constructed and operated at the proposed location Of without undue risk to the health and safety of the oublic."
necessity, this 10 CFR 50.35(a) determination ell entail an incairy into.whether the staff review satisfactorily has come to grips with any unresolved generic safety oroblems which might have an imoact upon operation of the nuclear facility under consideration."
"The SER is, of course, the crincioal document before the licensing bcard which reflects the content and outccme of the staff's sa#ety review. The board should therefore be able to iock to that document to ascertain the extent to which generic unresolved safety problems which have been previously identified in an FSAR item, a Task Action olan, an ACRS report or elsewhere have been factored into the staff's analysis for the carticular reactor--and with what result. To this end, in our view, each SER should contain a summary description of those generic problems under continuing study which have both relevance to facilities of the type under review and potentially significant public safety icolications."
"This summary description should include informaticn of the kind now centained in most Task Action Plans. More specifically, there should be an indication of the investicative program which has been or will'be undertaken with regard to the problem, the program's anticioated time span, whether (and if so, what) interim measures l
have been devised for dealing with the oroblem cending the completion i
of the investigation, and what alternative courses of acticn micht be available should the program not produce the envisaged result.'
l "In short, the board (and the public as well) should be in a cosition l
to ascertain from the SER itself--without the need to resort to xtrinsic documents--the staff's perceotion of the nature and extent of the relationship between each significant unre3olved j
generic safety question and the eventual operation of ;.S reactor l
under scrutiny. Once again, this assessment might well have a direct bearing upon the ability of the licensing board to make the l
safety findings recuired of it on the construction cer-nit level l
even though the generic answer to the cuestion remains in the offing. Among other things, the furnished information would likely shed light on such alternatively imoortant considerations as whether:
(1) the problem has already been resolved for the reactor under study; (2) there is a reasonable basis for concluding that a satisfactory solution will be obtained before the reactor is put in oceration; or (3) the problem would have no safety implications until after several years of reactor operation and, should it not be resolved by then, alternative means will be available to insure that continued oceration (if cernitted at all) would not pose an undue risk to the public."
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This appendix 's specifically included to respond to the decision of the Atomic Safety and Licensing Appeal Board as enunciated in ALAB 444, and as applied to an operating license proceeding Virginia Electric and Dower Ccmoany_-(North Anna Nuclear Power Station, Unit Nos.I and E ALA3491,NRC245(1978).
C.3 " Unresolved C fety Issues" 3
In a related matter, as a result of Congressional action on the Nuclear Regulatory Comissicn budget for Fiscal Yew 1978, the Energy Reorganization a t of 1.'74 was amended (PL 95-209) on December 13, 1977 to include, c
among other things, a new Secticn 210 as folicws:
" UNRESOLVED SAFETY USUES PLAN" "SEC. 210. The Commission shall develop a plan orovidi i for-specification and analysis of unresolved safety issue: relating to nuclear reactors and shall take such action as may be necessary to imolement corrective measures with respect to such issues. Such-clan shall be submitted to the Congress.cn or before January 1 1973 and progress reports shall be included in the annual report of the Commission thereafter."
The Joint Explanatory Statement of the House-Senate Conference Comittee for the Fiscal Year 1978 Appropriations Bill (Bil1 S.1131) provided the followino a'dditional information regarding the Comittee's deliberations on this portion of the bill:
"SECTION 3 - UNRESOLVED SAFETY ISWES" 4
"The House amendment required developms.
' a plan to resolve s
generic safety issues. The conferees as to a requirement that the plan be submitted to the Congress &
afore January 1,1978.
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. this plan should The conferees also expressed the inten' identify and describe those safet? issue-relating to nuclear 4
pcwer reactors, which are unresolve a, t'.e date of enactment.
It should set forth:
(1) Comission action taken directly or indirectly to develop and implement corrective measures; (2) farther actions olanned cencerning such measures; and (3) timetables and cost estimates of such actions. The Comission should indiccte the criority it has assigned to each issue, and the basis on which priorities have been assigned."
i In response to the reporting rege' ements of the new Section 210, the i
NRC staff submitted to Congrt January 1, 1978, a report, NUREG-04?O, entitled "NRC Program foi che Resolution of Generic Issues Related j
to Nuclear Power Plants," describing the NRC generic issues program.
The NRC program was already in place when PL 95-209 was enacted and is C-3 W
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of considerably' broader scope' than the " Unresolved S fety Issues Plan" required by Section 210.
In the letter transmitting NUREG-0410 to the 30, 1977, the Commission indicated that "the orocress Congress on December reports, which are required by Section 210 to be included in future NRC annual reports, may be more useful to Congress if they focus on the.
specific Section 210 safety items."
- It is the NRC's view that the intent'of Section 210 was to assure that plans were developed and implemented on issues with potentially significant public safety imolications.
In 1978, the NRC undertook a review of over 130 generic issues addressed in the NRC program to determine which issues fit this description and cualify as " Unresolved Safety Issues" for recorting to the Concress. The NRC review included the development of procosals by ?'.e.NRC Staff and review and final approval by the-NRC
'Conmissioners.
This. review is described in a report NURSG-0510, " Identification of Unresolved Safety Issues Relating to Nuclear Power Plants - A Report to Conaress," dated January 1979. The report provides the following definition of an " Unresolved Safety Issue:"
"An Unresolved Safety Issue is a matter affecting a number of nuclear power plants that poses important questions concerning the adecuacy of existing safety requirements for which a final resolution has not yet been developed and tnat involves conditions not likely to be accepable over Lhe lifetime of the plants it affects."
