ML19345B404

Nonproprietary Version of, Nuclear Safety Task Force, Final Rept

ML19345B404
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Site:	Crystal River
Issue date:	11/26/1980
From:	Baynard P FLORIDA POWER CORP.
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O CRYSTAL RIVER UNIT 3 NUCLEAR SAFETY REVIEW TASK FORCE FINAL REPORT NON-PROPRIETARY VERSION i

Noveraber 26, 1980 O l e 8 012010 Ogg

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u EXECUTIVE

SUMMARY

Nuclear Saf ety Review Task Force November 26, 1980

1. Background Information The Nuclear Safety Review Task Force was established following the February 26, 1980, inc ident at Crystal Rive- Unit 3 to awamine the adequacy of various systems, procedures and training at CR-3, and maha recommendations for improvements and establish appropriate priorities for implementation. This ef fort was to assure that needed corrective actions were identified and implemented prior to restart, and that a longer term program for cI;anges could be formulated.

2. Task Force Activities The Task Force, with specialized expertise cons! sting of both Florida Power Corporation employees and others from throughout the Nuclear industry,, reviewed various industry-wide documents on experiences to-date. These evaluations were utilized to develcp recommendations to improve the overall sa f ety of CR-3. These recommendations meet the intent of Improving the overall saf ety of the plant in terms of the following:

. Reducing plant upsets and reactor trips.

. Reduce system response fetiew!ng reactor trip.

. Reduce the probability and consequences of severe transients.

. Human engineeri1g and post-trip transient analysis.

To achieve the overall safety improvements, recommendations were made that procedures be nodified, training be reinforced, and Improvements on sensitivity of plant response be implemented. The man-machine interf aces were evaluated to aide tne operator in gaining better understanding of what is happening in the plant and how to better Interact with it. Recommendat ions for nodification vera made as a result of this evaluation. For example, a recommendation was made that the Main Stea.- Line Rupture Matrix to nodified so that It does not close FWV-161 and FWV-162. This will allow a passive flow path for Emergency Feedwater, improving the reliability of that system.

The objective was not only to provide further assurance that translents would not lead to core damage, but also that transients would not result in significant economic penalty.

3. Justification for Restart it is the concensus of the Nuclear Safety Review Task Force that if the short-term changes it has rec-ommended are implemented, the Crystal River Nuclear Plant can be opereted without adversely af fecting the health and safety of the public or Florida Power Corporation employees.

4 Conclusions Both the short-term and long-term recommendations provide a further marg i n of overall safety for CR-3. The overall nanagement philosophy toward a proactive licensing approach of evaluating our own problems, identif ying solutions to these problems, and implementing the improvements in accordance with the overall saf ety significance, leads us to an aggressive approach of beirg out front in problem solving with technical bases for recommended changes.

pY File Code: Alpha-X-t-1980 PY8(ExecSumry)DN-93

TABLE OF CONTENTS I. INTRODUCTION - CHARGE OF THE TASK FORCE II. MEMBERSHIP III. TASK ~ FORCE ORGANIZATION /0PERATION

, IV. SUMMAr,Y OF ACTIVITIES V. ' EXPERIENCE T0-DATE REVIEW VI. LIST OF CONCERNS VII. OVERALL ASSESSMENT OF SAFETY IMPROVEMENTS VIII.

SUMMARY

OF SHORT-TERM RECOMMENDATIONS INCLUDING TRAINING IX. LONG-T RM RECOMMENDATIONS SCOPE AND F TURE WORK X. CONCLUSIONS

ATTACHMENT 1~ - List of. Items Recommended to be Completed Before Startup i

ATTACHMENT 2 - List of Concerns ATTACHMENT 3 - Review Documents i ATTACHMENT 4 - Experience To-Date Document List ATTACHMENT 5 - Plant Staff Documents ATTACHMEhi 6 - List of Abbreviations n

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CRYSTAL RIVER UNIT 3 Nuclear Safety Review Task Force Final Report I. INTRODUCTION - CHARGE'0F THE TASK FORCE Following the February 26, 1980, transient at Crystal River Unit 3 (CR-3), the Nuclear Safety Review Task Force was estab-

. lished by and to report to Mr. J. A. Hancock, Assistant Vice President of Nuclear Operations, Florida Power Corporation. The purpose-of the Task Force was to reexamine the adequacy of vari-ous systems, procedures and training at CR-3, identify and devel-op recommended changes and recommend appropriate priorities of implementation. The priority of implementation process assures that needed corrective actions will be accomplished prior to re-start following refueling operations. The restart of CR-3 was scheduled on or before June 1,1980. The Task Force's efforts were not intended to be an unbounded inquiry, but rather, a hard-hitting look at the issues that experience to-date identifies as having greatest potential risk.

The first priority areas of inquiry for the Task . Force were the subjects of power and power supply failures and the c~ lose cou-pling of the NSSS with the secondary side of the plant. The ob-jective was to strengthen identified weak points in order to de-crease the chances of a core damage incident. The concern for the " man / machine interface" emphasized the viewpoint of the oper-ator as a major factor in review on an equal status consideration with the hardware side. It was not intended for the Task Force to examine all safety considerations identified in the FSAR on a nonprioritized basis; however, the' task force was not precluded from any area of inquiry that analysis / strategy lead them.

The summary of recommendations made by the Task Force are includ-ed in Attachment 1, List of Items Recommended to be Completed Before Startup; and Attachment 2, List of Concerns, including all short- and long-tenn recommendations.

II. MEMBERSHIP 4

The Areas of Expertise and Membership of the Nuclear Safety Re-view Ta.k Force are as follows:

e Florida Power Corporation -- Owner and ope. ator

. Leadership and Overall Industry Perspective

. Plant Operation

().,, e Instrument & Control and Electrical Engineering l

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. Management Perspective e Reactor Physics e Computer Applications y e Nuclear Safety e Systems Performance Personnel I..cluded:

. P. Y. Baynard e G. P. Beatty, Jr.

. R. E. Clauson e M. B. Foley, Jr. (Part-time)

. A. E. Friend e J. C. Hobbs , Jr.

• P. F. McKee e J. D. Martin

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.- D. A. Morrison

. K. O. Vogel e Bah.ock & Wilcox -- Nuclear steam supply system vender

, Licensing

. Engineering e Pl a n.t e Instrument & Controls e Systems Personnel Included:

e - C. Bosted e G. J. Brazill e E. R. Kane

. M. E. Newlin e G. D. Quale 1 e Gilbert Associates Incorporated -- Architect / engineer 1 . Engineering

, o Mechanical i

e Instrument & Controls Personnel Included:

. E. J. Anselmi e T. C. Reitz e- Institute for Nuclear Power Operations -- Industry operating experience, performance, training, and procedures

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() Personnel Included:

. P. E. Dietz

, Nuclear Safety Analysis Center -- Industry-wide focal point for safety considerations e System Safety Personnel Included:

e B. Chexal e Energy Incorporated -- Risk assessment and safety analysis 4

e Probabilistic Analysis e Safety Analysis Personnel Included:

e R. Bertucio e J. Young (Part-time) e Toledo Edison -- Owner and operator of Davis-Besse Unit 1 e Operations Personnel Included:

e M. J. Derivan 4

Members of the Task Force also utilized the expertise of others within their own organization as needed for. further support as required. The diversity of technical backgrounds of the members aided the Task Force's efforts in gaining various industry-wide perspectives for Task Force review.

III. TASK FORCE ORGANIZATION /0PERATION The Task Force members were selected by the management of the designated organizations. The initial organizational meeting was held on March 13, 1980. Daily meetings . were held starting March 17, 1980, with the concluding review meeting held April 15, 1980.

General areas of concern were discussed to develop ideas leading to an ef fective method of organizing the Task Force effort. Task Force goals were discussed.

. 7s The Task Force decided to organize two Task Teams. Both would

(_) . devel op areas / items of concern for- review, evaluation, and

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O recommendation by the entire Task rocce dering speciei rev4ew meetings held for this purpose on April 3,14, and 15,1980.

The operation of each Task Team was carried out during a four-week period, with daily consultation between the groups for the first two weeks, and daily integration thereafter. Sumaries of the Task. Teams' activities, methods of operation, and scope of efforts are outlined in Sections IV and V.

IV.

SUMMARY

OF ACTIVITIES Task Team 1, Sequence Review, met on March 17, 1980, to develop a list of items for review. It was decided to work first on two items: (1) NNI/ICS Power Supply Failure Modes, and (2) Annuncia-tor Systems.

Task Team 2, Experience To-date, met on March 17, 1980, to develop a preliminary list of documents pertaining to the CR-3 transient, TMI-2 transient, and other related events that could provide a basis for identifying specific items for improvement in many areas. After development of the document list, assignments were made to obtain and review documents and prepare a review sheet with recommendations for consideration by the Task Team, and eventually, the entire Task Force.

V. EXPERIENCE T0-DATE REVIEW The Experience To-date group (Task Team 2) developed a list of documents related to the CR-3 transient, TMI-2 transient , and other plants' events, to provide a diverse source of plant infor-mation from which to establish areas of concern and to select specific items for further study. Detailed reviews were per-fomed on each document.

The recommendations were discussed within the Task Team. Once a concensus was reached, the recommendation was placed on the List of Concerns for review by the entire Task Force.

1 The Experience To-date Document List is contained in Attach-ment 4.

VI. LIST OF CONCERNS As the review work by both Task Teams proceeded, specific items that were identified as plant improvements (in any area) were placed on a composite listing, together with a reference to the originating document, and a preliminary priority was designated to guide the assignment of work efforts during the completion of the Task Force ef forts. Assignments were then made to individual Task Force members to perform further research on the item and q report to the Task Force, for review by other members and consid-V eration by the entire Task Force.

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O The ria 1 'ist or coaceras is Attaca= eat 2-VII. OVERALL ASSESSMENT OF SAFETY IMPROVEMENTS The Task Force has made a rumber of recommendations on improving the safety of the CR-3 Nuclear Power Plant. These recomenda-tions can be classified by a defense in-depth concept of plant protection. The defense in-depth concept provides for improve-ments in plant safety through various stages of prevention and mitigation. Listed below, along with a brief sumary of recom-mendations, are the various defense stages.

. Prevent initial plant upsets and reactor trips. Reconmenda-tions within this defense stage minimize the probability of reactor trip. These recommendations cover both system and control modification, as well as Operator and Technician training. They also carry the additional benefit of improv-ing plant availability.

, Reduce system response following reactor trip. Recommenda-tions in this area are directed at changes that minimize plant upsets following a reactor trip. They typically in-clude improving EFW control system response, minimizing pressurizer level changes, and providing operator training and improved procedures for such events as loss of power supplies.

, Minimize the probability and consequences of severe transi-

e. *. s . Recommendations in this area are directed at improve-ments which assure availabilty of systems related to main-taining core cooling. These include such reconmendations as modifications to the rupture matrix (SLRM), provisions for automatic actuation of EFW, and high pressure injection when high RCS pressure exists, and the addition of a third EFW pump.

. Hunan engineering and post-trip transient analysis. While more difficult to define in ter...s of measurable safety im-provements, the Task Force has made a number of reconmenda-tions wMch fall into a category called Human Engineering and Post Trip Transient Analysis. Recommendations in this area include such items as modifications to the Control Room to more expeditiously locate instruments, review of alarm systems and providing for additional post-trip infonaation to better evaluate transient sequences. The latter recom-mendation would assure adequate infonnation availability in any future trips 50 that appropriate lessons can be learned.

The above represents a summary of the general philosophy of de-fense in Jepth. More detailed discussions of how the Task Force rec omment itions will improve plant safety are provided in the O foiiowia9 six subsectioas-

O A. eROCE0uRc5 The Task For:e made approximately ten recommendations which, in some manner, apply to plant procedurer. Some involved revision of existing procedures while some involved creation of new procedures.

EP-104, Steam Generator Tube Failure, is being rewritten to reflect the results of Task Force analysis and B&W subcool-ing guidelines. The procedure was expanded to cover opera- '

tion when PZR spray is unavailable.

Reconmendations were made to revise LOCA procedures and overcooling procedures to provide the operator with addi-tional methods to distinguish LOCAs from overcooling trans-ients.

Both the Task Force and B&W recommended lowering the 50 i

subcooling guidelines to 20' subcooling for RCS pressures above HPI setpoint. B&W will also revise tN Small Break Guidelines.

A recommendation was also made to provide guidelines for in-itiation of HPI Cooling. Procedure guidance will also be

{ provided for restart of MFW in the event that EFW is lost. ,

i Prior to the Task Force, there was no guidance for the Oper-  ;

ator in this situation. This procedure .alona with the HPI i

._ cooling guidelines, will reduce the frequency of HPI cool-t ing.

These revised procedures and new procedures will provide the Operator with increased capability to properly and quickly 1 respond to a large class of transients. Those transients are undercooling, overcooling, and LOCAs. The Task Force i

believes this will lead to a positive and significant in-provement in safety.

B. TRAINING Many training recommendations resulted from the Task Force.

All plant operating personnel will be trained in all changes occurring as a result of the Task Force recommendations.

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The major additional training reconnended by the Task Force was thorough training of Operators in responding to NNI and ICS failures.

C. POWER SUPPLIES 4

The February 26, 1980, transient resulted directly from a loss of NNI power. As a result, a major area of Task Force

't investig stion was associated with the prevention and mitiga-tion of power supply initiated transients.

The Task Force performed an evaluation of problems associa-ted with Power Supplies. The review included investigations made by' NSAC, B&W, and the NRC. The Task Force believes that implementation of the recommendations it made wouk

, have the potenhial of eliminating 11 of the 18 identified

trips through equipment modification. While not specifical-ly identified as eliminating trips, the Task Force also be-lieves that recommendations for Operator and Technician training have the potential for eliminating some trips. In addition, a number of recommmendations are made which would significantly reduce the severity of trips if they would j occur.

In summary, the Task Force identified a number of recommen-j dations related to power supply failures. These recomenda-tions provided for:

1. Reducing the trip frequency.

2. Reducing the severity of the transient.

3. Providing operators with redundant instrumentation and

additional alams so that they have sufficient reliable instruments from which to identify and control the transient.

4. Providing operator and technician training and proce-dures to more ef ficiently handle a loss of power supply.

The Task Force believes that the implementation of these reccomendations would have, as a minimum, changed the CR-3 transient of February 26, 1980, to a typical reactor trip with no CSFAS actuation, no PORV activation, and no reactor coolant spillage to the containment floor.

D. SENSITIVITY The overall issue of plant response or " sensitivity" has been raised a number of times regarding plants with reactors designed by Babcock & Wilcox such as the CR-3 nuclear power pl a nt. This issue has been typically identified with gener-al headings such as:

o System Response to Plant Upsets at Power e ICS/NNI Control i

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3 O e Pressurizer Level Response o Emergency Feedwater and Shutdown Control Systems The Task Force has evaluated a number of recommendations in this area and has provided recommendations which, if imple-mented, will improve the reliability and safety of CR-3.

The recommendations are amplified in more detail in the par-agraphs below.

1. System Response to Plant Upsets at Power The CR-3 Nuclear Power Plant, and other reactors de-signed by Babcock & Wilcox, prior to the accident at i

Three Mile Island Unit 2 in March of 1979 had a high pressure reactor trip setpoint of 2355 psig and a PORV setpoint of 2255 psig. The resultant reactor trip fre-quency of plants designed by B&W with that design was less than the trip frequency for other pressurized water reactor designs for each of the prior 3 years of i

operation. Subsequent to the accident at TMI, these setpoints were changed to 2450 psig for the PORV and 2300 psig for the high pressure reactor trip. As re-ported in NUREG-0667, " Transient Response of Babcock &

4 Wilcox Designed Reactors," the trip frequency of reac-tors designed by Babcock & Wilcox has approximately doubled. The Task Force has evaluated this data and believes that the prime recommendations of the Task Force in the area of system response are to return the PORV and High Pressure Trip Setpoints to their original values, provide automatic isolation of the PORV block valve, and eliminate the anticipatory reactor trip on turbine trip. The Task Force believes that such a change would provide a significant improvement in plant safety by reducing the severity of many transients. In addition to the improvements discussed above, the Task Force also believes that other improvements in plant reliability and safety can be made with longer term investigations. A summary of some of these recommenda-tions is provided below:

. Recommendation: Identify and implement discrete changes in the feedwater system to improve relia-bility and reduce safety system challenges.

. Recommendation: Investigate improvements that can be made to the hotwell controller to provide bet-ter control over the 0 to 100% full power range.

. Recocinendation: Review the Turbine Control System for potential reliability improvements. '

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. other recommendetions associated with the iCS which have the potential for improving system re-l sponse are discussed more fully in ICS/NNI Control section below.

In summary, the Task Force believes that a significant

, cause of recent plant upsets it CR-3 resulted from the l

changes in setpoints as a post-TMI-2 action. The Task Force believes that these setpoints :hould be returned to their original values, the anticipatory reactor trip i

on turbine trip eliminated and that an automatic isola-tion signal be provided to the PORV block valve to in-crease PORV isolation reliaullity. The Task Force also believes that additional improvements could result fron the other long-term recommendations identified above.

2. ICS and NN! Control A major area of investigation by the Nuclear Safety Task Force dealt with the nomal, transient, and upset functioning of the Integrated Control System (ICS) and the Nonnuclear Instrumentation System (NNI). While l

these are not safety systems and they were not designed to safety-grade criteria for redundancy and resistance to failure, the Task Force review has shown that there are improvements that can be made M these systems to reduce the consequences of their failure and to improve their nomal functioning.

The recommendations for improvements are in three broad areas. First, several recommendations improve the man-ner in which the ICS and NNI function during nomal plant operation. A second area includes several recon-mendations to improve the perfomance of these systems during transients in the plant. The last category in-cludes recommendations which make these systems more

" robust" and less prone to internal failures or which reduce the plant response to failures in these systems which do occur.

3. Pressurizer Level l The design of the CR-3 Nuclear Steam Supply System in-cluding pressurizer is such that during a nomal reac-tor trip, the level in the pressurizer will decrease to 30 inches from its nomal operating value of 200 inch-es. However, (1) the failure of relief valves on the l secondary system to rescat within their blowdown range, l

i (2) a slight overfeeding of the steam generators after a reactor trip, or (3) rapid ' refilling of the steam generators by emergency feedwater to their control set-O point (particularly the 507. level on the operate range l

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l O required for the natural circulation) could result in cooling the RCS down an additional 5'F, enough to cause pressurizer level to go off scale on the low end.

Known loss of pressurizer level has occurred 5 times at CR-3 and 23 times at all B&W designed plants. While loss of pressurizer level indication is not in itself a safety issue, since 75 effective inches of H O 2remain in the pressurizer at zero level indication, it is cer-tainly an operational blind spot in that a valuable pa-3 rameter might be off-scale for up to a minute during a transient.

[ The Task Force has investigated a number of recomenda-i tions relating to maintaining pressurizer 9evel on scale. These recomendations have been broken down i into short-term and long-term recommendations. Listed

] below are those recommendations and a brief discussion I associated with each:

Recomendations for Implementation as Soon as possible

. Recommendation: Adjust secondary steam relief valve blowdown to less than 5%.

{ Experience has indicated that blowdown of secondary re-lief valves has been as much as 8-9%. The adjustment

< of the blowdown setting to less than 5% will result in between 20 and 25 inches greater level in the pressur-izer during a typical reactor trip. This change also has the advantage of reducing reactor coolant system 1

pressure changes subsequent to a reactor trip.

. Recommendation: Add control function for positive 4 main feedwater reduction upon reactor trip.

. Reco,mendation: Automatically isolate main stean i to the moisture separator reheaters to minimize the potential for an overcooling transient.

Recommendations for Further Evaluation and/or Long-Tern Implementation:

The Task Force recomends performing further evalua-tions for potential long-term implementation of the items listed below. The Task Force does not believe

, that immediate implementation or resolution is required since the loss of pressurizer level is not in itself a significant safety issue.

