ML20010D611

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Minutes of ACRS Subcommittee on B&W Reactors 810506 Meeting Re Varius Aspects of B&W Reactor Operations.Related Fr Notice,Detailed Schedule of Presentations & Lists of Attendees & Documents Provided Encl
ML20010D611
Person / Time
Site: Crystal River Duke Energy icon.png
Issue date: 07/24/1981
From:
Advisory Committee on Reactor Safeguards
To:
Advisory Committee on Reactor Safeguards
References
ACRS-1857, NUDOCS 8108280359
Download: ML20010D611 (44)
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Text

{{#Wiki_filter:f.j k... -,# E,d. i ! J [ Ab 'I 1 t,. i M 3 l'! ', I ' l M#J _ //g 7 d L C.2,' PM 8 13-rl .\\D V L_... - J m CERTIFICATION DATE: 081 a Q\\ s MINUTES OF THE ACRS SUBCOM.MITTEE Ok BABCOCK AND WILC0X REACTORS 9 4 /t ..""24 O Y WASHINGTON, D.C. - MAY 6,1981 p E a 3013 0BF o The ACRS Subcommittee on Babcock & Wilcox (B&W) Reactors met in o a-s w,. on May 6,1981 in Washington, D.C. to review various aspects of B&W // ey I_I I 2 reactor operations. The primary p'urpose of this meeting was to review the application by the Florida Power Corporation to increase the maximum power rating of Crystal River Unit 3 from 2452 MWt to 2544 MWt. A secondary pur-pose was to review the February 26, 1980 incident at Crystal River Unit 3 in which an electrical failure in a portion of the reactor control system resulted in a challenge to the safety systems and the spilling of 40,000 gallons of primary coolant water into the reactor containment building. The final purpose was to review the recently reported inaccuracies in the reactor pratection ' system instrumentation. Notite of this meeting was published in the Federal Register on April 21,. l 1981 ( Attachment A).. A copy of the detailed schedule of presentation is attached ( Attachment B). No written comments nor requests to make oral comments were received from members of the public. No written reports were issued or approved by the Subcommittee at this meeting. A list of attendees at the meeting is attached ( Attachment C). A list of documents provided to l the Subcommittee during the meeting is also attached (Attachment D). Executive Session (0 pen to Public) l Mr. Ray, Subcommittee Chairman, opened the meeting at 8:45 a.m. with a statement regarding the conduct of the meeting in accordance with the provisions of the Federal Advisory Committee Act and the Govert. ment in the h Su inhet. C. McKinley was the D.signated Federal Employee for / th!s mee}fng. N fj I l j ' . \\. ! ! i,' l l, f i-8108280359 810724 PDR ACRS 1857 PDR

Meeting With the NRC Staff, Florida Power Corporation, and Babcock and W1cox Company Florida Po er Corpt, ration Reorganization Mr. Hancock, Assistant Vice-President of Nuclear Operations for the Florida Power Corporation (FPC), stated the company's philosophy regarding $he operation of Crystal River Unit 3 as: 1. Operate the plant safely without regard to electrical production. 2. Operate the plant within all legal constraints. 3. Af ter meeting the first two requirements, make as r.uch electrical power as possible. He mentioned that the Three Mile Island accident had caused his company to reconsider its ability to operate Crystal River Unit 3 (CR3) and to handle emergencies. The Company made a commitment to excellence and profession-alisr. and initiated' 4 major reorganization in order to achieve those object-ives. Mr. Ray noted the negative impression made by Florida Power's early perfor-mance and expressed appreciation for the cnange in attitude. Mrs. Baynard (FPC) described the new crganization in greater detail. She pointed out that because of the emphasis being placed by FPC on improving its nuclear program, Mr. Hancock chose to locate his office on the CR3 l I site in November 1979. In December 1979 the training staff was increased significantly. In January 1980 the training staff was reorganized and placed under the direction of the nuclear plant manager. In March 1980 there was a significant increase in staffing in all functional areas with additional ~ changes following later in the year and all changes essentially completed by the end of 1979. The staffing of the nuclear department ((

increased to about 400 people including all plant operating shift pe rs onnel., Mrs. Baynard is Manager of Nuclear Support Services and rcports directly to the Assistant Vice President for Nuc1 car Operations as does the. Plant Manager of CR3. As Dr. Zudans pointed out, the Assistant Vice President has to de a great deal of day-to-day coordination among the various managers. One of the objectives of this organizational structure was to provide the Plant Manager with the resources needed without the administrative burden. In order to address key problems, project teams are organized and headed up by one of the functional managers. This has proven to be an effective technique for FPC. FPC has a plant review committee that reviews safety issues and an inde-pendent nuclear review committee that looks into potential unreviewed safety issues and major design changes. In addition there is some nuclear expertise in some of the non-nuclear segments of the FPC organization. The discussion of the FPC organization brought out the equivalence of the nuclear operations with the other major departments of the company. Mr. Ray suggested that, if a presentation on this is made in the future, the organization charts be drawn to better show this equivalence. Mr. Hancock pointed out that the FPC management is not pleased with how well its nuclear plant has done to date. Management's commitment is not to have a just everage plant but to have it be the best. This can't be done overnight and requires a long tann commitment. i l ^

Mrs. Baynard reported that the Institute for Nuclear Power Operations made an intensive two and one-half week inspection of the CR-3 plant and the report of that inspection is due to be issued in the next few days. Action has already been initiated to correct some of f.he deficiencies not d by the inspection team. Mr. Robert Martin, Deputy Director of Region II, Office of Inspection and Enforcement (I&E), reported his findings regarding FPC's operation of CR3. Prior to FPC's reorganization the operation of CR3 was acceptable but at the poorest level o' any plant in Region II. The change in attitude of the FPC management and the organizational changes made have raised the perform-ance level up to the average in Region II. Although substantial improvement has been made to date, I&E will continue to monitor FPC's performance closely. Florida Power Corporation Training Frogram Mr. Kemper (FPC), Manager of Nuclear Operations Training presented the com-pany's training plans, and capabilities. The training program extends from management development to new hire radiological indoctrination. There is l a commitment to provide comprehensive quality training to maximize the utilization of human resources. Two management development courses are offered, one aimed at supervisory personnel with salary midpoints at about !2800 per month, and the other for executives with salary midpoints above $2800 per month. Three nuclear oper-ations management courses have been developed. These courses are in addition l to a complete program of operator training for both licensed and non-licensed l l l

m operators, including requalification training. General employee training is providid in at least nine general areas such as industrial safety, security, and emergency plans. More specifics of the training are included in Attachment E. ~ The training program for licensed operators takes about 2,000 hr. before he is given the license examination. Training includes simulator and other hands-on experience. Candidates for the various programs are screened for specific educational and experience requirements before being admitted into a course. Most of ~ the courses are presented by permanent staff members of FPC (fourteen FPC instructors and five contract instructors). Mr. Kemper briefly ' described each of the training courses. The Subcommittee and its consultants were favorably impressed by.FPC's dedication to a major training effort. February 26, 1980 Control System Failure Event Mr. Kemper (FPC) Jescribed the February 26, 1980 event in which a small power supply in the non-nuclear instrumi.atation system failed and caused a major plant transient, deprived the operators of significant information, and re-sulted in the spilling of about 43,000 gallons of primary coolant onto the containment floor. At approximately 14:23 (2:23 p.m.) the +24 volt bus in the Y non nuclear in-strumentation system failed due to an apparent short circuit. The loss of ..,.-m

