ML20080Q518
| ML20080Q518 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 10/04/1983 |
| From: | Kniel K Office of Nuclear Reactor Regulation |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20079F427 | List:
|
| References | |
| FOIA-84-35, TASK-A-45, TASK-OR ALAB-444, NUDOCS 8310130378 | |
| Download: ML20080Q518 (26) | |
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.'o, UNITED STATE' s
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NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D. C. 20555
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OCT 0 4 1983 l
Docket Nos. 50-400 50-401 MEMORANDUM FOR:
George Knighton, Chief Licensing Branch No. 3 Division of Licensing FROM:-
Karl Kniel, Chief.
Generic Issue.s Branch-Division of Safety Technology
SUBJECT:
SER INPUT - SHEARON HARRIS NUCLEAR STATION, UNITS 1 AND 2
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PLANT NAME:
Shearon Harris Units 1 and 2 DOCKET NUMBER:
50-400 and 50-401 LICENSING STAGE:
OL LICENSING BRANCH AND PROJECT MANAGER:
LBf3, B. Buckley DST BRANCH INVOLVED: Generic Issues Branch DESCRIPTION OF REVIEW:
Unresolved Safety Issues REVIEW. STATUS: Complete.
REQUESTED COMPLETION DATE: NS The Generic Issues Branch, DST, input to the Shearon Harris Units 1 and 2 Safety Evaluation Report is enclosed.
This appendix to the SER addresses the status of Unresolved Safety Issues pertaining to these facilities, and is in response to the ALA8-444 decision on this subject.
That decision specified that "...each SER should contain a sunnary description of those generic problems under continuing study which have both i'
relevance to facilities of the type under review and potentially significant public safety it. plication."
This appendix references NUREG reports providing proposed generic resolution cf seven of the Unresolved Safety Issues.
The Shearon Harris SER sections discussing the plant-specific infonnation for the generic programs are not available at this time.
However, the " standard" SER sections were assumed and should be verified by the Project Manager.
The Project Manager should also assure that.the plant-specific implementation of resolved USIs is addressed in the body of the SER.
As stated in Section C.4 of the Appendix C, we conclude that Shearon Harris can be operated prior to the ultimate resolution of all USI's applicable to this facility. However, we recognize that there are open items identified XA Copy Has Been Sent 10 PDR
OCT 0 4 1983 2.
l by the line branches for the systems related to USIs A Thus, the listed A-45 and A-47 as confirmatory items rather than open items.
i conclusion for these two USIs states that the closure of these i subject to the satisfactory resolution of the open items.
It should be noted that no direct reference is made in this appendix to NUREG-0800, since the Standard Review Plan does not directly address US However, for.'many opert issues,.this appendix refers to functional branch reviews referenced in the SER for'.the licensing basis. :These functional branch reviews are addressed by NUREG-0800.
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Generic Issues Branch Division of Safety Technology Enclosurest Shearon Harris Appendix C-USIs-
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.F. Schroeder T. Novak N. Anderson P. Norian R. Silver w/ enclosures
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, APPENDIX C.
' NUCLEAR REGULATORY C0fetISSION (NRC)
C.I Unresolved Safety Issues The NRC staff evaluates the safety requirements _used in its reviews against new infomation as it becomes available.
Infomation related to the safety of nuclear power plants comes from a variety of sources.
including experie~nce frour operating reactorsi research results; NRC staff and Advisory Comittee on Reactor Safeguards (ACRS) safety reviews; and vendor, architect / engineer and utility design reviews. After the accident at TMI the Office, for Analysis and Eviluation of Operational Data was' established to provide a systematic and continuing review of operating experience.
Each.
, time a new concern or safety issue is identified from one or more of these
..* -r-sourcesisthe nose femismettetes acties=turassure safe
- operattom. istassessed w.m This. assessment includes consideration of the generic implications of the issue.
In some cases, imediate action is taken to assure safety, e.g., the derating of boiling water reactors as a result of the channai box wear problems in
~1975.
In other cases, interim measures, such as modifications to operating
., procedures, may be. sufficient.to allow further study,of the. issue prior to
-making licensing" decisions.
In most cases', however, the. initial' assessment '
indicates that imediate licensing actions or changes in licensing criteria are not necessary.
If the issue applies to several or a class of plants the issue is evaluated further as a " generic safety issue." This evaluation considers the safety significance of the issues, the cost to -implement any changes in plant design. or operation and other significant and relevant factors to establish a priority ranking of the issue.
Based on this ranking.
resolution of the issus is scheduled for near tem resolution deffered j
until resources become avaiTable or dropped from further consideration.
j The issues with the highest priority ranking are reviewed to determine whether they should be designated as " unresolved safety issues" (NUREG-04.10, "NRC Program for the Resciution of Generic Issues Related to Nuclear Power Plants," dated January 1,.1978).
However, as discussed above, such issues are considered on a generic basis only after the staff has made an initial detemination that the safety significance of the issue'does not prohibit continued operation or require licensing actions while the longer tem generic review is under way.
C.2 ALAB-444 Requirements l
These longer-term generic studies were the subject of a Decision by the Atomic Safety. and Licensing Appeal Board of.the Nuclear Regulatory j
Comission.
The Decision was issued on November 23,1977(ALAB-444)in connection with the Appeal Board's consideration of the Gulf States Utility Company application for the River Bend Station, Unit Nos.1 and 2.
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'In the view of the Appeal Board, -
4 "In short, the board (and the public as well) sh'ould be in a position' to-ascertain from the SER itself--without the need to resort to extrinsic documents--the staff's perception of the nature and extent of the relationship between each significant unresolved generic safety question Once again',
and the eventual operation of the reactor under scrutiny.
this, assessment might well have a direct bearing upon the ability of the licensing board to make the. safety findings required of it on the construction permit level even though the generic. answer to the question remains in the offing.
Among other things, the. furnished infomation would likely shed light on,such alternatively important considerations (1) the problem has already been resolved -for the reactor as-whether:
under study; (2) there is a reasonable basis for concluding that ali b
'. %.hsstisfactoutsolution wi operatiodiorWtheVproMas useM have nom. safety imp
.~.2 by then, alternative means will be available to insure that continued operation (if pemitted at all) would not pose an undue risk to the publie."
- .This appendix is specifically includpd to respond to the d applied to an operating license proceeding Virginia Electric 245 0978).