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. Recomendation: Change pressurizer level compen-sation from the use of a temperature element to a pressure sensor.

The pressure sensor is a faster responding instrument.

If the change is deemed feasible, the existing error of i

as much as 20 inches of level could be significantly reduced. Since the error is in the directon of showing

' lower levels than actually exist, level would be less likely to go off-scale with this change.

. Recommendation: Review the EFW flow control sys-tem to ensure reliability and reduce systera response.

In sumary, it is the view of the Task Force that the implementation of the recomendations will reduce the

p. abability that pressurizer level indication will be lost at CR-3.

4. Emergency Feedwater and Shutdown Control Systems The Task Force recomendations associated with Emer-gency Feedwater and Shutdown Control fall into two cat-egories.These are recommendations which (1) reduce sys-l tem response following a reactor trip, and (2) reduce the probability and consequences of severe transients.

These recomendations fall into the general areas of

) improving EFW control to avoid overcooling, assuring EFW is actuated when required and improving steam gen-

, erator rupture matrix response. The Task Force identi-fied several additional items for implementation or further evaluation. Some of the important ones are:

. Recommendation: Modify the Emergency Feed Pump auto start circuit so that any power failures will not prevent actuation on low steam generator level.

. Recomendation: Review the Steam Generator Rup-ture Matrix Design.

. Recomendation: Provide main feedwater overfill protection circuitry.

. Recomendation: Review the EFW flow control sys-tem for modifications to ensure reliability and reduced system response.

. Recomendation: Install a self-contained, diesel-O driven Emergency Feedwater Pump.

. Recommendation: Install an automatic start signal to Emergency Feedwater and High Pressure Injection on a prolonged high RCS pressure signal.

In summary, the Task Force believes these recommenda-tions, when implemented, will improve the reliability l

and operation of post-trip control systems and thus contribute to improved plant safety.

1 E. HUMAN ENGINEERING AND POST TRIP ANALYSIS

One of the important lessons learned from the TMI-2 accident i was the need to improve the human engineering aspects of

, nuclear power plant control rooms. This lesson was rein-forced by the CR-3 transient and many of the Task Force recommendations are for changes and investigations leading to modifications to improve the man-machine interface so that the Operator can better understand what is happening in the plant and better interact with it. Some of the impor-j tant recommendations in this area are:

. Recommendation: Identify all branch circuits on elec-

trical distribution panels so the operator can readily q identify the equipment fed from each circuit.

. Reconmendation: Review the annunciator system for pos-sible improvements.

. Recommendation: Investigate use of zone of control re-sponsibility on the main control boards.

. Recommendation: Relocate the emergency FW bypass valve controllers closer to the OTSG level displays.

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. Recommenda tion: Perfona a general human factors review l of the control board layout.

. Recommendation: Investigate automating the immediate action steps of emergency procedures.

Another of the lessons learned in the TMI-2 accident was the need to improve the collection and recording of data during transients to facilitate subsequent review of the event.

Both during the TMI-2 accident and the CR-3 transient, prob-lems were encountered with the events recorders missing events that occurred. The review of the TMI-2 accident was greatly facilitated by the availability of data recorded by the B&W reactimeter. The review of the CR-3 transient was

, . somewhat more difficult in that only the main Control Roon strip chart records, the values of parameters that went into

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O alam, and operator recollection were available. Task Force recommendations provide for modifications in this area to improve collecting information for the post-trip analysis.

F. OTHER IMPROVEMENTS The Task Force has identified a number of recommendations that do not fall into one of the above categories. These

. recommendations typically fall into the categories of (1) reducing system response following reactor trip, (2) reduc-

ing the probability of and consequences of severe transi-ents, and (3) preventing other potential operational or safety problems related to support and safety systems.

. Recommendation: Check all 820 Signal Monitors for seal-in problems.

4 . Reconnendation: Examine atmospheric dump valve and turbine bypass valve circuits for potential improve-ments. Install manual closure of the ADVs from the Control Room.

. Recommendation: Reconsider containnent isolation modi-fications for multiphase isolation system.

. Recommendation: Evaluate the instrument air system for potential improvements.

. Recommendation: Provide steam line radiation monitors for quick identification of the affected steam genera-tor during a steam generator tube rupture accident.

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. Reconmendation: Addition of makeup pump protection circui try to ensure adequate suction pressure available.

. Recommendation: Upgrade of remote shutdown panel.

. Recommendation: Modification of MSIV system to prevent water hamer.

VIII.

SUMMARY

OF SHORT-TERM RECOMMENDATIONS, INCLUDING TRAINING The purpose of the Task Force was to identify items and actions which, when implemented, would lead to increasing the overall sa fety of CR-3 during both normal operations and transient events. The Task Force developed a philosophy for the improve-ment of sa fety. The Task Force thought that significant safety improvements can be achieved by the addition of systems, con-trols, interlocks and- training which serve to reduce transient Q frequency or mitigate transient severity. Provision of major

[~') equipment aMftions to mitigate consequences of core damage acci-dents was nat the primary concern of the Task Force.

The imediate result of the Task Force activit%s was to recom-mend approximately 50 corrective actions to FPC management for implementation by the plant staff and Nuclear Engineering. Each one of these items was recommended because it was thought to pro-vide an incremental measure of increased safety. Some of the items were identified as a direct result of the February 26, 1980, transient, while the remainder were identified during a General Plant Review.

The items generally fall into one of five categories:

A. Maintenance or component inspection as a direct resul t of the February 26th transient B. Upgrading the EFW C. Improving power supplies D. Improving reocedures E. Improving training Each item is di;;ussed in the text of this report. A sumary of the actions in the five categories is given here.

A. MAINTENANCE AND INSPECTION The steam generators, PORV, pressurizer and CRDMs have been inspected and analyzed for damage resulting from the Febru-a > y 26, 1980, transient. No indications of any damage were found. The NNI and the Events Recorder System were diagnos-tica'ly tested to determine the cause of their failures dur-ing the transient.

B. UPGRADING THE EFW The reliability of the EFW was improved by establishing a passive FW path to the OTSGs and modifying the auto start circuit to provide more reliability. Also, the blowdown on the main stean relief valves was lowered to reduce overcool-ing of the RCS after a normal trip.

C. IMPROVING POWER SUPPLIES Many recommendations were made to improve the operability of the plant through more reliable and diversified power sup-plies. Failure of control power and motive power to the NNI, ICS, PORV, ADVs, EFW, and MFW were examined for their ef fects on the plant. Recomendations were made to provide selectable sources of indication and control and to elimi-nate some single failures which can cause power loss to NNI o

G O or ICS. Operator training will be implemented to enable the Operator to more effectively respond to power supply fail-ures in the future.

D. IMPROVING PROCEDURES Many procedures were revised to provide the operator addi-1 tional information during overcooling transients, undercool-ing~ transients, and LOCAs. They also provide him with guid-ance to handle HPI cooling modes and solid RCS modes.

E. TRAINING Independent reviews of the CR-3 transient of February 26, 1980, made by several organizations (FPC, NRC, B&W, INP0, NSAC, etc.) have shown that proper Operator response to the transient mitigated the potential consequences that could have resulted. It is evident that greatly increased Opera-tor training, resulting from TMI-2 Lessons Learned, yielded 1

substantial benef:ts in correct operator response, and re-sulted in safe operations to cold shutdown and no damaged 4

equipnert.

Prior to the CR-3 transient, plans had been developed and approved, and implementation had begun to increase the staf-fing of the Training Department by approximately twofold.

The reporting level of the Training Manager was also elevat-ed to the Plant Manager.

The specific recommendations regarding training activities resulting from the work of the Task Force are as follows:

. Train all Operators and I&C Technicians in res)onse to NNI and ICS failures.

. Train all Operators on CR-3 transient sequence of events , concentrating on ICS response to failed NNI, and how lessons learned actions and equipment changes from TMI-2 affected the transient.

. Train all Operators on equipr'ent and procedural changes

, that are presently being made as a result of the CR-3 transient. Two major examples of this are training on Emergency Operating Procedures and Small Break Guidelines.

. Improved Operator training was perfonned and Emergency Operating . Procedures were modified to include additional Operator guidance.

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bd Future training activities should include the following recommended items:

. I&C Technicians' work practices and their potential impact on plant safety should be reviewed in plant '

training sessions.

. Expand plant training manual to include all plant systems.

. Provide additional periodic on-site training for Technicians in the area of overall plant systen knowledge.

. Provide additional on-site and offsite vendor training for Technicians. A formal vendor t raining plan is being developed and will be followed.

IX. LONG-TERM RECOMMENDATIONS SCOPE Af.D FUTURE WORK In addition to the short-term recomendations that will be imple-mented prior to plant startup, the Task Force also identified many other potential items of concern. These items were reviewed by the Task Force and were considered not immediately necessary for the continued safe operation of the facility and were, there-fore, classified into two categories of long-term recomenda-tions. These categorier 2re those that should be accomplished "as soon as possible," t,uu not a restraint to restart, and those that were to be "further_ analyzed" and, if determined feasible and desirable, would be prioritized and implemented at some fu-ture time, after restart of CR-3. These long-term recomenda-tions have been grouped into areas of improvement and are listed below:

. Upgrade and Improvements for Emergency Feedwater Improvements in this area are being considered to upgrade this system to safety-grade, or equivalent, with additional modifications to improve indication, control, and overall system reliability. There are also recomendations cor.cern-ing rupture matrix improvements and for ESFAS control of Emergency Feedwater.

. Improvements in Procedures The majority of. procedure revision is being conducted in the short-term but some long-term procedure revision is underway in the areas of drawing control, engineering field changes, and Small Break Guidelines. A review is also being conduc-ted on emer>;ency and abnormal operating procedures to im-Q-

prove the content and format utilization.

for better Operator

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( . Improvements in Training Programs and Manuals All Operator training related to the 2/26/80 transient and resulting modifications is being conducted in the short-term. The long-term recommendations in this area are to provide additional on-site and offsite training for Techni-cians. Also to update and expand the Plant Training Manual to include all plant systems and equipment.

. Human Factors Improvements for Operators

, Improvements in this area are intended to improve Operator recognition of plant conditions and to provide controls which are more reliable and conveniently located for ease of response to plant transients. Some of the areas being con-sidered are control board identification and color coding, annunciator system improvements, and power supply identifi-cation. Improvements are also being considered on control board indicators and the events recorder system. Procedures and drawings are also being improved to aid Operators in re-cognizing and controlling plant transients.

. Controls and Instrumentation Improvements This is the largest area where improvements are being con-sidered and should produce many changes which will result in improved plant control and greater system reliability. Much of the investigation going on in this area is to improve controls and control responses to system power failures.

Also under review are the failures of interlocks, alarm functions, 6nd indication utilized by the Operator. There are also recomnandations being considered for the addition of more instrumer.tation, as well as improving the existing instrumentation and controls. Consideration is being given to the need for additional anticipatory trips and to auto-mating some of the Operator's ranual immediate actions dur-ing plant transients.

. Upgrade and impravements for Main Feedwater Improvements in this area are concerned with items which affect the response and sensitivity of the main feedw3ter system. Changes are also being considered for main FW Control during runbacks to prevent main FW overfill and to improve overall systen reliability.

. Balance of Plant Items Other improvements considered by the Task Force are main

{) steam isolation valve control, waste gas system, Control

f uJom emergency lighting, decay heat, and high pressure in-jection systems. Improvements in this area would be to im-prove overall operation and reliability.

X. CONCLUSIONS In fulfilling its charge, the Task Force had a uniqae opportunity to review many areas of plant desig", and operation which affect the operability and safety of the plant. The considerable amount of expertise assembled, combined with the experience gained in operating the plant, provided an e<cellent overview of the inte-grated performance of the plant. Actual or potential problems were reviewed objectively and corrective actions were formulated to improve plant safety and reliability. In formulating correc-tive actions, care was taken to minimize adverse effects to other systems and operations.

The long-term recommendations outlined in this report should pro-vide a good starting point for future improvements to the plant.

The short-term recommendations presented. when impleme ,ed, will previde a measurably increased level of safety.

It is the concensus of this Task Force that when the short-term changes outlined in this report are implemented, the Crystal River Nuclear Plant would add more margin to its already adequate safety margin for operation without adversely affecting the health and safety of the public or Florida Power Corporation empl oyees.

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LIST OF ITEMS PECOMMENDED TO BE COMPLETED BEFORE STARTUP

1. Thorough testing of the NNI-(X) System to detemine cause of initial failure.

2. Establish procedural controls of selectable sources for in-dication and control.

3. . Initiate a more extensive surveillance program on the Events Recorder System.

4. Develop and perfom a functional test for rendant instru-mentation modifications.

5. Visually inspect the lower portions of the steam generator support skirts and anchor bolts. Remove any corrosive resi-due observed.

6. Inspect the Pressurizer heater bundles for seal leakage.

Electrically check pressurizer heater elements for continuity.

7. The CRDMs will be checked for proper insulation resistance prior to their return to service.

8. Pressurizer:

a. Perfom a visual inspection of the pressurizer relief system (i.e. , PORV, both code safety valves, and the discharge piping). Inspection of the discharge piping system, including hangers, should be perfomed to en-sve that no gross distortions have occurred.

b. Confirm by calculation that structural loads on valves and pressurizer, as a result of the extended period of discharge to the quench tank, are acceptable.

c. Disassembly, ins pection, and refurbishment (as neces-sary) of the PORV and the Code Soreties.

9. Review procedures covering when HPI flow can be cut back and secured during a small break or overcooling transient.

10. Review procedures covering OTSG tube rupture in accordance with revised B&W tube rupture guidelines and Small Break ,

Guidelines. '

11. Review procedures concerning proper OTSG tevel at HPI and manual RCP. Trip in accordance with revised B&W Small Break c Guidelines.

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O 12. Review procedures concerning nen to initiate nel cooiing upon total loss of secondary heat . removal capability. l

13. - Provide procedures and training for recovery from EFW actua-i tion to avoid 0TSG overfill.

14.r Establish minimum conditions for voluntarily entering de- I graded modes of operation.

15. Revise procedures to require ICS high reactor demand limit reset upon RPS reset.

16. Train all operators and I&C technicians in response to NNI and ICS failures.

17. FPC shall train Operators on CR-3 sequence of events , con-

, centrating on ICS response to failed NNI, on how lessons

learned from TMI-2 affected transient, and on plant changes that are being made as a result of the CR-3 event.

18. Repair Events Recorder System.

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- 19. Review the present design of computer alarm printout to help i eliminate overload and printout time delay.

20. Adjust secondary steam relief valves blowdown settings.

21. Provide around-the-clock I&C Technician coverage.

I 22. Perform corrective action regarding potential shorting in safety system from fiber clamps, if installed improperly.

23. Provide temporary backup air system fur main feedwater startup control valves.

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! 24. Review and revise procedures for immediate Operat'or action to trip DH pumps upon spurious closing of the DH dropline valves.

25. FPC management shall perfona an evaluation of the number of exempt personnel that should hold operators licenses. i

26. Establish administrative controls to minimize access to con-

, tainment in Mode 1.

27. Include in operator training and plant procedures methods of isolating letdown in the event of loss of ICS or NNI power supplies.

28.. Provide capability to facilitate operator action in event of

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29. Visually inspect EFW nozzle collars and repair as required.

30. Review restart of critical items on loss of offsite power without ESFAS actuation and revise applicable procedures.

31. Field changes to NNI/ICS systems should be performed in ac-cordance with design control requirements.

32. All field changes should include specific reference (s) to

, installation and maintenance precautions identified by the equipment supplier.

33. Check 820 Signal Monitor output for s tal-in problems.

34. Provide Operator with redundant indications of main plant parameters.

3S. Provide Operator control of atmospheric dunp valves upon loss of ICS power.

36. Provide automatic transfer switches for NN!-(X) and ICS AC power, normally on a vital bus with auto transfer to regula-ted instrument bus. Dual 24 VDC supplies will be powered from a vital bus and regulated instrument bus.

37. Provide subcooling monitors with reliable backup. One noni-tor thall be operable on loss of either Inverter A, B, C, or D or offsite power.

38. Provide positive indication of position of all three pres-surizer relief valves.

I 39. Modify vital bus panels for quick fuse replacement.

40. Modify EFW Pump Autostart Circuit and Reactor Trip Circuit so that any power failure will not prevent activation on low OTSG level (Control Grade).

41. PORV position indicat* ng lights for solenoid will be added.

42. A fatigue analysis will be performed on the pressurizer heaters to demonstrate that the 40-year design life was not adversely impacted by this transient.

43. The relief valve loadings on the pressurizer relief nozzles should be determined and their effect will be assessed.

44. Provide diverse containment isolation.

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45. Electrically interlock motor-driven EFW pump to start on loss of MFW.

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1 O 46. Review PORY circuitry to assure that credible power failures do not cause the PORV to open when it is not required to open.

Review power supply independence between PORV and PORV iso-lation block valve to assure that failure which would affect the PORY does not eliminate the possibility of PORY isola-tion block valve action.

47. Modify pressurizer spray valve control circuit so that NNI power failure will close valve.

48. Move 120 VAC ICS-(X) power source to vital bus.

49. Individually annunciate loss of AC power to NNI-(X) and (Y) buses and ICS.

50. Install indicating lights on all vital bus feeds.

51. Evaluate the OTSG Rupture Matrix with the intent to remove the signal fron FWV-161 and FWV-162 to assure a passive EFW flow path to both OTSGs on initiation of EFW.

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NUCLEAR SAFETY REVIEW TASK FORCE 1

i LIST OF CONCERNS I

i INDEX i 1.0 Procedures / Test l 2.0 Training  !

) 3.0 ICS/NNI l

4.0 Events Recorder / Data Collection i

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5.0 Not~ Classified

.f 6.0 Power Supplies 7.0 Tedesco Report on CR-3 Recommendations

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i PRIORITIES

1 Action to be accomplished prior to restart of CR-3 IB Action to be accomplished as soon as possible .

l 3 Continue to analyze. Implementation, if feasible, would be some time after restart of CR-3 ASSIGNMENT TO REPORT-- Persons Initials i

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A V U LIST OF (DNCERNS 1.0 Pf0CEDURES/ TEST Page I CONCERNS REFERENCE PR1(RITY, ASSIGemENT, RECCM4Eli)AT191 1

1.1 Thorough testing of the NNI-(X) sys- FPC Cbrrective Action List, Rec. I. l/KOV This activity has been completed.

tem to determine cause of Initial fallure.

l 1.2 Establish procedural controls for FPC Corrective Action List, item 5. 1/PFM Procedures will be developed and implemented tefore pleat selection of sources of Indication IPP0/NSAC Crystal River Task Force, startup.

and control. Rec. I I. A, and I l l. A.2.

B&W (R-3 Translent Assessment Re-port , Rec. .

1.3 initiate a more extensive periodic FPC Corrective Action List, Rec. 9. 1/PFM SP-505, " Operability end Functional Check of the Evoets

> test program on Events Recorder and Recorder Annunciator Sys+em" was developed to wookly test 4 Annunciator System. the system.

1.4 Develop and perform a functional FPC Correc+1ve Action List, item 12. 1/KOV Pr erequisites, Test Methods, and Acceptence Criterle have test for redundant Ins trumentation IPP0/NSAC &ystal River Task Force, been defined. Test Procedures will be written and laple-mod i f ica tions. Rec. Ill.A. . mantud bef ore plant startup.

1.5 Visually Inspect and clean the steam FPC Corrective Action List, Rec. 17 1/JDM Will be completed bef ore plant startup.

generator support skirts and anchor bolts.

1.6 Inspect the pressurizer heater bun- FPC Correc+1ve Action List, Rec. 18. 1/JD4 Will be completed before plant startup.

, dies for seal leakage.- Electrically check pressurizer heater elements for cont Inul ty.

1.7' The CROMs will be checked f or proper FPC Corrective Action List, Rec. 21. 1/JDM This activity has been conpleted.

Insulation resistance prior to re-turn to service.

