ML20248E354
| ML20248E354 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 10/18/1972 |
| From: | Giambusso A US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Andrea George POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| Shared Package | |
| ML20248A375 | List: |
| References | |
| NUDOCS 8910050188 | |
| Download: ML20248E354 (23) | |
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DSR# 6693
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- '- # OF PAGES /V UNITED STATES
[* p, I ' ^ / ,' M ATOMIC ENERGY COMMISSION
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- I WASHINGTON,0.C. 30H4
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ditt DEC 1 f,1972 Docket Nos. 50-333 Power Authority of the State of New York ATTH: Mr. Asa George Cencral Manager & Chief Engineer 10 Columbus Circle New York, New York 10019 Centlemen The Regulatory staff's continuing review of reactor power plant safety indicates that the consequences of postulated pipe failures outside of the containment structure, including the rupture of a main steam or feed-water lino, need to be adequately documented and analyzed by licensees
~ and applicants, and evaluated by the staff as soon as possible. Criterion
'> No. 4 of the Commission's General Design Criteria, listed in Appendix A of 10 CFR Part 50 requires that:
" Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal opera-tion, maintenance, testing and postulated accidents, including loss-of-coolant accidents. These structures, systems, and com-ponents shall be appropriately protected against dynamic effects, i
including the effects of missiles, pipe whipping, and discharging fluids, that may csult from equipment f ailure.s and from events and conditions outsido the nuclear power unit."
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' Criteria Nos. 40 and 42 of the previoue (1967) issue of the Commission's l Cencral Design Criteria reflected similar requirements.
Thus, a nucient plant chould be designed so that the reactor can be shut-down and maintained in a safe shutdown condition in the event of a postuinted rupture, outsido containmcat, of a pipe containing a high energy fluid, including the doubic ended rupture of the largest pipe in the main steam l and fcerlwater nyctcms. Plant structurcs, systems, and component o important I to satety should be designed and located in the facility to accommodate the effecto of such a postalsted pipe failure to the extent necesaary to assurc that a saf e shutdown condition of the reactor can be accomplished and maintained.
8910050188 ADOCK O8 h33 ppg PDR G
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Power Authority of the 2 CEC 10 FR2 State of New York Based on the information we presently have available to us on the James A. FitzPatrick Nuc1 car Power Plant, we understand that the main steam and feedwater lines are routed through a reinforced concrete tunnel from the primary containment penetration to the turbine building. The tunnel is open to the turbine building and is not close to vital equipment rooms. Although the steam lines are designed to Seismic Class II criteria beyond the outer isolation valve, these lines receive high quality control ,
under designation Group Q1.
Although it appears that the plant design is capable of withstanding the effects of a postulated rupture in the steam or feedwater lines, we request that you provide us with analyses and other relevant information j needed to determine the consequences of such an event, using the guidance provided in the enclosed information request. The enclosure represents our basic information requirements for plants now being constructed or operating. You should determine the applicability, for the James A.
FitzPatrick Nuclear Power Plant, of the items listed in the enclosure.
. If the results of your analyses indicate that changes in the design of structures, systems, or components are necessary to assure safe reactor shutdown in the event this postulated accident situation should occur, please provide information on your plans to revise the design of your facility to accommodate the postulated failures described above. Any design modifications proposed should include appropriate consideration of the guidelines and requests for information in the enclosure.
We will also need, as soon as possible, estimates of the schedule for '1 design, fabrication, and installation of any modifications found to bc necessary. days _after rec when we may Picagfpym expect to receive us vigi,n an 7_dment amen witTyourcipt_o_f,,this,,lotter._.
analysis of this ,
t postulated accident situation for the James A. FitzPatrick Nuclear Power i
Plant, a description of any proposed modifications, and the schedule l estimates described above. Sixty copies of the amendment should be provided.
A copy of the Commission's press announcement on this matter is also enclosed for your information.