C.3 " Unresolved Safety Issues" In a related matter, as a result of the CongressionaT action on the Nuclear Regulatory Consiission budget for Fiscal Year 1978, the Energy Reorganizati 12, 1977 to include, among Act of 1974 was amended (PL 95-209) on December other things, a new Section 210 as follows:
)
UNRESOLVED SAFETY ISSUES PLAN
)
SEC. 210 The Commission-shall develop a plan providing for specification and analysis of unresolved safety issues relating to nuclear reactors and shall take such action as may be necessary toSuch plan implement corrective measures with respect to such issues.
'sha7T be submitted to tt'e Congress on or before January 1,1978 and progress reports shall be included in the annueT report of'the Commission thereafter The Joint Explanatory Statement of the House-Senate Conference Committee for the Fiscal Year 1978 Appropriations Bill (Bill S.1131) provided the following additional infomation regar. ding the Committee's deliberations on this portion of the bill:
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'SECTION 3 - UNRESOLVED' SAFETY ISSUES The House amendment required development of a plan to resolve generic The conferees agreed to a requirement that the plan be safety issues.
The conferees submitted to the Congress on or before January 1,1978.
also expressed the intent that this plan should identify and describe those safety issues, relating to nuclear power reactors, which are It should set forth:
(1) unresolved on the date of enactment.
Commission actions taken directly or ir.directly to develop and implement corrective measures; (2) further actions planned concerning such.
The measures; and (3) timetables and cost estimates of such actions.
Commission should indicate the priority it has assigned to each issue, and the basis on which priorities have been assigned.
L, response,ta the reporting requirements of the new Section 210, the NRC t-1staff 'submftted toPCoep.aone dannery l.,,1978 a report NUREG-0410, entitled "NRC Program for the Resolution of Generic Issues Related to Nuclear.,
The NRC program Power Plants," describing the NRC generic issues program.was already broader scope than the Unresolved Safety Issues Plan required by Section 210.
30, 1977, In the letter transmitting NUREG-0410 to the Congress on December the Commission indicated:
"the progress reports, which are required by Section 210 ta be included in. future NRC. annual imports, nay be more useful to Congress if they focus on the s'pecific Section.210' safety items."
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It is.the NRC's view that the intent of Section 210 was to ensure that plans
.E were developed and implemented on issues with potentially significant public In 1978, the NRC under.cok a review of more than 130 safety implications.
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. generic issues addressed in the NRC. program to determine The NRC review included the development of proposals by the NRC Congress.
staff and review and final approval by the NRC Commissioners.
This review is described in a report NUREG-0510. " Identification of Unresolved Safety Issues Relating to Nuclear Power Plants - A Report to Congress,." 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 adequacy of existing safety requirements fc? which a final resolution has not yet been developed and that involves conditions not likely to be acceptable over the lifetime of the plants it affects.
Further the report indicates that in applying this definition, matters that pose "important questions concerning the adequacy of. existing safety (require-1) ments" were judged to be those for which resolution is necessary to compensate for a possible major reduction in the degree of protection of the public health and safety, or (2) provide a potentially significant decrease
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in the risk' to the. pubife health 'and safety - Quite. simply, 'an " Unresolved Safety Issue" is potentially significant from a public safety sta'ndpoint arid its resolution is likely to result in NRC action in the affected' plants.
All of the original issues addressed in the NRC program were systematically As a result, evaluated against this definition as described in NUREG-0510.
17 Unresolved Safety Issues addressed by 22 tasks in the NRC program were identified.
l An in-depth and systematic ' review of generic safety concern The if'any of these issues should be designated as Unresolved Safety Issues.
candidate issues originated from concerns identified in NUREG-0660, "NRC Action Plan as a Result of the TMI-2 Accident"; ACRS recommendations;The staff's rts; and other operating experience.
proposed Tist' was ewed W co m em,b S the Office of
' " + 4 ahmeemek. occurrence, Analysis and Evaluation of Operational Data (y,the.ACR o AE00), and the' Offfee of"Pelicy" t e-The ACRS and AE00 al",o proposed that several additional The Comission
_ Evaluation.
Unresolved Safety Issues be considered by the Consission.
considered the above information and approved the four Unresolved Safety l
A description of the review process of candidate l
Issues A-45 and A-48.
. issues, together with a list of the issues considered is presented in NUREG-0705,'" Identification of New' Unresolved Safety Issues Relating to I
An Nuclear Power Plants,. Special Report to Congress," dated March 1981.
expanded discussion of each of the new Unresolved Safety Issues is als l
contained in NUREG-0705'.
the Cessiission approved another issuc., A-49, Pressurized Thennai Shock, as an l
l Unresolved Safety Issue in December 1981.
The number (s)- of the generic task (s) (for The issues are listed beTcw.
example, A-1) in-the NRC program addressing each issue is indicated. in l
I parantheses following the title. Unresolved Safety Issues (Applicable Task Nos.)
(1) Waterhanner (A-1)
(2) Asymmetric. Blowdown Loads on the Reactor Coolant Syst BWR Mark I and Mark II Pressure Suppression Containments (A-6,. A-7, A-8, (3J (4)
A-39)
(
. Anticipated Transients Without Scram (A-9)
BWR Nozzle-Cracking. (A-10),
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Reactor Vessel Materiais Toughness (K-II)
Fracture Toughness of Steam Generator and Reactor Coolant Pump Surwts l
(A-12)
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'9)) Systems Interactions in Nuclear Power Plants (A-17) Envir
{10 (A-2.)
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) Reactor Vessel Pressure Transienk;-Protecticu.(A-26).
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) Residual Heat Removal Requirements (A-31)-Control of Heav
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(1 Seismic Des'gn Criteria (A-40)
, Pipe Cracks at BoiTing Water Reectors (A-42))
) Containment Emergency Sump Reliability (A-43 Station Blackout (A-44)
)) Shutdown Decay Heat Removal Requirements (A-4 l
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[ f Safety Implications of Control Systems (A-47)). Hydroge f
Equipment (A-48)
(22) Pressurized Thermal SSock (A-49) f "**+3ssthegviens of..tha' staff, the unresolv6d safety issues lis
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l spoke of "...those generic problems under continuing study whfcfV have.m.%
Ten of the tasks potentially significant public safety implications."
licable to Shearon Harris,, and six of these ten tasks (y issues are not appA-6, A-identified with the unresolved safet Tasks A-4-and A-5 address steam peculia: to boiling water reactors.
generator tube problems in Combustion Engineering and B l
i PWR plants with Tee-condenser containments or'BWR piants withWith regard plants.
pressure-suppression type containments. tasks that 6re applicable to l
Each of reports providing its proposed resolution of seven o addressed in a future supplement.
sectfort of this, Safety Evaluation Report in which they are discussed.