' this bus caused much of the control room indication to 90 to a midscale position and caused the power operated relief valve (PORV) to open. The midscale indications caused the reactor's rods to withdraw a::d the feedwater to shutdown. The protection system scrammed the reactor on high pr, essure. The "A" steam generator boiled dry. At approximately 14:26 the PORY indicated closed but was in fact fully open. The plant conditions indicated a small LOCA and based on TMI experience the operators isolated the PORY and high-pressure injection was automatically initiated on low pressure. Preactor building pressure began to rise and eventually peaked at about 4 psig. At about 14:32 the plant operators started the auxiliary feedwater pumps to backup the main fee,dwater pump which continued to operate. At about 14:33 the primary system was filled solid with water and one pres-surizer safety valve opened. The primary coolant pumps were stopped and natural circulation began. Although radioactive water was being released into containment and contain-ment radiation levels were increasing and the reactor building purge was l running, there was no appreciable release to the environment as a result of i this event. l At about 14:44 the non nuclear instrument power was restored and the operators l were able to accurately determine the plant conditions and confirm that it was in e safe mode of operation and the core hcd not been uncovered. The reactor building isolaticn actuated and ECCS actuated on high containment E

C ^ pressure (4 psig). At 14:46 the operators bypassed high-pressure and low-pressIre injection signals in order to divert cooling flow to other essential equipment. By 14:56 the situation was well understood and actions were being taken to minimize the amount of primary coolant teing spilled into containment. Within another hour all releases to containment were stopped, and by 16:00 pressurizer heatup was begun to restore pressure control by this time the transient was under control. A more detailed chronology is included in Attachment F. The plant operators used the TM?-2 experience extensively in managing this event and specific training had been perfomed on the simulator. i In response to a question from Mr. Etherington, Mr. Kemper said that he thought that the operators could have controlled this transient even if the non-nuclear instrument power restoration h6d been delayed indefinitely. Mrs. Baynard described the extensive investigation that followed this incident. An indepth reexamination was made of the adequacy of the various systems and procedures to assure protection against transients like those that initiated TMI and the February 26, 1980 incident. Part of the investigation focused on power and power supply failures, the close coupling of the nuclear systems with the secondary side steam system, and the man-machine interface. Many outside experts were called upon to help with the study. Fifty-one 1.tems were identified for resolution prior to plant restart. Another list was made of longer range changes or concerns. The NRC reviewed the study and the 51 items and focused on four particular l items. On5 of these was the determination of the cause of the transient which

,. was found to be a short circuit in a buffer module circuit board card. The second wah the means by which the CR3 operators can tell which instruments are receiving reliable power (indicator lights show which channels are receiving proper power). The third item was the upgrading of the auxiliary feedwater system to safety grade. The final item of NRC examination was the PORY control circuitry to improve its reliability. The NRC was satisfied with the actions taken by FPC and CR3 was permitLJ to resume power operations. Power Level Increase Mrs. Baynard reviewed the history 'of the request and analysis of the CR3 power level increase. The initial request was made in January 1979 but in April the NRC deferred all actions on power level increase requests. The incentive is financ'ial and is tied to a State of Florida requirement for FPC to reduce its oil consumption by five percent by 1985. The increased power output of CR3 is expected to save FPC's customers about $11 million per year in fuel replacement costs. The current request is to increase the maximum authorized power level from 2452 MWt to 2544 MWt, a 3.75% increase. No new codes or techniques were used in the thermal-huydraulic analyses, and operating margins a're'largely unaffected. Using the B&W-2 correlatio'n, the minimum DNBR is 1.98. 'Oper-ation at 2544 MWt requires the addition of a pump monitor trip to assure a minimum departure from nucleate boiling ration (DN3R) for various pump related accident analyses. The installation of this equipment is complete but testing and activation must wait until the load demand drops after June 1. [~

B&W had recently identified a generic problem related to instrument accuracy used in the reactor prctection system setpoint calculations. The problem is in the reactor protection system modules and results from miner changes and improved measurement techniques. All modules have been calibrated and tested and about 1/3 have been recalibrated and mtested. It is estimated that the high flux trip setting may have to be reduced about 0.6% to 104.9% while the high reactor coolant outlet temperature may need to be reduced 0 about 1 F to 618 F to compensate for the inaccuracies. Mr. Erickson (NRC) reported that the NRC had reviewed the application for the power level increase and had compared it with approved analyses and approved operation of other reactors of similar design and had found that, with the activation of the reactor coolant pump power monitors, it could be operatedat2544Mitsafely. Conclusion The Subcommittee commended Florida Power Corporation for the improved attitude toward nuclear power expressed in its statements and reorganization. The Subcommittee also noted favorably the extensive training plans proposed l by FPC. Finally the Subcommittee concluded that. it would recommend that there is no need for W full ACRS to review the proposed power level increase for l, Crystal River Unit 3. Florida Power and the NRC Staff were advised that they would not need to be present at the May 1981 ACRS meeting. If the ACRS did not accept the Sebcommittee's recommendation then another meeting would have to be scheduled.

= o 10 - A complete transcript of the meeting is on file at the NRC Public Docu- ~ ment Room at 1717 H Street, N.W., Washington, D.C. or car. be obtained j from Alderson Reporters, 300 7th St. S.W., Washington, D.C. (202) 554-2345. e< i I ' e f

. discuss proposed Reguhilory Guides end umt.s-ACA5 5utwesm the repod re be held so es to minit uze ivconveniened Regublions. No' ' a of this meeting we, pmposed pow er les el incre.uc. 12 members of the pubhc in attendancs

  • l M"'"J "i!* N!iC C#"'ror ced other The egends for the subject rt.eeting

.pu'.hshed M. rch 27 NitC Sclct> Rcscarch Prrgror 2. luly S. commissioners 4scsse safety relats d yy y,, gg 1901. Wa shing'un. DC. The Subcurnmststc will discuss the ACRS

  • M Axident Miti.cntion rentures fer w,ge,,a,,,uay,1 sat g

Nuc!cor Plan:s-NEC Staff report re Report to the Corr. mission on the NRC j Fy-83 Research Program and Budget. proposed additiocal features for Zion and I Notice of this meetmg was publi*hed I" d' * " P"i"' ""'I' *' P(' * ' P P * "' *

  • N. Emersency Pionniar--<eport by NRC During the initiat portion of the meeting M*rch 27.

Start re consideration of natural disasters in the Subcommittee. along with an> of its ,Permi 2. luly 16.1981. Detroit. MI. consutiants who ma> be present. will The Subcommittee will review the amer 3ency planning enchange prehminar> views regarding O. Anticipated Transient 1%thout eatiers to be considerd danna the ba ance of apphcation of Detroit Ed; son for an Scr m-9 upo ed ACRS repert re actions ne muurg Operating license.

  • Emergency Core Cooling Systems.