Sincerely, f
h ,Q . .m A. Giambusso, Deputy Director
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for Reactor Projects Directorate of Licensing page)
I(Secnext _________________-__ _ _ _ ___ _ _ __-__ _
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Power Authority of the 3 IG 16 YJ72 State of New York
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Enclosures:
As stated
,c : Scott B. Lilly, Esq. Ecology Action Power Authority of the c/o Richard Goldsmith State of New York Syracuse University College of Law Arvin E. Upton, Esq. E. I. White Hall, Campus Eugene B. Thomas, Jr., Esq. Syracuse, New York 13210 Lex K. Larson, Esq.
LeBenuf, Lamb, Leiby & MacRae Secretary of the Commission 1821 Jefferson Place, N.W. U. S. Atomic Energy Commission Washington, D.C. 20036 Washington, D.C. 20545 i Lauman Martin, Esq. Miss Juanita Kersey, Librarian
! Scotor Vice President & Oswego City Library
.I Genersi Counsel 120 East Second Street Niagara Mohawk Power Corp. Oswego New York 13126 s -300 Erie Boulevard, West Syracuse, New York 13202 J. Bruce MacDonald, Deputy Cot:missioner and Counsel New York Statu Department of Commerce a x! Counsel to the Atomic Er,ergy Council 112 State Street h Albany, New York 12207
't Ms. Suzanne Weber 78 West Seneca Street Oswego, New York 11126 Chairman, Atomic Safety and Licensing Appeal Board U. S. Atomic Energy Commission Washington, D.C. 20545
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! ' Chairman, Atomic Safety and j Licensing Board Panoi )
U. S. Atomic Energy Commission t'
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Wanhington. 1).C. 20545
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l Mr. Frnnk U. Karas l [
Chwi . Public Procer dings Staf f Office of the Secretary of the ,
Comrnisaion U. S. Atomic Energy Commission l
Washington, D.C. 20545 adelmanysNeseen .
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LO Coneral Information Required for Consideration of the Ef fects of a Piping System Break Outside Containment no following is a general list of information required for AEC review of thn ef fects of a piping nys tem break outside sntainment, including the double ended rupture of the latgest pipe in the main steam and feed-water systems and for AEC review of any proposed design changes that may be found necessary. Since piping layouts are substantially difforent from p5 ant to plant, applicants and licensees should determine on an individual plant basis the applicability of each of the following j i
items for inclusion in their submittals. !
- 1. %c systeert (or portions of systers) for which protection against pips whip is required should be identified. Protection from pipe whip need l >
not he provided if any of the following conditions will exist:
(a) Both of the following piping system conditions are mett (1) the service temperature is less than 200' F; and (2) the destgn pressure is 275 psig or less; or (b) The piping is physically separated (or isolated) from structures ,
sys tems , or components important to safety by protective barriers ,
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[ or res trained from whipping by plant design features, such as concrete encasement: or r
(c) Following a single break, the unrestrained pipe movement of either end of the ruptured pipe in any possible direction about a plastic hinge formed at the nearest pipe whip res traint cannnt itepact any
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s t ruct ure , nys tem or compos nt import an t to safety; or 4damnetMWWheuu>
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'O 2 (d) The internal energy isvel associated with the whipping pips can be demonstrated to be insufficient to impair the safety function of any structure, system, or component to an unacceptable level.
- 2. The criteria used to determine the design basis piping break locatirns '
- in the piping systema should be equivalent to the following:
(a) ASME Section III Code Clags I piping breaks should be
' postulated to occur at the following locations in each piping run or branch run (1) the termhal ends; (2) any inte1 mediate locations between terminal enda where the primary plus secondary stress intensities s (circum-ferential or longitudinal) derived on an alast'.cally The internal fluid energy level associated with the pipe break reaction may take into account any line restrictions (e.g., flow limiter) between the pressure source and break locetion, and the effects Theofenergy either single-isvol ended or double-ended flow conditions, as applicable.
in a whipping pipe ray be considered as insufficient to rupture an impacted ,
pipe of equal or greater nominal pipe size and equal or haavier wall thickness.
Pip S g is a prest.are retaining component consisting of straight or curved pipe and pipe fitting = (e.g., elbows, tees, and reducers).