Safety Evaluation Report Section
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NUREG Report and Title Task No 3.9.2.3 NUREG-0609, "Asystemetric Blowdown Loads A-2 on PWR Primary Systems"
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If.3.8 NUREG-0460, Vol. 4. " Anticipated Transients Without Scram for Light Water Reactors" A-9 5.3 NUREE-0744, " Resolution of the Task A-11 l
A-11 Reactor Vessel Materials Toughness Safety Issue," Vols. I and II, Rev.1, October 1982 3.11 NUREG-0588, Rev. 1 " Interim Staff A-24
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Position on Env'ronmentai Qualiffcationof Safe 5.2.2 NUREG-0224. " Reactor Vessel Pressu d RS8 A-26 BTP 5-2 5.4.3 SRp 5.4.7 and BTP 5-1 " Residual HeatRemov 31
- A-31 9.1.4 NUREG-0612. " Control of Heavy Loads at l
A-36 Nuclear Power Plants" listed in the following T e reYe'fiefiuf tssuescappliqable., to this facility are
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A,1 ' Water HannerWestinghouse Steam Generator Tube Integrity Tearing on A-IZ Potential for Low Fracture Toughness A-3 ts l '
A-17 Systems. Interaction in Nuclear Power P an
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A-40' Seismic Design Critertia.A-43 containment Eme A-44 Station Blackout A-45 Shutdown Decay Heat Removai Requirements A-47 Safety Implicaticns of Control Systems A-49 Pressurized Thermal Shockfor the gen.aric tasks up With the exception of A-12. all task actions plansNUREG-0649,"
to and including A-40 above are included infor Un Plants."
Task which represents staff resolution of USIThe NUREG contained d
November 1979. action plans fer later tasks were issued indivi ua table below..
Task Action Plans for Selected USIs_
Issue Task Task Action Plan Number-01/81 A-43 07/80 A 10/81 A-45 06/82 (Rev. 1) 05/82 A-46 b
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_., A-48 12/82 A-49 03/82 Each task action plan provides a description of the protlem; the staff's approach to its resolution; a general discussion of the bases upon which continued plant licensing or operation can. proceed pending completion of the task; the technical organizations involved in the task and estimates of the manpower required; a description of the interactions with other NRC offices, the Advisory Connittee on Reactor Safeguards and outside organizations; estima'tes 'of funding required for contractor supplied technical assistance; prospective dates for completing the task; and a description of potentia'l problems that could alter the planned approach or schedule.
In addition to the task action plans, the staff issues the " Office of Nuclear Reactor Regulation Unrosolved Safety Issues Sunnary, Aqua Book" (NUREG-0606) uom.aigearterly basis.swMch,penuidet,surrent. schedule information.for each.of,,., a~
the Unresolved Safety Issuer.
It also includer information relative to the m.
implementation status of each Unresolved Safety Issue fur which technical
-resolution is complete.
The staff has reviewed the ten Unresolved Safety Issues listed above as they l
relate to Shearon Harris. Discussion of each of these issues including
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references to.related discussions in'the Safety Evaluation Report are previded below in'Sectforr C'.4.
Ba' sed on its review of these items,.the' staff concluded, for the reasons set forth in Section C'.4, that, with exception of A-45 and A-47, Shearon Harris can be operated prior to the ultimate resolution of these generic issues without endangering the health and safety of the public Tasks A-45 and A-47 are accepted subject to the resolution of those cpen items identified in Section C.4.
The resolution of these open items wiTT be~ reported in a supplement to the SER C.4 Discussion of Task as They Relate to Shearon Harris Units I and 2 This section provides the NRC staff's evaluation of Shearon Harris' Units 1 and 2 for each of the applicable unresolved safety issues.
This includes the staff's bases for licensing before ultimate resolution of these issues.
The staff's conclusions are based in part on infomation provided by the applicant in his letter of July 30,1982 from M.,A. McDuffie to H. R. Denton.
l A-1 Water Hammer i.arge hydrauTie' Toads carr be caused by water' hanner events which are-initiated by steam bubble collapse, steam-driven slugs of water, pump startup into voided. lines, and rapid or improper valve closures.
Of the approximately 150 water hanner events which have been reported in nuclear power plants since 1968, the reported damage has been principally confined to pipe hanger and snubber damage, and associated pipe movement.
Of these
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occurred during preoperationaplant operationl i l operation thus suggesting a lf.have reported events.,about hathe first year of commerc a Safety Issue (USI.) A-1 is near ng completion.
i d stress the need toleedring steam learning process.
are reported in NUREG-0927 ant rs to plant The' resolution of Unresolved ts (e.g., use of J-tubes in toplead to fenerators,) jockey pumps and views The technical findings maintain proven design concepoperating condition 3
to d revisions to SRP Sections 3.9., h rev In conjunction wit l
47 are reflected in the propose 6.1. 9.2'.1, 9.2.2, 10.3, and 10...
and system
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- nerators, s Westinghouse SideKIFsteam geThis typ w.m.
avoiding water hammerWPW ;
SRP 10.
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The Shearon Harris plant util ze occurrence (as discussed inconducting a preo h at design.
l which are a bottom-feed, pre eless suceptible to wa to. verify transfer heater nozzle applicant has committed to
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operation to main feedwater/ pre-This type of 19(10.4.7)
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..Nonetheless,.the per response.to Question 410.without incurrin
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top-feed' auxiliary feedwater ipe break occurring (due to' a veredund ibed 02 operationneeds outlined in BTP ASB-1..
j In the unlikely event of a large p t is provided as described inains is assured by the water hammer) core cooling in Section 6.3, and protection ag inside, and outside containmen t the A-1 evaluations, we calculate tha hammer event is very low and the Shearon Harris FSAR.
Thus, based on our concluding significantly damaging waterl nt design and operat to final probability of a i
ris plant can be operated pr orhealth and sa be required.
therefore-concludes that p a being planned will not likely e without undue risk to the concluded that the Shearon Harresolution of this tubes to maintain of the public.
ity A-3 Steam Ge'nerator J p e Integr ident conditions.
capability of steant generatorl oper i n in several I
The primary concern is thetheir integrity during norm have experienced tube degradat o stress corrosion cracking,
i n are discussed below, an k
wastage, intergranular attac,
WestYnghousesteamgenerators Each of these forms of degradat o their occurrence at Shearon These are forms.
and denting. specific measures to prevent
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. (1) Wastage is characterized by general loss of metal from the tube wall due to a chemical corrosive reaction. Wa' stage has occurred only -in' steam generators'which used sodium phosphate as a chemical additive to: control the pH of steam generator secondary water.