Ch O u LIST OF CDtCERNS 1.0 PROCED@ES/ TEST Page 2 CONCERNS REFERENCE mimlTY, ASSIGMENT, REC 0peEM)ATION 1.8 Pressurlzer FPC Corrective Action List, Rec. 22. a . 1/JDM Will be completed before plant startup.

a. Vi stmil l y inspect the pres-surizer relief system. b. 1/EJA Will be empleted before plant startup.

b. Confirm by calculations that structural loads on valves, pressurizer and RC Drain Tank are acceptable.

c 1/JDM Will be cepleted before plant startup.

c. Di sa ssemble, Inspect and re-f urbi sh (as nece?sary) the PORV and Code Safetles.

1.9 Rowlew procedures covering when HP8 IPP0/NSAC Crystal River Task Force. l/ERK A recmmendation has been nede to B&W that small break g flow can be cut back and secured Rec. II.E. guidelines be revised. Upon issunnee of revised guide-4 during a small break or an overcool- Crystal River 3 Nuclear Safety Task lines by B&W Plant Procedures will be revised.

Ing transient. Force.

1.10 Field changes to NN 1/ICS Systems' B&W CR-3 Transient Assessment Re- 1/ REC This will te done. Engineering Procedures require this should be performed in accordance port, Rec. 12.

of all design changes.

with design control requiroments.

1.11 All field changes should Include ' B&W CR-3 Transient Assessment Re- I for implementation / REC specific reference (s) to Installa- port, Rec. 13 tion and maintenance precautions 3 for procedure changes / REC Identifled by the equipment Engineering Procedures will be reviewed and revised If supplier.

necessary.

1.12 Review EPs & APs for common lan- CR-3 Nuclear Safety Task Force 18/ DAM All EPs and APs will be revlowed to determine procedure g uage. Do EPs and APs stand on I&E TM I-2 Investigating Team, adoquacy and to ensure cmmon language.

their own? Institute a special In- . Rec. C.3.d.

s pection program to determine whe-ther procedures are adequate.

LIST nr rm: ERNS 1.0 PROCEDURES / TEST pogo 3 COPCERNS RFFEREPCE FR KR ITY, ASSIGMENT, REC 0bOENDATION n

1.13 Investigate making administrative CR-3 Nuclear Safety Task Force. IB/GDQ A recommendation has been unde for a system of color cod-requirements for power supoly iden- Ing Control Room Instruments, and Identifying poeor sup-tification and color coding schemes plies.

on Control Board.

3 Essex portion.

1.14 At another Nuclear faci l ity, the CR-3 Nuclear Safety Task Force. 1/ERK Several recommendations were nede concerning:

ECCS requirement to maintain 50*F subcooling Is in conflict with OTSG 1. Changes in subcooling requirassents for . ttrottling tube ruptures mitigation. The gulde- WI .

> ellnes require that HPl be initiated e and maintained any time hot leg con- 2. Changes in steam generator tube rupture guldelines.

ditions fall below 50*F. Subcooling the tube rupture mitigation sequence requires that the RCS be depressur- IB/ERK 1. Radlation monitors on steam lines.

Ized below the secondary pressure, i f the tube rupture (leak) Is of the 2. Analyze ATOG for f urther guidance.

"right" size, 50*F subcooling cannot be aehleved, the pressurizer will go solid and cont in ual feed of the steam generator (through the tube) w!!! occur. Does this apply to CR-37

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LIST OF CD TERNS 1.0 PROCEDURES / TEST page 4 COTERNS REFERECE PHIORITY, ASSICNMENT, REC 0ieENDATION 1.15 At another Nuclear facil i ty on a CR-3 Nuclear Safety Task Force. IB/CJ8 A recommendation has been mode tot small steam !!ne break with an B&W Recommendation ESFAS, the Operator must trip to 1. Establish guidelines in procedures so that the Opere-RCPs and rai se the OTSG level to for can determine if the E*FAS is due it, overcooling 955. This is counterproductive in (steam line breek) or a LOCA.

that:

2. Change steam line break procedures so that the of foc-

a. Filling the OTSGs will mask ted OTSG i s bolled dry tfrough the atmospheric dump which generator has the leak. valves or ttrough the M51Y to the condenser instead of through the break.

b. After determining af facted gen-erator, the Operator must boll 3. Change the small steam break procedure b include a

> away all the extra inventory I that he just added.

reference to actions in the event of ESFAS actuotlon.

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c. Filling to 95% with the main CR-3 Nuclear Safety Task Force. IB/ DAM feed pumps also increases the odds of overcooling the RCS.

Does this apply to CR-37 1.16 At another Nuclear Facility, their ' B&W Recommendation procedure requires operator to con-1/CJB A recommendation has been nede to B&W that Small Brook Guidelines be revised. Upon issuance of revised Gulde-trol OTSG level at 50% to 95% opera- lines by B&w plant, procedures will be revised, ting range following RCP manual trip after a low pressure ESFAS. Is 5C% '

adequate to sa t i sf y Small Breaks Guidelines? Does this apply to CR-37

O V (vl LIST OF CDPCERNS 1.0 PRCCEDURES/ TEST

• Page 5 COPCERNS REFEREtCE PRimlTY, ASSIGNENT, RECOBSENDATION 1.17 At Another Nuclear Facility refer- CR-3 Nuclear Safety Task Force. 1/CJB ERK A recmmendation has been made that operator giside-encing the Small Break Guldelines; lines be establi shed as to what steam generator level and the port ion calling for opening the pressure can be before HPl cooling is initiated. IAstil PORY and starting HPl as mon as a that time operator should be Interested in starting EFW.

total LOFW i s gerceived (i.e., wi th-out any other criteria), seems like an Invitation for a LOCA d ur Ing events In which EFW is delayed but not lost. .Another facility's data shows that EFW can be delayed a min-ute or two and can be recovered.

Premature Initiation of the steps In

[3= the procedure would cause a signIfI-a cent degradation of plant conditions unnecessari l y. Perhaps addition of subcooled margin criterion should be considered. Does this apply to CR-37 1.18 At another Nuclear facility, low ' CR-3 Nuclear Safety Task Force. 1/GJB Th i s i tem does not appl y = > CR-3. P4) f urther action re-pressure ESFAS Isolates injection. q ui r ed.

Present procedures require the Oper-ator to tr ip the RCPs on low pres-sure ESFAS. This conbination almost guarantees RCP seal damage, unless the Operator immediately bypasses

+he ESFAS Signal and reinstater seal injection. Similarly, a 4 psig RB pressure ESFAS isolates wel injec-tion, Intermediate cooling water, and sea t return; thus, the pump mo-tors are running without any cooling water. Does this apply to CR-3?

%) J LIST OF (DNCERNS 1.0 PROCEDL.RES/ TEST Page 6 COPCERNS l RETERENCE FRIORITY, ASSIGMENT, REC 04(NDATl0N 1.19 Review small break LOCA/ Reactor IPP0/NSAC Crystal River Task Force, 3/ERK A recommendat ion has twm made that FPC Pursus one of coolant pump trip procedures, (con- Rec. II.D several possible approac;es for providing e?ometic reec-corn *or tripping pumps - Consider CR-3 Nuclear Safety Ta sk Force WH tor coolant pump trip.

coincident pump trip network) Report, Rec. II .7.

1.20 Review proced ures for. starting MN CR-3 Nuclear Safety Task Force. 1/KOV A recommendation has been made for a procedre for recov- ,

pumps after loss of feedwater tran- ery from energency feedvetor actuation. Additional re-sients to prevent overfilling OTSG, commendations were ande for revisions to ocisting emer-gency procedures.

1.21 Electrical Distribution Panels. The Operators do not know what equipment CR-3 Operator Recmmendations. 18/ REC A recommendation has been nede to provide controlled is fed from each branch circuit. Go drawings in each panel listing panel loeds.

M in and label or provide lists.

[n 1.22 Better definition of responsibil.?y CR-3 Operator Recommendations. 1/PFM Al-500, " Conduct of Operations." will be changed to more of Operators and Shif t Supervisor in clearl y speci f y the oper

• tr and shif t supervisor deal-dealing with the PRC during transi- Ings wi th t he PRC.

ents.

1.23 Drawings suppiled by B&W/BMCo do not CR-3 l&C Technician Recommendations. ' 18/ REC Production Engineering has pwchsed emplete sets of EMCo (a) reflect the letest design chang- L drawings on aperture cards and are having new mylars nede es, and (b) in general are poor up for f uture drawing updates. New drawings will be 1s-prints. Reconcile with St. Pete, sued upon empletion.

B&W, and BMCo.

1.24 All drawings should be updated when jCR-3 I&C Technician Recomendations. 18/KOV A recommendation was made for changes in plant procedures a rodification is Installed to on- f or draw ing control .

sure adequate troubleshooting.

(Drawings are not up-to-da^e bncause MARS are not quickly factored into drawings.)

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LIST OF CONCERNS 1.0 PFOCEDURES/ TEST Pege 7 CONCERNS REFERENCE PRlORITY, ASS 1GPMENT. RECOMENDATIGi 1.25 Set up a su veillance program which CR-3 Nuclear Safety Task force. 3/ DAM A recommendation has been mode for the plant staf f to has a good probability of detecting provide input to Engineering to develop a survollisace NNI/lCS f a ilures which would normal- program.

ly go undetected until needed.

1.26 . Check all 820 signal monitors for CR-3 Operator Recommendations. 1/KOV A reconrendation was made to perform an enelysis of the seal-in problems. ICS and NNI to identif y and fix potential problems.

1.27 Rewrite EP-104 OTSG tube f a ilure us- CR-3 Nuclear Safety Task Force 1/PFM This will be done.

. Ing the recomendations provided by B&W. Expand the of f-normal sequenc-es. Develop a section covering how y to reduce primary pressure if spray y is not available. 1 9 \

1.28 Establish administrative controls to CR-3 Nuclear Safety Task Force. 1/PfM Changes in the Containment entry procedure will be mode.

e!!minate acmss to Containment in Mode 1. (Industrial Saf ety Consid-era t ion) .

1.29 Improve #Ps & EPs with new pro - 3/IB/PFM A recommendation has been made to study this matter dures for following events: f urt her. Procedures wlil be developed as Indicated by this study.

a. ATWS

b. LOOP and Failure of I Diesel

c. LOOP and Fallure of 2 Diesel 1.30 (Deleted)

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LIST OF CONCERNS 1.0 PROCEDLRES/ TEST Page 8 COPCERNS RET ERENCE FRimlTY, ASSl@ MENT, REEOMNDATION 1.31 Determine whether mininium conditions ISE - Dil-2 Investigation To m, 1/PFM Guidelines were suggested. Al-500 will be changed.

need to be established before volun- Rec. C.3.a.

tari!y entering degraded modes of operation.

1.32 Determine whether specific addition- I&E -

TM I-2 Investigation Team, 1/ DAM A review was done end it was determined that no further al guldance con.orning the blocking Rec. C.3.b. action was required on this concern.

and/or bypa ssing of ESFAS systems needs to be grovided.

1.33 FPC should have a debriefing follow- l&E - TM I-2 Investigation Team, IN FM Thi s i s al read y being done. Pb f urther action required.

Ing each transient so that lessons Rec. C.5.a.

g can be . learned and record s pre-g served. Consider an admin i strati ve req ui rement.

1.34 Do procedures need to be upgraded or CR-3 Nuclear Safety Task Force 18/ DAM A s econmendation was made to study this concern.

precautions identified regarding availability of redundant systems for loss of decay heat system.

1.35 The ICS can cause rod withdrawal lCR-3 Nuclear Safety Task Force 1/GDQ A recommendation was made to Institute procedure changes during certa in events.

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to require the ICS high reactor demand limit to be re-duced when the reactor will be operated for extended per-lods at less than 100%.

f 0 x v LIST OF CONCERNS 2.0 TRAINIC ringe 9 CONCERNS REFERENCE PRIORITY, ASSIGPMENT, RECO*DOATlG4 2.1 Train all Operators and l&C Techn!- FPC Otrrective Action List, Rec. 6. 1/DM A reconmendation has boon mede that the plant training clans In response to NNI and ICS IPP0/NSAC Crystal River Task Force, department work with Engineering to develop a course to f ailures, including Operator actions Rec. 1.0, l l .C. , a nd l i . A. 7. be presented to all licensed personnel and techniciens.

to close valves, etc., that may B&W 01-3 Transient Assessment Re-cause overftll, etc. Oevelop plant port, Rec. 5.

procedures for Operator response.

2.2 Operator training needs to be CR-3 Nuclear Saf ety Task Force. 1/DM A recorpendation has been made that the plant training r evi ewed. department work with Engineering to develop a course to be presented to all licensed operators.

1. Train ali Operators on CR-3 so-quence of events, concentrating 3 on ICS res ponse to f a i led NN I 7 and how isssons learned from

• TMI-2 af facted translent.

2. AlI plant changes that are being made as a result of the CR-3 event.

2.3 Instrument Technician work practices itP0/NSAC Crystal River Task Force, 3/DM A reconnendation has been made to f urther develop traln-and their potential Impact on plant Rec. 1.C Ing sessions for technicians to review:

safety should be reviewed in plant tral.ilng sessions. i 1. Work practices and their impact on plant safety.

2. Operator / technician interface to discuss potential events and their of f act on plant control.

3. Manual plant operations to minimize offects of maintenance on plant control.

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LIST & CDEERNS l

2.0 TRAINING

% 10 COCERNS REFEREEE fHl0RITY, ASSIGe8ENT, RfE0mEMMT10N 2.4 Expand Training Manual to incl ude CR-3 1&C Technician Recommendations. IB/KOV A recommendation has been ande to the Plant Training De-all plant systems and keep up-to-partment regarding molntenance of Training Manuels and date.

materials to be included.

2.5 Provide periodic on-site traln ing CR-3 1&C Technician Recommendations. 3pFM Plant Staff is in the Irocess of upgrading trefning for for Technicians In the area of over- all technical di sc ipl ines (18C, Electr ical , Mechenical all system knowledge. and ChenRad.)

2.6 Provide of f site vendor training for ,CR-3 I&C Techniclan Recommendations. 3/PFM A formal training plan, including vendor tretnIng, is be-Technicians or have Bailey come on- Ing developed and will be followed.

site to cond uct training (Balley, Diamond, Fisner, etc.)

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LIST OF CONCERNS i

3.0 ICS/NNI Page 11 CONCERNS REFERENCE PRimlTY, ASSIGNMENT, RECateqEM)ATICpl '

- 3.1 Provide Operator with redundant FPC Corrective Action List, Rec. 10. 1/ REC This will be done prior to plant startup.

Indications of main plant parame- IPP0/NSAC Crystal River Task Force, ters. Review power sources for aux- Rec. I i 1. A. 2 a nd i 1 1. A. 7.

Illary relays to consider redundant D&W m 's Transient Assessment Re-power or have (X) parameter on (X) port, Rec. I and 6.

I nd icator NUREG-0560, Rec. 17.

MPR Report, Rec. Ii1.1.

3.2 Examination of circuitry to ADV and itP0/NSAC Crystal River Task Force 18/TCR Recomendations were made to satisfy this concern.

TBV should be made to ensure that a Report, Rec. Ill.A.6.

f allure will not prevent their clo- B&W m-3 Translent Assessment Re-sure or as a minimun closure in the port, Rec. 4 3 steam generator that is to be used ICS FMEA Report, Rec. 6.

N for cooling. Provide manual closure 1/TCR Circultry will be added to allow closure of ADVs from

[ of ADY f rom Control Room. Control Room.

3.4 Confirm the capability to Isolate IPP0/NSAC Crystal River Task Force 1/EJA A recommendation has been made to include in operator letdown on loss of NNI or ICS power Report, Rec. 1 I I. A.6. training sessions and by procedures various methods of s upp l i es . B&W m-3 Transient Assessment Re- Isolating letdown In the event -f loss of ICS or NNI port, Rec. 7. power suppilos.

3.4 Examine ICS 22% power interlock to CR-3 Nuclear Safety Task Force. 1/TCR A recomendation has been made that a plant see whether LOOP will prevent start- modificationbe Installed to eliminate this problem.

Ing of RC pumps upon restoration of power.

3.5 a. Review NNI/lCS systems and asso-- lPP0/NSAC Crystal River Task Force 1/ REC H>dif ication w il l provide ABT's for NNI-(X) and ICS AC cl ated A.C. source and D.C. sup- Report , Rec. l i l . A. I . normally on a Vital Bus with auto transfer to Regulated ply system Fault Protection B&W CR-3 Transient Assessment Re- Instrument Bus. Dual 24 VDC supplies will be powered (1.E. ABT) make recorvnenda t ions port, Rec. I I. f rom a Vital Bus and Regulated Instrument Sus. See which will minimize fault propa- Oak Ridgo Report on ICS Rollability item 6.7.

gat ion (i.e., f us ing). Considor Analysis, Rec. 1.

separate ABT for (X). CR-3 thclear Saf ety Task Force.

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O O LIST OF CONCERNS 3.0 ICS/NNI Page 12 CONCERNS REFERENCE PRimlTY, ASS!GNMENT, RECO*ENDATION

b. Examine fusing (+)24,(-)24 V B&W Recommendation. 18/ REC A recommendation was made that an REl be generated for a '

from <ertical cabinet bus to MPR Report, Rec. 1.1. comprehensive review of this concern.

each rock of equipment in the CR-3 l&C Technician Recommendations.

ICS and NNI. (Would signif i-cant ly reduce the propagation of failures).

3.6 Rulew loss of ICS and NNI for dif- ItP0/NSAC Crystal River Task Force, IB/GDQ A number of recommendations were made regarding operating i

forent failures and their et f act on Rec. I l l . A. 4 and I l l . A 6. strategy and nodifications to ICS, NNI, control rod the component s oeing control led. Tekekron Report, Rec. 1. drive, emergency feedwater atmospheric dump valves, and Examples incluje: CR-3 Nuclear Safety Task Force, turbine bypass valves.

1. Decreas ine, FW flow while drawing 'B&W Recommendation.

3 . rod s.

7' 2. Mid-scale failures.

y 3. Examine modifica tion of ICS to, on loss of powers

a. " Fall" MFW pump demand to 0.

b. "Falla MFW block closed.

c. Fall low load valve closed.

d. Addition of new manual sta-tions on demand to TBVs, i ADVs, .and SUFW valves.

e. To start EFW ptsnps.

4 Rod withdrawal limi tat ions.

- 3. 7 The capability cf stopping main B&W m-3 Transient Assessment Re- IB/KOV A recommendation has been made that main FW pumps be feedwater to prevent excessive main port , Rec. 3. tripped and main and low load FW valves be closed on loss feedwater addition on loss of power ICS. FNEA Report, of ICS DC power or loss of NNI-(X) or NNI-(Y) AC or DC.

supplies should be reviewod.

i

R

' f)

N./

LIST OF CONCERNS 3.0 ICS/NNI Page 13 CONCERNS REFEFENCE PRI(RITY, ASSIGNMENT, RECOMENDATlOf 3.8 What is the best point to control MPR Report 1/3/ MEN Recommendation was made to review LOCA Procedure to on--

steam generator level at after a sure cperator has criterla to distinguish between I.0CA -

trip? and overcooling, and also to ensure proper guidelines for raising OTSG level. A reconenendation was also made to investigate an automatic mechanism for raising OTSG level f rom 50% to 95%.

3.9 Analyze 'a

/lCS for failures B&W Recommendation. 10/ REC A recommendation has been made that a study be mode of which wili p undetected until need- CR-3 Operators Recommendations. undetected failures. Modifications will be made to pro-ed. Annanciate as " trouble". Pro- vide positive Indication of loss of ICS or NNI power.

vide clear Indication that a control or indication has been lost.

3.10 investigate whether transferring CR-3 Nuclear Safety Task Force. 3/GDQ A reconsnendation has been made that a study be undertaken w Tay, control from Reactor control to to investigate the advantages and disadvantages of trans-feedwater control improves system ferring T ave control to feedwater.

