ML18081A534
| ML18081A534 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 10/29/1979 |
| From: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Schneider F Public Service Enterprise Group |
| References | |
| NUDOCS 7911150078 | |
| Download: ML18081A534 (12) | |
Text
e e
UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I 631 PARK AVENUE KING OF PRUSSIA, PENNSYLVANIA 19406 Docket No. 50-272 Public Service Electric and Gas Company ATTN:
Mr. F. W. Schneider Vice President - Production 80 Park Place Newark, New Jersey 07101 Gentlemen:
The enclosed IE Bulletin 79-17, Revision 1 is forwarded to you for action. A written response is required.
If you desire additional information regarding this matter, please contact this office.
Enclosures:
Sincerely,
~Grier f -- Di rector
- 1.
IE Bulletin No. 79-17, Revision 1 w/Attachment
- 2.
List of IE Bulletins Issued in the Last Six Months CONTACT:
L. E. Tripp 215-337-5282 cc w/encls:
F. P. Librizzi, General Manager - Electric Production E. N. Schwalje, Manager - Quality Assurance R. L. Mittl, General Manager - Licensing and Environment H. J. Midura, Manager - Salem Generating Station 7911150~ 7 'ii Q
ENCLOSURE 1 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 SSINS No.:. 6820 Accession No.:
79082201$7
.3 IE Bulletin No. 79-17 Revision 1 Date:
October 29, 1979 Page 1 of 5 PIPE.CRACKS IN STAGNANT BORATED WATER SYSTEMS AT PWR PLANTS Description of Circumstances:
IE Bulletin No. 79-17, issued July 26, 1979, provided information on the Rl cracking experienced to date in safety-related stainless steel piping Rl systems at PWR plants.
Certain actions were required of all PWR Rl facilities with an operating license within a specified 90-day time Rl frame.
After several discussions with licensee owner group representatives and Rl inspection agencies it has been determined that the requirements of Item 2, Rl particularly the ultrasonic examination, may be impractical because of un-Rl availability of qualified personnel in certain cases to complete the in-Rl spections within the time specified by the Bulletin. To alleviate this Rl situation and allow licensees the resources of improved ultrasonic inspec~
Rl tion capabilities, a time extension and clarifications to the bulletin have Rl been made.
These are referenced to the affected items of the original Rl bulletin.
During the period of November 1974 to February 1977 a number of cracking incidents have been experienced in safety-related stainless steel piping systems and por-tions of systems which contain oxygenated, stagnant or essentially stagnant bor-ated water.
Metallurgical investigations revealed these cracks occurred in the weld heat affected zone of 8-inch to 10-inch type 304 material (schedule 10 and 40), initiating on the piping I.D. surface and propagating in either an inter-granular or transgranular mode typical of Stress Corrosion Cracking.
Analysis indicated the probable corrodents to be chloride and oxygen contamination in the affected systems.
Plants affected up to this time were Arkansas Nuclear Unit 1, R. E. Ginna, H. B. Robinson Unit 2, Crystal River Unit 3, San Onofre Unit 1, and Surry Units 1 and 2.
The NRC issued Circular No. 76-06 (copy attached) in view of the apparent generic nature of the problem.
During the refueling outage of Three Mile Island Unit 1 which began in February of this year, visual inspections disclosed five (5) through-wall cracks at welds in the spent fuel cooling system piping and one (1) at a weld in the decay heat removal system.
These cracks were found as a result of local boric acid buildup and later confirmed by liquid penetrant tests. This initial identification of cracking was reported to the NRC in a Licensee Event Report (LER) dated May 16, 1979.
A preliminary metallurgical analysis was performed by the licensee on a section of cracked and leaking weld joint from the spent fuel cooling system.
Rl - Identifies those additions or revision to IE Bulletin No. 79-17 IE Bulletin No. 79-17 Revision 1 Date:
October 29, 1979 Page 2 of 5 The conclusion of this analysis was that cracking was due to Intergranular Stress Corrosion Cracking (IGSCC) originating on the pipe I.D..
The cracking was localized to the heat affected zone where the type 304 stainless steel is sensitized (precipitated carbides) during welding.
