ML19323A932
| ML19323A932 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 04/24/1980 |
| From: | Hanrahan E NRC OFFICE OF POLICY EVALUATIONS (OPE) |
| To: | Ahearne J, Gilinsky V, Kennedy R NRC COMMISSION (OCM) |
| Shared Package | |
| ML19323A931 | List: |
| References | |
| NUDOCS 8005070355 | |
| Download: ML19323A932 (31) | |
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NUCLEAR REGULATORY COMMISSION E
WASHINGTON, D. C. 20555 s
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- +k April 24, 1980 MEMORANDUM FOR:
Chairman Ahearne Commissioner Gilinsky Commissioner Kennedy Commissioner Hendrie Commissioner Bradford FROM:
EdwardJ.Hanraha[
SUBJECT:
OPE EVALUATION OF tie IMPACT OF POST TMI-2 NRC REQUIREMENTS ON B&W REACTORS AND THE CRYSTAL RIVER TRANSIENT I.
INTRODUCTION On February 28, the Commission directed the Office of Policy Evaluation to make an independent evaluation of the impact on B&W reactors and on the Crystal River event of the post-TMI-2 requirements.
In making its appraisal, OPE considered all of the post-TMI-2 requirements -- including the Short Term Lessons Learned Task Force (NUREG-0585) as well as bulletins, orders, and letters related to the TMI-2 accident.
In the course of performing this evaluation, we had discussions and meetings with the NRC staff and B&W engineering staff. On a site visit, we also met with Florida Power at the Crystal River Unit and Duke Power at their headquarters in Charlotte.
We received substantial technical assistance from Dr. M. A.
Schultz, a professor in the Department of Nuclear Engineering at Pennsylvania State University. There is clearly some overlap between the scope of our review and that of the recent B&s Task Force Report:
Transient Response of B&W Reactors (NUREG-0667).
Accordingly, I have commented on the B&W Task Force recommendations in several instances where they are directly pertinent to the Crystal River transient.
We prepared a comprehensive list of post-TMI-2 NRC requirements, the status of implementation of the requirements at Crystal River on February 26 and brief statement of our assessment of the impact of each requirement on the February 26 transient.
This compilation is presented in Appendix A.
The OPE staff encountered difficulty in compi'ing this comprehensive list of requirements since no single individual o organization knew all that had been required of B&W reactor licensees.
Several independent groups were responsible for analysis and development of new NRC requirements after the TMI-2 accident -- the Bulletins and Orders Task Force and the Short-Term Lessons Learned Task Force, and more recently, the B&W Task Force formed in response to the Crystal River event.
In addition, TMI-2 Contacts:
Bill Travers, OPE l
Dennis Rathbun, OPE 63-43276 i.,
I Commission
-2 l
l I
related requirements were set forth in various letters from the Office of l
Nuclear Reactor Regulation.
Appendix A is subdivided into six sections, representing five different sources of requirements to licensees.
After the TMI accident, attention to new TMI-related requirements has tended to diminish the role of the operating reactor project managers.
i While the purpose of this report is not to examine organizational alternatives, our experience in preparing it leads me to recomend that the Commission consider strengthening the existing project structure (as was done for the TMI-2 cleanup program) by assigning the NRC manager the responsibility for coordinating the implementation of all NRC requirements, license and technical specification amendments, etc., on reactors.
This would consist of an individual or organization who would be completely familiar with the status of a reactor and who would coordinate NRC requests i
and requirements on the licensee.
If the project manager were also organizational 1y or functionally connected with regional office staff, then there would be greater likelihood that NRC would speak with one voice to licensees.
It should be noted at the outset that many of the post-TMI-2 requirements were not directly relevant to the Crystal River transient because of the particular sequence of events.
However, some of the new requirements would have provided for protective measures that would have been effective l
if a different sequence of events resulted in an accident of serious con-sequences (i.e., high radiation release following fuel failure).
The overall net effect of the approximately 80 post-TMI-2 requirements was l
a positve contribution to improved safety.
In particular, operator
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training programs which addressed the handling of TMI-type transients directly contributed to mitigating tr.: Crystal River transient. A direct negative contribution of the requirements was the initiation of the transient by an improperly installed, newly required saturation meter.
l The following sections sumarize the highlights of Appendix A.
Section II discusses operator training and procedures, simulators, and the potential impact of NRC requirement changes on operator training.
In Section III the effect of instrumentation and hardware requirements are reviewed.
Comments relevant to the overall picture on post-TMI-2 NRC requirements are included in section IV.
II.
OPERATORS 1.
Operator Training and Procedures.
The post TMI-2 requirements of licensees in the areas of operator training, qualification, and improved procedures were of material assistance in managing the reactor during the transient and bringing it to a safe shutdown l
Commission,
condition.