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  • 3s, guyeom,g,,,, w;y then se r udcar p July 21-22.1981. Idaho Falls. ID. The

}=ns Fe presentations by and hold discussions with ,,f7 7 Subcommittee will review the NRC np enntativn e ta he abcock NuclearPlants-se ort by ACRS Research Program for the Semiscale and W1com Corrpany, their consultants. ana Subc*mm'""- other interested persons. Facilitbrbility andProbabilistic ejt9y81Asnd' Furthee information regarding topics

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3 da to be a o n ed. Assessment. july 2A29.1981. Los to be discussed, whether the meeting Angeles. CA. The Subcommittee wiH Dated. April 1uset has been cancelled or rescheduled, the l review some of the techniques being John C Ho3 e, Chairman's ruling on requests for the esed and will discuss the future of risk Advisory Coaunitzee Manogement off,cer. opportunity to present oral staterrents r assessment in the nuclear power p p,,,,_ 3.n m ra.d e-s-et. aes==1 and the time allotted therefor can be licensing process. Notice of this meetinf asu.s.c com neems obtained by a prepaid telephone call to was pubbshed March 27. the cognizant Federal Employee.Mr.

  • Ind.an Point 2/bfetal Cornponents.

Garry Young (telephone 202/63s-1414) y date to be determined. Washington. DCh Advisory Committee on Reactor between 815 a.m. and 5:00 p.m EDT. e i The Subcommittee will resiew the Saf eguards, Subcommittee on ne Designated Federal Employee for l possible reactor pressure vessel Babcock and Wilcox Water Reactors; this meeting is Mr. john C. McKinley. 4 deFradation caused by the flooding Meeting I have determined. in acc6tdance with incident which flooded the outside of Subsection 10(d) of the Federal v the lower portion of the pressure vessel Tbe ACRS Subcomm.ttee on Babcock Advisory Committee Act, that it may be i assembly. and Wilcox Water Reactors will bold a necessary to close sorce portions of this I ACKS Fut Coruminee Meetings meeting on May 6.1981, in Room 1167, meeting to pre'act proprietary 11 H E'**- infonnatior.he authority for such May 1-0. sars--Inema are tentatively review the Crystal River Nuclear Power closuna is rumption (4) to the Sunshine sche hf ed 1

  • A. Proposed NRCinterim Rule (10 CRg Plant. Unit 3 power upgrade. Notice of Act. 5 U.S.C. 552b(c)(4).

$0/ on Nydmgen Controland Cerio,n this meeting was published March 27. Dated. April 15.1981^ DcEroded Co e Conditions-wLseds In accordance with the procedures N proposed rufe. outlined in the Federal Register on ANsdommmemmgemenMacen

  • B Shoreham Nuclear Power Station Unit ncgny,7 y 1933 (45 gg ggg353,,,,3,7 J-proposed plant cperataon fTentatisc)
  • C Susquehanno Siece tlectric S:otion written statements may be presented by p o,,i. as-unar ra.on ses mal

""#'"**U""'*" Umts J onds-proposed plant operataon members of the public, recordings will (Tenta bwe). be permitted only during those portions

  • D IciprosedDecoyNeotRemoval of the meet'ng when a transcript is being Advisory, Committee on Reactsr Systems--proposed chae.ges at North Anna kept, and questions may be asked only Safeguards; Subcommittee on Safety Nuclear Power Station Unat 2 and other by members of the Subcommittee.ita Phiiosophy, Technology end Criteria; fe ofNRCSiting Policy-ACRS consultants, and Staff. Persons desiring UIIDE Subcorr.miti+e repo-t_

to make oral statements should notify The ACRS Subcommittee on Safety

  • F. Re vised Clodding Swellirt or ! Rupture the cegnizant Federal Employee as far in Philosophy. Technology and Criteria Models (NURECm')--cf anfy ACAS report ads ance as practicab!e so that will hold a meeting on May 6.1931.in i

appropriate arrangements can be made Rcom 1MG.1717 H Street. N.W., C on 15 s m fa ures IAct Could Be 3 to allow the necessary tim ? during the Washmgton, DC to discuss matters Couse of En ocerbote Nuclear Po wer Plant Accident.s--<!.scuss proposed ACRS report to meeting for such statement t. relating to the development of NRC. De entire meeting will b t open to requirements for new (beyond Near.

  • H DOENuc/corfacilitics-report by pfulic attendance except fc r those Tenn Construction Permit) plants and l

sessions during which the Subcommittee methads of developing requirements for DOE re review of TMI-2 lessons learned.

  • 1 Quontitatise Ris A Criteria.-report by finds 11 nece:sary to discuss proprietary new plants.

informatior One or more closed In accordance with the procedures q ant t at e sk na.

  • ]. Probabi stic Assessment ofNuc/coe sessions may be necessaiv to discuss outh,ned in the Federal Repster on Torilities--etpcrt by NRC Sta!T re use of such information. (SUNSlitNE ACr October 7.1980. (45 FR f4535), oral or probabilistic assessment in the NRC EXEMFilON 4). To the extent WTitten statements may be presented by practicable, these closed sessions will members of the pt blic, recordings will regufato y licensing process.
  • K. Crystal River Nuclear Po wer Station be permitted only during those portions a

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TENTATIVE SCHEDULE ACRS SUBCOMMITTEE MEETING ON B&W REACTORS ~ May 6, 1981 Washington, D.C. 8:30 a.m. Opening Statement - J. Ray 8:45 a.m. Introduction and Discussion cf Operations History and Management at Crystal River Unit 3 (CR-3) - Florida Power Corp. 10:00 a.m. NRC Assessment of Plant Operations at CR NRC Staff BREAK 10:30 a.m. 10:45 a.m. Discussion of February 26, 1980 Event and Follow-up Action at CR Florida Power Corp. 11:45 a.m. NRC Assessment and Action Plan Resulting from February 26, 1980 Event - NRC Staff 12:30 p.m. LUNCH 1:30 p.m. Discussion of Power L<. vel Upgrade at CR Florida Power Corp. 2:30 p.m. Discussion of RpS Instrument Inaccuracy Issue - B&W I BREAK 3:30 p.m. 3:45 p.m. NRC Review of Power Level Upgrade at CR NRC Staff 4:45 p.m. Subcommittee Discussion i: 5:00 p.m. ADJOURN i if r s I t ~

o B & W REACTORS '.5L'!'M.ITTEE MEET]NG: _ Roon 1167,1717 H St..NW, Washington, D.C. . LOC.ATION: ATTENNNCE t!S7 Tt' EASE T2rcr ~ I5._ f eiiug w, __ . v'el d'Y l Caws.v, /cRS _SueJonnirfer ~ s', WM Mrroi.s yr,,ur,, ~

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P. Y hnna! FPC - k r. hG en r5.M e4 7. // 20 -. !> 6 L. .v p, ~/~~41 Nr w r F, r3. Er > c e s I'^ x.m c o e r3 - y NEC-0 2 0 4 /_D L . f. D TOL E 10. /!SGC J 6. 'S'fAm PGLc5 JA(fucf/C 11n Porn (br> ~ # $$UrY" 11. 7 32. 7 L.M.LESJI(K Uwscoce y Uk. coy 31 Em o s' < W._ c.,'- m. / B % r y jui,>s '$wn Omn:-nd l 35 D. $ R au-1 hl_ E. E'E M PE e ptoeim "Powm. doep y. lfloti$' k'~'ek 0?/ j 17. $ b [/ein_ v3_h'!ocws.u.,: 29.#4 &#- l//(W ~ '"" tomT~./4 Taytoe l FssW x.l.C & Em x.d f & [VWDW 21- $ i, F Stef/cn usw.gc. ;- @gg .tw C.6)nnd_ NRc DR R-IcsB Wi e e r/ w h cog g ~ g !4 B. PA lLT7LE _ra. m-