A piping run interconnects components such as presourc vessels, pumps, and rigidly fixed valves that may act to restrain pipe movement A brantJa beyond run differs fromthata required for design thermal displacement.a piping intersection, as a piping run only in that it originates at
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'O calculated basis under the loadings associated with one -
half safe shutdown earthquake and operational plant
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conditions exceeds 2.0 5 , for ferritic steel, and 2.4 S , for austenitic steel; (3) any intermediate locations between tet tiual ends where the cumulative usage factor (U) derived from the piping l l
fatigue analysis and based on all normal, upset, and testing plant conditions exceeds 0.1; and (4) at intermediate locations in addition to those determined by (1) and.' (2) above, selected on a reasonable basis as necessary to provide protection. As a minimum, there should be two intermediate locations for each piping run or branch run.
(b) ASE Section III Code Class 2 and 3 piping breaka should be I postulated to occur at the following locations in each piping run or branch runt (1) the terminal ends;
' Operational plant conditions includa normal reactor operation, upset I c onditions (e.g. , anticipated operational occurrences) and testing conditions.
'S,is the design stress intensity as specified in Section III of the ASE Boiler and Pressuro Vessel Code, " Nuclear Plant Components."
6U is the curaulative usage factor as specified in Section III of the AS!W. Boiler and Pressure Vessel Code, "Nucicar Power Plant Components."
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(2) any intermediate locations between terminal ends where either the circumferential or longitudinal stresses derived on an elastica 11y calculated basis under the loadings associated with seismic svents and operational plant conditions exceed 0.9 (S h+S)Ig or the expansion sLusses i exceed 0.8 Sgg and (3) - intermediate locations in addition to these determined by 4
(2) above, selected on reasonable basis as necessary to provide protection. As a minimum, there should be two intermedists locations for each piping run or branch run.
- 3. The criteria used to determine the pipe brekk orientation at the break I
locations as specified under 2 abovs should be equivalent to the followings (a) Longitudinal breaks in piping runs and branch runs, 4 inches nominal pipe size and larger, and/or S is the stress calculated by the rules of NC-3600 and ND-3600 for h
Class 2 and 3 coniponents, respectively, of the ASE Code Section III Winter 1972 Addends.
S is the allowable stress range for expansion strces calculated by the A
rules c f HC-3600 of the ASE Code,Section III, or the USA Standard Code for Prtssure Piping, ANSI B31.1.0-1967.
0 Longitudinal breaks are parallel to the pipe axis and oriented at any point around the pipe circumference. The break area is equal to the ef fective cross-sectional flow area upstream of the break location.
Dynanic forces resulting from such breaks are nosumed to cause laterni pipe er.ovements in the direction normal to the pipe axis.
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(b) Circumferential' breaks in piping runa and branch runs exceeding 1 inch nominal pipe size.
- 4. A nummary should be provided of the dynamic analyses applicable to the design of Category I piping and associated sopports which determine l'
the resulting loadings as a result of a postulated pipe break including:
(a) The locations and number of design basis breaks on which the dynamic analyses are based.
(b) The postulated ruptur6 orientation, auch as a circumferential and/or longitudinal break (s), for each postulated design basis break location.
(c) A description of the forcing functions used for the pipe whip dynamic analyses including the direction, rise time, magnitude, duration and initial conditions that adequately represent the jet stream dynamics and the system pressure difference.
(d) Diagrams of mathematical models used for the dynamic analysia.
(e) A nummary of the analysse which demonstrates that unrestrained motion of ruptured lines will not damage to an unacceptable degree, structure, systems, or components important to safety, such as the control room.
'Circumferential breaks are perpendicular to the pipe axis, and the break area is equivalent to the internal cross-sectional area of the ruptured pipe. Dynamic forces reoulting from auch breaks are acouced to separate this piping axially, and cause whipping in any direction normal to the pipe axis.
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O 5. A description should be provided of the measures, as applicable, to protect againre pipe whip, blowdown jet and reactive forces including (a) Pipe restraint design to prevent pipe whip impact; (b) Protective provisions for structures, systems, and components required for safety against pipe whip and blowdown jet and reactive forces; (c) Separation of redundant features; (d) Provisions to separate physically piping and other components of redundant features; and (e) A description of the typical pipe whip restraints and a sucnary of number and location of all restraints in each system.