While the effectiveness of operator response during a reactor transient is difficult to assess quantitatively, both the industry and NRC staff have stressed the positive value of the TMI lessons learned in the operator management of the Crystal River transient.
During the loss of a major fraction of control room instrumentation, operators maintained adequate core cooling by continuing operation of the high-pressure injection system and by prompt isolation of the power-operated relief valve (PORV).
A comparison (Table 1) of operator response during the Crystal River transient vis-a-vis TMI-2 shows the times involved in some of the key actions in both accidents. Although the initiating causes of the two accidents were substantially different, there were similarities in their early stages.
For example, both accidents involved reactor trips within seconds with associated over-pressurization of the reactor coolant system.
However, operator actions, following revised procedures required since TMI-2, provided for effective management of the Crystal River transient.
In addition, NRC-required operator training paid off.
Specifically, the Crystal River operators recog-nized the need to isolate the PORV in response to a high drain tank alarm signal and, by doing so within a few minutes after reactor trip, eliminated the possibility of a loss of reactor coolant through the PORV. The Crystal River operators also acted to ensure reactor core coverage by keeping the high-pressure injection pumps operating at full flow until instrumentation was restored and reactor subcooling could be verified.
The requirements that licensees perform transient analyses and develop relevant emergency procedures and training (from the Short Term Lessons Learned) contributed to operator competence during the Crystal River transient.
In particular, the operators used the new procedures to determine and track saturation conditions and to verify establishment of natural circulation in the reactor coolant system.
Because of the hiah payoff from more operator training in off-normal procedures, I st rt the B&W Task Force recommendation that generic guidelines be de- ;oped to help the operators in the event of loss of non-nuclear instrumentation and integrated control system power.
These should be available to assist the operators in the event of power bus failures even if a safety system actuation or reactor trip does not occur.
i
l-Commission l TABLE 1 COMPARISON OF HIGHLIGHTS OF CRYSTAL RIVER EVENT (FEB. 26,1980)
TO CORRESPONDING EVENTS DURING THE TMI-2 ACCIDENT Timel 0
Initiating event at Crystal River was the loss of an instru-mentation and control electrical bus caused by a short circuit in a saturation meter.
About half of the normal control room instrumentation indication was lost.
(Note:
Initiating event at TMI-2 involved blockage'in the condensate polisher system.)
0 Feedwater pumps cut back because of an improper instrumentation signal created by instrumentation power bus failure.
14 seconds Automatic reactor trip on high pressure followed by turbine trip.
(8 seconds)
(Note:
Reactor trip occurred 6 seconds earlier at TMI-2.)
2 mins 29 sec PORV isolated by operator in response to a high drain tank alarm (2-1/2 hours)
(Note:
Slow operator response was a direct contributor to the severity of the accident at TMI-2.)
3 mins 20 see HPI actuated upon low pressure signal.
Operator continues HPI (2 mins) with no throttling. Reactor building sump isolated manually.
(Note:
HPI throttled by operator at 4-1/2 minutes into the accident at TMI-2 causing a loss of needed core coolant.)
9 mins Operator starts emergency feedwater pumps.
(Note:
At TMI-2 (8 mins) blocked feedwater valves were opened at this time.)
20 mins Power to instrument bus restored at Crystal River.
Instruments now operative.
20 mins Natural circulation verified.
(1 month) 21 mins 8 sec Reactor building isolated.
(30 mins) lTime for corresponding operator action or event at TMI-2 sequence is shown in parenthesis ( ) beneath the cime at Crystal River.
Commission -
2.
Simulator Training and Operator Training - The simulator training requirements proved especially beneficial in giving the operators experience in r:anaging the reactor under a wide variety of conditions, including unN r-and over-cooling events, solid system operation, and the establishment and verification of natural circulation.
In addition, curing the Crystal River transient, the operators through a communica'. ion link to B&W at Lynchburg were able to utilize the B&W simulator to predict plant response to proposed operator actions.
These predictions proved to be reliable and helped to guide the operators through the transient. Thus, in the light of the experience of the Crystal River transient, I support the B&W Task Force's recommendatic that additional simulator training be required of operators at B&W plants.
3.
NRC Requirement Changes - Despite the apparent improved performance of the operators, the changes in NRC requirements related to operator training may be a source of confusion.
For example, IE Bulletin 79-05 issued immediately after the TMI accident required that, in the event of high-pressure injection initiation, procedures were to be changed to assure that at least two main reactor coolant pumps were running.
At Crystal River, operator training programs were altered to reflect this new N6.. tequirement and the bases for the change were explained to the oper Ators.
IE Bulletin 79-05C, issued last July and based upon analyses submitted by vendors, required that in the event of high-pressure injection the operation of all main cooling pumps was to be immediately terminated.
The Crystal River plant superintendent reported that, in response to the latter bulletin, operator retraining ensued and a new technical justification was given to the operators.