The Shearon Harris steam i
generators will use a water treatment consisting of hydrazine and asunonium hydroxide (this is called aTT volatfle treatment or AVT).
Wastage has not been observed in steam generators using all volatile chemistry control.
1 (2)
Intergranular attack is a chemical reaction wherein the grain boundaries l
of the.Inconel.600 tube are attacked by either acidic or caustic
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solutions.~ Significant intergranula'r attack has occurred only in steam generators which have used sodiunt phosphate as a chemical additive and i
have an open crevice in th~ tube to tubesheet area. ' In the Shearon e
Harris steam generators, there is no open crevice in the tubesheet area, and there will be no sodium phosphate' addition.
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(3) Stress corrosion cracking has occurred in a number of' Westinghouse,, steam generators.
If one excludes stress corrosion cracking due to denting
-(to be discussed next), the most significant instances of stress corrosion cracking have occurred in the narrow radius V bend area of the tubes on the bundle interior. This problem has been restricted to Westhinghouse Model 51 steam generators l constructed in the early 1970's.
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.These steam, generators have part'icularly. tight U bends.which apparently has created; excessive residual ' tress s The Shearon Harris steam.
s generators are Westinghouse ModeT D which have smaller diameter tubes bent on a larger bend radius. Thus, the residual stress at the U bend area is lowered, and stress corrosion cracking will be precluded.
4.
Denting is the most serious degradation problem encountered in Westinghouse steam generators. Denting is caused by. rapid corrosion of the tube support pTates at the holes where the tubes pass through the support plates. Denting is known to be caused by operation of the steam generators outside the allowable range of water chemistry control parameters, specifically during times of major condenser leakage.
It appears that the-use of copper materials in the feed and condensate systems contributes to the severity of denting.
The following actions will be taken to prevent denting:-
(a) Under the sponsorship of the Steam Generator Owners Group, a set of chemistry guidelines for the satisfactory operation of steam generators has been written. These guidelines proy % based.on the best available research data on the steam generator tube corrosion problem, limits which will ensure that problems such as denting do not occur.
The applicant commits to operate in accordance with these guidelines.
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avoided in the condenser and feed systems.
blowdown been an important design feature of the con ble l
l during I
blowdown flow from the steam generator.
system has.been provided to permit proper chemistry contr ill layup periods.. Full flow condensate polish d to tion.-
maintain chemistry, within required limitis during opera reliability t
There is no industry experience available to assess the long-er l
CP&L will maintaint to ensure that d l D steam ge.nerator.
i communications wftft utfiftfes whicts operate similar equ pmenlconsid
"**op the hiestinghouse. Mo e i
their experience is properly factored ints the operat onaCP&L Included in the work i
and fully supports the objectives of that organiz Shearon Harris.
tions of the chemistry of the SGOG are Mng-tenu evaluations of the applica l tical guidelines, imp a degradation mechanics, and development of ana y causes of knowntechniques to%a:t steam generator performance.
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^been,a problem at
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d Tube vibration in the region near the feedwater nodle tas Modifications tor.
earlier units with the Westinghouse Model D steam i
blem. The staff's evaluation of these modifications is contained in NUR blished).
Model D4/D5 Preheat Steam Generator Modifications" (to be pu i
f cility should Pending completion of Task A-7, the measures taken at th s a
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d Further the minimize the steam generator tube problems encoimtere.
ill assure inservice inspection and Technical Specification requirements w b
that the applicant and the NRC staff are alerted to tu e d more frequent inspections and power derating coul ssary. Since cted to be occur.
the improvements that will result from Task A-3 are expe ators, they can be procedural, i.e.... improved inspection of the steam generimp ility begins, if t
Harris plant can necessary.
Based on the foregoing, we have concluded tha ithout undue risk to the health and safety of the public.
A-12 Steam Generator and Reactor Coolan't Pump Support _
Power Station Units During the course of the licensi'ng action for Nor tial for lamellar
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tsaring and low fracture' toughness of the s e Two different steel
) covered most of f.he m tcH al pump support materials for those facilities. speci heats for which h
used for these supports.in the relevant ASTM specifications, excess material was available.relatively poor at an operatin ther nuclear plants.
f 80*F.
It Similar materials and designs hav'e been used on o fracture toughnsss of'steami ls for al was, therefore, necessary to reassess the
" generator and reactor coolant pump support mater a plants and those undergoing CP and OL review.
t system component supports l
3.9.3.4 provides results
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of the piping and supports stress evaluation.
ts of Table NF-2331-1 of Supports were purchased and tested to the requiremen Carolina Power and Light Company maintain Requirements, which is ASME Section III of S&PV Code. membership on the AIF l
. currently:doing,research and tes
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materials.
designed to the provisions The staff has completed all technical work on US t
bears directly on Shearon Harris is that suppor sdequate fracture of ASME Code,Section III, Subsection NF, ha were designed in j
t i
n with respect to the accordance with the requirements of Subsect o adequacy of these components.
d in a recommended ported in NUREG-0577, The completion of work on USI A-12 has resulte implementation plan based on the technical findings reis, l
i nts and Revision 1, and the associated value-impact ana ysTh
) Section 5.3.4d into the drawn into a proposed new Standard Review Plan as Appendix C.
NRC SRP and be ap(plied only to Construction Perm h
(NUREG-0800). PDA). applications (providing the PDA includes irements will not apply tofor c structures) for PWRs* whichFhave not been gran e Design Approval publication date. Therefore, the proposed requ t
licensees of operating PWRs or to current a t
The staff has concluded that the Steam Generato fracture toughness and therefore, the ultimate resolution of t
supports at Shearon Harris have adequa e h and safety.of the public.
with regard to USI A-12 can be operated before this issue withough undue risk to the healt
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Plants _
A-17 Systems Interaction in Nuclear Powerwas initiated in May 1 The concern arises because the design,ana The staff's systems interaction program such as civil, electrical, questions d2finition of USI A-17.
h installation of systems are frequently t e Experience at operating plants has led to i
engineers with, functional specialit es--
l specialists is sufficiently e interactions among systems.
mechanical, or nuclear.
i ast might have been prevented.,1f of whether the work of these funct ona d
integratad to. enable them to minimize a vers ll Some adverse eye.nts that occurred in the pindepe the teams had assur2d the necessary t described a comprehensive program condi.tions of operation.
systems, and components i i
The Carolina PoweridgMCompany bas < no m.sv.m n
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that separately evaluates all structure,
d t d without endtngering the health"an tions.
t licensing to safety for adverse systems interacthe Shearon The plant has been evaluated against curren rinciple of defense-in-depth.