- 6. The procedures that will be used to evaluate the structural adequacy of Category 1 structures and to design new seismic Category 1 structures should be provided including (a) The ciethod of evaluating streoses, e.g., the working stress method and/or the ultimate strength method that will be used; (b) The allowable design stresses and/or strains; and (c) The load f actors and the load combinations.
- 7. The design loads, including the pressure and temperature transiento,
- the dead, live and equipment loads; and the pipe and equipment static, there.al, and dynamic reactions should be provided.
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'O 8. Seismic Category I structurni elemento such as floors, internt walls, exterior walls, building penetrations and the buildings as a whole should be analyzed for eventual reversal of loads due to the postuinted accident.
- 9. If new openings are to be providad in existing structures, the capabilities of the modified structures to carry the design loads f should be demonstrated.
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- 10. Verification that failure of any structure, including nonseismic Category I structures, caused by the accident, will not cause f ailure of any other structure in a manner to adversely affects (a) Mitigation of the consequences of the accidents; and (b)
Capability to bring the unit (s) to a cold shutdown condition.
- 11. Verification that rupture of a pipe carrying high energy fluid will not directly or indirectly result int (a) Loss of redundancy in any portion of the protection system i (as defined in IEEE-279), class IE electric system (as defined cable pens-in IEEE-308), engineered safety feature equipment, trations, or their interconnecting cables required to mitigate
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(b) Loss of the ability to cope with accidents due to ruptures of pipes other than a sessa line, such as the rupture of pipes causing a steam or water leak too small to cause a reactor accident but Isrgs enough to cause electrical failure.
- 12. Assurance should be provided that the control room will be habitable and its equipment functional after a steam line or feedwater line break or that the capability for shutdown and cooldown of the unit (s) will be available in another habitabis area.
- 13. Environmental qualification should be demonstrated by test for that
- electrical equipment required to function in the eteam-air environ-manr. resulting from a steam line or feedwater line break. The in-formation required for our review should include the followings (a) Identification of all electrical equipment necessary to meet requirements of 11 above. The time after the accident in which they are required to operate should be given.
(b) The test conditions and the results of test data showing that i
the systems will perform their intended function in the environ-ment resulting from the postulated accident and time interval of the accident. Environmental conditions used for the tests nhould be selected from a conservative evaluation of accident conditions.
(< ' The results of a study of steam systems identifying locations where barriers will be required to prevent steam jet impingment from dia-abling a protection system. The design criteria for the barriers should be stated and the capability of the equipcent to surviv..
within the protected environment should be deocribed.
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(d) An evaluation of the capoliity for safety related electrical equipment in the control room to function in the environment that may extat following a pipe break accident should be provided. Environmental conditions used for the evaluation should be selected from conservative calculations of accident conditions.
(c) An evaluation to assure that the onsite power distribution system and onsite sources (diesels and batteries) will remain operable throughout the event.
- 14. Design diagrams and drawings of the steam and feedwater lines
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including branch lines showing the routing from containment to the a
turbine building should be provided. The drawings should show elevations and include the location relative to the piping runs of safety related equipment including ventilation equipment. intaken, and ducts.
- 15. A discussion ahould to provided of the potential for flooding of safety related equioment in the event of f ailure of a feedwater line or any other line carrying high energy fluid.
- 16. A description should be provided of the quality control and inspection prograrr.s that will be required or have been utilized for piping systemw outside cont ainment.
- 17. If leak detection equipment is to be used in the proposed modi fic:st iom ,
a discusalon of its capabilities nhould he provided.
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- 18. A summary should be provided of the emergency procaduras that would be followed af ter a pipe break sceident, including the automatic and manual operations required to place the reactor unit (s) in a cold shutdown condition. The estimated times following the accident fa. ,11 equipment and personnel operational actions should be included in the procedure summary.
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- 19. A description should be providad of the seismic and quality classi-
, fication of th's high energy ' fluid piping systems including the staas
, . . and feedwater pipin5 that run naar structures, systems, or componects impsrtant to safety.