This reversal of position not only makes training difficult but undermines operator confidence in the validity of the procedures.
During the February ?6, 1980 incident, the latter requirement was in effect, and the main cwknt pumps were tripped upon high-pressure injection initiation.
The B&W Task Force now recommends that this point be studied further by NRC and industry, and it is possible that the instructions might again be changed. Such changes have the potential for creating confusion in the minds of the operators regarding the proper course to follow in a future emergency.
In any event, new or changed procedures should be thoroughly and systematically analyzed before they are required to avoid this undesireble situation.
There-fore, I recommend you require new or changed requirements be subjected to a rigorous analysis before implementation by staff.
4.
Shift Technical Advisor - The shift technical advisor played a helpful role during the course of the transient. At Crystal River, each shift technical advisor is a senior reactor operator respected by the l
i r"
Commission.
reactor operators and is active in the operators' training program.
For the advisor to be an effective member of the control room staff, he must be respected and must help train operators.
As a result of the TMI-2 related changes, the Crystal River operators are now receiving regular refresher courses and training every fifth week.
III.
INSTRUMENTATION AND HARDWARE Many of the new NRC requirements for supplementary instrumentation and hardware changes did not contribute significantly to the manageability of the Crystal River transient.
The reasons were that either (a) newly installed instrumentation and hardware which would have been important in a more severe accident (e.g., like TMI-2) were not called upon because of successful operator actions at the outset or (b) equipment which could have been useful at Crystal River had not yet been ir, stalled.
An example in the former category is the high-range radiation monitoring instrumentation which could have been needed following an accident involving failed fuel rods.
An example in the latter category is ultrasonic instrumentation which would have provided control room indication of PORV and safety valve position by monitoring downstream flow.
1.
Saturation Meter - The operators used the four operative saturation meters during the course of the transient.
In contrast to TMI, the meters indicated temperature and pressure regimes that could be fol-lowed at all times to ensure that the reactor maintained an adequate subcooling margin.
Recognizing the advantage of monitoring reactor subcooling, plant staff had built a special output display unit using an oscilloscope to clearly show in analog form the thermodynamic path of the transient.
Nonetheless, it should be noted that the immediate cause of the Crystal River transient was short circuit of the non-nuclear instrumentation (NNI) power buss caused by improper installation of a buffer card which isolated signals to the saturation meter.* The short circuit "ine card was installed into a receiving slot at a slight angle such that mating pins on a connector were misaligned with a set of flexible socket sleeves.
The resulting misalignment created extremely small clearances between pins, and between an end pin and a grounded bolt head.
The instrument technicians at Crystal River were familiar with similar buffer cards and installatio'n instructions were avail-able.
However, the installation apparently had to be made blindly; that is, the l,
module into which the card was inserted was not completely withdrawn from the l
containing rack, and the installation technician could not see that the pins and their sleeve sockets were improperly mated.
The actual short circuit did not manifest itself until approximately 2 weeks after the card installation.
Possibly as a result of vibration of creep, the pin shorted against the chassis bolt head.
~ -
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Commission resulting from the misalignment of the circuit illustrates the extreme vulnerability of the "X" and the "Y" NNI busses.
In this regard, the B&W's Integrated Control System Reliability Analysis (BAW-1564) of last August recommended corrective measures be taken on NNI power supply busses. Moreover, the recent NRC B&W Task Force recommends dividing the power supply busses into smaller power blocks.
I support the Task Force's recommendations in this area.
2.
Safety System Challenges - Two post-TMI requirements, which were not directly pertinent to thi Crystal River transient, may in the long term have safety implications for Crystal River and other PWRs.
The Short Term Lessons Learned requirements include the inversion of the set points of the PORV and the primary loop high pressure reactor trip.
With the original settings, the PORV would relieve pressure first and, if the pressure relief were insufficient, high-pressure trips would shut down the reactor and further reduce system pressure.
Because of the "stickin' open" problem with PORVs, the set points were g
required to be inverted, i.e., the high-pressure reactor trips would actuate first.
This change in effect alters the original operational and design philosophy of the B&W plants and causes additional challenges to the reactor safety systems.
A related issue is the initiation of an anticipatory reactor trip following a turbine trip, which was a requirement from the Bulletins and Orders Task Force.
Turbines may trip for various reasons, only some of which might result in the need for tripping the reactor.
According to the B&W Task Force report the number of reactor trips now being experienced as a result of turbine trips has substantially increased since the implementation of this NRC requirement. The industry representatives with whom we met believe that B&W reactors can safely " ride through" a number of types of turbine trips without the necessity of a reactor trip.
I believe that challenges to the reactor safety systems should be minimized and used as a last line of defense rather than for routine transient control.
Thus, I recommend that these post-TMI requirements -- inversion of PORV and high-pressure trip set points and anticipatory reactor trip following turbine trip -
- should be re-evaluated in the light of recent experience, considering all the implications of these requirements.