irements such as physical safety of the public. requirements that are founded on t e p h
t safety systems, and-protectionm
, ' ' Adherence to..this principle results in requ separation and independence of-redondanTine ruptures,d'sab
~
d against hazards such as high energy' reviews of safety-grade equipment an factors,an flooding seismic events, fires, human Also, the quality assurance program provisions are subject to disciplinaryaddress po for construction and operational phases of introducing adverse systems i
i which is. followed during the des gn,.each' plan d procedures i
f ty.
provide an adequate degree of plant sa e rganizational units and a interactions.
i The NRC staff's current procedures ass gn here there is a functionaldures a of various technical areas to specific o secondary responsibility to other units wDesigners fol Task A-17 has been developing ystems interactions which may i ws.
analyses of systems and interface rev e interface.
methods that identify potential adverse shave dures.
study began in.May.1978 and was completeL ff.
to identify areas where interaction the potential of negating or degrad
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The Phase I investigation was structured cause The study concentrated on coninonThe i are possible between systems and have fety function.
the performance of safety functions. failures of s L
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. was;to :then. identify where NRC review procedures may.not have properly accounted'for these interectionn
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The sindia Laboratories used fault-tree analysis on the selected plant design to identify component failure combinations (cut-sets) that could result in loss of a safety function. The cut-sets were further reduced by incorporating six linking systems failures into the analysis.
The results of the Sandia effort indicated a few potentially adverse systems interactions within the limited scope of the study. The staff reviewed the interactions
.for safety. significance and generic implications. The staff concluded that no. corrective measures needed to be' implemented insnediately'except for the '
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potential interaction between the PORY and its block valve.
This' interaction had been separately identified by the evaluations of the TMI-2 accident while Sandia was studying the selected plant.
Since corrective measures were already being implemented no separate measures, were needed under USI A-17.
-
- Resed en the fbregofsp dtscussfee,;.we conefude..that Sheeren Harris /can
.mm operate safely pending resolution of UST A-17.
After the resolution of UST A-17, we will determine whether Carolina Power and Light Company must perform further evaluations for adverse systems interaction.
A-40 Seismic Design Criteria - Short Term Prograar f
NRC... regulations require.that., safety-telated nuclear power pl. ant, structures, systems and components be ~ designed to withstand the effects of natural phenomena such as earthquakes. Detailed requirements and guidance regarding i
the seismic design of nuclear plants are provided in the NRC regulations and in Regulatory Guides issued by the Connission.
However, there are a number of plants with constructiorr permits and operating licenses issued before the l
NRC's current regulations and regulatory guidance were in place.
For this reason, further reviews, of,the seismic design of various plants are being undertaken-Task A-40 fr,. irr effect,. a compendiunt of short-tenn efforts to support such reevaluations by the NRC. staff in particular, those related to older operating plants.
Safety-related structures, systems, and components for Shearon Harris are designed to withstand the effects of earthquakes in accordance with current NRC. regulations, regulatory guides, and SRPs, as discussed in Section 3.7 of the FSAR. Specifically,the subjects identified in the NRC's problem description for A-40, i.e., magnitude of earthquakes (SSE), free-field motion (SSE), motion of plant equipment, and load combinations are discussed therein.-
As discussed in Section 3.7 of this Saf.ety Evaluation Report the seismic
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design basis and seismic design of the facility have been evaluated at the operating license stage and have been found acceptable.
We do not expect the results of Task A-40 to affect these conclusions because the techniques under consideration are essenttally whose utilized in the review of this facility.
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. Should.the resolution of Task A-40 indicate.a change' is needed.in licens.ing requirements, all operating. reactors, including Shearon. Harris will. be reevaluated on a case-by-case bas'is. Accordingly, we have concluded that' this facility can be operated prior to the ultimate resolution of this generic issue without undue risk. to the health and safety of the public.
l A-43 Containment Emergency Sump Performance Following a postulated LOCA, water would be collected in the containment emergency sump for use in the long-term recirculation mode, thus maintaining.
core cooling. Th.is water could also be circulated through the containment spray cooling system for removal of heat and fission products within 1
containment. The principal safety concern is loss of the ability to draw water from the containment emergency sump under post-LOCA conditions--thus leading to the degradation of, or disability of, the long-tenn recirculation safe,ty tr_ain.,, and. impainment of d,.ecay heat r.emoval.
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Two major concerns have been postulated, namely:
(a) adverse hydraulic conditions in the sump (e.g., air ingestion, break flow effects, vortex formation,. etc.) thereby leading to RHR pumping loss,end (b) severe sump screen blockages resulting from LOCA-cenerated insulation debris, which could cause loss of NPSH requirements.
The ' evaluation.of such-safety concern's hassoeen carried.out and th'e technical <
' findings are repo'rted in NUREG-0897. The result has been a reconenended
-revision to RG I.82 which reflects these findings. The destruction of plant insulation by the LOCA jet is viewed as a pote'ntial safety concern relative.
to screen blockage. The. evaluation of debris blockage is a plant specific requirement due to desige differencer and types of insulation employed. Air ingestion and vortex formation, are not as serious as previously hypothesized. NUREG-0897 and NUREE-0869 (which includes the proposed RG 1.82, Revision 1) and SRP 6.2.Z', were issued for public consnent in May 1983.
The requirements that may result from A-43 'are expected to be primarily procedural, i.e., an assessment of sump blockage following a postulated LOCA.
Plant modifications,. if necessary, can be implemented after operation of this facility begins.
The staff has reviewed the applicant's sump design; the results of this review are documented in Section 6.2.2 of this Safety Evaluation Report.
Based on the above, we have concluded that there is reasonable assurance that Shearon Harris can be operated prior to the ultimate resolution of this
' generic issue without undue risk to the health and; safety of the public.
~ A-44 Station Blackout Electrical power for safety systems at nuclear power plants must be supplied by at least two redundant and independent. divisions.
The systems used to
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remove decay heat to cool the reactor core following a reactor shutdown are included among the safetf systems that must. meet these requirements. Each electrical division for safety systems' includes an offsite alternating current power 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 accommodate a complete loss of all alternating current power, i.e., loss of both the offsite and the emergency diesel generator alternating current power supplies. This issue arose because of operating experience regarding the reliability, of alternating current power supplies. A number of operating plants have' experienced a total loss of offsite electrical power, and more occurrences are expected in the future.
In almost every one of these loss-of-offsite-power events, the onsite emergency alternating current power supplies were available immediately to supply the power needed by vital safety equipment. However, in some instances, one of the redundant emergency
< " power'suppYfes hai"beervufravaffabTe.*Ws:few cases there has been a complete loss of alternating current power, but during these events, alternating current was restored in less than a few minutes without any serious consequences.