Further the report indicates that in applying th'is definition, catters that pose "important questions concerning the adequ:cy of existing safety reouirements" were judged to be those for which resolution is necessary to (1) compensate for a possible major reduction in the degree of protection of the public health and safety, or (2) provide a potentially significant decrease in the risk to the public health and safety. Ouite simoly, an " Unresolved Safety Issue" is potentisily significant from a public safety standpoint and its resolution is likely to result in NRC action on the affected plants.
I All of the issues addressed in the NRC program were systematically evaluated against this definition as described in NUREG-0510. As a -
l result, seventeen " Unresolved Safety Issues" addressed by twenty-two j
tasks in the NRC program were identified. The issues are listed bel w.
i Progress on these issues was first discussed in the 1978 NRC Annual l
Repcrt. The number (s) of the generic task (s) (e.g., A-1) in the NF program addressing each issue is indicated in parentheses following tne title.
i "UN9ESOLVED SAFETY ISSUES" (APPLICABLE TASK NOS.)
1 1.
Waterhammer - (A-1) 4 Asymmetric Blcwdown loads on the Reactor Coolant System - (A-2)
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3.
Pressurized Water Reactor Steam Generator Tube Integrity - (A-3, A-4,A-5) 4.
.BKR Mark I and Mark II Pressure Suppression Containments - (A-6, A-7,A-8,A-39)
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Anticipated Transients Without Scram - (A-9)~
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BWR Nozzle Cracking - (A-10) 7.,
Reactor Vessel Materials Toughress - (A-11)..
1 8..
Fracture Toughness.of Steam Gertrator and Reactor Coolant Pump
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Supports - (A-12)
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Systems Interaction in Nuclear Power Plants - (A-17).
- 10. Environmental ~ Qualifi:ation of Safety-Related Electrical Equipment -
l (A-24)
- 11. Reactor Vessel Pressure Transient Protection - (A-26) 12 Residual Heat Removal Requirements - (A-31)
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- 13. Control of Heavy Loads Near Spent Fuel - (A-36) 14 Seismic Design Criteria - (A 40) l L
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Pice Cracks at Boiling Water Reactors - (A-42)
- 15. ; Containment Emergency Sump Reliability - (A 43) 17 Station Blackout - (A-44)
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'n the view of the staff, the " Unresolved Safety Issues" listed above are the substantive safety issues referred to by the Appeal Board in.
ALAB-444 when it spoke of "... those generic problems under continuing study which have.... potentially significant public safety implications."
Eight of the 22 tasks identified with the " Unresolved Safety Issues" are
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not apolicable to Virgil C. Sumer Nuclear Station, Unit 1 and six of these eight tasks (A-6, A-7, A-8, A-39, A-10 and A 42) are peculiar to boiling water reactors. Tasks A 4 and A-5 address steam generator. tube problems in Combustion Engineering and Babcock and Wilcox plants. With regard to the. remaining 14 tasks that are applicable to thi.s facility',
the NRC staff has issued NUREG reports providing its-proposed reso]ution of five of these issues. Ea'ch of these have been addressed in this Safety Evaluation Report or.will be addressed in a future supplement.
The table below lists those issues and the section of this Safety Evaluation Report in which they are discussed.
l Safety Evaluation Task Number NUREG Recort and Title Reoort Section A-2 NUREG-0609, "Asytrcetric 3.9.3 31owdown Loads on PWR Primary Systems" A-24 NUREG-0588, " Interim Staff 7.7.2 Position on Environmental cualification of Safety-Related Electrical Equipment" l.
A-26 NUREG-0224, " Reactor. Vessel 5.4.2 Pressure Transient Protection i
l for Pressurized Water Reactors" l
and RSB BTP 5-2 A-31 Regulatory Guide 1.139, Will~be addressed Guidance for Residual Heat in a future Removal" and RSB STP 5-1 supplement.
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Safety Evaluation Task Number _
NUREG Recort and Title Recort Section A-36 NUREG-0612. " Control of 9.2.4 Heavy Loads at Nuclear Power Plants" The remaining issues applicable to this facility are listed in the following table:
GENERIC TASKS ADDRESSING UNRESOLVED SAFETY ISSUES THAT ARE APPLICASLE TO THE VIRGIL C. SLMMER. UCLEAR 5TATION, UNIT 1 N
A-1 Waterhammer 2.
A-3 Westinghouse Steam Generator Tube Integrity 3.
A-9 Anticipated Transients Without Scram 4
A-11 Reactor Vessel Materials Toughness 5.
A-12 Potential for Lcw Fracture Toughness and Lamellar Tearing on PWR Steam Generator and Reactor Coolant Pump Supoorts 5.
A-17 Systems Interactions in Nuclear Power Plants 7.
A-40 Seismic Design Criteria 3.
A 43 Containment Emergency Sump Reliability 9.
A 44 Station Blackout With the exceotion of Tasks A-9, A 43, and A-44, Task Action Plans for the generic tasks above are included in NUREG-C649, " Task Action Plans for Unresolygd Safety Issues Related to Nuclear Power Plants." A technical resolution for Task A-9 has been proposed by the NRC staff in Volume 4 of NUREG-0460, issued for comment. This served as a basis for the staff's proposal for rulemaking on this issue. The Task Action Plan for Task A-43 was issued in January 1981, and the Task Action Plan for A 44 was issued in July 1980. Draft NUREG-0577 which represents staff resolution of USI A-12 was issued for comment in November 1979. The Draft NUREG contained the Task Action Plan for A-12.