Another matter is the particular sequence of events at Crystal River which led to lifting the code safety valves.
After high-pressure injection had begun, and the PORV had been isolated the only primary system pressure relief was through the code safety valve. Although the code valves are designed to pass steam, water was forced through them during the transient.
The possible valve seat erosion by water
Commission.
or two-phase flow (steam and water) through the code safety valves is not known.
I believe that the staff is planning to study this possibility. This sequence results in part from new PORV set points and the requirement to isolate the PORV.
I believe the lifting of code safety valves should be minimized if for no other reason than the inability to isolate them shoulo they fail to reseat.
Thus, I believe analysis of this matter by the staff is warranted.
IV. ADDITIONAL COMMENTS During the course of preparation of this report, certain issues surfaced which while not directly relevant to the Crystal River incident, are nonethel.ess germane to post-TMI NRC reactor safety requirements.
1.
Implementation Schedules for NRC Requirements - In the effort to expeditiously put in place new NRC requirements following the TMI accident, the staff appears to have given insufficient consideration to setting implementation schedules that would allow for the highest quality engineering practice.
In our discussions with industry representatives, we were informed that the accelerated implementation schedule for many of the post-TMI requirements has precluded a thoroughgoing systems evaluation.
Limited architectural engineering assistance was available to Florida Power Corporation to make the required changes in time.
Manpower constraints also caused B&W to be hard-pressed to perform all NRC-required analyses.
As a consequence, both Florida Power and B&W missed some of the implementation deadlines.
A specific example of the scheduling problems with the saturation meter and the associated buffer card is given in Appendix B.
The con-straints on manpower to carry out all the required work appears to have been endemic throughout the utility, A/E, and reactor manufacturing industry.
Of course, the need for quick NRC action is understandable given the urgency following the TMI accident both to assure public safety and to avoid unnecessary shutdown of the Babcock and Wilcox power reactors.
An incidental complaint expressed by utility representatives was that the NRC staff, while insisting on prompt licensee action, did not itself provide expeditious action on licensee submissions (e.g., staff reactor analysis required of licensees often on short schedules).
The licensees also stated that the regional staff and resident inspectors had a better appreciation for operational matters and licensee ability to implement new requirements than did the NRC headquarters staff.
I recommend that the Commission instruct the staff to set implementation schedules for NRC requirements which allow licensees as well as the NRC staff enough time for in-depth engineering evaluation and design, for procurement of high quality equipment, and for its proper installation.
Close coordination by headquarters with regional and resident inspection staffs should also be required.
Commission.
2.
Need for Systems Approach - Since the THI-2 accident, many individual requirements have been placed on licensees without the benefit of an integrated systems analysis.
Each new requirement appeans beneficial by itself, but no systems analysis of the totality of the requirements has been made.
In addition, the basis for a new requirement has not always been given and the requirement itself is often prescriptive --
the cure is given without explaining the problem being treated.
For instance, the Short Term Lessons Learned requirement that licensees at Crystal River and other plants provide high point vents for both the reactor coolant system and the reactor vessel may be an illustration of where NRC requirements are overly prescriptive and do not allow sufficient flexibility. The industry representatives with whom we met said that the venting requirements are an example of an instance where the staff would be better served by defining those accident conditions which the staff wants venting to mitigate and then having licensees propose the specific venting arrangement applicable to their plants.
They are concerned that the additional valving on the reactor coolant pressure boundary is not warranted.
Venting requirements for noncondensible gases may reduce rather than increase reactor safety margins.
Industry representatives are of the opinion that remotely controlled vents on the B&W " candy cane" piping to the steam generators may be adequate during the majority of potential transients and that vents in the reactor head may not be needed and, indeed, may be undesirable.
I believe that NRC needs to make greater use of system engineering techniques, probabilistic analyses, event trees, and fault trees and less use of deterministic methods.
Also NRC requirement of licensees should be functional rather than prescriptive.
3.
The Use of Non-Safety Grade Equipment - The new saturation meter that was required to be installed quickly was to be control-grade equipment rather than safety-grade e,quipment.
Considerable discussion has taken place within the Commission staff and industry as to the merits and feasibility of using safety-grade equipment for non-safety functions.
Interestingly the new hardware installations recomended by the B&W Task Force all specify-safety grade equipment.
Philosophically, this position appears correct -- if a change is being ordered because of a safety deficiency or implication, the new change should be manufactured to the highest safety-grade standards (except where seismic qualificatio,ns render safety-grade specification impractical).
For instance, if the saturation meter had been designated as safety-grade, the Crystal River transient might have been avoided altogether.
If the saturation meter had been safety grade, a proper Failure Mode and Effects Analysis may have been required to be performed on this equipment and its power supply.
A good analysis would have discovered the design deficiency -- no fusing of the buffer amplifier card rack -
and corrected it ahead of time.