- 20. A description should be provided of the assumptions, methods, and
.results of analyses, including stes.s generator blowdown, used to -
calculate the pressura and temperature transients in compartments, pipe tunnels, intermediate buildings, and the turbine building following a pipe rupture in these areas. The equipment assumed to function in the analyses should be identified and the capability of systems required to function to meat s. single active component failure should be described.
- 21. A description should be provided of the methods or analyses perforr.wd -
to demonstrate that there will be no adverse effects on the primary and/or secondary containment structures due to a pipe rupture outside thece structures.
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! V in.the event of a pipe break. Hot' lines which must sustain large 12 thermal-and mechanical stresses are designed with combinations of penetration sleeves and flued fittings.
Based ~on our review of the information contained in this' application and similar designs we conclude tnat the' primary
- ..,. containment design basis is acceptable,
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5.2.2' Missile and Pipe Whip Protection Several locations on the main' steam lineci .and feedwater lines p are not restrained to prevent pipe whip in the event of pipe failure at 'these. locations. The applicant has stated that the physical layout with?n the drywell precludes restraints at these points. For all. ,
un other lines and locations, restraints have been provided where ,,;
a break could result in containment impact. The applicant has.
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- identified the unrestrained high stress areas in these lines where
, breaks could result in pipe whip such that the pipe could impact the prisary containment wall. At those locations which are accessible the applicant has provided 1-1/4 inch thick impact plates as sup-plementary protection for . the drywell. In additian, he has. agreed to perform augmented inservice inspection of these weld locations gfd during each inspection period. At the remainder of these identified M' '
areas the physical layout precludes installation of impact plates.
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M{ighd Here, the applicant will perform augmented inservice inspection of u xW e,
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.the weld.s during each inspection period. The requirements of this augmented. inspection will be set forth in the' Technical Specifications and will call f or 100; rather 'than 25% inspection during each period.
The applicant has also considered the effects of pipe whip on the emergency core cooling systems. The systems are redundant and
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physically separated such that'a ruptured pipe could impact and e
affeet only one of the redundant ECCS. The remaining ECCS components were shown to limit peak fuel clad temperature to 1370*F following the most severe postulated break sequence.
The applicant han considered the effect of missiles ranging in size from nuts and< bolts to valve bonnets, and concludes that no missile would have sufficient energy to penetrate the containment.
In. addition, where possible, components are arranged so that the 6 direction of flight of potential missiles is away f ro:r. the contain-ment wall.
l The ef fects of pipe whip and steam jet impingement on the shield and vessel support structute resulting from a LOCA occurring within i the sacrificial shield area were analyzed and found to be acceptable.
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We-conclude'that the applicant has.provided adequate measures to.
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5.2.3 Containment Isolation m.-
Mu u The ability to isolate the primary containment provides the
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G>. ,-u or tae.. containment atmospnere, and tav environs 11. the e ven t of L
accidents.or other non-nominat conditions. Isolation is
. accomplished by'means of valves. The numbers, types and locations
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of these valves in the various lines depend on the manner in which-theflines penetrate the reactor vessel and the containment. Where necessary, the valves are equipped with operators and close automatically when sensors detect certain accident or fault conditions.
The consequences of postulated pipe failures both inside and outside of the containment have been evaluated and are described in Section 10. 'The isolation valves and their control systems have been reviewed to' assure that no single f ailure can result in la loss' oficontainment integrity. An exception exists in the
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m case of instrument lines connecting to the reactor coolant n
. system which penetrate the containment and dead-end at instruments located in the reactor building. Such lines are provided with manually operated isolation valves and excess flow check valves, both.of which are outside the containment. A break in the line between the containment and the outer check valve
-- would result in blowdown directly into the reactor building.
. [p r The applicant has installed 1/4 inch diameter orifices in each hbkl Ils lwa: 4
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A.- of . tiiese lines lasive tne p ri ma ry . c on t a inmen t to prevent over-ca.
pressurization of'tne reactor building and limit offsite doses to substantially below tue 10 CFit Part 100 values in the event of the postulated instrument Jine break. Based on our review of the desien we conclude that the provisions for instrument-lines penetrating the primary containment are adequate and satisfy the supplement to Safety Guide 11.