The information provided in NUREG-0649 l
reets most of the informational requirements of ALAB-444 Each Task tction Plan provides a description of the oroblem; the staff's approaches to its resolution; a general discussion of the bases upon which continued plant licensing or operation can proceed pending ccepletion of the task; the technical organizations involved in the task and estirates of the mantcwer required; a deecription of the interactions with other NRC offices, the Advisory C.amittee on Reactor Safeguards and outside organi:ations; estimates of funding required f ar contractor supplied technical assistance; proscective dates for completing the task; and a description of potentia?
oroblems that could alter the planned approach on schedule.
In addition to the Task Action Plans, the staff issues the " Office of Nuclear Reactor Regulation Unresolved Safety I; sues Summary, Aqua Book" (NUREG-0606) on a quarterly basis which provi;es current schedule information for each of the " Unresolved Safety Issues." It also includes information relative to the implementation status of each " Unresolved Safety Issue" for which technical resolution is complets.
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We.have reviewed the nine " Unresolved Safety Issues" listed above as they
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relate to Virgil C. Summer Nuclear Station,- Unit.1. - Discussion of-each of these issues including references to related discussions in the Safety Evaluation Report are provided-below in Section C.S.
Based on our review of these. items, we have concluded, for the reasons set forth
'in Section C.5, that there is reasonable assurance that this facility can be operated prior to the ultimate resolution of these generic issues without endangering the health and safety of the public.
.C.4 New " Unresolved Safety Issues" An in-depth and systematic review of generic safety concerns identified since January 1979 has been perforced by the staff to determine if any of Lthese issues should be designated as new " Unresolved Safety Issues."
.The candicate issues originated frem concerns identified in NUREG-0660,.
"NRC Action Plan as a Result of the TMI-2 Accident;" ACRS recommendations; abnormal cccurrence reports and other operating experience.
The staff's
-proposed list was reviewed and commented on by the ACRS, the Office of Analysis and Evaluation of Operaticnal Data (AE00) and the Office of' Policy Evaluation. The ACRS and AE00 also proposed that several additional
" Unresolved Safety Issues" be considered by the Commission. The Cocmission considered the above information and ' approved the following four new " Unresolved Safety Issues:"
A 45 Shutdown Decay Heat Ramoval Reqcirements
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A 46 Seismic Gualification of Equipment in Operating Plants A-47 Safety Implications of Control Systems A-48-Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment A description of the above process together with a list of the issues censidered is present in NUREG-0705, " Identification of New Unresolved Safety Issues Relating to Nuclear Power Plants, Special Report to Congress," dated March 1981. An expanded discussion of each of the new
" Unresolved Safety Issues" is also contained in NUREG-0705.
The applicability and bases for licensing prior to ultimate resolution of the four new USIs for Virgil C. Summer, Unit 1 are discussed in Section C.5.
J C.5 Discussion of Tasks as they Relate to Virgil C. Summer Nuclear Station, Unit 1 A-1 Waterhammer Waterhammer events are intense pressure pulses in fluid systems
. caused by. any one of a number of mechanisms and system cunditions..
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Since 1971 there have been over 100 incidents involving waterhammer in pressurized water reactors and boiling water reactors.
The water-hamers have involved steam generator feedrings and pipino, decay heat removal systems, emergency core cooling systems, containment spray lines, service water lines, feedwater lines and steam lines. However, the systems most frequently af fected by oaterhammer effects are the feedwater systems. The most serious waterhammer events have occurred in the steam generator feedrings of pressurized water reactors. These types of waterhamrer events are addressed in Section 10.4.3 of this Safety Evalua-tion Recort.
With regard to protection agains: other ootential waterham er events cur-ently provided in olants, piping tesign codes recuire consideration of imcact leads. Apprcaches used at the design stage include:
(1) increasing valve closure times, (2) piping layout to creclude water slugs in steam lines and vapor formation in water lines, (3) use of snubbers and pice hangers, and (4) use.of vents and drains.
In addition, as described in Section 3.9.2 of this Safety Evaluation Report, we require that the applicant conduct a preoperational vibration dynamic effects test program in accordance with Section III of the ASME Code for all ASME Class 1 and Class 2 piping systems and oiping restraints during startup and initial operation. These tests will provide adequate assurance that the piping and piping restraints have been designed to witnstand dynamic effects due to valve closures, pump trics and other operating redes associated with the design operational transients.
acnetheless, in the unlikely event that a large pipe break did* result i
from a severe watarhamer event, core cooling is assured by the emergency core cooling systems described in Section 6.3 of this Safety Evaiuation Recort and protection acainst the dynamic effects of such pipe breaks inside and outside of containment is provided as described in Section 2.6 of this Safety Evaluation Report.
Task A-1 may identify some potentially significant waterhamer scenarios that have not explicitly been accounted for in the design and operation i
of nuclear power plants.
The task has not as yet identified the need for recuirin, any additional measures beyond those already required in the short term.
Based on the foregoing, we have concluded that the facility can be operated prior to ult mate resolution of this generic issue without undue risk i
to the health and safety of the public.
A-3 Westinghouse Steam Generator i~ube Integrity The primary ccncern is the capability of steam cenerator tubes to maintain their integrity during normal operation and postulated accident conditions.