• ransient respoase.

3.11 Review the reliability of input sig- ICS FMEA Report 3/GDQ Two suggestions were made as to possible ways to desensi-nals f rom the NI/RPS system, specif- B&W Recomendat ion, tirs the ICS to RC flow signal failures.

leally the RC f low signal.

3.12 Study whetrer single failures can be MPR Report 18/GOQ Ref er to item 3.6 for recomendations.

handled so they do not cause loss of both automatic and manual control of a single component.

3.13 Remove mixtures of power supplies IPP0/NSAC Crystal River Task Force, 3/ REC This will be done for redundant Indications. A where possible. The signal condi- Rec. lil.A.5. recomendation was made to identif y where cross-powering tioners should be powered from the MPR Report. exists and reevaluate the feasibility of making chrnges.

same souros as indicator and trans-m i tte r.

A U h d

LIST OF CONCERNS

3. ICS/NNI .Page 14 CONCERNS REFERENCE PRamlTY, ASSIGNMENT, REC 0804EWATim 3.14 Add control furection for positive B&W Recommendation. IB/GDQ A recormendation was made to develop an ICS modification MFW reduction upon Reactor trip. to accomplish this.

3.15 Consider additional ICS runbacks on B&W Recommendation. 3/ MEN A reconrnendation was made to evaluate several proposed ,

Secondary System upsets (Desereator runbacks and to install those deemed desirable.

le vel ) .

3.16 The condensate control is all (Y), CR-3 Operators Recomendations, m-3 3/TCR consider change. Safety Task Fe,rce.

3.17 Consider ICS change to independently B&W Recormendation. 1/CDQ An evaluation was made and it was reconssendej that no interlock TBVs to prevent single changes in the present design be made. No f urther action 3 f ailure; e.g., steam pressure trans- . required.

mitter from spuriously opening TBVs.

4 3.18 Tripping both +24 Volt power sup- NRC Recommendations (B&W/NRC Licen- 1/GDQ An evaluation was made and it was recomended that no plies on reduced voltage at either sing Telecon). changes in the present design be made. No further action bus must be reviewed for overall required.

Nuclear P'9te saf ety. Concerned about 22 Y .rlp when some boards might operate as low as 17 V. Can/

should the 22 Y trip be changed?

3.19 Make B&W Subcooling monitors redun- NRC Meet ing on CR-3 Tronsient. 1/KOV Will be completed before plant startup, da nt. One monitor shali be operable on loss of Inverters A, B, C, or D or of f si te power.

3.20 Single valves in NSCCCW supplying RB IREP on CR-3 Review. 3/TCR cooling units. Should redundant valves be installed powered / con-trol led f ron redundant channels?

v -

LIST OF CDCERNS 3.0 ICS/NNI Page 15 COPCERNS REF EREPCE FRICRiTY, ASSIGMENT, RECOMNOATION 3.21 in order to assure having means of B&W CR-3 Transient Asse ssment 18/KOV A recommendation was made to add redimdent letdown con- -

relieveing pressure dur ing loss of Report. trol capabili ty.

one channel of NNI, the PZR spray control and the letdown control should be powered / controlled from d iverse rnurces.

3.22 Concern item 1.2 recommends setting CR-3 Nuclear Safety Task Force. 18/GCQ A recommendation has been unde for specific changes in NNI signal select sw i tches to make the NNI and ICS to make this possible, the A Loop (X) and the B Loop (Y).

Certa in features of the present NNI arrangement prevent this and must be g changed.

M c1 3.23 Change pressur izer level componsa- CR-3 Nuclear Saf ety Task Force.

tion from temperature compensa tion 34JB A recommendation was nede to perform an eval u1 tion of to pressure empeuation. This con- this change.

cern is based on sevel error caused

• by Inherent time delays in the RTD sensors measuring prossur izer temperature.

3.24 Review Interlocks which can fall on CR-3 Nuclear Saf ety Task Force.

loss c' a power supply tw determine I f bypas:,es are needed to allow nu. IB/GJB A recommendation was nede that this review be done and mal shutdown or to accmmodate emer- that bypass capability and/or bypass procedures be pro-gency procodures. vided where necessary.

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O O LlST OF CONCERNS 4.0 EVENTS RECORDERA . COLLECTION Page 16 i

COTERNS PEFERECE PRICRITY, ASSICNMENT, REC 0h#ENDATION 4.1 .tepair Events Recorder System. ItPO/NSAC Crystal River Task Force 1/AEF This Item has been completed.

Report. Rec. lil.3.1.a.

FPC Corrective Action List, Rec. 8 4.2 Consider plant translent monitor INP0/NSAC Crystal River Task Force 3/ REC A MAR has been written to provide this.

as part of Tech. Support Center. Report, Rec. 111.B.2 4.3 a. Expand capabflity of events re- IPP0/NSAC Crystal River Task Force 18/KOV A recamredation has been made that the ennunclaior/ event

. corder. Report, Rec. Ill.B.1.b. recorder be upgraded to a state-of-the-art system.

CR-3 Nuclear Saf ety Task Force.

b. Review the present desi gn of i 1/AEF This item has been completed.

alarm printout to help eliminate g overload and printout time delay .

L m

(line printer).

4.4 Review recorder adequacy (speed of CR-3 Nuclear Safety Task Force. 3/ DAM A recommendation has been nede that the gresent Bailey chart recorders). Consider changing reconfers be replaced.

recorders.

_w - _ _ _

/s LIST OF CDtCERNS 5.0 NOT CLASSIFIED Pege 17 COPCERNS REFERENCE HIORITY, ASSICNMENT, RECOMPENDATION 5.1 Reverse PORY and high pressure trip Crystal River Nuclear Safety Ta sk - IB/GJD A rec;mmenostion was made that: 1. B&W proceef with work setpoints. Eliminate Reactor trip Force. Identlfled in task letter with an early report of recm-on turbine trip. mended hardwarO changes. 2. B&W should develop a basis for elimination o' the reactor trip on turbine trip.

5.2 Adjust secondary steam ret lef valves Crystal River Nuclear Safety Task 1/PFM, MJD This work will be completed before plant startup.

, . blowdown setting to less than 55. Force.

5.3 At another plant, on a LOOP with Crystal River Nuclear Safety Task I/TCR This item does not apply to Crystal River 3. Pts further subsequent failure of bot h diesels Force. action required.

to start, EFW is tmavailable, be-g cause the steam-driven EFW pump is L isolated from the OTSGs by motor-N driven Isolation valves t hat are normally shut. Does this apply to CR-37 5.4 At another plant, on loss of of fsite ~ Crystal River Nuclear Sa fety Task 18/GJB Short-term reccynmendations are made to provide automatic power, the makeup pumps stop and Force load ing of the MU pump on the diesel. Changes 'In LPl must be restarted manually. The B&W Recommendation loading seq uence are suggested to prevent overl oad ing seal injection valve is automatical- diesel with addition of the MU pump. Long term recom-ly loaded on the diesel. W i thout mendations are made to improve seal injection reliabil-seat injection (MU pump stoppod), Ity.

the demand signal will have the valve full open. On restart of MU pump (Operator action), the possi-bility of severe thermal shock to the seat exists. Does this apply to CR-37

O O LIST OF CX)NCERNS 5.0 NOT CLASSIFIED Page 18' CONCERNS REFERENCE PRIORITY, ASSIGNMENT, it .*EWATION 5.5 At another plant, OTSG EFW Crystal River Nuclear Safety Task 1/TCR This item does not apply to Crystal River Unit' 3. No control valves are shed from the Force. further action required, diesel af ter an ESFAS. They are motor-operated (i.e., f a ll "as is"). Theref ore, on a loss of of f-site power coupled with ESFAS, SG rderf i l l is as su red. Does this ap-ply to CR-3?

5.6 Consider increasing steam line break Crystal River Nuclear Safety Task 1/TCR A reconenendation was made that no changes be made to the matrix setpoint to greater than Force. present setpoint. No f urther action required.

600 psi.

7 5.7 Study feasibility of around-the- B&W-NRC meeting. 1/PFM Around-the-clock I&C coverage will be established as soon 5 clock I&C coverage. as additional people can be hired and trained. It is planned to have the specifled coverage by plant startup.

5.8 Provide PORV lsolation with ESFAS or Crystal River Nuclear Safety Task 1B/GJB A short-term reconsnendation was made for B&W to proceed low RCS pressure signal. Force with work identified in task letter, upgrading PORV to .

B&W Reconndation. requirements of NUREG-0667.

5. 9 Provide psitive Indication of posi- FPC Corrective Action List, Rec. 4. 1/J0M This item will be completed prior to plant startup.

tion of al .' three relief valves.

5.10 Modif y vital bus ,mnels for quick FPC Corrective Ac. ion List, Rec. 14 1/ REC MAR 79-6-71 will provide this capability.

f use replacement.

5.11 Modify EFW pump autostart circuit FPC Corrective Action List, Rec.15 1/ REC MAR 79-11-67 will provide assurar.co that upon NNi-(X),

and Reactor trip circui t so that no IPPD/NSAC Crystal River Task Force (Y), or ICS power failure the emergency feedwater pumps 4 control power fa ilure wil l prevent B&W Rec. 1. will start and the Reactor will be tripped automatical-EFW activation on low SG level. ly. This change will remain in effect until long-term changes are made to minimize NNI-(X) and (Y) power fall-ure. - Analysis will be done to dotermine severity of trans i ent .

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h O O LIST OF CONCE 6

5.0 NOT CLASSIFIED Page 19 CONCERNS REFERENCE PRICRITY, ASSIGNMENT, RECopWEM)ATlON 5.12 Provide PORY valve position lights FPC Corrective Action List, Rec. 16. 1/R:.C FMR 80-3-62 will provide this Indication.

for solenold.

5.13 Perform a fatigue analysis on the FPC Correc+ 1ve Act ion List, Rec. 19. 1/CJB This item will be completed prior to plant startup.

pressurizer heaters.

, 5.14' The relief valve % aings u.. the FPC Corrective Action List, Rec. 20. 1/EJA This item will be completed prior to piant startup.

pressurizer reelef nozzles shou'd be determined.

5.15 Provide diverse Containment isola- FPC Corrective Action List, Rec. 23. 1/.'OH This item wlil be completed prior to plant startup.

t ion.

7 '5.16 Pavlew steam generator rupture ma-ItP0/NSAC Crystal River Task Force 1/lB/3/TCR A review was made of rupture matrix activation logic

$ trix to minimize undesired Isolation Tekn6kron Report, Rec. I and several changes were recomended.

of EFW from OTSG. :NUREG C560 #G Crystal River Nuclear Safety Task l Force 8&W Recommendation.

MPR Report.

5.17 The pressurizer heators should not B&W Recommendation 9 ;10/GDQ A review was made of pressurizer Instrumentation f all on when pressurizer level is and several possible modifications were suggested. i low.

5.18 Services to the RC pumps should be' B&W Recommendation 10. IB/GJB Recomend f urther evaluation of four identif ied RCP maintained to allow continued opera-services. Refer to item 3.4 tion of at least me pump per loop and also provide Information to tho '

Operator which would preclude soal damage.

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( Da LIST OF COPCERNS 5.0 NOT CLASSIFIED Page 20 COPCERNS REF ERENCE MllGtlTY, ASSl(NMENT, REC 0D04ENDAT10N 5.19 baW review of containment Isolation NLREG-0560 #10. 18/GJB A review was made of Conta!nment .' Isolation strategy and recommendations. 'several possible short-torm and long-term modifications were tuggested.

5.20 Review and suggest Improvements on ' Crystal River Nuclear Safety Task 18/AEF A revlaw was made of the annunciator systen and a recam-annunciator system. Force. mendafion was mad, to temporarily block many noncritical alarms 9n reactor +rlp.

5.21 At another plant, SLRM system shuts Crystal River Nuclear Safety Ta sk I/TCR This item does not apply to Crystal River No. 3. Nofw-the MSIVs and MFIV on the af fected Force. ther action required.

OTSG, plus SLRM opens the steam sup-ply valve to the steam-driven emer-g gency FW pump. However, the steam 4 supply valve that SLRM opens is from 0 the af fected OTSG. Is this proper?

(Tnis OTSG will boli dry.) Does this apply to CR-37 i

5.22 At another plant, SLRM starts the Crystal River Nuclear Safety Task IMCR This item does not apply to Crystal River No. 3. ft) fur-steam-driven EFW pump while the ICS Force. ther action regulred.

controls OTSG level. Does this cause the affected OTSG to be fed continuously? Does this apply to ,

CR-3 7 5.23 At anothea plant,- Excessive Feed wa- ' Crystal River Nuclear Safety Task 1/TCR This item does not apply to Crystal River No. 3. No fur-ter Event Tree shows cases where HPl . Force. ther action required.

is automatically actuated, overrid-den by the Operator, ano subsequent-ly is needed, due to s.asi l LOCA (e.g., FORV sticks open). FPI t hen is no longer available for automatic operation. Should an autcmatic re-set % included if RCS pressure goes a

bach above 1600 psi? Does this ap-

pl y to CR-37

LIST OF (X)tCERNS 5.0 NOT CLASSIFIED Page 21 CONCERNS REF ERENCE l PRICRITY, AS$1CNMtJ.1, RB 0Mp(NDATION 5.24 At another plant, automatic contain- Crystal River Nuclear Sa fe ty Ta sk 1/TCR This item does not apply to Crystal River No. 3. No fir-ment Isolation of the MSIVs by ESFAS Force. ther action is required.

(Iow RC pressure or high building

' pressure) wilI deny access to con-denser via the condenser dumps.

Thus, radioactive fluid will be forced to the atmosphere through

. steam safeties or ADVs until the RCS pressure is reduced below the sec-ondary pressure. The of f si te dose accumulation will be greater than if the condenser, were used to trap and g partition the ef fluent.

da

~ 5.25 Partial collapse of EWST Dome due to ' B&W Recommendation. 1/JDM This item does not apply to Crystal River ho. 3. No fur-vacuum breaker fa i l ure. Site in- ther action is required.

struction sent to B&W to FPC on 3/29/79. Verify action.

5.26 PZR sampling system thermowells do ' B&W Recommendation. 1/JDM Thermowells have been blocked off. No further safety i not meet pi pel ine design criteria. concern. No f urther cetion is required.

Verify replacement.

5.27 Modification relating to CR-3 Task Crystal River Nuclear Sa fety Ta sk 1/GJB A subcommittee of the Task Force has been appointed to Force recommendations should be re- Force. review modifications as deemed appropriate.

viewed by the Task Force.

5.28 EFW Nozzle collar fal: r a Field B&W Recommendation. 1/JDM This item will be empleted before plant startup.

Change Authorization was vapared.

Verify corrective action s

o A a

LIST OF COTERNS 5.0 NOT CLASSIFIED Page 22 COPCERNS REFERENCE PRIORITY, ASSIG4 MENT, RECOMMENDATION 5.29 Temperature Induced errors on steam B&W Recommendation. 3/CJB This item is currently being studied by B&W.

generator wa ter level measurement. [

Concern Is overfIIi because of tem-perature-Ind uced errors when level i s raised to 95% on operate range.

5.30 Potential shorting in sa f ety system B&W Recommendation. 1/JDM This item will be completed before plant startup.

from improperl y Installed fiber cIamp. Verify corrective actions were taken.

-5.31 Reactor vessel seal plate gasket ' B&W Recommendation. 1/JDM Action has been completed.

g leakage. Site Inst. sent to B&W to ,

4 FPC 12/6/79. Verify action.

.ru 5.32 Recorder cable deterioration. Re- B&W Recommendation. 1/JDM Action has been completed.

placement of cables had been prevl-ously recommo.eded. Verify action.

5.33 Potential overpressurization of the B&W Recommer.fation. IB/ REC decay heat removal system. Proce-dures should be reviewed to assure this concern has been covered.

5.34 Potential for stress corrosion B&W Recommendation. 1/GJB Incore guide tube piping is 304L. No further action is cracking of SS304 Incore guide tube req ui red .

piping. Not a concern if 304L.

Verify that piping is 304L or in-spect.

V

O O LIST OF CONCERNS 5.0 NOT CLASSIFIED page 23 COPCERNS REFERENCE PRIORITY, ASSIGiMENT, RECOMMENDATION -

5.35 Investigate multiple channel Crystal River Nuclear Safety Ta sk 3/ DAM A recommendation has been sede that mdt fications belm- '

cross null capability for Identify- Force. piemonted to accomplish this.

Ing failed signals before transfer-ring to ICS control.

5.36 a. Eval uate Inst. air system for Crystal River Nuclear Sa fe ty Ta sk I/JCH Retaln dimi tntil a permanent Installation can be ac-capacity and verify I f adeq ua te Force. comrn i s'ved , dependent upon evaluation of Instrument air accumulator capacity. syste..

b. . Provide a diesel backup air cm- Crystal River Nuclear Sa f e ty Ta sk IB/JCH A recmmendation was nede that a permanent backup diesel pressor (considering loss of Force. air empressor be added.

of f site power event).

N 4 5.37 Investigate additional antic i patory Crystal River Nuclear Sa f ety Ta sk 3/ MEN A recommendation was made to study additloril trips and W . trips from sec. Plant upsets. . Force. to implement those deemed desirable.

5.38 At another plant, on loss of condon- Crystal River Nuclear Sa fety Ta sk I/TCR This item does not apply to Crystal River No. 3. No fir-ser or SLRM actuation, the NSSS can- Force. ther action is required.

not be cooled below approximatel y 550*F because secondary pressure is set at approximately 1065 psi (steam safetles). If the . ADVs are kept Isolated, they are un ava il able on LOOP. Does this apply to CR-37 5.39 investigate possible rupt ure Crystal River Nuclear Sa f ety Ta sk 3/ME Recommended further eval ua tion of drain tank level disk Indication. Force. transmitter problems.

5.40 Perform a management review to as- B&W Recommendation 14 3/JCH The NCRC will perform this review.

sure that proper change and mainte- Teknekron Rcport nance practices are in place ( i .e.,

i B AW -1564 Indicates 32.5% of trips are caused by Operator /Techn ic ian action - Teknekron Indicates 50% of LERs result from adm in istrati vo

, problems).

-O O LIST OF CONCERNS 5.0 NOT CLASSIFIED Page 24 COPCERNS REFERENCE PRIORITY, ASSIGNMENT, REC 0h04ENDATIG4 5.41 Evaluate feasibility of upgrading NUREG-0560 #4 3/EJA A review was made and several suggestions for a Selsmic condensate storage tank to Seismic Class I water supply were discussed and forwarded to En-Categorr I. gineering for evaluation.

5.42 Emergency feedwater autostart equip- NUREG-0560 #5 IB/ REC A MAR has been written to provide a saf aty-grade actua-ment does not meet single fallure tion system. In the short term, the con 1 ol orade actua-cri terlon. (Safety-grade). tion system will be upgraded to prevent ICS/NNI power f ailures from negating an autostart function.

5.43 Recorrnendation on zone of control Crystal River Nuclear Safety Task 3/DN4 A recorrnendation was made that Operations Superintendent board responsibility. Force. should discuss with his pe ronnel methods of obtaining maximum utilization of personnel during translem s.

7 5.44 OTSG tube ruptures require Operator Crystal River Nuclear Safety Task IB/TCR A recommendation was made to add additional radiation

.$ identification of the af fected Force. monitors on steam line.

OTSG. Are radiation moni tors re- B&W Recorrnendation.

quired on the outlet of the OTSGs?

5.45 Review main feedwater overfill con- B&W Recommendation. IB/GJB Recorrnendation made to follow B&W ef fort on this item cern for potent ial f Ixas, wIth Intent to implement MFW overfIII prevention system.

See items 3.6 and 3.7 for related discussion.

5.46 Review EFW overfill resulting f rom B&W Recommendation. 3/GJB Recommendation made to follow B&W ef fort on this item single f allures for potent ial fixes. with Intent for long-term implementation.