In addition to the main through-wall crack, incipient cracks were observed at several locations in the weld heat affected zone including the weld root fusion area where a miniscule lack of fusion had occurred.
The stresses responsible for cracking are believed to be primarily residual welding stresses in as much as the calculated applied stresses were found to be less than code design limits. There is no conclusive evidence at this time to identify those aggressive chemical species which promoted this IGSCC attack.
Further analytical efforts in this area and on other system welds are being pursued.
Based on the above analysis and visual leaks, the licensee initiated a broad based ultrasonic ~xamination of potentially affected systems utilizing special techniques.
The systems examined included the spent fuel, decay heat removal, makeup and purification, and reactor building spray systems which contain stagnant or intermittently stagnant, oxygenated boric acid environments.
These systems range from 2 1/2-inch (HPCI) to 24-inch (borated water storage tank suction), are type 304 stainless steel, schedule 160 to schedule 40 thickness respectively.
Results of these examinations were reported to the NRC on June 30, 1979 as an update to the May 16, 1979 LER.
The ultrasonic inspection as of July 10, 1979 has identified 206 welds out of 946 inspected having UT indications characteristic of cracking randomly distributed throughout the aforementioned sizes (24 11-14 11-12 11-10 11-8 11-2 11 etc.) of the above systems.
It is important to note that six of the crack indications were reportedly found in 2 1/2-inch diameter Rl pipe of the high pressure injection lines inside containment.
These lines are attached to the main coolant pipe and are nonisolable from the main coolant system except for check valves.
All of the six crack indications were found in two Rl high pressure injection lines containing stagnated borated water.
No crack Rl indications were found in high pressure injection lines which were utilized for Rl makeup operations.
Recent data reported from Three Mile Island Unit 1 indicates that the extent Rl of IGSCC experienced in stainless steel piping at that facility may be more Rl limited than originally stated above.
Of the 1902 total welds originally Rl inspected 350 contained U.T. indications which required further evaluation.
Rl These 350 welds have been reinspected with a second U.T. procedure which pur-Rl portedly provides better discrimination between actual cracks and geometrical Rl reflectors.
Hence, the licensee now estimates that approximately 38 of the Rl 350 welds contain IGSCC and the remaining welds, including those in high pressure Rl injection and decay heat lines, contain only geometrical reflectors.
Further Rl metallurgical analysis of these welds is required to verify the adequacy of the Rl U.T. procedures and to determine the nature of the cracking.
Rl IE Bulletin No. 79-17 Revision 1 Date:
October 29, 1979 Page 3 of 5 For All Pressurized Water Reactor Facilities with an Operating Li~ense:
- 1.
Conduct a review of safety related stainless steel piping systems within 30 days of the date of this Bulletin (July 26, 1979) to identify systems Rl and portions of systems which contain stagnant oxygenated borated water.
These systems typically include ECCS, decay/residual heat removal, spent fuel pool cooling, containment spray and borated water storage tank (BWST-RWST) piping.
For this review, the term 11 stagnant, oxygenated borated water systems" refers Rl to those systems serving as engineered safeguards having no normal operating Rl functions and contain essentially air saturated borated water where dynamic Rl flow conditions do not exist on a continuous basis.
However, these systems Rl must be maintained ready for actuation during normal power operations.
Where Rl your definition for stagnant differed from the one given above please supple-Rl ment your previous response within 30 days of this Bulletin revision.
Rl (a) Provide the extent and dates of the hydrotests, visual and volumetric examinations performed per 10 CFR 50.55a(g) (Re:
IE Circular No. 76-06 attached) of identified systems.
Include a description ~f the non-destructive examination procedures, procedure qualifications and accep-tance criteria, the sampling plan, results of the examinations and any related corrective actions taken.
(b) Provide a description of water chemistry controls, summary of chemistry data, any design changes and/or actions taken, such as periodic flushing or recirculation procedures to maintain required water chemistry with respect to pH, B, Cl-, F-~ o2.
(c) Describe the preservice NOE performed on the weld joints of identified systems.
The description is to include the applicable ASME Code sec-t ions and supplements (addenda) that were fo 11 owed, and the acceptance criterion.