In addition, the recuirements for increased steam generater tube inspections and repairs have resulted in signiff-cant increases in occupational exposures to workers. Corrosion resulting in steam gencrator tube wall thinning (wastage) has been observed in C-8
several Westinghouse plants for a number of years. plants operating r
exclusively with an all volatile secondary water treatment. process have not experienced this form of degradation to date. Another major corrosion-
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related phenomenon has also been observed in a number of plants in i
recent years, resulting from a buildup of-support plate corrosion products l'
in.the annulus between the tubes and the support plates.
This buildup eventually causes a diametral~ reduction of the tubes, called " denting,"
and deformation of the tube support plates. This phenomenon has led to other problems, including stress corrosien cracking, leaks at the tube / support plate intersections, and U-bend section cracking of tubes which were l
highly stressed because of support plate deformation.
acecific reasures such as steam generator design features and a secondary water chemistry control and monitoring program, that the ' applicant has employed to minimize the onset of steam generator tube problems are described in Section of this Safety Evaluation Repcrt.
In addition, l
Section of this Safety Evaluation Report discusses the inservice inscection requirements. ' As described in Section
, the applicant has met all current requirements regarding steam generator tube integrity.
The Technical Specification will include requirements for actions to be taken in the event that steam generator tube leakage occurs during plant operation.
Task A-3 is expected to result in improvements in'our current ry;Jirements for inservice inspection of steam generator tubes. These impretements will include a better statistical basis for inservice inspectisn program
' requirements and consideration of the cost / benefit of increased inspection, pending completion of Task A-3, the measures taken at this facility should minimize the steam generator tube problems encountered.
Further the inservice inspection and Technical Specification requirements will assure that the applicant and the NRC staff are alerted to tube degradation should it occur. Appropriate actions such as tube plugging, increased and more frequent inspecticns and power derating could be taken if Since the improvements that will result from Task A-3 will necessary.
be procedural, i.e., an improved inservice inspection program, they can be implemented by the applicant after operation of this facility begins, if necessary.
Based on the foregoing, we have concluded that this facility can be ocerated prior to ultimate resolution of this generic issue withcut
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undue risk to the health and safety of the public.
A-9 Anticicated Transients Without Scram Nuclear plants have safety and control systems to limit the consequences of temporary abnor;nal operating conditions or " anticipated transients."
Scre deviations from normal operating conditions may be minor; others, occurring less frequently, may impose significants demands on plant equipment.
In some anticipated transients, rapidly shutting down the nuclear reaction (initiating a " scram"), and thus rapidly reducing the generation of heat in the reactor core, is an important safety measure.
-If there were a potentially severe " anticipated transient" and the l
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reactor shutdown system did not " scram" as desired, then an " anticipated ~
transient without scram," or ATAS, would have occurred.
The anticipated transient without scram issue and the requirements that l
must be met by. the applicant prior to operation of the facility are l
discussed in Section 15.3.5 of this Safety Evaluation Report.
The ATWS issue is currently scheduled for rulemaking in mid-sumer 1981.
The' applicant will be required to comply with any further requirements l'
on AP45 which may be imposed as a result of the rulemaking.
Based on our review, we have concluded that there is reasonable assurance that this facility can be operated pricr to ultimate resolution of this i
generic issue without endangering the health and safety of the public.
A-11 Reactor Vessel Materials Touchness Resistance to brittle fracture, a rapidly propagating catastrophic l
failure mode for a component containing flaws, is described quanti-l tatively by a material property generally denoted as " fracture toughness."
l Fracture toughness has different values and charactersitics depending l
upon the material being considered. For steels used in a nuclear reactor pressure vessel, three considerations are important.
First, fracture toughness increases with increasing temperature; second, fracture toughness decreases with increasing load rates; and third, fracture toughness decreases with neutron irradiation.
In recognition of these considerations, power reactors are cperated within restrictions imposed by the Technical Specifications on the l
cressure during heatup and cooldown operations. These restrictions assure that the reactor vessel will not be subjected to a combination of pressure and temperature that could cause brittle fracture of the vessel if there were significant flaws in the vessel materials. The effect of neutron radiation on the fracture toughness of the vessel material is accounted for in developing and revising these Technical Specification limi tations.
For the service times and vperating conditions typical of current operating plants, reactor vessel fracture toughness for most plants provides adequate margins of safety against vessel failure under operating, testing, maintenance, and anticipated transient conditicns, and accident conditicns I
over the life of the p V. Mcwever, results frem a reactor vessel surveillance program aac cualyses performed for up to 20 older ocerating pressurized water reactors and those for some more recent vintage plants will have marginal toughness, relative to required margins at normal full power after comparatively short periods of operation.
In addition, results from analysas performed by pressurized water reactor manufacturers indicate that the integrity of some reactor vessels may not be maintained-in the event that a main steam ifne break of a. loss-of-coolant accident l
occurs after approximately 20 years of operation. The princioal objective l
of Task A-11 is to develop an improved engineering method and safety.
criteria to allow a more precise assessment of the safety margins that C-10 l
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e are available during normal operation and transients in older reactor vessels with marginal fracture toughness and of the safety margins available during accident conditions for all plants.
Based on our evaluation of this' facility's reactor vessel materials toughness, we have conciuded that this unit will have adequate safety margins against brittle failure during operating, testing, maintenance and anticipated transient conditions over the life of the units.