5.47 Review EFW flow control system for B1W Recommendation. 3/ REC A recomendation was made that an REI be generated for modifications to ensure reliability this review.

and reduce system response.

5.48 Review the design for potential of B&W Recommendation. ERK A review was made and it was suggested that further in-Increasing the range of pressurizer Operator Recommendations. vestigation te done.

level Ins tr unent a t io n.

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,O q) O U/

LIST OF CONCERNS 5.0 NOT CLAS$1FIED

. J 25 CONCERNS REFERENCE PRIORITY, ASSIGNMENT, REC 0mENDATION 5.49 Consider. expanding the range of RC ' Rec. C.1.2.9 f rom I&E TSMi-2 18/GDQ A recomendation was arje to increase wide-rangw RC pressure Indication to approscimately Investigation Team, pressure Instrumentation to 0-3000 psig.

0-3000 psig. Consider expanding B&W Recommendation.

ranges of other instrumentation, such as Thot*-

5.50- The potential for a Mu injection B&W Recommenhelon. 1/GJB Potential safety concern from B&W 20$FA units. Not ap-Ilne break resulting in Hi f low and plicable to CR-3. See item 5.65. No further action is Lo MU pump suction pressure causing required.

pump damage has to be assessed.

5.51 EFW line break in Containment could B&W Recormendation. 3/TCR A study was done and it was determined that an EFW line 3 result in loss of DH Removal capa- break would not result in a loss of DH removal capabil-7 bility through OTSG. Review and ity.

@ a sses s.

5.52 SLB upstream of MSiv could lead to B&W Recommendation. 18/TCR See item 5.16.

an aggravated condition because of .

continuous feeding of EFW. Review for applicabiIity to CR-3.

5.53 Evaluate increasing se.condary reIIef Crystal River Nuclear Safety Task 3/ MEN Refer to item 5.2. Defer any action until results of 5.2 valve setpoints and post trip con- Force. are assessed.

trol pressure.

5.54 Identif y and implement discrete B&W Recommendation. 3/EJA A study of present FW system circuitry was made and sev-

, changes in FW system to improve re- Operator Recomendations.. eral engineering studies and modifications were recom-IIability and reduce safety system monded.

challenges. '

1 M'0/NGAC Rec. l l 1.B.1.C.

5.55 Review autoinitiation of Incore tem- IB/JDM A recomendation was made that incore temperature perature printout . Cons i de r . I n l t i a-tion on HPi or RX trip.

LIST OF CONCERNS 5.0 NOT CLASSIFIED Page 26 C0tCERNS REFERENCE PRIORITY, ASSICNMENT, REEOMMENDATION '

5.56 Emergency feedwater bypass valves, Operator Recommendations.

valve controllers, should be placed IB/ REC A MAR has been generated to nove the EFW bypess alvo on the corresponding engineered control the PSA section of the neln control boord near safeguards panel.

the pump coontrols and flow Ind icators.

Associated SG level and EFW flow Indicators should be locatad near the controls.

5.57 . Put Ind icators and controls for Operator Racommendations. IB/ REC A MAR has been generated for this modl fication.

emergency feedwater on the same panel.

5.58 The control board totwell level In- Operators Recommendations.

3/REE A recommondation was riede to gener e.'e a MAR to investi-g strumentation only reads +4 Inches -

gate this capability.

4 Operators want +12 Inches - Opera-m tors want control and Indication f rom the hotwel l .

5.59 Reactor Building purge fans and flow Operator Recommendations.

controls recorders are spilt between 3/ REC A recommendation was made to generate a MAR to implement this change.

the front of the board and the back. It takes two Operators to get purge started.

5.60 All extraction alarms are on one ' Operator Recommendations.

3/ REC A recommendation was made to generate a REl to study this alarm. Consider remote panels (from concern.

Control Room) to provide quic ker Identification of the problem.

5.61 NaOH valves - should open on coincl- ' Crystal River Nuclear Sa fety Ta sk 18/GJB A rectrnmendation was made to study two alternative enlu-dont HPI and 30 lbs. RB pressure ra- Force. tions, ther than 4 lbs.

t-D D d V LIST OF 00PCERNS 5.0 NOT CLASSiFlED Page 27 COPCERNS REFERENCE FRl(RITY, ASSIGMENT, RECOP#ENIMTION 5.62 Diesel Jacket heater size is Operator Recommendations. 18/ REC A recommendation was nede to generate an REI to review questionable - and tube oli heater sizing and recommend changes as necessary.

would tu a problem during cold wea-ther loss of offsite power event.

Investigate modifications.

5.63 Moisture Separator Reheaters - Operator Recommendations.

If 18/ REC A recommendat ion wa s made to generate a MUt to use the Opera +3r forgets to isolate main existing relays to close steam supply valves on turbine steam to the moisture separator re- trip.

heater, an overcooling could occur.

Consider automatic closure of MSR steam supply valves on a turbine g trip or low load operation.

N N 5.64 For single valves (A&B Valves) Operator Recommendations. IB/ REC A recommendat ion was nede that these modi f ications be that receive both ESFAS . A&B, local done.

control of affacted equipment is required (sometimes in high radiation level areas). Consider Installing controllers in Control Room or, as a minimum, place valve controls in a low radiation environ-ment. (Approximately 20 valves.)

5.65 An Inadvertent actuation of B ESFAS ' Operator Recommendations. 18/GJB A study of present circuitry was made and recmmendations with A pump running could cause were made for both short-term and long-term destruction of a MU pump (closure of mod i f ications.

MUV-64) unless Operator acts.

(Could cause fa i l ure of two pumps) .

Consider a suction pressure sw i tch which will be defeated on a full ESFAS signal.

5.66 Deleted

O. O LIST OF CDNCERNS

'5.0 NOT CLASSIFIED Pege 28 COM: ERNS REFERENCE PRIOR 4WY, AS$104 MENT, RECOMMENDATION '

5.67 investigate adding -: a 'self- Operator Recommendations. 18/ REC A reconmendatt was made b generate an REl' to add en

. contained diesel-driven emergency additional pump.

feedwater pump.

.5.68 Review . proposed design for remote Operator Recommendations. 18/ REC A recommendation was made that a review be perfornied b shutdown panel . I .e., can't reset satisfy this concern.

some items (HPI, LPI, R3 Isolation)

- In addition, we cannot bypass or reset rupture matrix and we don't have pressurizer heater control at the remote shutdown panel.

g 5.69. Upgrade present shutdown panel Operator, Recommendations. IB/ REC A recommendation was made that a MAR be generated to pro-4-

Co Instrumentation for redundancy in vide this.

case of. loss of power supplies.

5.70 increase the effective range of Operator Recommendations. 1B/TCR Th i s wi l l be done.

Reactor Building sump ' level (i.e.,

+9').

5.71 Spurious closure of either of the DH Operators Recommendations. 1/IB/GJB A recommendation was nede that decay heat pumps be '

dropline valves during DH operation tripped anytime their suction supply valves are not fully can damage both DH pumps miess the open.

Operator. promptly trips the pumps. "

Automatic protection of the pumps (overridden by. ESFAS) should to i provided.

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D v J LIST OF (DNCERNS 5.0 NOT CLASSIFIED Page 29 CONCERNS REFEREPCE fRIGilTY, ASSIGMENT, RECOMNDATION 5.72 investigate providing automatic Crystal River Nuclear Safety Tark 18/CJB A recommendation was made to implement this concept.

delayed- emergency . feedwater . flow Force.

l and HP! Injection on high RC pres-sure. The concern Is for a total loss of feedwater ' and pressur izing to the ' safeties and the Opera-tors not recognizing the loss of feedwater.

5.73 Condensate control . system is Crystal River Nuclear Safety Task 3/ REC A recommendation was made that a MAR be generated to In-not completely adequate. The . hot- Force. vestigate this problem.

well controller . will not control g well over 0-100% range. ,

'I to S 5.74 The . steam line rupture matrix I&C Technicians Recommendations. 3/KOV - A recunmendation was made that the system be modified to does not. have built-in surveillance Improve testabil ity.

testing capabilities.

Switch setpoint and deadband cannot be verfiled on line. Care should be taken on new system design to ensure t hat power failure will not Isolate emergency feedwater. The new MAR ,

should be reviewed for power failures.

5.75 investigate various means for assur- l&C Technicians Recommendations. 3/ REC A recommendation was made that an REI can be generated to Ing reference legs for steam genera- Crystal River Nuclear Safety Task study this problem.

tor levels - pressurizer levels, RC Force.

drain tank level, and makeup tank level. Provide pa st documentation on the subject or addltlonal work if n ecessar y.

i 'l

O O LIST OF CONCERNS

~

5.0 NOT CLASSIFIED 'Page 30 CONCERNS REFERENCE PR1(R1TY, ASSIGNMENT, RECO*EtcATIGi

.5.76 Replacing BMCo BY transmitters with I&C Techniciars Recomendations. 3/ REC A reconenendation was made then an REI be generated to Rosemount SQ transmitter. (Have study this replacement.

high failure rate with pressurizer level, all BYs tend to shif t zero).

5.77 Modify deaerator level transmitter l&C Technicians Recomendations. 3/ REC A recomendation was made that a MAR be generated to cor-so that it will read accurately dur- rect this problem.

Ing and af ter vacuum degassing.

Provide temperature compensation n et wor k.

5.78 Single failure can cause failure of Nono 10/KOV A recomendation was made that the annunciator / events re-

.3 annunciator panel. Investigate whe- corder power feed should have an automatic bus transfer '

7 ca ther a modification should be made. to another power source.

o 5.79 Core exit thermocouples. Depend ing Nono ID/ REC A recommendation was made to add a recorder to the main on exit thermocouples requires a control board, bstter display system. Investigate.

5.80 investigate the feasibility of going None. 1/GJB The in-dspt!, investigation of this item is beyond the to cold shutdown with only safety- scope of the charge of this Task Force. Reg. Guide 1.139 grade equipment. (mC Question, was not part of the design basis for CR-3. No f urtiver Reg. Guide 1.139). action is required.

l 5.81 Perform control board layout design NUREG-0660 #1.d 3/PFM A recommendation was made for a comprehensive Control review taking into consideration hu- Room Hunan Factors Evaluation, with changes Implemented man factors. as determined by this study.

5.82 Investigate placing bypass valve CR-3 Nuclear Safety Task Force IB/PF'4 A study was made and several alternative solutions were or slow opening f or equalizing Operator Resommendation recorrended f or implementation.

MSt ys to prevent steam hamer.

_. . .- - _ . -- .- . -_ _ .. .-. .~ ~ .- - - --. . .. . , .

LIST OF CONCERNS 5.0 NOT CLASSI'FIED Page 31

~

CONCERNS REFERENCE PRIORITY, ASSIGNMENT, RECORNENDATION

.5.83 Analyze steam rupture matrix for Crystal River Nuclear Safety Task 18/KOV An analysis of the steam rupture matrix was mode end failure modes such as loss of con- Force. several deficiencies were discovered. It was recosunended trol / motive power. these def iciencies be *:orrected. .

5.84 Renaw steam rupture matrix " Rupture Crystal River Nuclear Safety Task IB/KOV A recommendation was made to make the changes described.

Matrix Actuation Channel 1 and Rup- Force.

ture Matrix Actuation Channel 2"

- (RMAC-1, PNAC-2) . Change design drawings and plant proosdures. - (To eliminate potential conf usion caused by using "A" and "B" designations.)

3 5.85 Add redundant . radiation monitors - It /TCR See Item 5.44, 7 on each steam line just outside Con-

$ t a inment. Under most events, this will be the first alarm received.

This w!Il alert Operator to OTSG tube leak, so he can take appropri-ate action to ensure that PZR pres-sure/ level control is not lost.

5.86 Consider adding some means of spray- CR-3 Nuclear Saf ety Task Force. 1B/GJB Recommendations were made for modifications to allow ing the PZR so that pressure can be spraying the pressurizer from HPl.

easily reduced during natural circu-l a t io n. Consider a line tapped into WI.

5.07 investigate adding reset control of CR-3 Nuclear Safety Task Force. 3/ REC A recommendation was made that an REI be generated to '

EFP-2 on the main control board. provControl Room operator -ith remote reset capability of EFP-2 5.88 investigate r ilng motor-operated Operators Reconwnendations. 1/PFM A study was made and it was determined that the methods equallring va, se (approximately CR-3 Nuclear Safety Task Force. out linnd in item 5.82 were pref erable to an equalizing 2-1/2" around MISVst to avoid in- line, stant repressurization of MS lead up to turbine stop valves.

A LIST OF CDPCERNS 5.0 NOT CLASSIFIED Page 32 COPCERNS REFEREtCE PRimlTY, ASSIGNENT, REColedEK)ATION 5.89 investigate giving Operator the Operators Recommendations. IB/TCR Recmmendations nede inder item 5.16 will provide menuel ability to manuall) Initiate steam Initiation. Another recommendation was nede 10 provide rupture matrix for either OTSG. This operator guidance on use of this feature.

should assist the Operator in Iden-tifying which OTSG has rupt ured t ube s.

5.90 Could gas be transferred from stor- NGRC 8.ction item List following TMI. 3/JCH Conduct a study to determine equipment, procedures, etc.

age tank (s) to Containment (to pre- needed, and feasibi l i ty of this type of trans fer. A vent overpressurization; reIIef study was conducted and this was determined to he valve and/or rupture dlse opera-feasible; a potential problem of safe personal access tion)?

exists.

N L 5.91 Determine equipment, procedures, ' N@C Action item List following TMI.

N 3/REE This will be done by January 1, 1981.

personnel, logistics, e tc . , that would be needed to Install /use hy-drogen recombiners (post-accident).

5.92 Review application of type of ' NGRC Action item List following TMI.

3/REJ Pressure sw itches are on hand; valves needed are on swliches used in steam line rupture order. A MAR exists for this.

matrix circultry.

5.93 Determine feasibility of electrical ' Crystal River Nuclear Sa f ety Ta sk 1/ REC Th i s wi l l be done.

By Interlocking motor-driven EFW force.

pump to start on loss of MFW.

5.94 CR-3 HPl setpoint is 1500 psi. NCRC Action item List Following TMI.

i 1/ERK An investigation wa". made and it was determined that no Other B&W plants use 1600 psi . What change in the pr9sent setpoint is warranted. No f urther i s the basis for 1500 psi setpoint, act ion is required.

and should a change be considered?

?

N 0

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LIST OF CDPCERNS 5.C NOT CLASSIFIED .Page 33 CONCERNS REF ERENCE PRIORITY, ASSIG4 MENT, RECOMMENDATION 5.95 Plant electrical auxiliarles are NCRC Action item List following TMI. IB/PFM A reccrnmendation was made to conduct a study to determine powered from startup transformer best bus source, considering fail ure modes.

during operation. If loss of of f-site power event occurs, plant trips because FW, CD, etc., pumps trip.

Should auxillaries be powered from plant auxiliary transformer, so that on loss of of f site power, plant re-mains on-line (since auxillaries would not be af fected)?

5.96 Determine feasibility of electrical- NGRC. IB/CJB Not reconmended.

g ly Interlocking HPI to start on loss

of all FW (main & EFW).

'N 5.97 Confirm the adequacy of emergency ' Essex Report. 3/REE Covered by item 5.81.

Iighting in the Controf Room.

5.98 investigate automatic isolation ' Crystal River Nuclar Sa f e ty Ta sk IB/ REC Conduct feasibility study.

of letdown on a Reactor trip. Force.

5.99 RB Spray is not cooled. What is NCRC. 1/TCR FSAR took no credit for RB spray cooling analysis and ac-effect on RB cooling re!! ability ceptable results were shown. No action required.

during small break LOCAs?

5.100 - 5.103 Deleted 5.104 Review immediate action stops of Rec. C.1.e.5 f rom l&E/TMl-2 investi- 1/KOV Several actions were identified and were referenced to procedures to determine what manual gation Team. other items for action. No further action is required.

actions the Operator is required to take. Consider making those actions a utoma tic .

_ _ _ _ .__._______I

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O O LI5T OF CONCERNS

5. .NOT CLASSIFIED Page 34 CONCERNS REFERENCE PRI011TY, ASS lGNMENT, REC 0804ENDATIOf 5.105 Evaluate the sump systems f or alI . Rec. C.1.e.7 from I&E/TMI-2 18/TCR safety-related rooms, to ensure that investigation Team.

flooding f rom Internal or external sources Is adequately addressed.

5.106 Perform a detailed review of the Rec. C.1.e.8 from l&E/TMi-2 3/KOV Engineering to conduct a study.

corinunication paths between redun- Investigation Team.

dant safety-related equipment to assure against common mode f ailure of steaming, fire, and f loo 1 5.107 FPC Management should perform an Rec. C.I.e. from l&E/TMl-2 1/PFM This has been done.

g evaluation of the number of exempt investigation Team.

{

A personnel that should hold Operators Licenses.

5.108 FPC Legal Department t hould review Rec, Col.e. from I&E/TMl-2 3/PYB A recommendation was made to have FPC Legal Department Recorrriendation C.5.b of l&E/TMI-2, Investigation Team. review'Al-500.

regarding the preservation of evi-dence to determine if further action is required.

5.109 investigate past ESFAS actuations on CR-3 Nuclear Safety Task Force. 3/ERK A reconenendation was made to more completely evaluate the B&W plants. Make recommendations past ESFAS actuations, where appropriate, and/or show that past actions have reduced the number of actuations.

5.110 Are there improvements that can be CR-3 Nuclear Saf ety Task Force. 3/GDQ A recomendation was made to study this concern, made in the turbine control system to reduce the number of turbine trips. Co sider evalt.ating opera-tional history at Crystal River and other utillties with the same type turbines. -

v)

LIST OF (DCERNS 5.0 NOT CLASSIFIED Page 35

. C0tCERNS REFERENCE FRIGilTY, ASSl9 MENT, REC 08WENDATION 5.111 Failure of . a single valve in the CR-3 Nuclear Saf ety Task Fcree. IB/GJB Several possible modifications were suggested. (Refer to

. decay heat drop line (DH ' Suction 1.34.)

from the RCS) prevents establishing cooling with the DH system. A by-pass line could be provided.

5.112 Review HP! design to provide ' CR-3 Nuclear Safety Task Force. 3/GJB A recommendation was nede to study this concern f wther.

either parallel suction valves or normal l y ' open suction valves to en-sure HPl flows from either Train A or Train B.

R 5.113 The existing design of the ESFAS ' CR-3 Nuclear Safety Task Force. 3/ERK A recommendation was nede b investigate the reasoning L

m system causes start of the decay supporting the existing setpoints. If appropr iate, heat pumps at 1500 psi. Di pumps Initiate action to change setpoint to a more appropriate cannot pump Into the system until value.

approximately 400 psig is reached.

The pumps will sit on recirc. for a significant length of time. Should there be a separate setpoint for DH pump actuation?

5.114 -Fallure of the ADV, TBV, and other ' CR-3 Nuclear Safety Task Force. 3/GDQ A recommendation was mode to conduct a study to identify equipment on the main steam headers devices whose fa il ure could cause a smell steam line

, can result In a small steam IIne break. This study should evaluate the af fects and break transient. probabit'ty of t hese breeks and make approor late reconmenda t ions.

C~D o U

LIST OF CONCERNS 6.0 POWER SUPPLIES Pege 36 CONCERNS REFERENCE PR1(RITY, ASS 1GNMENT, RECOWEM)ATI(M 6.1 Review PORY CIRC. to assure (1) PORY FPC Corrective Action List, Rec. 2. 1/ REC MARS have been generated which will assure that these doesn't open when It Is not required itP0/NSAC Crystal River Task Force concerns are satistled.

to open; (2) the PORY does not also ii1.A.3a af fact a failure which would af feet B&W Recommendation 2.

the block valve; and (3) assure the capability to open the PORV, on a loss of NNI Power, f rort main control

[ board.

.I 6.* Modify pressurizer spray valve so FPC Corrective Action List, Rec. 3. 1/ REC A MAR has been generated which will accomplish this.

that NNI Power failure wilI close B&% Reconnendation d.

valve.