ATTACHMENT D Documents Available to ACRS Subcommittee on B&W Reactors May 6,1981 1. Memo for Mr. Jeremiah Ray from Garry G. Young dated April 22, 1981 Subj : Status Report for the May 6,1981 ACRS Subcommittee Meeting on B&W Reactors 2. Tentative Schedule ACRS Subcommittee on B&W Reactors May 6,1981 Washington, D.C. 3. Viewgraphs used hy Mrs. Patsy Baynard'(FPC) (21 alides). 4. Viewgraphs used by Mr. Bill Kemper (FPC). (25 slides) 5. Sequence of events of Crystal River Unit 3 transient on 26 February 1980 (Attachment F to these minutes) 6. Viewgraphs used hy Mr. Stetka (NRC) (6 slides). 7. List of FPC concerns following investigation of 26 February 1980 incident (44 pages) 8. Viewgraphs used hy NRC Staff (6 slides) 9. Viewgraphs used hy B&W (J.H. Taylor & K.D. Tuley) (7 slides). t

.( ~ FLORIDA POiTER COR.PORATION MANAGEE per!' OPMEh7 DE.P ARTMEh7 The philosophy of Nnagement Developmect at Florida Power Corporation reflects strong comeitsent to provide cceprebetsive quality resources for expanding the co:petencies of supervisory and mar.ag+ rial personnel, and in so doing, maximize the e f f ec tive ut iliz a tioc of ham.at resocces to the besefit of both the individual and the organization. Based on the premise that asarewss precedes choice and choice precedes change, all programming focuses oc beighteting svareness and exploring choices for maximizing effectiveness it. met.agerial roles. Phase I programs address tbc needs af first and second level supervisory personnel whose salary midpoint is less than $2BDC per mocth. CO"RSES BOURS Interaction k nagement 40 Science of Skillf ul Thinking and Communicating 16 Technical Report Writing 16 k nagement Adcinistrative~ Skills 48 Management and Motivation 32 Phase II is designed for zid to executive level managers (salary midpoint above $2800 per month) and features a highly reff ned approach to

  • lividual developmental prograncing. Assessment Center methodology is an integral part of Phase II &nagement Development. Courses in Phase II are tailored te needs en an individual basis.

i l I i ( u,3,m Wsual Products Division '3V. St. Paul, MN 55101 Made in USA T-w

f 1 h [ NUCLEA1 OPERATION! BJAGEMIh"I DEVELOPET WY - Recognition that Plorida Power Corporation ManaEement Developeent did not { neet the specialised needs of Doclear Operations. p a WAT-A tasm force was assigned' to review development needs of Nuclear Operatiens and make recoer,endations. Additional courses determined to be beneficial were: Awareness and Stress Management Training Program This is a 3 day course by a consultant. It is desigt.ed to aid personnel in beco:ing aware of their environ =ent and how it may be creating stress. Kepner - Tregoe Program This is a 5 day course by a Plorida Power Corporation instructor. It is designed to aid personnel in Problem Analysis, Devision Analysis, and Potential Problem Analysis. Managerial Grid This is a 5 day course by a consultant. It is designed to increase organizational productivity and individual ef fectiveness. l l Visus 1ProductsDivision 3' w x,s,,, C4 D%..I & AM RRini k Aodo in I l' I

f ~ BOT LICT.KSr Tt.11NING I. FURPOSI A. Meet requirements fer operator licensing in accordance with NUREG-0737 and 10 CTR 55. i P licessed operators for tte plant staff.

5. Provide competent II. TRAINING METHODS A. Formal classroo: instruction and examination B. On-the-job Training (in plant) under instruction C. Operation of simulatoe controls D. Sinciated NRC exacinatioes III. PROGRAM 1.EKGTH Yifty weeks (2000 hours)

L-dC ~ VisualProducts Divisio'- St. Paul,MN 55101 Made r R

ertngfrter woumsua Classroo= Lab. Tctal PHASE I Hrs. Ers. Ers., Section 1 - Math & Physics 4S 0 _ 48 ' Section 2 - Reactor Principles 86 30 116 Section 3 - Heat Transfer & Fluid 88 16 104 Flov Section 4 - Tacility Design (syste=s) 82 58 140 Subtetal 304 104 408 l PHASE 2 Section 1 - Pool Reactor Training 40 0 40 i 160 160 Section 2 - In-plant OJT Subtotal 40 160 200 PHASE 3 Section 1 - Facility Design (systems) 28 26 54 Section 2 - Operating Characteristics 20 20 40 Section 3 - Small Break Analysis 34 32 ' 66 Section 4 - Instruc>entation/ Controls 84 72

156, Subtotal 166 150 316 PHASE 4 160 160 Section 1 - In plant OJT Subtotal 0

160 160 PHASE 5 Section 1 - Standard Technical Spees. 54 36 90 Section 2 - Procedures 64 44 108 Section 3 - Radiatior. Protection 40 38 78' 40 Section 4 - Recogtition/Mitigatioc 40 Core bmage Sobtotal 198 118 316 PFASE 6 Section 1 - Final Exacinations 9 31 40 Section 2 - Simulator Operation 60 60 120 200 _200 Section 3 - It plant OJT Subtotal 69 291 360 PHASE 7 SectioT1 - Audit Exacitatioes 9 31 40 200 Section 2 - Overall Review 200 24T Subtotal 239 K TOTAL 946 1014 2000 Visual Products Division '3M

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St. Paul, MN 55101 Made in USI- [

a I I NON-LICENSE OPERATOR TRAINING I i 1. FUPJOSE Provide the entry level 5.:1ur Fouer Flast Operator with 'necessary l academic background to enter Ect License Tra,ining. l' II. TRAINING PETHODS A. Formal classroom instruction aM exacinations i 3. On-the-job training (In plant) t III. PROGRAM LEETE Six Veels (240 hours) NON-LICENSE OPERATOR TRAINING (PROGRAM DETAILS) I Classroom Lab. Total t j Phase 1 Hrs. Ers. Hrs.. Section 1 - Mathematics 72 0 72 Section 2 - Electrical Theory 22 0 22 I i Section 3 - Rest Transfer L 40 0 -40 Fisid Flow Sectic,n 4 - Fecility Design 88 18 106 (systems) ? Total 222 18 240 l l l l. l l 1, 1 I i \\ VisualProducts Division '3r/. m e:is iox, St. Paul,MN 55101 Made in US NI