Since Task A-11-is. projected to be cor:pleted well in advance of this facility's reactor vessel reaching a fluence level which would noticably reduce fracture resistance, acceptable vessel integrity for the postulated accident conditions will be assured at least until the reactor vessel is j
l reevaluated for long-tern acceptability.
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- n addition. 'the surveillance program required by 10 CFR 50, Apoendix H l
will afford n opportunity to reevaluate the fracture toughness pericdically "uring the first half of design lifc.
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Therefore, based upon the foregoing, we have concluded that this facility-l can be operated prior to resolution of this genaric issue without undue risk to the health and safety of the public.
l A-12 Fracture Toughness of Steam Generator and Reactor Coolant Pumo Succorts Curing the course of the licensing action for North Anna Pcwer Station Unit No.1 and 2 a number of questions were raised as to the potential for lamellar tearing and low fracture toughness of the steam generator and reactor coolant pump support materials for those facilities. Two different steel specifications (ASTM A36-70a and ASTM A572-70a) covered most l
of the material used for these supports.. Toughness tests, not originally l
soecified and not in the relevant ASTM specifications, were made on those i
heats for.vhich excess material was available. The toughness of the A36 steel was found to be adequate, but the toughness of the A572 steel was 0
relatively poor at an operating temperature cf 80 F.
Since similar materials and designs have been used on other nuclear plants, the concern: regarding the supports for the North Anna facilities are applicable to other P'4R olants.
It was therefore necessary to reassess the fracture tcughness of the steam generator and reactor coolant puro support materials for all operai.ing PiR plants and those in C? and OL review.
I NUREG-0577, " Potential for low Fratture Toughness ar.d Lamellar Tearing on pWR Steam Generator and Reactor Coolant Pump Supports," was issued l
for ccmment in November 1979. This report surrari:es work performed by
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the NRC staff and~ its contractor, Sandia Laboratories, in the resolution of this generic activity. The report describes the technical issues, the technical studies performed by Sandia Labcratories, the NRC staff's technical positions based on these studies, and the NRC staff's plan for implementing its technical positions. As a part of initiating the implementation of the findings in this report, letters were sent to all applicants and licensees on May 19 and 20, 1980.
In these letters a revised proposed imolementation plan was presented and specific criteria for material qualifications were defined.
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organizational units and assign secondary respcnsibility to other units where there is a functional or interdisciplinary relationship.
Cesigners folicw somewhat similar procedures and provide for interdisciplinary reviews and analyses of systems. Task A-17 provided an independent study of methods that could identify important systems interactions adversely impacting safety; and which are not considered by current review procedures. The first phase of this study began in May 1978 and was completed in February 1980 by Sandia Laboratories under contract to the NRC staff.
The Phase I investigation was structured to identify araas where inter-actions are possible between and amcng systems and have the potential of negating or seriously degrading the performanue of safety functions.
The study concentrated on common cause en linking failures ameng systems that could violate a safety function. The investigation then icentified where NRC review procedures may not have properly acc0unted for these interacticns.
The Sandia Study used fault-tree methods to identify component failure combinations (cut-sets) that could result in loss n# a safety function.
The cut-sets were reduced to minimal combinaticns by incorporating six common or linkiog systems failures into the analysis. The results of the Phase I effort indicate that, within the > cope of the study ani; a few areas of review procecares need improvement regarding systems interaction.
Howeve, the level of detail needed to identify all examples of potential system interaction candidates observed in some operating plants was not within the Phase I scoce of the San,dia Study.
It i; expected thet the development of systemat5 ways to identify and evaluate systems interactions will reduce the likelihood of common cause l
failures resulting in the loss of plant safety functions. However, the studies to date indicate that current review procedures and criteria supplemented by the application of post-TMI findings and risk studies orovide reasonable assurance that the effects of potential systems interaction on plant safety will be within the effects on plant safety creviously evaluated.
Therefore, we concluded that there is reasonable assurance that Virgil C. Summer, Unit I can be operated prior to the final resolution of this generic issue without endangering the health and safety of the public.
A a0 Seismic Cesian Criteria - Short-Term Procram NRC reguiations recuire that nuclear power structures, systems and components important to safety be designed to withstand the effects of natural phenomena such as earthquakes. Detailed requirements and guidance regarding the seismic design of nuclear plants are provided in the NRC regulations and in Regulatory Guides issued by the Commission. However, there are a number of plants with construction permits and operating licenses issued before the NRC's current regulaticns and regulatory guidance were in place. For this reason, rereviews of the seismic design of various plants are being undertaken to assure that these plants do not present an undue risk to the public. Task A 40 is, in C-13
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i effect, a compendium of short term efforts to supcort such reevaluation efforts of the NRC staff, expecially those related to older operating
- plants.
In addition, some revisions to the Standard Review Plan sections and Regulatory Gudes to bring them more in line with the state-of-the-art will result.
As discussed in Section 3.7 of this Safety Evaluation Report the seismic design basis and seismic design of the facility have been evaluated at the operating license stage and have been found acceptable. 'de do not expect the results of Task A 40 to affect these conclusions because the techniques under consideration are essentially those utilized in the review of this facility. Should the resolution of ask A 40 indicate a change is needed in licensing requirements, all operating reactors, inclucing Su=er will be reevaluated on a :ase-by-case basis. Accordingly, we nave concluced that this facility can be operated prior to the ultimate resolution of this generic issue without endangering the health and safety of the public.
A 43 Containment Emergency Sumo Reliability
. ollowing a postulated loss-of-ccolant accident, i.e., a break in the reactor coolant system piping, the water ficwing from the break would be collected in the emergency sump at the low coint in the containment.