N O 6.3 Move 120 VAC ICS (X) Power to vital FPC Corrective Action List, Rec. 7. 1/ REC A MAR has been generated which will accomplish this.

• bus.

6.4 Annuntiate loss of AC power to (X) FPC Corrective Action List, Rec. 11. 1/ REC A MAR has been generated which will accomplish this.

and (Y) buses and ICS. B&W Recommendation 6.

6.5 install Indicating lights on all vi- FPC Corrective Action List, Rec. 13. 1/ REC A MAR has been generated which will accomplish thlr.

tal bus feeds.

6.6 Review au?o start of critical items CR-3 Nuclear Safety Task Force. 1/18/KOV A recomendation was made to review this and related on LOOP without ESFAS actuation. Concerns.

I 6.7 Design / Install backup power sources CR-3 Nuclear Safety Task Force. 18/ REC A recommendation was made to prepare an REI to review

( AC and DC) for NNI 'Y) channel. this problem.

6.8 Review f use coordination for all vi- CR-3 Nuclear Safety Task Force. 18/KOV A recommendation was made to do en In-depth review of tal bus feeds. vital bus fuse coordination.

O O  ;

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LIST OF (I)tCERNS 6.0 R)wER SUPPLIES p.g. 37 COtCERNS REFERENCE Rt1GtiTY, AS$1G#8ENT, REC 08eENDATION 6.9 Cause of the loss of regulated INST CR-3 Nuclear Safety Task Force. 3/ REC BUS 2. for 2 seconds, approximately 6 minutes . into the CR-3 translent, should be determined and corrective action taken.

CR-3 Nuclear Safety Task Force.

- 6.10 Cause of the low voltage alarm on 3/ REC Inverter 3C at the beginning of the CR-3 translent should be determined and corrective action taken.

I CR-3 Nuclear Saf ety Task Force.

6.11 Cause of numerous fire alarms during

> the CR-3 translent should be deter- 3AtEC N

s La mined and corrective action taken.

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4 i

i 4

i 4

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O O LIST OF 00tCERNS 7.0 TEDESCD REPORT - 88W REACTm TRANSIENT 1ASK FORCE page 33 C0tCERNS REFERDEE FHimlTY, ASS 19esENT, RECO*0EENDATION

'7.1 Classify EFW system as engineered 3NFM Reccrvnendations have been made which sake the system more safety feature. like an ESF AS system, as well es recommending the addI-tion of a diesel-driven EFW pump.

A. Upgrade to meet safety grade criterla (relsmic design re-q uirements being evaluated by PAS).

B. Where upgrade may not be feasi-ble, consideration would be giv-

, en to the addition of a dedica-ted EFW system (i.e, a separate g traln). ,

I w

to 7.2 Automatically Initiated and control- 3#FM A recommendation was ande that these Improvements be led by engineering safety fea- Impl emen ted.

tures which are Independent of the 6CS, NNI, and other non-safety sys-

, tems.

4 A. Reevaluated autostart signal selection to preclude OTSG dry-out (i.e., EFW autostart on an-ticipatory loss of feedwater).

B. OTSG level controlled to revent overcooling during recovery from ,

f eedwater translents.

C. Terminate EFW flow to OTSG be- ,

fore overfilling.

7.3 Not Applicable.

x) k LIST OF CDPCERNS 7.0 TEDESCO REPCRT - Bad REACT 01 TRANSIENT TASK FORCE Pege 39 CotCERNS REFEREtCE PRIORITY, ASSIChMENT, RECOMMENDATION 7.4 Steam line break detection and miti- 3/PFM The Task Force recommended several leprovements % the gation systera. steam line rupture matrix that amet these concerns b some degree. Part B will need f ather study and is not A. Eliminate adverse Interaction covered fully by our reemmendation.

with EFW system.

B. Reevalunto and modify such that system is capable of differenti-ating between steam line break and overcooling and edercooling transients.

R 7.5 Improve reliability of instrumenta- 1/340Q Numerous short-term and long-term recommendations were g tion and plant control: (l.D.1/ made to address these concerns.

. So ll.F.4).

A. Separate and channelize power buses and signal paths for NNI and associated cx>ntrol systems.

B. Reconsider "mid-sca l e" f a il ures as preferred fa il ure mode for instrumentation.

C. Multiple Instrument f a il ures should be mambiguousl y Indl- ,

cated to Operator.

J

, D. Control systems should Mve the I

capability to detect gross fall-ures and take appropriate defen-sive action automatical ly.

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5 s.

LIST OF (X) TERNS 7.0 TEDES(D REPORT - BW REACTOR

,- TRANSIENT TASK FORCE Page 40 COtCERNS RU ERENCE PRICRITY, ASSl(MMENT, REDOWENDATION E. Revlow and rearrange, as neces-3 sary, NNI power buses to provide I

red undancy of Indication for each coolant and secondary sys- ,

tem loop.

F. Prompt followup actions shou *d be taken on:

. .BAW 1564 (ICS Reliablfity j Anal ysi s)

, g . NSAC-3/lff0-1 Recommenda ,

s a tions (Evaluation of CR-3 o incident)

. IE Bulletin 79-27 (Loss of '

j NNI Power Supplies).

7.6 Establish prompt Implementation 3 safety grade / REC 4

of select data set of principal plant parame*ars for Operator (safety-grade) (l.D.2).

I Reconmended set:  ?

3 1

A. Wide-range RCS pressure.

B. Wide-range pressurizer level.

C. Wide-range RCS temperatures: t

. . Hot leg (each loop) b.

I

- - - - - . .- - . -- ,-, . _ - - . . , . - - - --, - , . . - - - - ~ . - - - - -

,- _ . . - _ . . . - . . .. _. . _ . . . - - ._._ . ~ . . _ _ . - . . . . . .

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O O LIST OF CONCERNS

~ 7.0 TEDESCO REPORT - B&W REACTOR TRANSIENT TASK FORCE page 43 .

CONCEANS REFERENCE PRIGt1TY, ASS 1GNMENT, REConMEIOATlON

. Cold leg (each loop)

. Core outlet (two or selectable).

D. Makeup tank level l

E. Reactor Building temperature and pressuri zer.

F. Wide-range OTSG level (both).

2 G. Wide-range OTSG pressure (both).

2=

N i H. Source range Nl.

I. Intermediate range Nl.

i 7.7 increased usage of incore thermo- 3/ REC couples: (1.D.1) ,

l A. Provide for flexibility to sub-stltute incore thermocouples as i an Input to saturation meter.

B. Provide the capability of having a continuous or trending display 4 of incore thermocouple output.

. 7.8 ' Provide a safety-grade Containment 3/GJB A recommendation was made to determine if there are sco- ,

high radiation signal to initiate narios in which our system might be Inadequate, and to

'c Containment vent and purge isota- Initiate changes as Indicated by the results of this i t lon. (lI.E.4.2). i study.

1

O O U LIST OF (DPCERNS 7.0 TEDESCD REPORT - BM REACTOR TRANSIENT TASK FORCE Peg, e COFCERNS REFEREtCE FRIORITY, ASSIG#4ENT, REcone4ENDATION 7.9 Plant operating and control f un c- t/3/ERK The Task Force has recommended that several items d tions should be modified to maintain rectly ref ated to this issue be investigated.

pressurizer level on-scale and pres-sure above HPI actuation setpoint (assuming no fa i l ures) . h4eeting these objectives should be Indepen-dont of all manual Operator actions.

7.10 Perform sensitivity studies of pos- 3/ERK The Task Force has ident i t led several items which have slble modifications which would re- the potential for mitigation of feedwater upsets.

duce the response of the OTSG to feedwater flow perturbations. (Con-g sider active and possive measures). '

1 ro (ll.E.5) 7.11 Modifications should be made, to the 3/ DAM All EPs and APs were review.d to determine what manual extent feasible, to reduce or elimi- Immediate actions are regulred. Automation of these

] nate manual imed iate actions for act ions Is to be analyzed under several Ta sk Force emergency procedures. (ll .E.5) Items.

7.12 Provide a qualified l&C Technician 1/PFM This is covered by item 5. 7, No further actlon is on duty with each shif t. (l.A.1.3) recommended.

7.13 Operator training on CR-3 event, as ' 1/ DAM Same as items 2.1 and 2.2 well as plant-speci fic loss of NNI/

ICS analysis and proced ures.

(l.C.5) 7.14 B&W develop generic guidelines for 1/ DAM A recommendation has been rede to develop a procedure, loss of NNI/ICS. (l.C.1). review EPs and APs to assure validity and upgrade Operator training.

7.15 Mandatory one-week simulator 1/PFM Th i s i s al read y done. No f urther action is required.

tralning for Operators as part of requalification program. ( B .A.3)

i

.JbT OF COPCERNS 7.0 TEDES00 REPORT - Baw REACTm TRANSIENT TASK FORCE Page 43 C0tCERNS REF EREtCE PRImiTY, ASSIG4 MENT, RECOMMENIAT10N i ,

7.16 Staf f evaluation of B&W RCP restart 1/ERK A recommendation was made to request B&W to orpeditiously criteria dur ing small break LOCA. release nodifications to the existing guldelines in this (ii.K.1) area which recognize the new subcooling limits. No f ur-ther action is required.

7.17 Staf f review alternative solution to 1/UJB This is addressed by items 5.1 and 5.8. No f urther ac-POR Y unreliability/ safety system tions were reconmended.

chal lenge rate of concerns. (l .C.1) 7.18 Expeditious completion of CR-3 1 REP i/T;PB Pb recommendations . were nede. No ftrther action is study: ( l l .C.1 ) req ui r ed .

g. A. Examine scope, methods, etc.,

g for rurther IREP work.

w B. Consider possible modifications to CR-3, based upon results.

7.19 Staf f develop plant performance crl- 3/GJB Several recmmendations were made regarding this Item. A teria for anticipated translents for recommendation was cede that ind ustr y take + lead to all light water reactors. (V.1) develop these criteria.

.7.20 Continue studles of need to trip RCP 3/ERK This is addressed in item 1.19.

during small break LOCA. (Conducted jointly by Industry and PRC) .

(ll.K.3) 7.21 Draf t NUREG-0667 recommends that the 1/GJB This capability already exists at CR-3. A r ecomendation location of EFW Into the OTSG (i.e., was nede that B&W evaluate the preferred injection path.

the auxillary feedwater nozzles) be No f urther action is required, reeva l uated . Specifically, to mini-mize the overcooling ef fect is re-commended for consideration.

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O ^1T^c""c"1 3 REVIEW DOCUMENTS TT1-1 An nunciator- Sys tem TT1-2 ICS and NNI Power One Line and Redundant Instrument Lay'ut TT1-3 Sumary of Reconvaendations from Review of Loss of Power to ICS, NNI-(X), and NNI-(Y) Events A3-1

l O ANNUNCIA10R svS1EM 111-1 1.0 ANNUNCIATOR SYSTEM PROBLEMS General discussion of the annunciator system (AS), led to identi-

fication'of the following problems

A. For the short time interval after a Reactor trip, the AS overwhelms the Operator with infonnation. Alarms sound and

windows light up at a rate faster than the Operator cen re-s pond. During the February 26th trip, the Operator received 957 alams in 150 minutes. ,

B. The infomation is not prioritized with respect to essential Reactor parameters.

[ C. It was found that many of the alarms are of no value to tne Operator in recovering from a transient.

D. Many of the alanns provide confirmatory (redundant) infoma-

- tion. A single event may trigger multiple alarms. In a transient situation, this is unnecessary.

E. This deluge of infonnation on the Operator creates a nui-sance, and makes it difficult to pick out the important alarms. The usefulness and efficiency of the system is re-duced to the point where the Operator often ignores the AS.

From this point on, the alanas become a nuisance.

F. During normal operation, where the Operator has time to re-s pond to individual alarms, the AS provides a usfeul and necessary function.

G. The hard copy printer is easily overwhelmed during a Paattor trip. It runs minetes behind and is of no use.

These problems fonned the basis for tha proposed modifica-tions to the AS.

2.0 SOLUTIONS AND MODIFICATIONS 1 Modifications which could be carried out in t!.ree timeframes were J

examined.

A. Modifications to be implemented before June 1st start-up.

B. Modifications to be completed - in a 12- to 18-month time period.

O C. ton 9-ter modificetions.

A3-2 i

_ _= _ _, _ . . -. . _ - .

4 O It was determined that modification of AS was not necessary prior to startup. The AS is not the primary source of any infomation, but only alerts the Operator to changes in plant conditions. The Operators are already accustomed to not rely on it during periods of high annunciator activity. It is, therefore, not necessary to restrain startup for AS modi fications.

2.1 -SHORT-TERM SOLUTIONS The only short-tem solution (i.e. , before June 1st) deemed feas-ible was to change the color of the annunciator windows. Alarms could be grouped by priority and assigned different colors. The overall effectiveness of this solution (including Operator famil-iarization) was not known and could not be determined by Therefore, we did not recommend it and did not pursue June 1st.

it further.

2.2 INTERMEDIATE-TERM SOLUTIONS The inteme%te solution was detemined to be rewiring of the annunciator ,.onel to praide a power cut-off to the noncritical alams (both window lights and horn). The power d1 ruption would be initiated upon trip and could be reinstated at the discretion of the Operator. Infomation to the hard copy printer would not change. This idea received most of the attention of the Task Team.

2.3 LONG-TERM SOLUTIONS Long-te m solutions potentially involve so many options that they were beyond the scope of this Task Team. These will likely in-volve purchasing new computer equipment and infomation process-ing hardware. This creates enumerable options and should be dealt with separately.

2.4 TASK TEAM EFFORT The Task Team proceeded to identi fy critical and noncritical alams. The intent was to provide the Operator only with criti-cal infomation shortly after a trip.

3.0 ELIMINATION OF ANNUNCIATOR ALARMS Alams were eliminated on the basis of their being in one of the following categories:

Category 1 - Those alams which are essential to either,

a. Cooling the core.

Mitigating transients.

O b.

4 A3-3 i

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. O c. Indicating conditionc .aich are hazardous to personnel .

Category 2 - a. Alarms for equipment protection.

, b. Confirmatory alarms (i.e., autostart).

c. Alams indicating any condition of a non-safety '

system. ,

d. Anticipatory alams (lube oil level low).

Category 3 - Alarms which the Operator has no control over and often simultaneously indicate in another location.

It was determined that Category 1 alams should not be changed.

Category 2 alarms should be eliminated after a trip. Category 3 a'ams should be eliminated entirely from the Control Room.

A copy of AP-102 was marked-up to indicate which alams fall into eaA category. Category assignments were made according .c the foi awing guidelines. The importance of the alarm after Reactor trip was the determining factor.

A. Alams which were primary indicators of essential system conditions. Primary indicators are level, flow, pressure, temperature. Essential systems are those involved with:

. reactivity control

. primary heat removal l . secondary heat removal

. fordwater flow

. electrical power, cooling water, or air to these systems

. containment overpressurization.

( B. High radiation alarms were Category 1.

i i C. Fire alarms were Category 1.

4 D. Equipment protective alams were Category 2. A few excep-tions were made for the turbine due to its large capital .

cost.

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E. All first-outs were assigned Category 1.

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F. Practically the entire P, Q, R DG-A, and DG-B panels were assigned to Category 1, even though this is a violation of other guidelines. This infonnation was considered to be

' useful after a trip, in addition to which, these alarms are not generally nuisance alanns. They don't nonnally alann after a trip. Numerous alarms on these boards would be an indication of trouble.

G. In instances where a single event causes multiple alarms, only the primary alam was put in Category 1.

H. Any alam references in an EP or AP was originally put in Category 1. This was later revised to eliminate the ones that didn't fall under one of the other guidelines.

4.0 EVALUATION OF RESULTS The effectiveness of this proposed modification was examined with respect to the February 26th transient. The annunciator printout for two hours after the trip was marked up, eliminating the Cate-gory 2 and 3 alarms. The reduction in alanns is shown in the following table. It is also interesting to note that the peaks of the total alanns do not correspond to the peaks of the essen-tial alanns.

5.0 RECOMMENNTIONS The AS Task Team recomends the following items:

A. A system be deployed to eliminate nonessential alarm 3 after

, a trip. This system should eliminate the horn and lights to i nonessential alanns. The hard copy printer should continue to print all alams. All alarms not marked in red in the attached copy of AP-102 are considered nonessential alams.

This systen should be designed to automatically activate on i Reactor trip. It shc ald not be possible to activate this system while the Reactor is at power. The Operator should

- have the capability to bring back the lights for intermit-tent periods on one panel at a time. A flashing light some-where in the Control Roan should be activated whenever the alanns are suppressed. The entire system should be deacti-vated by the Operator when he feels the Reactor is under control. This system should be implemented by June 1981.

B. No modifications to the hard copy printer are recomended.

C. No recomendations were made for changes to the CRTs.

D. No changes are recommended before June 1, except #8 and #9.

A3-5

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, . () E. Any changes to the annunciator System will require changes to the APs, EPs, and Operator training. Operators must be familiarized with the System before it is deployed.

F. Priority panels were not recommended. Rearranging of con-trols and panels was not recommended at this time. This will require wholesale changes to the Control Room and should be coordinated with new hardware purchases.

J The following two problems were found to be minor shortcom-ings of the window lights. The proposed modificatior.s can be easily effected and will slightly improve the effective-ness of the system. These should be implemented before

startup.

G. The flashrate of the annunciator light should be increased to facilitate detection by the Operator. The present flash-rate is often equal to the Operator's scan rate, thereby never giving him a flashing indication.

' H. The First-out color should be a darker / deeper red.

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SUMMARY

OF RECOMMENDATIONS FROM REVIEW 0F TT1-3 LOSS OF POWER TO ICS, NNI-(X) and NNI-(Y) EVENTS 1.0 Investigate steam line rupture matrix modification such that ma-trix actuation does not isolate emergency feedwater. This would improve response to undercooling events. Any side effects of this change on other transients should be assessed prior to mak-ing the change. (See Integrated List, Item 5-16.)

2.0 The 'CS 22% RC Pump start interlock fails to open to prevent RC Pump start on loss of power. On a Loss of ICS Powet Zvent which leads to HPI initiation and securing of RC Pumps, this could in-terfere with restoration of forced flow RCS cooling. The mode of operation of these interlock relays should be revised so that power failure of the ICS does not prevent pump restart. (See In-tegrated List, Item 3-4.)

3.0 The loss of power to the ICS causes the turbine bypass and atmos-pheric dump valves to fail halfway open. This results in a Small Steam Line Break Event. Actuation of the steam line rupture ma-trix isolates the turbine bypass valves, but there is no conveni-ent way to isolate the atmospheric dump valves. A change to pro-vide isolation, and, if possible, control of these valves should be provided. 'chemes

should include

A. Solenoid energized from ICS AC (DC?) power which dumps con-trol air on loss of power (valve fails closed on loss of air).

B. Put EMO controlled from Control Room on isolation valve in line with atmospheric dump valve.

C. Put manual station with signal generator from other power

! source downstream of ICS with relay to transfer to this man-ual station on loss of ICS power. (See Integrated List, Item 3-2.)

4.0 Condensate pump control is all NNI-(Y). Failure of NNI-(Y) leads ,

to a plant transient causing Reactor trip, after which the 50%

condensate flow is excessive. The existing DFT overflow line may or may not handle the excess condensate flow. Investigate chang-ing the condensate pump control. (See Integrated List, Item 3-16.)

! 5.0 It appears that there is a relay in the ICS control circuits for the FW main block valves and low load block valves, which will fail closed on loss of power and cause these valves to close on loss of ICS. This design should be verified. It should be de-l termined that this provides positive valve closure, and if not, I

(]~ appropriate circuit changes should be made. (See Integrated List, Item 3.6.)

A3-16 i

6.0 The loss of power event studies reinforce the need to modify the OTSG low level start of EFW so that it is not defeated by a sin-gle failure in ICS, NNI-(X) or NNI-(Y). (See Integrated List, Item 5-11.)