(d) Facilities having previously experienced cracking in identified systems, Item 1, are requested to identify (list) the new materials utilized '
in repair or replacement on a system-by-system basis.
If a report of this information and that requested above has been previously submitted to the NRC, please reference the specific report(s) in response to this Bulletin.
- 2.
All operating PWR facilities shall complete the following inspection on the Rl stagnant piping systems identified in Item 1 at the earliest practical date Rl not later than twelve months from the date of this bulletin revision.
Fa-Rl cilities which have been inspected in accordance with the original Bulletin, Rl Sections 2(a) and 2(b) satisfy the requirements of this Revision.
Rl IE Bulletin No. 79-17 Revision 1 Date:
October 29, 1979 Page 4 of 5 (a) Until the examination required by 2(b) is completed a visual examination Rl shall be made of all normally accessible welds of the engineered safety Rl systems at least monthly to verify continued systems integrity.
Sim-Rl ilarly, 'the normally inaccessible welds, shall be visually examined Rl during each cold shutdown.
Rl The relevant provisions of Article IWA 2000 of ASME Code Section XI Rl and Article 9 of Section V are considered appropriate and an acceptable Rl basis for this examination.. For insulated piping, the examination may Rl be conducted without the removal of insulation.
During the examination Rl particular attention shall be given to both insulated and noninsulated Rl piping for evidence of leakage and/or boric acid residues which may Rl
- have accumulated during the service period preceding the examination.
Rl Where evidence of leakage and/or boric acid residues are detected at Rl locations, other than those normally expected, (such as valve stems, Rl pump seal~, etc.) the piping shall be. cleaned (including insulation Rl removal) to the extent necessary to permit further evaluation of the Rl piping condition.
In cases where piping conditions observed are not Rl sufficiently definitive, additional inspections (i.e., surface and/or Rl volumetric) shall be conducted in accordance with Item 2.(b).
Rl (b)
An ultrasonic examination shall be performed on*a representative sample
,Rl of circumferential welds in normally accessible portions of systems Rl identified by 1 above.
It is intended that the sample number of welds Rl selected for examination include all pipe diameters within the 2 1/2-Rl inch to 24-inch range with no less than a 10 percent sampling being Rl taken.
The approach to selection of the sample shall be based on the Rl following criteria:
Rl (1) Pipe Material Chemistry - As a first consideration, those welds in austenitic stainless steel piping (Types 304 and 316 ss) having 0.05 to 0.08 wt. % carbon content based on available material certification reports.
(2) Pipe Size and Thickness - An unbiased mixture of pipe diameters and actual wall thickness distributed among both horizontal and vertical piping runs shall be included in the sample.
(3) System Importance - The sample welds shall focus the examination primarily on those systems required to function in the emergency core cooling mode and secondly, on the containment spray system.
Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl The U.T. examination sample may be focused on noninsulated piping Rl runs.
The evaluation shall cover the weld root fusion zone and a Rl minimum of 1/2 inch on the pipe I.D. (counterbore area) on each side Rl of the weld.
The procedure(s) for this examination shall be essentially Rl
- Normally accessible refers to those areas of the plant which can be entered Rl during reactor operation.
Rl IE Bulletin No. 79-17 Revision 1 Date:
October 29, 1979 Page 5 of 5 in accordance with ASME Code S~ction XI, Appendix III and Supplements Rl of the 1975 Winter Addenda, except all signal responses shall be ~val-Rl uated as to the nature of the reflectors.
Other alternative examination Rl methods, combination of methods, or newly developed techniques may be Rl used provided the procedure(s) have a proven capability of detecting Rl stress corrosion cracking in austenitic stainless steel piping.
Rl For welds of systems.included in the sample having pipe wall thickness Rl of 0.250 inches and below, visual and liquid penetrant ~urface examina-Rl.
tion may be used in lieu of ultrasonic examination.
Rl (c) If cracking is identified during Item 2(a) and 2(b) examinations, all Rl welds in the affected system, shall be subject to examination and repair Rl considerations.
In addition, the sample welds to be examined on the Rl remaining normally accessible noninsulated piping shall be increased to Rl 25 percent using the criteria outlined in paragraph 2(b).