This water would be recirct.iated through the reactc-cystem by the emergency core cooling pumps to maintain core coling.
This water would also be circulated through the containment spray system to remove heat and fission products from the containment. Loss of the ability to draw water frcm the emergency sump could disable the emergency core cooling and containment spray systems.
One postulated neans of losing the ability to draw water from the emergency sump could be blockage by debris. A principal source of such debris could be the thermal insulation on the reactor coolant system piping.
In the event of a piping break, the subsequent violent release to the high pressure water in the reactor coolant system could rip off the insulation in the area of the break. This debris could then be swept into the sump, potentially causing blockage.
Currently, regulatory positions regarding rump design are presented in
'legulatory Guide 1.32, " Sumps for Emergency Core Cooling and Containment icray Systems," which address debris (insulation).
Regulatory Guide 1.32 rec:ncends, in addition to providing redundant separated somos, that two orotective screens be provided. A low approach velocity in the vicinity of the sump is required to allow insulation to settle out 4
before reaching the sump screening; and it is required that the sump remain functional assuming that one-half of i.he screen surface area is blocked.
A second postulated means of losing the ability to draw water from the emergency sump could be abnormal conditions in the sump or at the pump inlet such as air entrainment, vortices, or excessive pressure drops.
These conditions could result in pump cavitation, reduced ficw and possible damage to the pumps.
Currently, regulatory positions regarding sump testing are containedin Regulatory Guide 1.79, "Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors," which addresses the testing of the recirculation function. Both in-plant and scale model tests have been performed by applicants to demonstrate that circulation through the sump can be reliably accomplished.
As indicated in Section 6.3.3 of this Safety Evaluation Report, the applicant will perform out-of-plant scale model tests of the containment sump design. The applicant will be required to demonstrate that there is reasonable assurance that the sump design will perform as expected fallowing a loss-of-coolant accident.
The near tem implementation of Task A-43 for this facility is excected to be crecedural in nature and assure adequate housekeecing and emergency accordingly, l
- rocedures to succlement the sumo tests discussed above.
we have concluded that this facility can be operated prior to ultimate i
resolutien of this generic issue without endangering tne health and l
safety of the public.
l A Ja Station Blackout Electrical power for safety systems at nuclear pcwer plants must be supplied by at least two redundant and independent divisions. The systems used to remove decay heat to cool the reactor core following a i
reactor shutdcwn are included among the safety systans that must meet these requirements. Each electrical division for safety systems includes an offsite alternating current pcwer connection, a standby emergency diesel generator alternating current power supply and direct current sources.
Task A-44 involves a study of whether or not nuclear power plants should be designed to accomodate a complete loss of all alternating current
- cwer, i.e., loss of both the offsite and the emergency diesel generator siternating current power supplies. This issue arose because of operating experience regarding the reliability of alternating current pcwer supplies.
A number of operating plants have experienced a total loss of offsite l
l electrical power, and more occurrences are expected in the future.
":uring each of these loss of offsite pcwer events, the ensite emergency l
alternating current power supplies were available to scoply the pcwer needed by vital safety equipment. However, in scme instances, cne of t
l the redundant emergency poaer supplies has been unavilable.
In addition, there have been numerous reports of emergency diesel generators failing l
to start and run in operating plants during periodic surveillance l
tests.
A loss of all alternating current power was not a design basis event for
-he Sumer facility. Nonetheless, a combinaticn of design, operation and testing requirements that have been imposed on the applicant will assure that these units will have substantial resistance to a loss of all alternating current and that, even if a loss of all alternating current shculd occur, there is reasonable assurance that the core will be cooled. These are discussed below.
A loss of offsite alternating current power involves a loss of both the preferred and backup sources of offsite pcwer. Our review and basis for C-15
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- acceptance of the design, inspection, ~and testing provisions for the offsite power system are described.in Section 8.2 of_this Safety Evaluation
/
-Report.
If.offsite power is lost, two diesel generators'and their associated
~
L distribution systems will. deliver emergency power to safety _-related equipment. - Our -review of the design,. testing, surveillance, and maintenance provisions for_ the onsite emergency diesels is described:in Section 8.3
. of the:SER. Our requirements include preoperational testing to assure the reliability of the. installed diesel generators in accordance with our
. requirements discussed in the SER.
In addition, the applicant has:been reouested to implementLa program for enhancement of diesel ' generator L
l reliability to better. assure the long-term reliability of the diesel l
generators. This' program resulted from reccmmendations of.NUREG/CR-l-
C660, " Enhancement of Cnsite Emergency Generator Reliability."
Event if both offsite and onsite alternating current power are' lost, cooling water can-still be provided to the steam generators -by the l
- auxiliary feedwater system by employing a steam turbine driven pumo l
..that does not rely on alternat ng current power for operation. Our i
review of the auxiliary feedwater system design and oceration is described in Section of the Safety Eva?uation Report.
The issue of station blackout was also considered by the Atomic Safety and Licensing Appeal Soard (ALtB-603) for the St. Lucie Unit No. 2 facility.
In addition, in view of the completion schedule for Task A-44 (October 1982), the Appeal Board recommended that the Commission take expeditious action to ensure that other plants and their operators are equipped to accommodate a. station blackout event. The Cemission has reviewed this recommendation and determined that some interim measures i
should be taken at all facilities including Surme while Task A 44 is being conducted. Consequently, interim emergency procedures and operator L
training for safe operation of the facility and restoration of alternating l
current powcr will be required. The staff notified the applicant of these requirements in a letter from D. Eisenhut, NRC, to the apr s lant dated February 2S, 1981. We wil; condition the crarating license for Suc=er that their procedures and training be completed by fuel load date.