7.0 Loss of NNI-(X) or NNI-(Y) results in a fairly severe plant tran-sient. No readily apparent signal selection made will eliminate this transient, although its effects can be reduced. An objec-tive should be defined, both for short- and long-term. In the short-term, it appears that providing an (X) Th to the A FW Btu limit and a (Y) Th to the B FW Btu limit, in conjunction with other " preferred" selections so that the A OTSG is all (X)and B is all (Y) will help. In the long-term, other techniques should be studied. These could include:

A. High (or low) select between NNI-(X) and (Y) signals in the ICS to try to obtain a good signal.

B. Average signals from the NNI in the ICS.

C. Combination of the above, as is done on NI flux.

D. Use 2 of 3 auctioneer (on T ave and steam pressure).

E. Provih automatic transfer so loss of (Y) automatically picks O the (X) signal. (See Integrated List, Item 3-6.)

8.0 A complete review of the emergency feedwater pump start circuit should be made to assure that it is " robust" with respect to other failures. In particular, is the low level signal the only interface with the ICS/NNI, or are there others? For example, on ICS failure, does the loss of RC Pump signal cause the proper ac-tion?

9.0 NNI-(X) presently has two sets of DC power supplies which will be fed from separate power (Vital Bus 3C and Regulated Bus 3B).

Further, NNI-(X) is being further modified to add an ABT switch so that its AC loads can be supplied from either of these two sources. NNI-(Y) is only fed from the Vital Bus 3D; it has no alternate supplies for either its AC or DC. A change to the NNI-(Y) should be investigated to add dual diode autioneered DC power supplies fed from 2 sources (Vital Bus 3D and a Regulated bus) and an ABT for the NNI-(Y) AC loads, as is being done for NNI-(X). ,

1 10.0 Reopening the MSIVs can put considerable mechanical stress on the valve positioners and steam lines. The addition of bypass lines to equalize steam pressure prior to opening the MSIVs should be investigated.

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-EXPERIENCE T0-DATE DOCUMENT LIST

- 1. B&W Rx Trip Analysis & Response (All B&W Plants)

2. Equipment Deficiencies to be Identified (CR-3 Transient)

3. NUREG-0565, Review 'of B&W ECCS Systems

4. ICS - FMEA (B&W 1564) 6 Recommendations

5. CR-3 Reactor Transients Experience - ,

6. NRC Action Plan - NUREG-0660, Includes B&W Task Force and NUREG-0578, and -0585

7. FPC Corrective Action List (23 Items, including Sequence of Events of February 26, 1980, Transient)

8. Teknekron LER Statistical Report Operators and I&C Techs Concerns List (CR-3) 9.

! 10. NUREG-0560,-Tedesco Report - NRC Staff Analysis of FW Transients

11. B&W Response .o 10 CFR 50 '54 Order of 25 October 1979, Over-i cooling Response - Backlog Const. Contracts

~

12. FPC Action List Post-TMI-2 4

4 1 ~. . Engineering Schedule for CR-3 Shutdown Items - Corrective Actions for Completion Long-Tenn (One Year)

14. INP0/NSAC CR-3 Transient Task Force Report

15. Systens Interaction - FLUOR LER Review for Zion Station

16. I&E Circul r 79-22 s

17. Personal Files of Task Force Members Fron B&W

18. ESSEX Report - Human Factors List 1

19. Arkansas Unit 2 Loss of Inverters - ISE Circular 79-02

20. B&W CR-3 Short Term Recommendation List 15 Item List

21. - NSAC Summary of Instrument and Control Bus Power Failures

22. INP0 Review of Bus Failures A4-1

7- s s) 23. B&W CR-3 Transient Assessment Report

24. MPR Associates Report on CR-3 Transient of February 26, 1980

25. Oak Ridge Report on ICS Reliability Analysis

26. 'NUREG/CR-1219 Analysis of TMI-2 Accident and Alternatives Sequences

27.  !&E/TMI-2 Investigation Team Recommendations

28. List of NRC Concerns VIA Telephone Call (CR-3)

29. B&W OTSG Tube Rupture Accident Analysis

30. EPRI NP-1118 Vol.1 November 79 Human Factors Methods for Nuclear Control Room Design

31. FPC Loss of NNI Power Analysis

32. Summary of Trips - Crystal River 3, Oconee 1, Oconee 2, Oconee 3, Davis-Besse 1, Three-Mile Island I, and Rancho Seco

33. FPC Instrument Availability Analysis

34. FPC Loss of NNI-(X) or (Y) Analysis

35. FPC Analysis of Power Supplies for NNI Relays

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SUBJECT:

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G .* Y.?e/Q To: D. C. Poole DATE: March 1, 1980 I ORPsyg

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On February 28, 1980, a committee was established to investigate the NNI-X power supply failure associated with the February 26, 1980, transient.

The members of this committee are as follcws:

T. C. Luckehaus, FPC (Chairman)

J. L. March, Bailey R. P. Cunningham, FPC F. W. Pluebell, FPC R. E. Clauson, FPC R. M. Queenan, B & W R. P. Schmiedel, FPC M. Hunt, NRC

(

The committee was directed to investigate the following:

A. Determine whether the spray valve opened on loss of NNI-X power.

B. Deternine whether the PORV opened on loss of NNI-X power.

C. Determine the cause of the NNI-X power failure.

1. Evaluate whether AC power was lost or a voltage transient occurred on 120V AC Vital Bus 3C feeding NNI cabinet #2.

2. Review overcurrent protection on Lambda DC power supplies.

3. Review possible causes for loss of DC voltage on bus.

D. Determine whether alare 14:23:26 OTSG 1A pressure low was a vital alarm.

Comments and Assessments:

A. Evaluation of the spray valve opening on loss of NNI-X power supply (Reference Enclosure 1):

l Upon a loss of +24V DC power in the hWI channel X with -24V DC still g available, the signal monitor module starts to open the pressurizer (f spray valve. When the -24V DC finally is lost, the spray valve stops 9Ce 20%$)

i e- . . _ .

D. C. Poole March 1, 1980 y'%

t Ni moving. Due to attempts to reenergize the NNI-X power supplies by the technicians and with the appt rent short on the +24V DC supply holding voltage down while restoring the -24V DC monentarily, the valve was jogged open. The total valve opening was less than the 40% open limit as it did not indicate during the event.

When NNI power was restored, pressure was above 2205 psig and the spray valve opened. The valve eventually closed when pressure dropped below 2155 psig.

Module tests verified setpoints and function as designed.

B. Evaluation of the PORV opening on loss of the NNI-X power supply (Reference Enclosure 2):

Upon a loss of NNI-X +24V DC voltage on bus, the PORV opened and sealed in.

Upon loss of the remaining NNI-X -24V DC supply, control circuitry is de-signed to keep the PORV open until RC pressure drops below 2380 psig. The NNI power monitor removed all NNI channel X power in less than 1/2 second.

Upon loss of the remaining power, the PORV remained opea until the NNI power was restored and the RC pressure dropped below 2380 psig.

Module tests verified setpoints and function as designed.

C. (1.) Evaluation of loss or voltage transient on the 120 volt AC vital bus

3C feed (Reference Enclosure 3, Dwg. SS-201-063 VB-S)

Enclosure 3 lists the systems which are powered by VBDP-5. This panel is the only load on a 25 KVA 120VAC single phase dual input inverter.

During the incident, a " clear"" inverter 3C AC out under voltage" was received at 14:25:01. This relay clears at ll4VAC. A " bus dead" alarm relay is mounted above the 120V vital bus panel. Product lit-erature indicates an alarm will occur at 80% or approximately 96VAC.

Testing performed 3/1/80 showed actual alarm would have occurred at 49.7VAC. This alarm was not received. All feeder circuits are pro-tected by circuit breakers and in series with fast acting shawmut form 101 fuses. None of these fuses or breakers were affected.

A test performed in NNI-X power supply referenced in C(3) proved on a loss of AC supply that upon restoration, power would be restored.

C. (2.) Evaluation of the overcurrent protection on the Lambda DC power supplies (Reference NNI/IC.C Instruction Book Volume IC, Tab 19):

In reviewing the instruction manual for the Lambda supplies, it was det- l ermined the input fuse (Sloblow F1) protects the AC input circuit. The overload of the DC output does not cause fuse failure. Since output of the supplies after clearing the fault was verified, no AC input fuses were blown.

(') An automatic current-limiting circuit limits the output current to approx-imately 110% of the 40*C rated current of 14.0 amps. In the short circuit l l

l

I .

D.'C.'Poole March 1, 1980 condition, the voltage and current decrease until current -is approx-imately,30% or less of rated current. Thus, the output of the supply

~

is protected upon shorting of the bus.

C. (3.) Evaluation of possible causes'for loss of DC voltage on bus:

A performance test procedure, PT-454, was developed to examine the NNI-t X Power Supply System. The following tests have been performed to date:

l l1. Verify auctioneering diodes and power supply capability to carry bus voltage-

2. Determined. AC voltage level to produce 22V DC on bus which trips power supply monitor - 86.6VAC

3. Test to determine if 51 and S2 could be closed with load on bus and power. supply monitor in service. Si and S2 time delay .5 seconds.

4. Interrupted AC power to Si and S2 at VBDP to determine power will

restore without tripping Si and S 2-

5. Visual Inspection ~and testing of all 820 modules in NNI-X with 24VDC relays.

Based on satisfactory testing of items 1 through 4, the most probable cause of +24VDC loss is a shorted component which has burned clear.

Item 5 testing is designed to locate this particular failed component.

Item 5 is continuing and.will require additional time to complete.

D. Evaluation of 14:23:26 OTSG 1A pressure low alarm (Reference NNI Instruc-tion Boo'i Volume 3, Bailey Dwg. D8034034C):

Steam generator outlet pressure loop A is alarm point 5308 fed from SP-6A-PTl per point identification summary.

The SP-6A-PTl string is all X power including the output voltage buff r in NNI-4-6-1. Since the voltage buffer output goes to mid-scale on loss of all DC power (600 psig) this would give alarm at 637 psig. Estimated time for signal to fall to 0 volts (600 psig) is less than 2 seconds.

• D  %

T. C. Lutkehaus

Technical Services Superintendent TCL/tb enclosures.(4)

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CASCACE' 53-C20M .

O Tirne , Evgn t_ g 14:23:00 RC Flow "A" 73 X 10 lbs/hr D (PIR) 6 (continned) RL Flow "it" 73 X 10 lbs/hr Letdown Flow 48 npm (Yl SG "A" Ivl (OP) 67%

01SG "B" lvl (OP) 65%

OTSG "A" FRLV 242 inches OTSG "B" FRIN 254 inches OfSC "A" Presuure 911 psig OTSG "B" pressare 909 psig Main Steam Prcesure 894 psig a

Main Steam Tor:p. 589 F.'

Co olenser Vacuum 1.76 psia V

Ctnerated KJ 834 1'

UFT level 12.7 ft.

i Feed Flow "A" S X 10 lbs/hr Feed Flow "B" S X 106 ;.bs/hr F.ed Pressure "A" 970 poig Feed Pressure "B" 968 psig; 14:23:21 4- 24V NNI-X BUS Fall (ANN)

+ 24V NNI-X Ph'R SUPP FAIL The cause of t his in i t int ing event was mi sal ignment between pins on the 4710 buf fer module and the T-SAT buf fer card which had been installed Fchruary 15, 1980. This pin misalignment re-Ii, sulted in a short on the positive 24 Vlic bus which dr.o;ged the i

)

4 i

I 4

(

PS-2' l CR-3 1RANSlENT OF 26 February 1980 h SI.Q111 NCE OF INI NTS l.Vl.N f S Y..N.o!'S I S.

.1ust prior to the transient, the reactor was operating at a p p r a< i ma t el y 100%

f ull pcwer wit h Int erya ted Cont rol Svsten (ICS) in autv utic. No tests were in prog re :s and minor n..inti nance v:n; being per f onwd in the Nen-Norlear Instrumentation (NNI) cabinet "Y".

At 14 : 2 3 on l'eb rua ry 26, 1980 Crystal R ive r-3 Nuclear St at ion c'-;.c. ient ed a rs act or t rip f rom approxirat ely 103% full power. The fcllowing synopsis of key events, the record 3.ource ( ), and plant parameters was obt.nined from the plant computer's Post-Trip Revieu (PTR) and the Sequence of Events Manitor v

(cmp ) , the Plant Alarm Summary (ANN), control room strip charts, the Ehift a

Supervisor's log, and Operator statem.nts.

Ti tac Event 14:23:00 T11 e following is a soinary of plant cont'i t i on s prior to the trip (PTR)

Flux 98.6%

rtC Pressure 2157 psig 0 PZR level 202 inches MU tank level 71 inches Tg "A" 599"F, Tg "P" 600 l'.

T C

• * ,I TC ,,g., 5 F.

k C

I D""D 'T

1. J '

Time Event

- 14:23:21 bus voltage down to a low voltage trip condition. This

• (ANN) -

(crentinued) caused AC power supply breakers S-1 and S-2 to open 1/4 to 1/2 necond af ter t he initiating event due to a built in time delay on the breakers. The tripping of S-1 and S-2 breakers was not recorded by the annunciator printout because f ailur e of t he printout i

from 14:23:21 t o 14:24:36. In addi t f on, there wc> no trip indication on the negat Ive 24 VDC bus.

23:21 PRFSS E1 EC'l RO:MT IC RLF VIS RCV-10 OPEN y,'N )

hn the positive 24 VI)C was lost due to the se pn-nre d is-cussed abovr- the signal nonit ors I, NN1 changpd st at e canning i

PURV and Pressori zer Spray valves t o op<:n. The PURV cireuitry is designed to seal in upon actuation and did so. The sub-sequent loss of the negative 24 VDC halted spray valve motor V operation and prevented PORV seal in from clearing on low i l' r e s s u r e . It is post ulated that the PORV opened fully and the spray valve stroked for approxtmately 1/2 second. The 40%

open indication on spray valve did not actuate, therefore, a L1e spray zalve did not exceed 40% open.

14:23:21 SC A 05 BTU LIM (ANN)

PL0fMA'irtl 1M OKji_I

)

RC TOTAL F1O 1.OW f

RC PZR LVL LOW The NN1(X) Foser Supply Failure resulted in many erroneous

) signals to the prinary plant control system. Teold failed to 570 F (normal indication was 557 F) producing several

/

7, l l

t I

(

g T,i nte_ ,rve n t_

14123:21 spurious alarms. Tave failed to 570 F (decreared). The (ANN)

(continu"d) resultant Tave error modifftd the react or demand such that control reds were witbdrawn to increase Tave and reactor

j. iwe r . hi- power loca .e.c was t ea r;i na #il at 10'C by t in- ICS a nil a "R. .ic t o r De :n d fl i gh I. i u. i t. " a l . r ni w. i s r .eived. Tiot i

failed to S/0 F (low) and RC flow failed to 40 X 10 l b:, /h r in each loop (low), r.ot h t hese failures cri a t "J i in U al .ir m and limit on f e edwat er which redu ed fe.uheat e r flow '2 both OTSC's to essentially zero. Turbine Ib..ider Pri.ssure indication failed to 900 psig (high) which caused the tarhine valses to ec>en clightly to regulerte header pressure thus increasing gtnerated negawatts. Thcse coc.hined failures resulted in V a loss of reactor heat sinks causing an excessively high RC pressure.

ii In addition pressurizer level li.dication failed to 160 resulting in increased makeup flow rate.

14:23:35 H QCTOR PROT. Cl!ANNEL A TRIP (COMP)

RFACTOR PROT. CllANNET. S TRIP RFACTOR PRor CliANNEI. C TRIP REACTOR PROT CliANNEL D TRIP CRD TRIP CONFIRMED Tl! RHINE ON ICS CONTROL (No)

AUTO Ir! OP (ll lit PJ

,m Tl!RhlNE ON COV VAIXE CONTROI (No)

e

's

(

(~

1

)

Event; Q II "'e.

14't 21: 35 React or t rip caused by high HCS pressure at 2300 psig.

(C,,n.p)

(ront inued) Turbine was t r ipped by t he TCS in respom;e to the Peactor t rip.

Pressure pc d.ed at 2330 psig as recorded by RC pri- sure ';a r r ow Range 1oop B charge recorder, s 14:24:02 . l_ii. Gli Pr:1.SSU R F._ i t. l EC'_l I ON R I.QU I R ED o

This was a comput er printout and indicates <50 sub-cooling. See att nzbed 1;raph of RC Prensu re/Ter.p . vs.

Time (Figure 1). This graph is based on Post Trip d.ita and actual incore therr.oco @le data. From the react. trip point (14:23) to 14:33, core exit t e:rperat ure data was V obtained by ext rapolation and calculated dat a. This is supported by two alarm data points plotted at 18 and 21 of subcooling during this period from the computer. It should be noted that the lowest average subcouling margin was prob.ibly 42 F for a very short period of tir.e.

• NOTE: This comput er program was initiat ed as a result .

of the TMI-2 incident.

14:24:02 I)E3 EPMOR 1.EVEh (PT)

(COMP)

Condensate pump tripped due to high DFT level. This is verified by "????" printed by the computer, indicaling the level instrua,ent was over ranged. In addit ion, a low flow indication in the gland tream condenser was also recorded by the c or..p u t e r a t 14:24:25.

'9, 4

~

• T "T))

D"h) w o Ju.2 N $

( .

y Ti.me- .E_ve_n_ t-14S25: 50 RC, DR TNK 1NI,111 (A:=N)

The high RC Drain T. ink lev '. al.irm was po:;i t ive i nd i r.it i on that the PORV h.> opened.

14:26 PORV HI OCK YAl.VE Cl.USED

~ ~ ~ ~ ~ ' ~ ~ ' ' '

(Ope r.it or actfor - t i ne Ope n ator at at ed that hKY Block Val ve (RC-11) w.:s closed due s

appr( i n.a t e )

' to .lecreasing RCS pressure and bi h RC drain tank level.

14:26:41 .C" N.I. R C - 2 1MiO. psi -i R_I_P M.T../_5 r '1.N_L_P_

14:26:48 r ilNT. _RC- 3 15 00 P.S I 'I R_I.P_ ",i_/.ST 1 R I P_

(

(ANN) i:S ACT A itP IN.1 ON

.E_S AC. T. .U H P T NJ .O_N._

14:26: 43 ttV I UP PL"'P A (HUN) v (Co:!P)

.!').P L_U P.. _P_U_M_.P.

._ .- . .C .0. " _. . '_._

llP1 ini t iat ed au t oinat ically Joe t o low HCS pr:rsure of 1500 psig. The low pressure coiali tion resulted from the PORV .

rerai.ning open af ter the plant was tripped. l'ull HPI was irii tia ted wi t h 3 pumps resulting in appr oxivat ely 1100 gpm flow to the RCS. At this time, all remaining non-essential R.B. isolat ion valves were closed it, accordance with TMI 1.c.ssons 1

1. earned Guidelines and approved procedures.

14.26:54 _RC_l'ly.1P A l (STOP)

(COMP)

)

RC Pl'MP B2 (STOP) a (' 14:26:57 RC PUMP B1 (ST01_')

(COMP)

MC PUMP A2 (S' LOP)

M s

C

I .

I ss Time. Ev . .n t.

y 14=: 26:5I. Op .r.,t or t urned KC pumps of f as required by the applicable (CO !P) emergency procedure and B A W sn il1 break guidelines.

14:26:57 (COMP)

(continued) 14:27:20 R B P N I .R.U R.1: -! :;c u.O s i NG.

(PfR)

This is the first indication .nat the RCDT rupture disc had rupt ured RB pressure increase data was obtained from Post Trip Ptview and Strip Chart indication.

14: 30 COMMENCED F II.l. I NG OTSG "B" (CHARf)

(Time np p r oxi:u t e) -

14: 31:32 .R. B _r..n S S_ _A f 2__ _P S I_G_ _

v (ANN)

This alarm was initiated by 2 psig in RB. This is attributed to steam release fi ca RCDT. Code cafetics had not opened at this time based upon tail pipe temperatures recorded at 14: 31:53 and 14:31:58 (COMP).