W------____________ .s. T REQUALIFICATION TPJINING = ll .I. Purpose t' il sequirements for Operater Requali,fication 'i A. Meet I in accordance with FIRIG-0737 and 10 CFR 55. I B. Maintain overall knevledge level of Operators i l through periodic review. I h II. Training Methods A. Formal classroom instructicn and examinations, f l 8 l B. Ser.i-annual oral exar.iuations. f 2. I C. Simulator Controls Operatio: ,<r I~~ III. Program Length Twelve months including 261 heurs of classroom and r....; "E lab instruction. ..y._ e r -,x . ?.f~. 7r.p .r !)5s, A y '.;.V ). 1 y.- g l.,2,g%Vi ' %. i. ,Q. $[:I. ? \\ %.g-%o 9 u. v. <.- <., ~ _ _. _.. Visua! Products Division '3M St. Paul,MN 55101 Made in USA i xxmu. I.~

t REQL' ALIFICATION TRAINING 1 k b I I PHASE 1 Classroor Mrs. Lab Ers. Total Ers. I f i Sectien 1 - Categories 228 A through N 226 l Section 2 - Se=i-z.nnual 4 4 Oral Section 3 - Annual 20 20 Simulator Trg. Seetion 4 - Annual Requalification 9 Exacication 9 TOTAL 237 24 261 lh Jr 9 I Vrsual Products Division '3f-St. Paul,MN 55101 Made in US oc os su. T

v 4 - ~ !A d" ldJ.AK OPLif1CES TECHNICAI. ADVISOR l E l I 1. Furpose to safe To aid in fulfilling Florida Feuer Corporations Comitment d d operation of Crystal River - Drit 3, and to implement our req irement un .j i E.iRIG-0578. j II. TraininS Methods l Forr.a1 classroom ins roetion and n==f nations a. On-the-job training (Ic-plant) under instruction. b. III. Prograt Length Eighteen months incloding 1424 hours of classrooma and lab. instruct I I i [ l l I l. l~ 1 VisualProducts Division 3 St. Paul, MN 5510' Made in U

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m (Details) Cicssroo= Ecurs lab Bours PROOF).M PRASE I (Acade=ic - Dtiversity ed Florida) j j / 40 NUCLEAR EN01sEERING 1 40 hTCLEAR ENGINEERINO 2 40 TEIRM3 DYNAMICS FOR ENGINEERS 40 HEAT TRANSFER AND FLUID PECEANICS 40 THIRM3, HEAT #C ESS TRANSFE1 IK KC. EAR SYS"' EMS 30 REACTOR MATERIA' S 20 REACTOR CHEMISTRY 30 40 hTCLEAR AND PRCCESS IKS ROMEnATIOE ' 40 RADIATION PRCCECTIOK 40 50 t l PRINCIPLES OF REACTOL OTIAATION C 90 M I PBASI II (Management, Administrative Controls) KGAGIMEU (ES) 40 Supervisory, Stress, Decisional Analysis AIMIKISTFJ.TIVE COCRSLS (FPC) Selected Administrative Instructions, 80 Energency Plan, CTL, Tech. Specs. l PRASE III_ (Plant" Systems - FPC/E S) Mechanical, Electrical, Instruoe tation 300 Syste=s PRASE IV (General Operating Procedures, Transient / Accident Assessment) i 1 GENETAL OPERATIM PROCEDURE (NUS) 24 f l Start up, Shutdown, ECP, SD Margin TRANSIEhT/ACCIDEC ANALYSIS (FPC) '100 Methods, EP and AP, perspective 1 t. F11ASE V (On-shift Training) l 1,20 As, extra operator NASE VI (Simulator - 34V) I I i 120 Lecture and Eands On VisualProductsDivision 3 St. Paul,MN 55101 Made in U m.s,.

I' GENERAL EMPLOYEE TRAINING 9 General Enployee Training Frogram 'n.e General E=ployee Training Program provides training to all perser.s regularly envicyed in the nuclear plant in the following areas at the frequer.:y specified. Persens regularly employed in the nuclear plant are defined as: 1. Any Florida Pcwer Corporation personnel whose work station is located within the CR-3 protected area. 2. Ary coctract personnel who kre egeted to be on site for a year or more and act in a supervisory capacity or have a large degree of freedoe to act without direct Florida Power Corporation supervision. General Employee Training Program (A) = Annual (B) = Every two (2) ye.ars (1) = Only required for those who use respirators. A. General Plant Description (B) B. Security (A) f l C. Emergency Plans (B) D. Procedures and Instructions (B) E. Industrial Safety (A) T. Quality Assurance (A) G. Radiological Health and Safety (B) E. Respiratory Protection (A) (1) I. Fire Protection Program (A) e i l 0 l I I g } r 'h t 3:

GENERAL ET:.DTEE - EA1.TH PHYSICS TRAINING t Radiological Realth and Safety Progra= (RP-101) 1. Radiation Theory b 2. Biological Effects 3. Exposure Ea:ards 4. Units of Measure a) Radiation Exposure b) Radioactive Material i 5. Measure =ent of Personnel Expossre a) External b) Internal c) Limits 6. Personnel Exposure Reduction a) Alara Concept b) Internal Exposure c) External Exposure 7. Radiological Postings a) General Postings 'o) Area Design 8. Control of Surface Contamination a) Area & Component Contanimatioe b) Personnel Contaminatioc - Protective Clothing i 9. Radiatiot Work Permits a) Purpose b) Standing Radiation Vork ",cuit j c) Radiation Work Permits t t l li I VisualProducts Division '3h} j yy, / ((

RESPIRATORY PROTECTION TEAIM5G PROGRAM i REASONS _FOR USE f 1.0 ~ 1.ici t Inhalation of Airborne Radioactive Contanisants 1.1 I 1.2 ~ Substitute For Engineering Controls ~.2.1 Design Features 1 1.2.2 Operational Control Methods 1.2.3 Containment Methods 2.0 AIR 30R'iE RAMOA~IIVE COCAMIRAES 2.1 Ftysical Properties _Ptaximum Permissible concentrations 2.2 2.3 Physiological Actions j 2.4 ~ Toxicity 2.5 Means of Detection 3.0 RESPIRATORY PROTECTIDF IEVICES IN USE AT CR-3 3.1 Constructice and Principles of Operation 3.1.1 Air Purif ying Full-Face Respirator 3.1.2 Air Supplied Full-Face Respirator 3.3.3 Air Supplied (hbble) Bood 3.1.4 Self-Contained Breathing Apparatus (SCBA) 3.2 Limitations 3.2.1 Full-Face Respirator 3.2.2 Air Supplied (kbble) Hood 3.3 Selection for Use 3.4 Special Problems Associated With Use 4.0 PROPER USE OF RESPIRATORY FR3TECTION BQUIPMENT 4.1 Pre-use Inspection 4.2 ~ Donning .i 4.2.1 Obtaining Fit 4.2.2 Checking Adequacy of Fit l 4.3 Renoval Af ter Use_ 4.4 ~Froper Disposal 4.5 Maintenance by User 5.0 EMERGENCY ACTIONS _ l 5.1 g1 function of Equi ee_gt_ t 5.2 other Energencies I r 4 1 6.0 USE WITE ADDITIONAL PR0fECTIVE EQCIPMENT_ 6.1 Frotective Clothing l' 7.0 FIT TESTING OF FULL-TACI RESPIRATORS _ ).1 pualitativeTesting 7.2 qaantitative Testint l Vs.elProducts DMsion '3Y St. Paul,MN 55101 Made in US' oc xu,m 5

T E CH' I C AL AND MAIETENANCE PERSONNEL SYSTEMS TRAININC ' Systems Training Progra: Listin; t 1. Reactor Theory and Operation t 2. R: System 3. Reactor Protective Syste: 4. Du: lear Instrumentatice 5. Makeup and Purification Syster 6. Decay Heat Removal System 7. Emergency Core Cocling System 8. Engineered Safegc.ards Systes 9. Steam, Peed and Ccedensate