In the event RI that cracking is i denti fi ed in other systems at this sampling l eve 1, Rl all accessible and inaccessible welds of the systems identified in RI item 1 shall be subject to examination.
RI
- 3.
Identification of cracking in one unit of a multi-unit facility which causes safety-related systems to be inoperable shall require immediate* examination of accessible portions of other similar units which have not been inspected under the lSI provisions of 10 CFR 50.55a(g) unless justification for con-tinued operation is provided.
- 4.
Any cracking identified shall be reported to the Director of the apppropriate NRC Regional Office within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of identification followed by a 14 day written report.
- 5.
Provide a written report to the Director of the appropriate NRC Regional RI Office within 30 days of the date of this bulletin revision addressing the Rl results of your review if required by Item 1.
Provide a schedule *of your Rl inspection plans in response to Item 2(b) in those cases in which the Rl inspections have not been completed.
Rl
- 6.
Provide a written report to the Director of the appropriate NRC Regional
- Rl Office within 30 days of the date of completion of the examinations required Rl by Items 2(a), 2(b), or 2(c) describing the inspection results and any car-Rl rective actions taken.
Rl
- 7.
Copies of the reports required by Items above shall also be provided to the Director, Division of Operating Reactors, Office of Inspection and Enforce-ment, Washington, D.C.
20555.
Approved by GAO, 8180225 (R0072), clearance expires 7/31/80.
Approval was given under a blanket clearance specifically for identified generic problems.
Attachment:
IE Circular No. 76-06
Attachment to*IE Bulletin No. 79-17, Revision 1.
Nove:nbel: 26, 1976 I'E Circ\\lla;- No. 16-06 STR!:SS COR..'io'tOSlON c*.cu..cxs IN ST}.GNk\\'T, LOW PRESSURE STAn"'LtSS PIPING CONTAINING "BOll:C ACID SOLUTION AT l'WR' s D~CllP'IIO:: OF CIR~!STANCES:
During the period No~teiber 7. 1974 to Ncve.mb~r l. 1975, Deve~al inc:!.den~s of t.hrougb*wall-cra~king have occ~rred ill the 10-inch. sr.hedule 10 ~e 304 st.ai:cless $teel pi;>illg of the ieac!:or Building Spray and Decay Ee.at Removal Sys"::i!l!:.$ at: J.rka.:lsas Nuclea: Plant No. l.
0:1 Ocr;ober 7, 1976, Virginia Electric aod Pc~er also ra-pot'tt!d ~r.-:-:Ju&il wal1 eracki.~g in the lO-inch schedule ~O type 304 stain1ess disc~e=s~
piping of the "A'1 J:"ec:r:rculat:ion spray h.eat e:u:.hanger at Sur-ry U:;i t No. Z.
A ~eceut inspection of Unit l Contain::ient Recireulaticn Sp~'Y Pip1ug re~ealed cr*ck:U:ig si:ii.J.ar t:Q Cnit 2.
On October 8j 1976, another incide~t of si~ila~ e~ack.ing in S-in:h schedule lO ty~e 304 stainless p~pi~g of the Sefcty !njectiou r'..::.?
Suct:iou Line..'!.t ::.be Ginua facility Yat:. reported b~ the licensee.
Inf~tmatio~ reccivee on t~~ ~ecallurgical analys1$ conducted to dat~
indicates that th~ fail.ur~s ~ere the rQs~lc of i~te=sr~nul3r st:~ess corrosi6n crac;.:i<.in~ that: initiat:cad on the ins.:i.de of t:he piping.
A.
cot:.'1~nal.ity of fac-co:-s-cbsl!!rved as::ociarcd ~ith the cor:rcsion ~-=:.h.;:.o.i~
weri!:
- l.
The c~acks -;ere adjacent: to and propagated alo~g t'leld zones of ::hi;;
thi~~~alled le~ pressure pip£~i, ~or. par~ cf th~ reacco~ coola:t s')*stem..
- 2.
Cracking occun_ed. in piping containing l.*el.a.tiv~ly stagnant =oric acid solu~~on ~ot :equired fer nonnal cpe~ating cc~di~ions.
- 3.