With proper fusing, the short on the NNI bus would not have occurred, nor would the ensuing Crystal l
River transient.
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Commission.
V.
CONCLUSION We have concluded that the overall net impact of the approximately 80 post-TMI NRC requirements was definitely positive. The most important improvements are those which deal with the operator.
Our review found that the post TMI-2 requirements of licensees in the areas of operator training, qualification, and improved procedures were of material assistance in managing the reactor during the Crystal River transient and bringing it to a safe shutdown condition.
Many of the new NRC requirements for added instrumentation and hardware changes did not contribute significantly to the manageability of the Crystal River transient because either the new instrumentation which would have been important in a more severe accident was not called upon due to successful operator actions at the outset or' equipment which could have been useful had not yet been installed.
Our review of the Crystal river transient indicates to us that additional improvements are needed in several areas:
A project-type organization should be considered for coordinating and issuing NRC requirements.
This would be an individual or organization who would be completely familiar with the status of a reactor.
If the project manager were also organizationally or functionally connected with regional office staff, then there would be greater likelihood that the NRC would speak with one voice to each licensee.
Routine challenges to reactor safety systems should be avoided.
Challenges to reactor safety systems should be used as a last line of defense rather than for routine transient control.
More realistic implementation schedules should be established for new requirements.
The Commission should instruct the staff to set imple-mentation schedules which allow licensees as well as the NRC staff sufficient time for in-depth engineering evaluation and design, procure-ment of high quality equipment and its proper installation.
Staff and industry should conduct further studies on the use of safety-grade versus non-safety-grade equipment.
The initiating event at Crystal River may have been 4.'1ded if a proper Failure Modes and Effects Analysis had been performeo on the installation of the saturation meter and its associated power supply.
k
Comission -
NRC needs to make greater use of systems engineering techniques, pro-babilistic analyses, event trees / fault tree methodology, and less use of deterministic methods.
Also NRC requirements should be functional rather than prescriptive.
Attachments:
As stated cc:
Leonard Bickwit Sam Chilk William Dircks Harold Denton Victor St.ello Robert Minogue Robert Budnitz Howard Shapar Carl Michelson i
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APPENDIX A:
POST-TMI-2 REQUIREMENTS This appendix presents a comprehensive listing of post-TMI-2 NRC requirements in summary form.
This summary was drawn from the work of several independent groups which had the responsibility after the TMI-2 accident for analysis and development of these new requirements. Source documents include the reports of the Bulletins and Orders and the Short Term Lessons learned Task Forces as well as various letters of the Office of Nuclear Reactor Regulation.
For each new NRC requirement, an implementation status at the Crystal River plant as of the February 26 transient and a short impact statement are presented. The 1
assessment of impact was made after discussion with NRC staff in the Offices of Inspection and Enforcement and Nuclear Reactor Regulation and industry repre-sentatives.
The organization of the Appendix is as follows:
Section I:
TMI-2 Lessons Learned Task Force (NUREG-0578, 7/79)
Section II:
Report of the Bulletins and Orders Task Force (NUREG-0645; Volumes I and II; 1/80)
Section III:
IE Bulletins Relating to TMI-2 Section IV:
Generic Evaluation of Small Break Loss of Coolant Accident Behavior in B&W Designed 177-FA Operating Plant (NUREG-0565, 1/80)
Section V:
Letter of September 13, 1979 from D. G. Eisenhut to All Operating Nuclear Power Plants Section VI:
Letters of September 28, 1979, and January 9,1980 from R. W.
Reid(NRR)toB&WLicensees e
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SOURCE: THI-2 LESSONS LEARNED ThSK f0RCC (fAREG-0578. 7/79)
Complete Prior Implementation to Crystal River Requirement Required by Transient (2/76/001 Impact on the Crystal River Transient
' 8, Design additional instrumentation and 1/1/80
. Procedures in place
- Positive tapact but additional new instrumentation might procedures for easy interpretation of for existing instru-have shortened event significantly.
Inadequate core cooling. (2.1.3b) esentation; but only some new instrumen-tation had been in-stalled.
9.
listall additional core coollag instru-1/1/80 Not all new instru-Same as item I.8.
mentation Installed.
- 10. Provide containment isolation r ' diverse 1/1/80 No but dedicated Positive tapact. Containment isolated by dedicated (Extension operator as lateria operator within 4 minutes into tie transient sequence.
signals. (2.l.4) to 4/14/00 measure.
outage granted.)
II. Review and revise as necessary contalsment 1/1/00 Will le coupleted If completed. would have released the dedicated operator isolation provisions for non-essential (Extension to during this outage.
for other dulles during the transient. (Note destgts systems and the potential for inadvertent 4/14/00 outage revleus have shown some isolated systems would reopen granted.)
uponreset.)
reopening (upon reset of the isolation signal.