10. Turbine EEC Ccmtrols and Oil 11.

Bail,ey 855 Ccweter, ID'-5100 Conguter ~ 12. Control Rod Drive System

13. Non-Nuclear Instrumentation i

14. Electrical Distribution

15. Air Handling 16.

Radiation Monitoring 17. Overt 11 Systems Diagram and Operation 16. Integrated Control System P 19. Condenser Air Removal l f 20. Extraction Steam e

21. Heater Drains y

Seismic System i 22. -23. Waste Disposal System 24. Meteorological Tower A P 1 9 o W~

/ l i . NUCLEAR TECHNICAL SUPPORT TECHNICIAN TRAINING Ecurs [ A. Safety 4 t 12 B. Print Reading Techniques C. Pneumatics 16 1. Actuators 8 2. Positicroers 24 3. Transmitters 16 4. Controllers D. Electronic Instruments 24 1. Transmitters 8 2. Indicators 8 3. Signal Converters E. Mechan,ical Instruments 4 1. Pressure Indicators 4. 2. Temperature Indicatore 8 3. 14 vel Indicators P. Specific P' int Systems 16 1. Reactor Protective / Nuclear Instruments 16 2. Engineered Safeguards 80 3. Turbine Electro-Hydraulic Control 40 4. Integrated Control System 12 5. Security System 300 Visual Products Division 3r 78 6969 3951 6 . t Paul. MN S5101 Made in US 9

cal 2BRATION LAB TRAINING 1. CR-3 Test E quip me n t 2. Calibration Lab Frocedures 3. On-The-Job Training l l i I i l Visual Products Division'3f.' 78.eas9. asst.s St. Paul.MN 55101 Made in US ' Y

1 ~ EiiANICAL MAIN ~E*tANCE TRAININC A. On going Mechanical Maintenance Traini:4 1. Mechanical drives, complings and aligament II. Pumps, Centrifugal and Camd III. Air Compressors 1 IV. Diesel Engines Y. Relief Yalves VI. R.igging l B. Other Mechanical Training 1. Valves, packing and repair II. Fork Lif t Operation Training I III. Svitchiung and Tagging Training i IV. Industrial Safety Training i t 1 1 1 h e a ) lh

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EEINT.D TRAIYlE PRDORAM t l-1 I I. Purpose l To provide exempt teckieal per-1 with trainisg about Crystal I. g River Unit 3 oc that they will be able to perfore their work more l effectively and efficiently. i II. Training Methods t Formal classroce and lai instructice including exaninations. III. Program 1.enEth The program is still under development. Bowever, the Systee.s Training I Course will last approxirately ten months, (one day per week). -{ i t I Courses being considered fee inclusion are: -{ >t Reactor and Steam Flatt Fu nfamental . a ~x f Engineering Documentaties kh a Codes and Standard i Switching and Tagging Administrative Procedures 1 .C;. .f Start-up Physics Testing ,,, f Engineering Econoury ,4,,40 i .. t." $*5f[g, ' +%, Computer Time Sharing Operation pg<py I \\ Ej?$l .. [Q,2 I

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I *. / ELECTRICAL TRAIKING 1. Reactor and Steam Plant Fundamentals, Systems Training 2. CR-3 Communications 3. Switchgear Equipment 4. Rigging and Electrical Safety 5. Fire System Mcintenance 6. Control Rod Drive System 7. Cranes and Boists 8. Annunciator Syster 9. Main Turbine Exciter 10. Limitorque Operators 11. Pressurizer Eeater Controls 12. Vital fystem Inverters l i i l I i I i l l \\W-Visual Products DMsion '3t/. - -... ~. l

o C SITE EtER0ENCY T1.111200

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A. E.d.ERGE'Q TEAM TRAIN 1EG I. Medical Emergency Team II. Emergen:y Repair Team III. Radiation Emergency Tern IV. Fire Emergency Team V. Environmental Survey Team VI. Accident Assessment Personnel e* B. EMER0ENCY FR30ED'.'RIS TRA!KIE 1. IM-100 Imergency Plan II. EM-201 Duties of an Individual Discovering an E=ergency C. EMER0ENCY DRILL.S & EXERCISES (Ref: Sect. 18-C'.-100) 1. Annual Requirements (+ 3 months) A. Site Emergency Drill (Note: Annual Requirements B, C & D may be satisfied by inclusion in the Site Emergency Drill). i 1. All major elements of the plan should be ( tested within 5 year period. i 2. Ivery 6 years: 1 drill mst start between 1 6 PM and Midnight. 1 defil mst start between i j Hidnight and 6 A:1. l } VisualProducts Division 3f St. Paul,MN 55101 Mi1"ifA

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SITI DER 0ENCY 'IRAIH15C B. Radiological Monitoring Drills C. Fxergency 1.iguid Sar;11r4 Frocedures D. Medical Incrgency Drill i II. Semi Annual Requirements (+ 6 hks) A. Responses to and analysis of off-site airborne and liquid sas;1es, and direct radiation measurements. (Note: One of these drills inay be satisfied by inclusion in the Site Emergency Drill.) } III. Quarterly Requirew nts (+ 1 month) l l A. Fire brills (Eote: Que of these drills may be as:isfied by inclusion in the Site Emergency Drill.) f i i i n 1 L I t ( VisualProducts Division 3h, St. Paul, MN 551 1 fArI in USA K

CEEMISTRY AND RADIATION PROTECTICE TRAINIMC r-I. CMIMISTRY II. CBIF./KAD EQUIPFINT III. MATE IV. FEYSICS V. PROCEDUFIS (As Required) VI. RADIATION BIOM' VII. RALIATIQN AND RADIATION DETECTION VIII. REGULATIONS IX. STATISTICS X. STSTD!S Escal Products Division '3f,*. oc on,m St. Paul, MN 55101 Made in USA

O S. C 74 SEQUENCE (AS OF 2300 3/1/80) l -( 26 February Transient CR-3 EVDC STNQSg At 14:23 on February 26, 1980 Crystal River -3 Nuclear Station experienced a reactor trip frc approxi=ately 1002 full power. A synopsis of key e', vents and parameters was obtained fro the plant computer's post-trip review and plant alarm su==ary, the sequence of events monitor, control room strip charts, and the Shift Supervisor's log. Thereactor>wasoperatingatapproxi=ately100%fullpowerwithintegrated Control System (ICS) in auto =atic. No tests were in progress and minor main-tenance was being perfor=ed in the Non-Nuclear Instru=entation (NNI) cabinet "Y". Time Event Cause/Cecnents 14:23:00 The following is a su==ary of plant conditions prior to the trip I' lux 98.6% RC Pressure 2157 psig ( PZR level 202 inches MU tank level 71 inches T "A" 599'F. 'TH "B" 600*F. TH "A" 557'F. TC "B" 556*F. RbFlow"A"73X106 lbs/hr RC Flow "B" 73 X 106 lbs/hr Letdown Flow 48 gpm OTSG "A" Iv1 (OP) 67% OTSG "B" 1v1 (OP) 65% OTSG "A" FRLV 242 inches OTSG "B" FRLV 254 inches OTSG "A" Pressure 911 psig OTSG "B" pressure 909 psig Main Steam Pressure 894 psig Main Steam Te=p. 589'F. Condenser Vacuum 1.76 Generated F.J 834 DFI level 12.7 ft. Feed.F1'w "A" 5 X 106 o lbs/hr Feed Flow "B" 5 X 106 lbs/hr l-Fced Pressure "A" 970 psig Feed Pressure "B" 968 psig g /g A/,M kf 14:23:21 +24 Volt Bus Failure (NNI Cause'IitIWE m

rently,

( power loss'. "X" supply) the positive 24 VDC bus shorted dragging the bus voltage down to a 1 1 m --1