Analysis of su~face p:oduc~s a~ this tim~ ina1ca~e ~ ch.lor~e ion intera~~ion \\ti.t:h o~ide fol""li:.ati~n in tha relativaly stagnant *bo~ic acid solu:;ior. as the riroba'ble Ct.:it't"oda.nt, ~it-h t:h!! staee of s:.:ess prcbab1y due to weld:Lng and/or fabrication.
nle soi.:rc::~ of :he chloride ion is Dot d.i.finitcly k~O~"':'I..
Bowe~er,..
~£:
.AN0-1 the chlctides a.:id. sulfide level observed i."l the su~face ta~ish film near ~elds is bslieved to have beeii i.~troducad into the pipins durt\\1S test.ing of the sodium t.h"iosulf.ite discharge vAlves, a:: valva leakage.
51.::il.!.:i:ly, a't Gin.:i* the chl.cri.cies and potential O)."Y&eu
- e.
e IE Circul.u No. 76-06
- 2 November 26, 1976 av.ai.lability were assumed to have been present since original conscruc~1on of the borated water scoraga eank which is vented to atmospher~.
Ca~osion attack at Sur~ is attributed to in-laakage of
. chlo't'id.es through recirculat:ion sp't'ay heat exchange t:ubing, allo""i11g bui1dup of contaminated wscer i~ ~n otherwise nc?:lnally dry sp~~y piping *.
AcTION TO I\\E TAlttN BY LIC%NSEE; l*
Provide a description cf your prcg~2l:ll fo~ assuring continu--~
integ%i::y of those safet:y-related pipins syscams 'Which are no~
frequeti.tly flushed, or t..*hich contain* ocl1flc~"'ing liquids.
thi.s
~~ogram sho~ld i:ic:.l.ude consideration of hydrostatic testing in acccrdanc:e wi~h ASHE Code Sect:ion XI rules (1974 Edition) !or al.1 ac~iva systems r~quirad for safety i~jactio~ ar.d containment sp-ray, includi::.g their recire.ulat.ion mode~, £:-om source of "'acer supply up to the ~acond isola:ion valve ci the ~=imary S)""Stem.
Si::U.lar tes~s should be considered for *ctha: safety-relaced piping syst:e!:!s.
- 2.
Yot.ar pros:ram should also c.onside~ volUl:let.i.*ic examination of a re:pr.esa:it:ative nu=ber of-circu~ereotial pipe wel.d.s by non-desi:ruc.t:ive e~am.in.a.t:i..cn techniques.
Such QXa.w.ina ticns sho-uld be' performed generally in ~cco:danc~ with Appandlx I of Section XI of the A.SM:: CQaa~ e~eept: :hat tha ~~amined area should cover a distane~ of ap?ro~:.tnately si~ (6) ti=es the p!pe "Wall 'thickness (':iu; r.ot Jess than 2 i:lches and need not exc.aed S inches). on each sio e of i:he -weld.
Sup~ lemen t:a.:Y ex~ation cmch:liques, such a.s radiog1:'aph~* t should be used where necessary for evaluation or c.o~fir:::a.tion cf ultrasouic indications resulting free such ~xamination.
- 3.
A repo~ describing your ~rcg:a~ an~ schedule for these 1uspec-*
t~cns should be suemittcd ~ithin 30 days after receipt of this C~r:;ular.
- 4.
nie 1'."ltC Regional Office should be infc:::med \\11thin.*2~ hours, cf any adverse findings resulting during ncndestruc~ive evaluaeion of the accessible piping welds ident1!iad abov~.
- 5.
A sw:::aiy 4~?or'C of the e~a=irJaticns and evalu.ation of r~sults should be !il.1tmitted within 60 days ft'Om. tho da'Ce of completion of prcposed t~ti.nz ancl examinations.
IE e
Circular No. 76-06 November 26, 1976 This su=:iary ~eport should also include a brief description of plant cond1t~ons, operacing pro~ed~~es o~ othe~ activi.t1es wh~ch prcv:ide assurance that the <?£fluent chem.i.s~ry wili ~~ail:
lo~ levels of pote:n:ial corrodants in such relacively stagnan:
reg1o?lS within the piping.