14:31:49 FWP A Sl1CTION VALVE UPEN (No)

(COMP)

At this t ime OTSG "A" Rupture Mat rix had act uat ed.

Thi s occurred due to <600 pai g on OTSG "A. " The low pres-sure was caused by less of water level in OTSG "A" because of the 1;TU limit and f aul ty OTSG level t ransmit ter signal .

This rup t ure ma t r ix t r i p resul t ed in the closure of all

- feedwater and steam block valves which service OTSG "'" and C canned main Fecdwater Pump 1A to trip.

11 C

l l

\

l

( ')

T i m. _e_ Event _

l 1%': 32 .ES A]B , BYPASS to 14:41 The operat or st at ed that ES was bypassed and llPI balanced am.mg (operator action-time all~4 nozzles (t ot al flow approxin:at ely 1100 npm) per sm: ell approximate) b rt ak guidelines. Actual flow was not balanced due to failure of 2 flow indicators to the mid l'osition. It is not clear at this tiiie whether or not this act. ion had any acpreciable effect on total llPI fl ow to the l<C syst em since flow in the two lines which were represented by failed flow meters lirob-ably increasc d subst ant f ally of f set ting, any reduct ion in the two metered lines.

14:32:35 CHEt1 Aln) PP 1.1 P-- 3 AU 10 OP

( Ar.N )

The s t eam d riven Eme rgency Feedwat er Pun.p was st arted a t v

this tiw to ensure feedwater was available to feed OTSG's.

14:33 STAR'I ED t10 TOR DR IVEN EMhRdi:SCY Fl.t'.DWATER PUMP to 14:44 The Motor Driven Ewrnency Teedwat er Pump was start ed at this -

(operator ,

action-time time to ensure feedwater was available to feed OTFC's, approxinate)

~

14:33:01 RC.,0UTI.If[ A PRIl.SS. (Ny M)'n')

I (Cot 1P)

The core exit incore thermocouples indicated the highest core f

outlet temperature value was 560 F. RCS pressure was 2361 psig o

at this time, therefore, the subcooling margin was 100 F.

I Mininum subcooline, margin for the entire transient was approx i ma t e ly 2,o.P. It is post olated that some localized 1, oiling a -

occurred in the core at this point.

'12

(

u_ _ _ _ _ _ _ _ _ _

/

~**D D *D T]

~, 66 o .S.. 6 i

Ti.me.

-- - -Ev-

.e n.t.

14:13:01 (cm1P)

(Continued) The pre ;o r i wr is " solid" and code safety valve (RCV-3) 1ifts. This is the highest PCS prec.sure recorded by the r o::.p u t e r .

14:34:43 JtP, Djp1E M.Il ifVL 111, 101C- 19 a l a rr..ed a t this point. lii ghes t level indicated during tourne of incident was 60 R/hr. Hi>;h r.diation levels in RB caur.ed by release of nan-condensable p.res and cool-ant from the pressurizer into the RB VTA PORV, RCV-8 and the RC drain tank.

14:35:33 .* -24. v NNI-X_ Rl'S FA I L (ANN) v

• _24 V NN T-X PI'R FJiPP FAIL 14:35:34 -24 V NNT-X HilS FAIL (ANN)

_ 24_ Yl'iI-X. ._13iR _SUPli FA I,I; At this time an a ttempt was 1..ade to repewer NNI-X without su? cess. " Spikes" were also observed on affected strip chart recorders.

14: 37:36 .CO:jl'_lir ER _ove r I oild to 15:11:44 Caused by overload of buf fer. This resulted in loss of (COMP) data stored in buffer during this time and until printer catc hes up with data input.

9 13

om mi ~Q'

$c o NL d :3 Time- .I?v.e n t-14t38:'S iSOL VLV LOOP B ITV -14 Cl.SD - EF REQ This valve was cle ed to prevent ove r f eed i ng Ol SG "t$"

beyond 100% indicated Operating Range.

14:44:11 *

.1 24 V ';NI X PWR SliPP. Fall.

14: 44:12 * -24 V ';NI-X PWR SllPP FAIL (ANN)

..* -24V NNI-X B!)S. F.AI.L.-

• +24V NN1-X P.itS FAIL NNI was restored by reriovint; the X-NNI Power Supply % nitor

%dule. This allowed the breakers to be reclosed. Is t this time, it was ohnersed that OfSG "A" was below the indicating range, the pce,surizer was " solid" (Indicated v off scale high), RC outlet temperature indicated 556"F (Loop A & B average), and RC average t en.pe ra t u re indi-cated 532 F (Loop A & B). The higehst core exit thermo-couple temperature at this tine was 531"F. RCS pressure

• was 2400 psig (saturation t e rap . at this pressure is 662"F.).

Thi s da t a ye ri f i_ed_ na.t u_ral c_i rcula t i on uns _in p ro' ress and the plant subcooling margin was 131"F. (based on core exit thermocouples). l l

l 14:44:31 ES ACT B RB ISOL ON (ANN)

At this tIve, HB pressure increased to 4 psig and initiated RB Isolat ion. The operator veri fled al1 iuiced late ac tIons occurred properly for llPI, LPI, and RB Isolation and Cooling.

The increasing HB pressure resul t ed f rom RCV--8 r_e lean _ing reactor coolant made up by the llPT.

14 e

1

,w y T I tiie l'.vi n t_

14 $ 4f> : 10

• ES ACT A Ril 1501, ON

~~ ~ "

(ANN)

Operat or hypac. sed llPI .1.PI, and RB Tsol;it lon and cooling.

The .e "FS yst.' , ui s e hyp.esed to again hal.nne lie t 11 w and re, tore t ool ing wat er toe 'antial .io v i l i a ry opii p-m.z n t (i.e., RCP's, letdown coolerc, C R ici's etc.). At this tiae the operator ha8 valid indication on all 4 ilPI flow p.i t hs .

)

14:51:57 FW SYS ACT B 'IRIP

(.i ';N)

The actuation of the Rupture Matrix re,ulted from a d e;; r a d a-ti.in of OfSB "B" preisnIe.

Cold emerg( ncy fced was being injected into OTSG "B". This

r. atrix actuation isolated all ferdwater and c,t t am block valves

)

t o O f SG "B" and t ripped the "B" rnain 147 p ump . Bot h Euergency FW pumps were already in operation. OTSG "B" level at this ,

ti.ne was 70% (Opesar.ing nange).

14:42 liPI iliROTTI.ED (Operator a t. t.i on- t i me The ope ator thi ot t led llPI and reduced RCS precsiire t o 2300 approximate)

At this time, the raaxinum core exit t he r.no c ou p l e temperature was 515 F, RCS pres ure was 2390 psig. Therefore, tl.e sub-cooling n.atnin was 147 F. Natural circulation was in effect. l

) ,- All prere<[uis i t e conditions had 1.een <trisfied, flow was j l

\

'15 c

1

'D P *]D

• T)) ' S A i We J\\ .o Ju A

.m '

T i m_e _Ev e. n.t-y 14': 52 throttled to approx.imately 250 gpm to reduce RCS pressure and (opt rator ac t i on-t irw to reduce the flow rate t hrough RCV--8 .md in t o the RB.

appros iant e)

((ontinued) 14:53 PE .EST.N!il I SH 1.1.'l u e.fij (operator action-tir.e At thi s t i:re , the ope cator was est ablish int; HCS pre sure a pp r o.:i t.a t e )

centrol via normal RC nal.c u p an d l e t d own .

14:56 OPND NU PPnP lad:lRC. val.VES (operator action-time This was done to as.ure the MU pumps would have riinimum appro.sinate) flow .>i al1 tines to prevent pump damage.

14:56:43 FW SYS ACT A BYP I.0T FJ SF.T (ANN)

• N(*RM SH!n DOWN FW SYS ACT A NOT BYP v

• FW SYS ACT A 'lRIP The operat or bypaved O rSG "A" Rup t ure ?! s t r i x and

^

re-established feed to the OTSG "A".

Feed.ea ter was slowly ashoi t t ed t o O'ISC "A" .chich was indicating low level off scale to this point. Feedwater was admitted through the Auxiliary FW l.cader via the EFW -

I t bypass valves. The feedrate was very slow in or<ler to minimize thernal shock to the OTSG and depressurization of the RCS.

RCS pressure control was unstable at this time.

t v

0

'16

a t 1

'm/-

T i.ne.

- .E.ven t.

14<57:09

• NORM SilVTDOWN FW SYS ACT E NOT hYP l'W SYS ACT B 101' N01 H l: SET

• FW SY S .^ T h 1 k l P The opetator hyp.insed OlSG "B" Rupture M.strix.

This was de:m to regain FW control of inSG "B". l. vel

v. i s s t i l l b i <;h in this OTSG (approxira t e ly 65",, ope rat ing T.m n:;e ) . Tinretore, fe.d was not necessary at this time.

The Main Steam Isolation valves were open in pieparation for bypass valve operation (when nece ary).

14:57: 15 MUV NOT FUI.L OPEN. ..

--.-253 - . . .- . - .

-RC . - PP - - .3A . CO:wT

. - - . _ SLI:ED . . _ . . OFF

- . . -111 14:47:53 A 14:59:03

• RC PP SEAL TOTAL Fl.0W l.0W The operator re-establistied RC Pemp Seal return.

This was done in preparation for a RCP start and t o mini:ai /c pump seal degradation.

15:00:09 SG 3A LVL 1.UW/!OW (ANN)

This verified feedwater was being admitted to OTSG "A" and made it available for core cooling via natural circulation.

Feed to this generator was continued with the intent of I l

proceeding to 957. on the Operating Range.

g

, i 3 /

Ms 4

I j O

Time. Event _

15:(?:O') '

At this time ll.e re w.is //"F of fnbeooling in loop "A".

(?.NN)

(continued)

This value was ha.<d upon "A" RCS loop par.nwt e rs. The "A" '160p was bei ng cooled down by OT4G "A" fi ll . nd the ope r.i t o r w.is a t i . ,pt i nn, t o equ.il i; e 1oop ti.pesatio ">

15:15 .NATUR AL C1 krtilai t on v.i.v l F i ED.-. .-

approxiuate) o At this time there was a 23 F Del t a- T and loop t e"pera t ur es were nearing citual',.arion. This Delta-T was ca l c u l .i t ed 1..m ed upon leop A 6 H T-cold's and core exit thetuocouple readions.

15:15 MANNED Tile TECHNICAL SUPPORT CENTER (T i :..e -

approxitiate) At this tine it uas evident that of f-si te personnel should he mw!c ;.wu.e of the transient that CR-3 had undergone and v

informed of the enrrent situation.

15:17 CLASS "B" EMERGENCY DFC1.ARED (time-

• approximate) A Class "B" emergency was declared as a inec..o'tionary raauure, even though ennditions for a Class "b" I:ve rgency we re not met as specified .in E'i-203, Revi sion 6, page 6, paragraph 7.3.1, (i . e . , reactor building sprays had not initfated).

This was done b.ead on the fact there was a loss of coolant through RCV-8 in the c o n t a i n;..e n t and llPI had been initiated.

All non-essential CI:# 3 personnel were directed to evacuate and contact of off-site agencies began. Survey team was

( sent to the Auxiliary Unilding.

18

(

99' y- 'S' D g 6e @ . .- :3 s 1, T iine Q F/. . n t_

15:19:22 .* I SOI. VI.V l .ouP B RN- 3 4. C. _l.S. D_ _ .- l i'11. Q.

(GN)

At this point ol SG "A" I t vel w :s inir.aning and the decision wen c.nle ta cei! nce f i 11 i ng 01 SG "13" s i mo! t :nn-ons1y. The intent was to:o 9 3 /, on bot h OTSC'- w i l lo ,o t . ve, . ling Hr:S cooldown li, it s (100" F/hr) whi le mai nt a ini rg; iwS pi . mure control.

15:?6:38 50!illH l1Y!H:0.*.J UL T.NR I.VL I.OW This rennited f rom t he tank supply valve open i ng w!n:n the 4 psig RB isolat ion and cooling ;ignal act uat ed. The sodium hydroxide was releaned ,to both 1.PI trains. Sodium l!y d ro x iile was admitted to the RCS via liPI from tlie hWST. (Approxi-rna t ely 2 ppm injected into the RCS.)

15:28:52 Ct':!PUTER overload to 16:26:56 Caused by overload of buffer. Resultinn in lors of (CO'1P) ,

con.puter dat a unt il printer catches up wi th data input.

15:50 T1.r<MINATED hPI (operator action .ime At this tire, all tenditions had been satisfied (per sr.all appru. .at e) bn ;ik operat ing guidelines) to ter ainat e llPI . FCS pressure cont rol had he.en est ablished using normal inakeup and letdown.

IlPI was it r:.,i na t ed and n'.ino rmal releaset. to the RR were discontinued.

%19

(

'l D""D *D ~3 A

r, Oo @ . . a l

1, Tim.e- Event 16:00 CO.MMrNCED PRESSURIZFR liFAIUP (ope t .it or ac t i on--l i ne At t h i n t i .ne , RCS pre- : ore imd t empei at ure were under cont rol .

ipp ros I. .it e)

N it on al c i r( ulatli.n was functioning as J. ;igurd (appi o . i m.it e l y 23 F delta-T). RCS t. i pe ra t ure as being :a i n t a i ord a t appr<wi :itely 450 . RCS pressure was approx i..at ely 2300 psig. The decision v,s n.a d e to eve -ence pre -;urim r in atup in preparation t o re-est ahl ish a st eam :,p. ice in the prossuri wr.

16:07 SURVI:Y TEAM REPORr (fine

.pproximate) The Friergency Survey Team reported that all offsite radiation survey r.sults were backgrriund, v 16:08:04 *1URh. E.M F RG . FW PUMP Fall ED 'IO START

( AN'.')

• Tim B . E." E RG . FW PUMP AUTO STARTED EMERG FEED PP TURB S'J M SUPP NOT READY The oper at or shut down the Steam Driven hFW pump. .

Tire not or d r iven E.ae rgency FW pursp was running. Therefore, the steam driven pump was not needed. The p l .m t rcu ained in this condition for approximat ely 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, while best irg up t he pressuriz' to saturation ter'perature for 1800 psig.

16:15 PRESS REl, EASE (Time ap p r o x i .na t e) Media was notified of plant status.

I

'20 C

J

t L. _ , T I .;c. Fv e n.t-18=: 05 f.STAHl.1 SIWD STFM1 SPACE I N ME",Slik 171'.R

( i . : .e

.s pp ro v. li .. i t e) At this point, p re. .o r t e e r t . o.:pi r a t u re s.'a s app ro:d o.a t e i y 620 F. Pre ;ur i zer 1. ve1 was et urned Io "on scale" by i n r i c a s i n:; letdown. Pre suri/er level w.<, then reduced to n o r, .:e l ( p. l .i t I n y, level ined norical pre- .o re eoutiol was i:stabliclied via pressur i xer lu at ers.

18:30 T hiell. N. :

..F D C! %SS . "li" W RC'NCY. r(c;fil T 10N 1: (t i rre a pp rox i ra.s t e) S t a t e .nid F. de ra l :qcu c i es we re not i f i ed .

21:07 Fuor. i:D. F.1.0W I N I 1 I A'l ED . I N . .R.C.S

( t irie

p p r t. >. i a.a t e ) The de c i s i on was c..ule tu re -es t abl i sh f orn d f l ow cool iny, in the RCS at this t i e.e . E!.W im d ';H C w e r e tonsulted.

se RCP-1B sind ID were started. At this point, krS p.iriin.e t e rs were stablited and maintainted at RC prersure-?OOO psir,, RCS t erapera t ure-420 F. Pressurfeer icvel-235 inches.

The plint was cons i de red in a noi r.a 1 con f i:;u r.,1 i on .

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'21 C

6 ATTACHMENT 6 0

, g LIST OF ABBREVIATIONS t

ABT Automatic Bus Transfer AC Alternating Current' l ADV Atmospheric Dump Valve AFW Auxilf ary Feedwater (Emergency Feedwater)

AI Administrative Instruction

. AP Abr nmal Procedure AS Annuciator System AT0G Anticipated Transients Operating Guidelines

ATWS Anticipated Transient Without Scram BMCo Bailey Meter Company

. BOP Balance of Plant

. Btu British Thermal Unit  !

BWST. Borated , Water Storage Tank B&W Babcock & Wilcox Company CD Condensate CPCo Consumers Power Company CR0 Contr:1 Rod Drive CRDCS Control Rod Drive Control System 3 CRDM Control Rod Drive Mechanism CRT Cathode Ray Tube CR-3 Crystal River Unit 3

C/A Corrective Action DC Direct Current 2 DFT Deaerating Feedwater Tank (Deaerator)

DH Decay Heat ECCS Emergency Core Cooling System EFP Emergency Feedwater Punp EFW Emergency Feedwater EHC Electro-Hydraulic Controller EM0 Electro-Magnetic Operator EP Emergency Procedure EPRI Electric Power Research Institute ESFAS Engineered Safety Features Actuation System i FMEA Failure Modes and Effects Analysis FPC Florida Power Corporation FSAR Final Safety Analysis Report FW Feedwater i FWV Feedwater Valve t

GPM Gallons Per Minute

HPI .High Pressure Injection IE, I&E Inspection and Enforcement Division (NRC)

~

ICS , integrated Control System INP0 Institute of Nuclear Power Operations 1

IREP Interim Reliability Evaluation Program 1 I&C Instrunent and Control l l KVA Kilovolt Ampere l l KW Kil owatt lh 1

LER Licensee Event Report i A6-1 l

l

I LOCA Loss of Coolant Accident LOFW Loss of Feedwater-LOOP Loss of Offsite Power LPI Low Pressure Injection MAR Modification Approval Record MFW Main Feedwater MFWP Main Feedwater Pump Main Steam Isolation Valve MSIV MSR Moisture Separator Reheater MSSV Main Steam Safety Valve MU Makeup MUV Makeup Valve Na0H Sodium Hydroxide NDT Nil Ductility Temperature NGRC Nuclear General Review Committee NI Nuclear Instrumentation NNI Non-Nuclear Instrumentation NRC Nuclear Regulatory Commission NSAC Nuclear Safety Analysis Center NSCCCW Nuclear Services Closed Cycle Cooling Water NSSS Nuclear Steam Supply System OTA Operations Technical Advisor 0TSG Once Through Steam Generator

PAS Probabilistic Analysis Section (NRC)

PORV Power Operated Relief Valve

- PSA Primary / Secondary Auxiliaries PSI Pounds Per Square Inch PSIG Pounds Per Square Inch Gauge PWR Pressurized Water Reactor PZR Pressurizer i P/T Pressure / Temperature  :

RB Reactor Building  ;

j' RC Reactor Coolant '

~

RCDT Reactor Coolant Drain Tank RCP Ret ' tor Coolant Pump RCS Reactor Coolant System I

. RCV Reactor Coolant Safety Valve i REI Request for Engineering Information RPS Reactor Protection System l RTD Resistance Temperature Detector RX Reactor SG Steam Generator SLB Steam Line Break i SLRM Steam Line Rupture Matrix SMUD Sacramento Municipal Utilty District SU Startup SUFW Startup Feedwater

' Tave Reactor Coolant Average Temperature

'TBV Turbine Bypass Valve Tc Reactor Coolant Cold Leg Temperature TDFWP Turbine Driven Feedwater Pump I

+

A6-2.

i 1

- - - - - - - - . . - . . . . - , ~

Th , Tat h

Reactor Coolant Hot Leg Temperature TMI, TMI-2 Three Mile Island Unit 2 Tsat Reactor Coolant Saturation Temperature T/C Thermocouple V Volts VAC Volts Alternating Curent VBDP Vital Bus Distribution Panel VDC Volts Direct Current g~ c,

's A6-3