In addition, there have been numerous reports of emergency diesel generators failing to start and run in response to tests conducted at operating plants.
A loss of;all ac power is not.a desig'n. basis event for Shearon Harris.
Nonetheless, the combination of design, operation, and testing requirements
~.
that have been imposed on the applicant will assure that these units will have substantial resistance to this event and that even if a loss of all ac power should occur, there is reasonable assurance that the core will be 1
- cooled. These are discussed below.
A Toss of offsite alternating current power involves a loss of both the' preferred and backup sources of offsite power. Our review and basis for acceptance of the design, inspection, and testing provisions for the offsite pow'er system are described in Section 8.2 of this Safety Evaluation Report.
If offsite ac power is lost, two independent and redundant onsite diesel generators per unit and their associated 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 8.3.1 of the SER. Our requirements include preoperational testing to assure the reliability of the installed diesel generators in accordance with the provisions of Regulatory Guide 1.108.
In addition,. the applicant has been required to implement a program for cnhancement o'f diemT generator reTiability to better assure the long-tenn reliability of the diesel generators. This program resulted from recommendations of NUREG/CR-0660, " Enhancement of Onsite Emergency Generator Reliability."
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Even if both offsite'and'onsite ac' power are lost, cooling. water from the c::ndensate storage tank can'still be provided to the steam generators by the auxiliary feedwater system by employing a turbine driven pump that does not A reliability analysis of the auxiliary Tely on ac power for operation feedwater system has been performed by the applicant and the results This analysis presented to the NRC in response to NUREG-0737, Item II.E.1.1.
Our review of the auxiliary considers the complete loss of ac power.
feedwater system design and op'eration is described in Section 10.4.9 of the,
l Safety Evaluation Report In addition to the above, the Commissi6n has detemined that.seme interim-measures should be taken at all plants to accommodate a station blackout pending resolution of the issue.
Consequently, the NRC reques,ted (Generic Letter 81-04, dated Feburary 25,1981) a review of plant operation to
.datermina the applicant's capability to mitigate a station blackout event and properly ~fspfementras necessary, emergency procedures,.and training programs.
Appropriate review of' the' procedures and ' '
m for station blackout events.
training programs for station blackout events will be completed before fuel load. ~
Based on the above, we have concluded that there is reasonable assurance that Shearon Harris can be. operated prior to the ultimate resolution of this
. generic issue without undue. risk. to.,the health and-safety of the public.
A-45 Shutdown Decay Heat RemovaT Requirements Under normal operating conditions, power generated within a reactor is removed at steam to produce electricity through a turbine generator.
Following a reactor shutdown, a reactor produces insufficient power to operate the turbine; however, the radioactive decay of fission products.
continues to produce heat (so-called " decay heat"). Therefore, when reactor shutdown occurs, other measures must be available to remove decay heat from the reactor to ensure that high temperatures and pressures do not develop It is that could jeopardize the reactor and the reactor coolant system.
evident, therefore, that all light-water reactors (LWRs) share two common (1) to provide a means of decay-heat-removal functional requirements:
transferring decay heat from the reactor coolant system to an ultimate heat l-sink and (2) maintain sufficient water inventory inside the reactor vessel to The reliability of a particular cnsure adequate cooling of the reactor fuel power plant to perfom these functions depends on the frequency of initiating Gwents that require crejeopardize decay heat removal operations and the probability that required systems will respond to remove the decay heat.
The TMI-2. accident demonstrated how a relatively comon fault, which the operatur should have been able to cope with easily, could escalate into a potentially hazardous situation, accompanied by severe financial losses to
~
b tihe u'tility, owing to difficuities arising-in..the decay. heat removal..(D.HR) process.
Other circumstances, of a more unusual nature (e.g., damage to systems by external events such as floods or earthquakesvor by sabotage), which could make removal of the decay heat difficult, can also be foreseen.
The question arises, therefore, whether current licensing design requirements are adequate to ensure that LWRs do not pose unacceptable risk as a result of failure to ' remove shutdown decay heat, and.whether,. at a cost consnensurate
'with 'the increase in safety which.could be achieved, improvements could be.
.I made in the effectiveness of shutdown decay heat removal in one or more l
transient or accident situations.
Resolution of this question is considered to be of sufficient importance to merit raising it to the status of a USI.
%.To.so.ie. extent,. the effectiveness, of the. DHR systems is linked to that of the cnsite and offsite eTectricirT suppTf3is-the intrfomance and reifabflity of -
those supplies is being considered in USI A-44, " Station Blackout."
Consequently, the scope of work required in relation to the decay-hea.t-removal systems is complementary to Task A-44 above.
The overali purpose of Task' A-45 is to evaluate the a'dequacy of current licensing design requirements, in order to ensure that nuclear power plants
'do not' pose'an unacceptable risk' because of'.: failure to remove shutdown. decay..
heat. This w1TT require the development of a comprehensive and consistent ~
set of shutdown cooling requirements for existing and future LWRs, including the study of alternative means of shutdown decay heat removal and of diverse
" dedicated" systems for this purpose.
This USE will evaluate the benefit of providing alternate means of decay heat
. removal which could substantiaTly increase'the plant's capability to handle a broader spectrum of transients and accidents. The study will include a number of plant-specific DHR systems evaluations and will result l
in reconsnendations regarding the-desirability of, and possible design requirements for, improvements in existing systems of an alternative decay-heat-removal method, if the improvements or alternatives can significantly reduce the overall risk to the public in a. cost-effective manner.
An integrated systems approach to the problem will be employed. Accordingly, quantitative methods will be used, where possible to define design requirements for future plants and to measure the effectiveness, and
'acceptabiYity ~ of the shutdown decay-heat-removat systems irr existing plants.
The principaT means for removing the decay heat in a PWR under nomal conditions immediately following reactor shutdown is through the steam generators, using the auxiliary feedwater system.
In addition to the WASH-1400 study.. later reliability studies and related experience from the c
9
Three Miie I'sland Unit'2 (TMk 2)* accident have reaffiiined.that~ the loss of a.
l capability to remove heat through the steam generator is a significant contributor to the probability of a core melt event. The staff's review of j
the auxiliary feedwater system design and operation is described in Section 10.4.g of this SER.