Yo~r ~esponsl!!r: should be subQitted to the D~~ector of thl.$ office,
~'ith a copy to the NRC Office of Inspection and Enfo~cemen.t, Division of Reactor Inspection ~rograms, ~ashing~on, O.C.
20555.
App~oval of NRC :equ~remeu~s for repcrts conce1""ning possible ge~eric
- problems ba.s been obtained under 44 U. s.c 3l5l from the U.S *. Ge'!:.et'a.l Accounting Office.
(CAO Approval B-1802.55 (?.0062), expires 7 / 31/77.)
Bulletin No.
79-10 79-11 79-12 79-0lA 79-02 (Rev 1) 79-13 79-14 ENCLOSURE 2 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Subject Date Issued Requalification Training 5/11/79 Program Statistics Faulty Overcurrent Trip 5/22/79 Device in Circuit Breakers for Engineered Safety Systems Short Period Scrams at 5/31/79 BWR Facilities Environmental Qualification 6/6/79 of Class lE Equipment (Deficiencies in the Envi-ronmental Qualification of ASCO Solenoid Valves)
Pipe Support Base Plate 6/21/79 Design Using Concrete Expansion Anchor Bol,ts Cracking in Feedwater 6/25/79 System Piping Seismic Analysis for 7/2/79 As-Built Safety Related Piping Systems IE Bulletin No. 79-17 Revision No. l Date:
October 29, 1979 Page l of 3 Issued To
.All Power Reactor Facilities with an OL All Power Reactor Facilities with an OL or CP All GE BWR Facilities with an OL All Power Reactor Facilities with an OL or CP All Power Reactor Facilities with an OL or CP All PWRs with an OL (for Action),
A 11 Other Power Reactor Facilities with an OL or CP (For Information)
All Power Reactor Facilities with an OL or CP
Bulletin No.
79-15 79-14 (Revision 1) 79-16 79-17 79-05C&06C 79-18 79-19 79-20 79-21 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS (CONTINUED).
Subject Deep Draft Pump Defi-ciencies Same Title as 79-14 Vital Area Access Con-trols Pipe Cracks in Stagnant Borated Water Systems at PWR Plants Nuclear Incident at Three Mile Island -
Supplement Audibility Problems Encountered on Evacuation Packaging Low-Level Radioactive Waste for Transport and Burial Same Title as 79-19 Temperature Effects on Level Measurements Date Issued 7/11/79 7/18/79 7/30/79 7/26/79 7/26/79 8/7/79 8/10/79 8/13/79 8/13/79 IE Bulletin No. 79-17 Revision No. 1 Date:
October 29, 1979 Page 2 of 3 Issued To All Power Reactor Facilities with an OL or CP Same as 79-14 All Holders of and Applicants for R~actor Operating Licenses All PWR Power Reactor Facilities with an OL All PWR Power Reactor Facilities with an OL All Power Re~ctor Facilities with an OL All Power and Re-search Reactors with OL, all Fuel Facilities (except Uranium Mills),
and certain Materials Licensees Certain Materials Licensees All Power Reactor Facilities with an OL or CP
Bulletin No.
79-14.
(Supplement) 79-02 (Rev 1)
(Supplement No. l) 79-13 (Rev 1) 79-22 79-14 (Supplement No. 2) 79-23 79-24 79-13 (Rev. 2)
LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS (CONTINUED)
Subject Date Issued Same Title as 79-14 8/15/79 Same Title as 79-02 8/20/79 Cracking in Feedwater 8/30/79 System Piping Possible Leakage of Tubes 9/5/79 of Tritium Gas Used in Timepieces for Luminosity Same as Title 79-14 9/7/79 Potential Failure of 9/12/79 Emergency Diesel Generator Field Exciter Transformer Frozen Lines 9/27/79 Cracking in Feedwater System 10/17/79 Piping IE Bulletin No. 79-17 Revision No. 1
- Date:
October 29, 1979 Page 3 of 3.
Issued To
.Same as 79-14 Same as 79-02 (Rev 1)
All Designated Applicants for OLs Each Licensee who Receives Tubes of Tritium Gas in Timepieces for Luminosity Same as 79-14 All Power Reactor Facilities with an OL or CP All Power Reactor Facilities which have