2.1.4)
- 12. Provide dedicated containment recombiner penetrations and isolation systests that meet the redundancy and single failure requirements of Camelssion regulations.
(2.1.5a) a.
Submit description and schedule to NRC.
1/1/80.
Actl'on completed.
No tapact since hydrogen was not generated during the Crystal River transient. (flote containment purge is b.
Complete installation.
1/1/81 Hot completed as the primary means of hydrogen control at Crystal River.)
of 2/26.
' 13. R;vlew shleiding requirements and procedures 1/1/00 Review completed.
Sammt as ites 12.
rclating to recombiner use and upgrade.
(2.1.5c) i e
e j'
m l
1.
SOURCE: IMI-2 LESSONS LEARNED TASK f0RCE (NUREG-0578.7/19)
Complete Prior Implementation to Crystal River Impact on the Crystal River Transient Required by Transient (2/26/80)
Requirement
- 14. Implement a program to reduce leakage or 1/1/30
. Extending and No impact during transient. no fuel failure occurred.
potential for leakage from systems that would formalizing the existing program.
cr could carry radioactivity following a transient. (2.1.6a)
- 15. Estibitsh a prevent (ve maintenance program to 1/1/80 Preventive maintenance No impact since preventive palatenance programs were program established, already in effect.
reduco leakage. (2.l.6a).
IE. Perf;re a review of plant snielding require-ments assuming accident conditions. (2.1.66)
No impact since no fuel failure occurred during transient.
a.
Complete review.
1/1/80 Yes b.
Complete necessary changes.
1/1/81 Future implementation.
- 17. Aust11ary feessater design shall be upgraded, if necessary. Lo provide automatic Initiation.
(2.1.7a) (Subsequently excluded from Lessons Learnedrequirements.)
NRC required automatic initiation signal failed to a.
Complete per control grade.
1/1/80 Yes function due to failed instrumentation (attributable to power supply failure). Consequence was dry out of one steam generator. Auxillary feedwater supply started manually about 3.5 minutes into transient.
I b.
Complete and upgraded to safety grade 1/1/81 future implementation.
if safety grade automatic auxillary feedwater initiation la place, would have prevented dry out of steam generator.
requirements,
- 18. provide safety-grade indication in the control room of aunl11ary feedwater flow for each steam generator. (2.1.7b) a.
Complete per control grade.
1/1/80 Completed.
positive effect because operator was able to verify (Eaception auxillary feedwater flow to steam generator.
to 6/1/a0) l-1 1
e
I.
SOURCE: IMI-2 LESSONS LEARNED TASK f0RCE (NUREG-0578,7/79)
Complete Prisc Implementation to Crystal River Requirement Required by Transient (2/26/00)
Impact on the Crystal Alver Transient Installation to safety Same as ites I.18.a (if completed) because uould have
- 18. b.
Complete to safety-grade.
1/1/81 grade is a future flee. removed single failure possiblitty.
Ig. Review and upgrade capablitty to.obtain and analyze samples from the reactor coolant system and the containment atmosphere under high radioactivity conditions. *2.1.8a) a.
Review complete.
1/1/00 Yes No impact since no major radioactivity release in Crystal Alver transient.
b.
Procedures for sample collection and 1/1/00 No analysis complete.
(Exception to4/l/80) c.
Description of plant modifications.
1/1/80 Yes d.
Plant modlfications complete.
1/1/81 Future action.
- 20. Provide high range radiation monitors in 1/1/81 Future implementation.
Sames as ites 19.
containment. (2.1.8b)
- 21. Provide high range lastreasentation to measure noble gases in eIfluents. (2.1.8b) a.
Develop lateria seasures to allow quantifi-1/1/80 Yes Same as ites 19.
cation of release rates up to 10,000 Cl/sec.
b.
Complete lastallation of annitors.
12/1/81 future implementation.
- 22. Provide system description and procedures for -
No impact since no major release of lodine occurred 1/1/80 Yes antif lag lodine gaseous effluent levels.
during Crystal River transient.
2.1.8b
- 23. Implement program to quantify lodine in gaseous 1/1/81 Future legalementation.
Same as fles 22.
effluents. (2.1.8b) 0 e
4 9
9 9
..I
9
't.
SOURCE: THI-2 LE550NS' LEARNED TASK TORCE (NUNEG-0578,7/Fg)
Complete Prior Implementation to Crystal River Requirement Re4utred by Transient (2/26/80)
Impact on the Crystal River Trar. stent
- 24. Provide equipment and associated training and 1/1/80 Actions complete slightly positive impact. Provided operators at Crystal procedures for accurately determining airborne (Exception (check).
River with confidence that Iodine was not being released inside the plant, tedine concentrations laside the plant where to 5/1/80) rsonnel may be present during an accident.
2.i.Sc)
- 25. Provide accident analyses, and upgrade.