It should be noted as discussed below that many improvements to the steam generator auxiliary feedwater system were required of the licensees by the 1
However,. the staff still believes that NRC following the TMI-Z=. accident.
providing an alternative means of decay heat removal 'could substantially.
i increase the plant's capability to deal with'a broader spectrum of transients and accidents and potentially could, ther. fore, significantly reduce the e
Consequently, this unresolved safety issue will overall risk to the public.
investigate alternative means of decay heat removal in PWR plants, including This study will but not limited, to ; using existing equipment where possible.
fncTude'Y representatfve sampfe of plant-specif.ic decay-heat-removal-systems m
It will result in recommendations regarding' the adeq'uacy of' evaluations.
.cxisting decay-heat-removal requirements and the desirability of,, and.
possible design requirements for, an alternative decay-heat-removal method, other than that normally associated with the steam generator and secondary coolant system.
~
The auxiiiary.feedwater (AFW). system.is. a. very important safety system in a PWR in terms of providing a heat ' sink 'via the steam generatdrs to remove core As mentioned above, the TMI-2 accident and subsequent studies decay heat.
As discussed have further highlighted the importance of the AFW systems.
below,, the NRC staff has required certain upgrading of the auxiliary feedwater systems for all LWRs following the TMI-2 accident. Although this USI.will investigate alternative means of decay heat removal, the NRC staff concludes that in general (not ort a plant-specific basis) if the licensees l
comply with the upgrading of requirements for the AFWS,. the action taken following the TMI-2 accident justifies continued operation and licensing
(
I pending completion of~this USI Further discussion and the bases for this view are provided below.
TMI-2 Accident 28, 1979 involved a main feedwater transient l
The accident at TMI-2~ on March i
coupled with a stuck-open pressurizer power-operated relief valve and a temporary failure of the auxiliary feedwater system, and subsequent operator l
The intervention to severeTy reduce. flom from the safety injection system.
l
-resulting severity of the ensuing events and the potential generic aspects of l
the accident on other operating reactor 1ed the NRC to initiate prompt l
(1) ensure that other reacter licensees, particularly those with action to:
plants similar in design to TMI-2, took the necessary action to substantially reduce the likelihood for TMI-2-type events, and (2) investigate the potential gendric implications of this a.ction on other operating reactors.
l i - -..
The Bulletins & Orders Task Force (B&OTF) was established Office of Nuclear Reactor Regulation (NRR) in early May 1979 and its work on December 31, 1979.
and directing the TMI-2-related staff activities associa generic evaluations of loss-of-feedwater transients and small-break loss-of-coolant accidents for all operating plants to ensure safe operation.
summarizes the results of the work perfomed.
Generic and Plant-Specific Studies For B&W-designed operating reactors, an initial NRC staff study was co t of
.and, published in NUREG-0560, " Staff Report on the Generic Assessmen FeedwaterTransients ffy Pressurized Water. Reactors Designed by the Ba This study conside' red the'particular design features and operational history of B&W-designed operating plants in light of the T Wilcox Company."
As a result; of this accident and related current licensing requirements. study, a n pursued.
Generally, the' activities involving the B&W-desig actions have been specified regarding transient an The results of the NRC staff review operator action, and operator training.
' of the B&W small-break analysis is published in NUR Wilcox-Designed Operating Plants."
Similar studies have been completed for op(erating plants designed by CE), and General Electric Westinghouse (W), Combustion Engineering Those studies, which also focus specifically on the predicted plant performance under different accident scenarios involving feedwater and small-break loss-of-coolant accidents, are publish
" General
- Accidents in Westinghouse-Designed Ope' rating Plants;" NUREG-0635, I
Evaluation.of Feedwater Transient and Small-Break Loss-of in Combustion Engineering-Designed Operating Plants;" and NUREG-0626,
" Generic Evaluation of Feedwater Transients and Small-Bre 4
Accidents in, GE-Designed Operating PTants and Near-Term Operating Licen Applications."
Based on the review of the operating plants in light of the TMI-2 accident, the NRC staff reached the following conclusions:
9
a s
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isacceptableprovided<
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' design and operation, and-The continued operation of the operating plants 0645 are implem i
that certain actiors related to the plants (1) training of operators identified in NUREG-consistent ting plants in response to tion schedules.
t (2).Theactionstakenbythelicenseeswithopera ified in NUREG-0623, Trip During Small the IE Bulletins (including the actions spec
" Generic Assessment of Delayed Reactor. Coolant Pumphe d Water Reactors") provide Break Loss-of-Coolant Accidents in Pressurize added assurance for the protection of the fety significane of public.
h In' addition,the B&OTF independently confirmed t e saNRR task fo those related actfons recommended..b,y other NUREG-0646.
Pressurized Water Reactors (PWRS1 pressurized water reactors The primary method fo'r removal of decay heat fro This energy is the main feedwater or auxiliary turbine condenser or the transferred on the secondary side to ef ther Following feedwater systems, and is rejecte'd to either theat f ty/ relief valveso AFW was highlighted and a number of f the AFW (see NUREG-0645,
the TMI-Z accident, the importance of theIt was also required that improvements were made to improve the reliability o) ired AFW
" Report of the Bulletins and Orders Task Force".
ower source; that is, if qu operating plants be capable of providing the re hrs from one AFW pump train independent of any ac p t
both offsite and onsite ac power sources are los.a least-one alterna This Some pressurized-water reactors potentially h'fe f feedwater is postulated.
the high-pressure injection of removing decay heat if an extended loss ore to the primary method is known as " feed and bleed" and uses l
gy is removed (HPI) system to add water coolant (feed) at high pre PI through the power-operated. relief valves (POIt should be noted t system.
thereby reducing (bleed), if necessary. pumps.that. cannot. operate at f Limited vendor ting range of the HPI.
tely cooled by this means, psi).the system pressure to within the operatrolled to a safe level.
d analyses have shown that the core can be a equ to achieve and maintain coldliability h
When -the primary system is at low-pressure, t e removed by the residual heat removal system r
RHR systems.
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shutdown conditions.and performance criteria and standards for
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i is described in of the. residual heat removal system' design and operat on Section 5.4.7 of the SER.
t em m 1 systems Conclusion In susunary, because of the upgrading of current decay-hea -rt ff c that was required following the TMI-2 accident, t t d before the ultimate resolution of this generic issue without endangeringHowev heat taff on an indivi h
removal system requirements must be examined by t e sFor is the public.
-. reviewing additional infonnation related to case basis.
findings in a Items identified fn Section.10.4.9, and will report itsConsequentT
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iliary feedwater.
supplement to this SER.
satisfactory resolution of the items regarding the i
an be operated prior to ultimate resolution of this generic issue w health and safety of the public.