Leakage .through the closed main steam line isolation valves fo11owing a postulated LOCA presently relies on the low leakage characteristic of the valves. The acceptability of present leakage limits and the need for an auxiliary sealing system are under study by the staff. There is nothing in the existing design which would preclude incorporation of an additional sealing
.,o feature if such is. determined to be necessary. The applicant v
will continue to study developments in this area.
Based on our review we conclude that the primary contain-ment' isolation provisions are adequate.
} (, 5.2.4 Leakage Testing Program Leakage testing of the reactor primary containment and associated systems is intended to provide initial and periodic
+ verification of the leaktight integrity of the containment.
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s$ The applicant has stated in Amendment Nos. 4 and 5 that the 9_; a y -
primary reactor containment and its components have been designed 4"hE:U.1 4Yh a
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personnel and equipiaent hatches and airlocks, and isolation valves, have been designed with the capability of being individually l
leak. tested at calculated peak pressure.
We coaclude that the containment system will permit contain-ment leakage rate testing in compliance with the AEC proposed
" Reactor Containment Leakage Testing for Water Cooled Power Reactors," 10 CFR S 50.54(o), Appendix J, and therefore is acceptable.
In addition'to agreeing to meet the requirements of proposed Appendix J, PASNY has agreed to perform a leak test of drywell .
to suppression chamber piping, headers, downcomers and vacuum ,
breaker valves at each refueling outage. They will also determine '
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'"i acceptable bypass leakage limits and other test criteria and will be required to perform frequent surveillance testing of the vacuum breakers. We have not completed our review of the details of the test and surveillance program. However, the applicant has indicated l
his intention to base it on the recently approved Browns Ferry leak check program. We find this commitment acceptable pending completion of our review.
J 5.3 Secondary Containment
[f The reactor building, together with the Standby Gas Treatment System (SGTS) and the main stack, form the secondary t
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4.0 REACTOR COOLANT SYSTEM 4.3 Fracture Toughness PASNY has established and submitted, in Supplement 20 to the FSAR, ;
operating pressure and temperature limitations during startup, shutdown, and hydrostatic testing of the reactor coolant system. These limitations reflect the recommendations of Appendix G, " Protection Against Non-Ductile Failure," of the 1972 Summer Addenda to the ASME Code,Section III.
The minimum temperature for the 1000 psig operating pressure leak test is 130*F. The minimum temperatures for the 1190 psig system pressure tests and the hydrostatic tests at 1563 psig are 150*F and 173*F, respectively. We conclude that these temperature and pressure
[ limitations are acceptable.
4.4 Reactor Recirculation System The following information is an update of that contained in Section 4.4 of the SER and Supplement 1.
In Supplement 20 to the FSAR, FASNY presented the results of a study to determine the effects of missile generation from recirculation pump
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overspeed following a LOCA. The study addressed the probability of lI'l destructive pump overspeed resulting in escape of high energy missiles EU from the piping system with these missiles causing damage to the primary containment or internal piping.
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I The referenced study indicates that.the probability of a recircu-1 lation pump missile ejecting from a broken recirculation pipe and
-8 causing loss.of containment integrity is about 3 x 10 . The staff concludec that the probability of this potentially damaging event is sufficiently small, i.e., within a safety objective for the prob-p ability.of a particular potential failure path of about 10~ per year.
l We conclude that the applicant's proposal to provide a decoupling device and not to add pipe restraints is acceptable. Our conclusions are based on the following: ;
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- 1. The decoupling' device between the recirculation pump and motor will prevent the formation of motor assembly missiles.
- 2. The staff's ' valuation of the applicant's probability analysis
- indicates the concurrent probability of fuel damage and recircu-lation pump missile ejection causing loss of containment integrity is acceptably small (i.e., less than 1 x 10~ per year). Accord-ingly, the delays and risks associated with installing additional
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pipe restrainta outweigh the incremental small gain in safety.
- 3. The applicant had previously agreed to install decouplers between the pumps and motors to prevent excessive motor overspeed. These 1
- were to be installed at the first refueling outage. However, the
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- (N-General Electric. Company.is currently having difficulty in procuring snd design testing of a prototype and has indicated that the installation of the decoupler may be extended to the second refueling. The staff finds that because of the low probability of the event, the addition of even another 2 years delay would not significantly affect-the staff's conclusion.