Staggered Anakysiscompleted.
Provided substantial positive Impact. Operators had emergency procedures and training as schedule effective procedures for maintaining core coollag needed. (2.1.9)
Identified despite loss of control room lastrument indications, in liUREG-0578)
- 26. Provide a continuous containment pressure 1/1/81 In place.
Ileipful to operators but a source of sose NRC confuslos.
Indication in the control room. (2.1.g)
- 27. Provide a continuous containment water level 1/1/81 Installation in about Would have helped if lastalled by providing indication Indicatten in the control room. (2.1.9) one year.
of containment water level -- especially relevant ta the possibility of water covering electr.lCal comlMments and/or machinery.
- 28. Provide a continuous indication of containment 1/1/81 Future implementation.
No impact since hydrogen not generated during Crystal River transient.
hydrogen concentration to the controt room.
(2.1.9)
- 29. Provide high point vents. remotely operable from the control room. in the reactor coolant system and reacter vessel. (2.1.9) a.
Complete design.
1/1/80 Florida Power design 110 impact because reactor coolant system dlJ not reach not yet accepted by saturated ccr.dttions. (Note: industry representatives NRC staff.
does not agree Laat vents should be placed in the reactor vessel head.)
b.
Complete lasta11ation.
1/1/81 1
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o II. ~50uRCE: REPORf of THE But.tETINS AND ORDERS TA5K TORCE (tamEG-0645. VOLS I AND II. I/80)
Complete Prior implementation to Crystal Alver Requireeent Required by Transient (2/26/80)
Impact on the Crystal River Transieht 1.
Develop and laplement procedures for initiating for items 1.2.3
. Procedures in place.
MRC requirement may not have gone for enough to assure and controlling emergency feedwater independent
- 4. and 5 of II:
emergency fee &ater to steam generators. Existing cf the lategrated Control System. (See 1.17)
Crystal River and requirement had positive effect in that operator had other B&W plants the ability to control water level in the steam generator were shutdown by independently of ICS. But this capability overrides Commission order by the automatic steam break isolation system during of April 79.
the Crystal Alver transient.
Completion of 2.
Upgrade the timeliness and reliability of these items was All upgrades to improve timeliness and reliability cllvery from the emergency feewater system.
required prior of aualliary feedwater supply were la place.
to restart.
a.
Provide for starting of motor-driven Completion was No impact on plant operation during this transient emergency feehater pumps from a vital accomplished by since no loss of non-vital aualliary feedwater buses, bus.
summer 73, b.
Station an operator at local valves No impact since transient did not occur during testing during testing in coanunications with of aualliary feedwater valves and piping.
control room.
c.
Verify emergency feeduater pumps are Potential positive impact since avalitary feedwater pumps were required during the transient.
d.
Provide for obtain'ing alternate sources No impact since alternate sources of water for aualllary of water for emergency feed, feedwater were not required during the transient.
G.
Provide for automatic start of motor No impact, since operators initiated motor driven driven aual11ary feedwater pumps..
avalliary feehater pumps manually, f.
Provide for timely operator notification No impact, same as item 2.e above, of emergency feedwater automatic Initiation.
g.
Verify failure position of emergency Potential positive impact verified open at Crystal feed flow control valves.
River.
h.
Remove interlock which prevents turbine Potential positive impact since operators initiated driven pump from injecting when motor both motor and steam drive emergency pumps simul-i driven pump is operating, taneously. But(asinII.1)thiscapabilitywas overriden by the automatic steam break isolation 1
system during the Crystal River transient.
d 0
I b
I
II. SOURCE: REPORT OF Tile BULLETINS AND ORIERS 1ASK FORCE (IAIREG-0645 VULS. I AND ll, 1/DU)
[
Complete Prior l
laplementation to Crystal River Requires-ent Required by Transient (2/26/80)
Impact on the Crystal River Transient j
3.
laplement a hard-wired reactor trip See item 1.
Control grade Installation completed. (Note industry actuated on loss of main feeddater representatives feel that tripping the reactor follow-and/or turbine trip.
Ing a turbine trip is not desirable.)
I a.
Complete control grade.
Control grade Installation completed. No impact since reactor trip was caused by high reactor coolant systene pressure.
b.
Complete safety-grade.
Future laplementation.
Not applicable.
4.
Complete analyses for potential small See item 1.
Revised procedures Positive tapact since revised procedures based upon small breaks and develop and laplement in place.
break analyses provided operators with effective scans operating instructions to define for managing the transient (i.e., keep on the lipi pumps).
operator action. (See1.25) 5.
Complete 1MI Unit 2 training, at the
See item 1.
Completed.
Positive tapact; inciease operator awareness of impor-88W slaulator, for all licensed Lance of maintenance of adequate core cooling.
reactor operators.
6.