A-47' Safety Implications of Control Systems idents.being'made.more This issue concerns the potential for transiier.:s or acc tions. Thess severe as a result of control system failures or m lt of the accident or transient under consideration. 'One concern t circuit, open
- a. single failure such as a loss of a power su f
tion of several i
i i ated operational l
more severe than those transients analyzed as, ant c pA se control features.
than analyzed.
system failures which could make the accident more severe by creating a harsh occurrences.
Accidents could conceiva. ly cause control system failures b
environment in the area of the control equipment orIt is generally be r result in contro'l system failures would not lead to serious events oDetailed the control equipment.
conditions that safety systums cannot safety handle.
been perfonned to evaluations on all nonsafety systems, however, have not h t could affect a verify this beliet The potential for an accident t aT l
l system failures, may particular control system, and effects of the contro l
tions are differ from plant to plant. answers to those concerns, ic l
tions.
criteria.that will, be used for plant-specific eva ua d with the goal of required.
Shearon Harris control and safety systems ha t matic or manual
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W equipment required for abc'id'ent' initiation and operation of any safety sys emin a safe shutdown con t
" accident."
mitigation and/or to maintain the plant following any anticipated operational occurrence ord physical s For the been accomplished by providin'g independence and nonsafety system rafety system trains and betweei safety an ided.
latter,. as a minimum, isolation devices were prov equipment faults to the Also, to ensure that the operation of safety sys t
preclude the propagation of nonsafety sys em
. protection systems. equipment is not impaired, the single fa design, as cont'emplated for the plant design.
d to determine whether A systemate evaluation of the control system trol system failures which
~ this' Unresolved Safety Issue, has not been performevere tha postulated accidents couTdt cause s,ignificant cond accidents i would malte the accident consequences more se uch as steam generator However, a wide range of bounding transients an t
are being In addition, systematic reviews of safety systemstrol system
. overfill and overcooling events wou performed,with the goal of ensuririg that conmultip tidn Notice 79e22r systems.
i Also,. the applicant has been required (NRC Informa 1979) to review the l system failures which exacerba 7
" Qualification of Control Systems," September 1,
-- possibility of consequential contro l t d events would be adequately effects of high energy line breaks (HELB the where needed, to assure that the postu a eAs part of.the rev T
.i t ly qualified or an adequate i
qualification program to assure that equ pmen mitigated.
1:
equipment to the limiting exposed to HELB environments has been adequ to h
Infonnation Notica 79-22 when submitted and the a2 2 and 3.1 hostile environment.
program will be reported in Sections 7.7..
Evaluation Report, respectively.
f post-accident instrumentation l
Cooled Nuclear Power With the recent emphasis on the availability o(
Following an Accident"); the h t control system failures Plants to Assess Plant Conditions During and required to maintain the plant staff reviews evaluate the designs to assure t a t d operational occurrence or i
will not deprive the operator of informat oni
/
t l systems and i
f interdependence of identify any control systems whose malfunction cou accident."
The applicant was requested to; document the degree o ify the use (if any) of connoncomm these identified control systems and identpowe
.n.
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The st'aff s'
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whose failure could have potential safety significance. Safety Evaluati evaluation is documented in Section 7.7.2.1 of the Instrumentation and i:
In addition, IE Bulletin 79-27 (" Loss of Non-Class IEd to ensure 0 1979) was issued to Control Power System Bus During Operation," November 3,
the applicant requesting that evaluations be perfo f power to any The results of d controls.
Report, Section 7.5.2.
electrical bus supplying power for instruments anthis i
.I tisfactory j
Based on the above, we have concluded that subject to the sa d 3.11, there is resolution of the items regarding Sections 7.7.2.2 an d prior to the
- l reasonable assurance that Shearon Harris can b isk to the health d
t and safety of the public.
A-49 Pressurized Theriaal Shock in pressurized The issue of pressurized thermal shock (PTS) arises becauseide i
. water reactors (PWRs) transients and acc severe overcooling-(thermal shock) of.the reactor pressureIn thes f
l' stress with a maximum
.with or followed by repressuritat orr.
the reactor vessel internal surface results in therinaThe magnitude of
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tensile stress at the inside surface of the vessel the reactor vessel wall as a function of time.
compounded.by pressure stresses.
ith or followed.by Severe reactor systest overcooTing events s sult from a variety of causes. These include system transients, some o t ck open valves instrumentation and. control system malfunctions (including in either the primary or secondary system) and postu asmall break loss-of-coolant acciden t m line breaks i
(MSLBs),. and feedwater' line breaks.
el has lost The~ PTS' issue is a concern for PWRs only after the reac tron irradiation.
its fracture toughness properties and is embrittled by neu ~
n Harris Units 1 h
The standard and regulatory. requirements to which the S ear and 2 reactor vessels were designed-and fabricated are 5.3 of the SER.
l material is As long as the fracture resistance of the reactor vessed to cause relatively high overcooling events are not expecte i ls decreases with However, the fracture resistance of reactor vessel mater a The rate l
power plant.
exposure to fast neutrons during the life of a.nuc earition of the of decrease is dependent on the metallurgical compos 7
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.7 If the fracture resistance of the vessel has bee.1 redu ld cause sufficiently by neutron irradiation, severe overcooling events cou and welds.
f e The propagation of small flaws that might exist near the inner sur ac.
l wall assumed initial flaw might be enlarged into a crack through the vesse re of sufficient extent to threaten vessel integrity, and therefore, co cooling capability.
For the reactor. pressure-vesseT to fail'and constitute: a risk to p these health and safety, a number of contributing factors d
ties and propagate; (2) a level of irradiation (fluence) and material proper fluence composition sufficient to cause significant embrittlement. (the e depends, on. materials (3) a severe. overcooling transient with to occur more rap'fdTy,4) the crack resulting from the propagation of initial repressurization; and (cracks must be of such size and location that the the As a result of the evaluation of the PTS issue, the staff recomme actions to revent PTS events in l
23,1982)d the staff recomendations and Consission in SECY 82-465 (November ht ould l
.. operating reactors. The Commission accepte has' directed' the staff ta deve).op a Notice of Proposed Rulemaking I
ttrforr'(below which. PTS. risk is considered
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arly analysis and implementation of" sui:h' flux reduction.
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acceptable),rehIree'eenngcr establish an RT scr ing l
programs as are reasonably practicable to avoid re l
t are i
including within three calendar years of reaching the screening criter on analyses of proposed alternatives to minimize the PTS problem.
h Shearon Such a rule is now being-developed by the staff. We beifeve-that t e Harris plants could easily meet the requ.irements of such a rule.
h Shearon On the basis of the above consideration,. the staff concludes that t e issue and Harris facility can be operated before complete lth and safety of the public l
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