Based on the above, we have concluded that chere is reasonable assurance that Succer can be operated prior to the ultimate resolution of this l
generic issue without endangering the health and safety of the public.
L45 Shutdcwn Decay Heat Removal Reouirements Under nonnal c 3ratino :.onditions, power generated within a reactor is removed as steam to produce electricity via a turbine generator.
Following a reactor shutdown, a reactor produces insufficient power to operate the turbine: however, the radioactive decay of fission products contir' es to produca heat (so-called " decay heat"). Therefore, when l
re3cm shutdown occurs, other measures must be available to remove.
deccy heat from the reactor to ensure that high temperatures and pressures do not develop which could jeopardize the reactor and the.aactor ccolant system.
It is evident, therefore, that all light water reactors (LWRs) share two common decay heat removal functional requirements:
(1)to
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provide a means of transferring decay heat from the reactor coolant system to an ultimate heat sink and (2) maintain sufficient water inventory inside the reactot vessel to ensure adequate cooling of the reactor fuel. The reliability of a particular power plant to perform these functions depends on :ne frequency of initiating events that require or jeopardize decay heat removal operations and the probability that required systems will respond to remove the decay heat.
This Unresolved Safety Issue will evaluate the benefit of providing l
alternate means of decay heat removal which could substantially increase the plants' capability to handle a broader scectrum of transients and l
accidents. The study will consist of a generic system evaluation and aill result in recommendaticns regarding the desirability of and cossible l
esign requirements for incrovements in existing systems or an alternative
- ecay heat remeval methed if the improvements or alternative can significant ecuce the overall risk to the public.
j The primary method for rem 0 val of decay heat from pressurized water reactors is via the steam generators 'a the secondary system. This energy is transferred on the secondary side to either the main feedwater or auxiliary feedwater systems, and it is rejected to either the turbine l
l condenser or the atmosphere via the steamline safety / relief valves.
Folicwing the TMI-2 accident, the importance of the auxiliary feedwater l
system was nighlighted and a number of steps were taken to improve the reliability of the auxiliary % dwater system. The staff's review of these items is contained in Section of this Safety Evaluation eport.
It was also stipulated that plants must be capable of providing i
the required AP4 flow for at least two hours from one auxiliary feedwater
- ump traf a, independent of any alternating current power source (that l
is, if both off-site and on-site alternating current cower sources are lost),
pressurized water reactors also have alternate means of removing decay heat if an extended loss of feedwater is postulated.
This method is l
known as " feed and bleed" and uses the high pressure injection system to
'dd water coolant (feed) at high pressure to the primary system. The decay heat increases the system pressure and energy is removed througn the pcwcr-operated relief valves and/or the safety valves (bleed), if recessary.
l l
At low orimary system pressure (belcw about 200 psi), the long-term cacay heat is re: roved by the residual heat remeval system to achieve cold shutdawn conditions.
Based en Jr foregoing, we have concluded that 'lirgil C. Sunrer, Unit 1
- an be operated prior to uitimate resolution of this generic issue without endangering the health ard safety of the public.
A 46 Seismic Oualification of Ecuitment in Cceratina Plants The design criteria and methods for the seismic quaiification of mechanical and e?ectrical equipment in nuclear pcwer plants have undergene significant change during the course of the comercial nuclear power program.
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Consequently, the margins of safety provided in existing equipment to resist seismically induced loads and perform the intended safety functions may vary considerably. The seismic qualification of the equipment in oparat.ng plants must, therefore, be reassessed to ensure the ability to
'oring the plant to a safe shutdown condition when subject to a seismic event. The objective of this Unresolved Safety Issue is to establish an explicit set of guidelines that could be used to judge the adequacy of the seismic qualification of mechanical and elec*rical equipment at all operating plants in lieu of attempting to backfit current design criteria I
for new plants. This guidance will concern equipment required to tafely shut down the plant, as well as ecuipment whose function is not required l
for safe shutdcwn, but whose failure enuld result in adverse conditions anich might impair shutdown functions.
i l
Virgil C. Summer Unit 1 was designed wing current seismic criteria and the design nas been reviewed and approved by the Commission staff in i
accordance with current design criteria and methods for seismic qualifica-ti on. Therefore, we conclude that Virgil C. Summer Unit 1 can be
- cerated prior to resolution of this generic issue without undue risk to the health and s'afety of the public.
l l
A 47 Safety Imolications of Control Systems l
l This issue concerns the potential for transients or accidents being made l
more severe as a result of control system failures or malfunctions.
l These failures or malfunctions may occur independently or as a result of the accident or transient under consideration. One concern is the otential for a single f ailure such as a loss of a cower supply, short circuit, open circuit, or senscr failure to cause simultaneous malfunction of several control features. Such an occurrence could conceivably result in a transient more severe than those transients analyzed as anticipated operational occurrences. A second concern is for a postulated accident to cause control syste.n failures which could make the accident ore severe than analyzed. Accidents could conceivably cause control system failures by creating a harsh environment in the area of the control ecuipment or by physically damaging the control equipment.