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JAF FSAR UPDATE l
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s-16.3.2.2 Excessive Pipe Movement Concern "The Applicant is studying design modifications to prevent violation of the containment by pipe whipping and generation of missiles in the unlikely The Committee believes event of a failure of the primary system piping.
that such design modifications as are practical should be implemented in a manner satisfactory to the Regulatory Staff. The Committee wishes to he kept informed about this matter."
Resolution 16.3.2.2.1 Introduction The possibility of violating the primary containment of a nuclear power The plant by the impact of a ruptured pipe has been of major concern. postulated pipe question of restraints, to reduce the consequences of a review of the ,
break, was raisedAtinthat connection with the construction permittime our opinio FitzPatrick Plant.
the piping systems of interest and that restraints were not credible in That opinion notwithstanding, we performed a limited evaluation necessary.
of the plant design which was curtsnt at that time and concluded that a I]
technically defensible system of pipe restraints or energy absorbing
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material could be designed and installed in the Therefore, FitzPatrickthe Plant following at a significant, but not entirely unreasonable, cost.
13 to the FitzPatrick Plant PSAR:
statements were made in Supplement d
"The Authority recognizes the AEC concern overcontainment.
postulated pipeTo ruptures meet this an subsequent pipe movement . within - the - p*1 mary design criteria were established for the TitzPatrick concern the following Plants
- 1. Where practicable, considering' available structurally inspection adequate and supports, available space, 'the 'need for inservicewill be ' restrained against pi
! the need for maintenance, pipes a
movement resulting from postulated pipe rupture so that the integrity of the primary containment will be maintained.>
g
+
- 2. Wherever it is not practicable to prevent pipe movements by
, restraints, the integrity of the ~ primary containment material applied will be to the an energy-absorbing
~
+ maintair.ed with at areas subject to the impact of the ruptured primary containment n
i pipe.
- 3. Fipes will be restrained, if shown necessary by analysis, to prevent the loss of' the entire emergency core cooling function.~
Single active component failure will be included in the analysis ara v u
in addition to those failures resulting from excessive pipe .*
w
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FSAR UPDATE m
movement. At least one core spray subsystem or one LPCI loop utilizing two RHR pumps for core cooling will be available for the limitation of core temperature rise. In any event, thereafter, sufficient flow from one RHR pump, in conjunction with one RHR heat exchanger and two RHR ' service water pumps, will be available for long term containment cooling.
Preliminary investigations indicate that optimum solutions to the pipe movement problem will involve a combination of restraints and energy-absorbing material. The engineering optimization will be determined during the course of design so that the requirements stated above will be met. In any event, we will retain the heavy restraints already included in the design for the purpose of resisting dynamic loads such as those resulting from seismic or hydraulic forces."
A system was developed which attempts to satisfy, in a practical manner, the above requirements for containment integrity, "considering available structurally adequate supports, available space, the need for inservice inspection and the need for maintenance". The physical separation of ECCS systems already incorporated in the basic design allows this part of the criterion to be met no matter what reasonable combination of occurrences may be assumed.
Based on the above conditions, the following two criteria have evolved.
- 1. Integrity of the drywell is paramount.
- 2. Protective devices will not be provided where their inclusion will introduce other difficulties which will have an adverse e f fect on overall plant safety such as seriously hindering in-service inspection.
On the basis of the considerations contained within this report, a system of pipe restraints was developed which resulted in a net increase in total plant safety.
16.3.2.2.2 Problem Definition The prol,ebility of a postulated pipe break accident is dependent on many factors which are best defined in a fracture mechanics context. However, even though it is possible to conclusively demonstrate the unlikelihood of such an accident, it is wise to consider the consequences and possible design alternatives. This includes the evaluation of pipe acceleration and the investigation of impact phenomena. It is not obvious, a priori, that a direct impact on the drywell will necessarily violate the primary containment.
A large defect in the pipe (if undetected) will most likely develop into a localized surface opening and leakage will ensue. Experimental results indicate that if the defect reaches a critical length, in the presence of 16.3-3 Rev. 0 7/82
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