[bv. M Pag 2 2 Evant Cau s 2 /Cc== tnt s Tim 2 ,( low voltage trip condition. There is a built-in k to second delay at which time all power supplies will trip. There was no trip indication on nega:.ive (-) voltage. This event was missed by the annunciator. Following the NNI power failure, much of the control room indication was los*t. Of the instrum- 'entation that remained operable transient conditions made their indic-cation qu'estionable to the operators. 14:23:21 PORV and Spray Open When the positive 24 VDC supply was lost due to the sequence discussed aben the signal monitors in h3I changed statt ceusing PORV/ Spray valves to open. The PORV circuitry is designed to seal in upon actuation and did so. The resultant loss of the negative 24 VDC-halted spray valve motor operator and prevented PORV seal in from clearing on low pressure. It is postulated that the POI'. ( opened fully and the spray valve stroked for approximately second. The 40% open indication on spray valve did not actuate, therefore, the spray valve did not exceed 40% open. 14:23:21 Reduction in Feedwater As a result of the "I" power supply failure many primary plant control signals responded erroneously. Teoid failed to 570'F (normal indication was 557'F) producing several spurious alar:r Tave failed to 570* F (decreased). The resultant Tave error modified the react: de=and such that control rods were withdrawn,.to increse Tave and reacter power. The power increase was ter=inate at 103: by the ICS and a " Reactor De anc High Limit" alarn was received. Thot failedtog70*F(low)andRCflowfailed to 40 X 10 lbs/hr in each loop (low). Both these failures created a BTU alar: and limit on feedwater which reduced feedwater flow to both OTSG's to essentially zero. Turbine Header Pressu: failed to 900 psig (high) which caused the turbine valves to open slightly to

Rev. 5 Pcg2 3 '( T1=e Event Cause/Ce==ents regulate header pressure thus increasing 1 generated megawatts. These combined failures resulted in a loss of hest sink to the reacte: initiating an excessively high RC pressure condition. 14:23:35 Reactor Trip / Turbine Rx trip caused by high " CS pressure at 2300 p: R Trip Turbine was tripped by the reac % r. 14:24:02 Hi Pressure Inj. This was a co=puter printout and indicates Req. (Flag) <50' subcooling.* See attached graph of RC 1 Pressure /Te:p. vs. Ti=e. This graph is Sased on Post Trip data and actual incore ther=:- ccuple data. From the reactor trip point (ll : to 14:33, core exit te=perature data was ( obtained by extrapolation and calculated data This is supported by.tvo alar = data points plotted at 18' and,21' of subcooling during this period from eta co=puter. It is i=portan to note that lowest level of subcooling was 8'F for a very short period of time. (

  • NOT2: This computer program was initiated as a result of the TMI incident.

14:24:02 Loss of Both Suspect condensate pump tripped due to hi p Condensate Pu=ps DFT level. This is verified by 7777 printe3 by computer, indicatitig the level instru=er.: was over ranged as well as a low flow indication in the gland stea= condenser as al indicated by computer. 14:25:50 PORY Isolated At this time a high RC Drain Tank level alar: gggyg;// was received. This was resultant fro = the g p,ffy g PORV remaining open and was positive indicati gM that the PORV was open. At this ti=e, the T operator closed the PORV block valve due to " - M r 7 NT 7-RCS pressure decreasing and high RCDT leve - t 14:26:41 HPI Auto Initiation HPI initiated automatically due to low RCS pressure of 1500 psig. The low pressure condition was resultant from the PORV re:aini full open while the plant was tripped. Full HPI was initiated with 3 pu=ps resulting in i approximately 1100 gpm ficw to the RCS. At l this time, all remaining non. essential R.B.

0., f. g isolation valves l

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R:v. 5 page 4 .( Time Event Cau se /Coc=ent s were closed per TMI Lessens Learned Guidelines i '14:26:54 RC Pu=ps Shutdown Operator turned RC pu=ps off as required by the applicable e=ergency procedure and B & W small break guidelines.' 14:27:20 ' RB Pressure Increasing This is first indication that RCDT rupture dise had Yuptured. RB pressure increase data was obtained from Post Trip Review and Strip Char indication. ,14:31:32' RB Pressure High This alarm was initiated by 2 psig in RB. Th: is attributad to steam release from RCDT. Cet safeties had not opened at this time based up tail pipe temperatures recorded at 14:32:03 (Computer). 14:31:49 OSTG "A" Rupture Matrix This occurred due to <600 psig in OTSG "A. Actuation The low pressure was caused by OTSG "A" boili: dry which was resultnt from the BTU limit an: I failed OTSG level transmitter. This resulted in the closure of all feedvater and steam blo: valves which service OTSG "A". 14:31:59 Main Teedvater Pump 1A Caused by suction valve shutting due to f Tripped matrix actuation in previous step. 14:32+14:41 ES A/B Bypass Manually bypassed and HPI balanced between s12< 4 nozzles (Total flow approximately 1100 gp= -small break operating guidelines). ./14:32:35 Started Steam Driven Started by operator to ensure feedvater vas E=ergency Teedvater Pu=p available to feed OTSG's. g 14:33 Core Exit Te=p. Verified The core exit incore thermocouples indicated the highest core outlet temperature value was 560*F. RCS pressure was 2353 psig atthis ti=c therefore, the subcooling margin at this ti=c was 100*F. Minimum subcooling margin for the entire transient was 8 T. It is l postulatad that n -localized boiling occurred in the core.-at-this-point--as- -tod ic a t ed-by-the-s e l f-powe red-n eu t ron- -detectorov l V14:33+14:44 Started Motor Driven E=er-Same discussion as " Started Steam Driven E=er-( gency Feedvater Pu=p gency Teedvater Pump." /14:33:30 RC Pressure High (2395 psig) At this point, pressurizer is solid and code l l safety lifts (RCV-8). This is t.!.e highest l RCS pressure as recorded on Post Trip Review. Apparently, RCV-8 lifted early due to seat