It is genenlly believed by the staff that such control system failures would not lead to serious events or result in conditions that safety systems cannot safely handle. Systematic evaluations have not been rigourously performed to verify this belief. The potential for an accident that could arfect a partisular.ontrol system, and effects of the control system failures, may differ from plant to plant. Therefore, it is not possible to develop generic answers to these concerns, but rather plant-specific evaluations are required. The purpose of this Unresolvcd Safety Issue is to define generic criteria that will be used for pien*-specific evaluations.
The Sumer control and safety systems have been designed with the goal of ensuring that control system failures will not prevent automatic or manual initiation and operation of any safety system equipment recuired to trip the p' ant or to maintain the plant in a safe shutdown condition C-18 i
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a' following any " anticipated operational occurrence" or " accident." This has been accomplished by either previding independence betaeen safety and non-safety systems or providing isolating devices between safety and non-safety systems. These devices preclude the propagation of non-safety system equipment faults to the protection system. This ensures that operation of the safety system equipment is not imoaf red.
A systematic evaluation of the control system design, as contemplated for this Unresolved Safety Issue, has not been performed to deternine whether postulated accidents could cause significant centrol system failures wnich would : ake the accident consecuences more severe than cresently anal
.ed.
Mcwever, a wide range of bounding transients f
and accidents is presently analy::ed to assure tnat the postulated events sucn as steam generator overfill and overcooling events would be adecuately mitigated by the safety systems.
In addition, systematic reviews of control system failures (single or multiple) goal of ensuring that safety ;ystems have been oerformed with the will not defeat safety system action.
Based on the above, we have concluded that there is reasonable assurance that t ? Su:ver Unit can be ocerated prior tc the ultimate resolution of this generic issue without enc;ngering the health and safety of the public.
A a8 Hydrogen Control Peasures and Effects of Hydrocen Burns on Safety Ecuioment Following a loss-of-coolant accident in a light water reactor clant, combustible gases, principally hydrogen, may accumulate inside the primary reactor containment as a result of:
(1) metal-water reaction involving the fuel element cladding; (2) the radiolytic decomposition of i
the water in the reactor core and the containment sumo; (3) the corrosion of certain construction materials by the spray solutiu1; and (a) any synergistic chemical, thermal and radiolytic effects of post-acciden',
envir;nmental conditions on containment protactive coating systems and electric cable insulation.
3acause of the potential for significant hydrogen generation as the result of an accident,10 CFR Section 50.44, " Standards for Combustible 3as Control System in Light Water Cooled ;cwer Reactors" and the General
- esign Criteria al, "Containnent Atmosphere Cleanup" in Aopendix A to 10 CR Part 50 require that systems be ::rovided to centrol hydrogen
- encentraticns in the containment atmosphere folicwing a postulated accident to ensure that containment integrity is maintained.
10 CFR Section 50.44 requires that the combustible gas control' system provided be capable of handling the hydrogen generated as a result of degradation of the emergency core cooling system such that the hydrogen release is five times the amount calculated in demonstrating compliance with 10 CFR Section 50.46 or the amount corresponding to reaction of the cladding to a depth of 0.00023 inch, whichever amount is greater.
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y The accident at TMI-2 cn March 28, 1979 resulted in hydrogen ceneration well in excess of the amounts specified in 10 CFR 50.44.
As a result of this knowledge it became apparent to NRC that specific design measures are needed for handling larger hydrogen re. leases, carticularly for smaller low pressure containments. As a result, the Commission deteaained that a rulemaking proceecing should be undertaken to define the manner and extent to which hydrogen evolution and other effects of a degraded core need to be taken into account in plant design.
An advance notice of this rulemaking proceeding on degraded core issues was published in the Federal Register en October 2,1980.
Recocni::ing that a num:er of years may be recuired to ccr:olete this rulemaking oroceeding, a set of sncet. term or interim actions reia.ive to nydrocen control recuirements were develcced and implemented. These interim -easures were described in a seccnd October 2,1980 Federal Register notice. For plants with large dry containments such as Virgil C.
3umer, Unit 1, no near-tern mitigation measures are required by the interim rule.
The Virgil C. Summer p.lant has about two :nillion cubic feet of net free
/olume. Assuming 30 to 50% metal-water reaction in the core, the resulting
.niformly mixed concentration of hydrogen in the containment will range f rca 5 to 10".
This is well below the concentrations for detonation and even belcw the limits for cor..b s' ion if there were more than 50% steam in the containment atmosphere.
Cesign pressure of the Virgil C. Summer plant is 57 psig. Analyses performed on the Zion and Indian Point plant; show that the failure pressures are greate-than twice the design pressures.
If the substantial amount of metal-water reaction were to occur shortly folicwing anset of a large LOCA and while the containment is still near its peak pressure, the pressure increase caused by the noncondensible hydrogen gas and its associated exothermic formation energy will be substantially less than the failure pressure.
If the metal-water reaction were to occur well af ter onset of the large LOCA, then the containment heat renoval system would have condensed much of the steam in the containrent and reduced the containment pressure.
This would provide a substantial margin for accomrrodating the hydrogen generated by the I
metal-water reaction.
In addition, the "Short Tern Lessons Learned" frcm the 111I-2 accident have been implemented on th6 Virgil C. Summer plant. This action will reduce the likelihood of accidents that could lead to substantial amounts of metal-water reaction.
Accordingly, pending resolution of this Unresolved Safety Issue and the rulemaking proceeding on hydrogen generation, the Virgil C. Summer olant can be operated without undue risk to the health and safety of the :ublic.
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