R:v. 5 Pags 5 Time Event Cause/Co==ents -i leakage prior to the transient and RCV-9 did not lift. v14:34:23 RB Do=e Hi Rad Level RMC-19 alarmed at this point. Highest level indicated during course of incident was 50 R/hr. High radiation levels in RB caused by release of non-condensable gases in the press-urizer and coolant. 14:35:33 Attempted NNI Repower With-This resulted in spikes observed dn de-ener-out Success sized strip charts. ~ 14:36:50 Compu:.er Overload Caused by overload of buffer. Resulting in no further computer data until buffer catches up with printout. 14:38:15 WV-34 Closed This valve was closed to prevent uverfeeding OTSG "B" b,eyond 100:s ' indicated Operating Rang 14:44:12 NNI Power Restored Success-NNI was restored by re=oving the X-NNI Power 'g fully Supply Monitor Module. This allowed the breakers to be reclosed. At this time, it var observed that the "A" OTSG was dry, the press-urizer was solid (Indicated off scale high), RC outlet temperature indicated 556*F (Loop A & B average), and RC average temperature indi-cated 532*F (Loop A & B). The highest core er. thermocouple te=perature at this ti=e was $31*~ RSC pressure was 2400 psig (saturation temp. a this pressure is 662*F.). This data verified untural circulation was in proeress and the plant subcooling margin was 131*F. (based on core exit thermocouples). /14:44:31 RB Isolation and Cooling Actuation At this time, RB pressure increased to 4 psis and initiated RB Isciat sn. The operator verified all im=ediate actions occurred proper.- i: for HPI, LPI, and RB Isolation and Cooling. T1 l! passing HPI at this time. increasing RB pressure was resultant from RCV-1 s 14:46:10 Bypassed HPI, LPI and RB These "ES" syste=s were bypassed at this ti:c Isolation and Cooling to again balance HPI flow and restore cooling t. vater to essential auxiliary equipment (i.e., ( RCP's, letdown coolers, CRDM's etc.). l-r

e R;v. 15 Page 6 'd Time Event Cause/Com=ents '14 :51:57 Rupture Matrix Actuation on The actuation was resultant from a deg-OTSG-B radation of OTSG-B pressure. Cold e=er-gency feed was being injected into the OTSO at this time. This matrix actuation isolated all feedwater and steag block valves to the B-0TSG and tripped the '!B" main FW pu=p. Both E=ergency FW pu=ps were already in operation at this time. B-0TSG 1evel at -this ti=e was 70 (Operation Range). -14:52 EPI Throttled and RCS At this time, the maximum core exit ther:c-Pressure Reduced to 2300 couple te=perature was 515'F RCS pressure psig was 2390 psig. Therefore, the subcooling margin was 147'F. Natural circulation was in effect as verified previously. All con-ditions had been satisfied to throttle H?l. Therefore, flow was throttled Sovn to approx-imately 250 gpm to reduce RCS pressure to 2300 psig in order to atte=pt to reduce the flow rate threugh RCV-8 and into the RB. /k4:53 Reestablished Letdown At this time, the operator was attempting ( to establish RCS pressure control via normal RC makeup and letdown. 14:56 Opened MU Pu=p Recire. This vac done to assure the MU pumps would Valves have minimum flow at all times to prevent possible pump damage. /14:56:43 Bypassed the A-0TSG Rupture Feedwater was slowly admitted Matrix and Reestablished - to the A-0TSG which was dry up to this point. Feed to the A-0TSG Teedwater was admitted through the f.uxiliary FW header via the EFW bypass valves. The feedrate was very slow in order to minimize thermal shock to the OTSG and resultant depres surization of the RCS. RCS pressure control was very unstable at this time. It is posh 'c. th.at some,localiz,ed'1,$iling.occured in core at this point ~as indicated by.self. neutron detectors. k =. W

Pcg2 7 ' f- .( Time Event Caus e /Co== ants 14:57:09 Bypassed the B-0TSG This was done to regain FW control of the Rupture Matrix B-0TSG. Level was still high in this OTSG e. (approximately 65 Operating Range). Therefore, feed was not necessary at this time. The Main Steam Isolation valves were open. in preparation for bypass valve operation (when necessary). 14:57:15 Established RC Pu=p This was done in preparation for a RCP start Seal Return (when necessar,) and to minimize pump seal degradation. 15:00:09 Reestablished Level This verified feedwater was being admitted to In A-0TSG the OISG and made it available for core cooling via natural circulation. Feed to this generator was continued with the intent of proceeding to 95% on the Operating Range. 15:00:09 77'T Subcooled A" Loop This value was based upon "A" RCS loop para =etersatthistime.The"A"lo;cpwas being cooled down at this time by the A-0TSG fill and the operator was attempting to equalize loop tempe atures. 15:15 23'F Delta-T/ Manned the At this time, loop temperatures were nearing Technicil Support Center equalization. This delta-T was calculated from loop A & B T 's and core exit thermo- ~ e couples. 15:17 Declared Class "B" E=ergency This was done based on the fact there was a loss of coolant through RCV-8 in the ~ containment and HPI had been initiated. All non-es sential CR# 3. personnel were directed ,te tvacuate 'and(cont,act off, site agencies ~be-gan. Sox 7ey" team was sent to Auxiliary Buildi: '15:19 Opened E=ergency FW Block At' this point the A-0TSG 1evel was increasing to B-0TSG and the decision was made to com=ence fillier the B-0TSG simnitaneously. The intent was to go 95% on both OTSG's without exceeding RCS cooldown limits (106'F/hr) while maintaining RCS pressure control. e e e enume e6

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Rsv. 5 Page 8 i '( Ti=e Event Cause/Co=:ents ,15:26 Lo Level Alar = in Sodiu= This was recultant from the tank supply valvc Hydroxide Tank opening when the 4 psig R3 isolation and coel-ing signal actuated. The sodium hydroxide was released to both LPI trains. ~ Sediu= Eydrixide was ad=itted to the RCS via HPI from.the E'45T. ('Approxitately 2.ppe injected into the RCS.) 4 At this time, all conditions had been satis-15:50 Terminated HP1 ~ fied (per small break operating guidelines) to terminate HPI. RCS pressure control had been established using normal makeup and letdova. HPI was terminiated and essentially all releases to the RB were discontinued. 16:00 Co==enced Pressurizer At this time, RCS pressure and te=perature Beatup were well under control. Natural circulatien was functioning as designed (approxi=ately 2T delta-T). RCS temperature was being maintain < at approxi=ately 450'. RCS pressure was apprc imately 2300 psig. The decision was made at this point to connecce pressurizer heatup in { preparation to re-establish a steam space in the pressurizer. 16:0/ Survey Tea = Report The E=ergency Survey Team reported no radiatic survey results taken of! site were above back-ground. 0 16:08 :04 Shutdown Steam Drive Emergency N Pump The motor driven Emergency W pump was runnin therefore, the steam driven pump was not need. The plant remained in this condition for app-roximately 2 hours, while heating up the pres: urizer to saturation temperature for 1800 psi, l 16:15 Press Release Media was notified of plant status. 18:05 Established Steam Space Pressurizer t.t this point, pressurizer temperature was approximately 620'F. Pressurizer level was brought back on scale by increasing letdown. From this point pressurizer level was reduced e to normal operating level and normal pressure was established via pressure heaters. 18:30 Terminated Class E Emergency State and Teder'al Agencies notified. k W*

O. R:v 5 Pega 9 -( Tine Event Cause/Co._ ents 21:07 Forced Flow Initiated The decision was made to re-establ'.sh forced in RCS flow cooling in the RCS at this ti=e. E s', and NRC vere consulted. RCP-13 and ID verc started. At this point,,RCS paraneters verc stabilized and maintained at RC pressure-20' psig, RCS temperature-420*F. Pressuri:er level-235 inches. The* plant was considered a normal configuration. i + I d e(L,.,A ~. ) s, ,fr s ~ ( lh psuG 13 % -! 8 g.g#a ~ %v/JT&TT.6/ t e l:

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