Make modifications to provide control Longer term Installation of con-Positive effect. Allowed operator to verify auxillary room verification of emergency feedwater laplementation trol grade modifica-feedwater flow to steam generators.
flow to each steam generator. (See 1.18) schedule: pro-tions completed.
bably 1/1/81 for items 6, 7, 8 and 9.
7.
Subelt a failure mode and effect analysis of Analysis by B&W No lapact since NRC review and potential follow-up the Integrated Control System.
completed, requirements have not been completed.
8, Continue reactor operator training and Actions completed Positive effect -- with operator training nodified as drilling of response procedures.
a result of the IMI accident.
~
9.
Develop and sulanit Technical Specifications Tecluitcal specift-IstC licensing has not yet coupleted its review of revised for Limiting Conditions of Operation cations were not technical specifications (e.g., no technical specifica-appropriate to the above. requirements.
tions in place now for auxiliary feedwater flow Indica-tors; turbine trip; reactor trip; steam generator level Initiated reactor trip; and auxillary feedwater auto-
.matic start as a result without survelllance requirement, operability of foregoing cannot be assured).
j e
III. SOURCE: IE BULLEillis RELATIter. TO IMI-2 Complete Prior Implementatloa to Crystal River Aequirement Required by Transient (2/26/80)
Impact on the Crystal River Transient Review 1HI-2 Prellainary Nottf tc4tton (FN11 4/11/79 Yes Positive impact since preliminary notif.lcation and detailed chronology of THI-2 accident.
following IMI-2 alerted other DW plant operators 79-05,79-05A (Item 1) to the potential for 1MI-2 type small break loss-of-coolant accidents.
Review transients siellar to 1MI-2 that have 4/11/79 Yes Positive impact same as item III.I.
0 occurred at BW facilities and review the ev:1uation of the 11/29/77 transient at Davls-Besse 1.
79-05,79-05A(Itan2)
Review operating procedures for recognizing.
4/11/79 Yes Positive impact; new procedures were relled upon.
0 preventing, and mitigating vald fonmation during transients and accidents. 79-05 79-05A(Item 3)
Review operating procedures and training lastructions to ensure that:
- c. Operators do not override engineered 4/11/79 Yes Positive impacts llPI automatic initiation was not safaty features actions unless con-Interfered with at Crystal River. IIPI was continued tinued operation will result in unsafe at all times during the loss-of-instrumentation period plant conditions. 79-05 Item 4?.
of the transient.79-054 (Itan 4a).79-058 (Liten 2J b.
IIPI system remains in operation (if actuated automatically) unless:
(1) Both LPI pumps are operating at a Positive isoact; same as item III.4.a.
flow rate greater than 1000 GPM and the situation has been stable for 20 mismtes, or (2)llPIhasbeeninoperatio.:for20 4/11/79 Yes Positive lapact operators throttled llPI after regaining mins. and the RCS is at least 50'F instrumentation which allowed subcoollag determinations subcooled.79-054 (Item 4b).
to be made.79-058 (Item 2) (Note. 20 minute llPI requirement inillally suggested by BW later dropped.)
~
awe
III. SOURCE: IE 80LLETilli RELATileG 'TO iHI-2 Complete Prior Implementation to Crystal River Requiremient Required by Transient (7/26/80) lepact on the Crystal River Transient
- c. Until automatic RCP trip is installed 7/28/79 Yes Performed action as required. RCP's were tripped manually upon llPI inillation. Unclear whetler impact and operational:
was positive. negilglble or negative.
(l) Upon reactor trip and HPI inttla-tion caused by low RCS pressure.
trip all operating RCPs, and (2) Provide two operators in the CR at all times to accomplish RCP trip and other required items.79-05A (Item 4c).79-05C (Short-tern Item la and Ib) d.
Operators are provided with additional 4/11/19 Yes Positive impact; provided effective guidance to infonnation and guidance not to rely operators on evaluating plant conditions.
on pressurizer level Indication alone in evaluating plant conditions.79-054 (item 4d) 5.
Review all safety related valve positions 4/11/19 Yes Potential positive impact; sinceenginenred safety features were required during* tim transient.
cnd positioning requirements and positive controls and all related test and mattenance procedures to assure proper ESF functioning, if required. Verify all Afil valves are in the ooen position. 79-05(Itas5).79-05A l
(Itan 5) l 6.
Review oper&5tng modes and procedures for 4/16/79 Rev,lews not Positive impact in that tisse procedures assured all systems designed to transfer potentially (Exception cospleted.
that tlm contalement was isolated during the radioactive gases and Ilquids out of con-to 3/31/80 transient. Also provided increased operator awarencss talament to assure the undesired pumping granted) of these systems.
f
.. cf radioactive gases or liquids will not occur inadvertestly. Ensure that this l
does not happen ce ESF reset. List all such system:. and list 1
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III. SOURCE: IE BULLETINS RELATIhG TO IMI-2 Complete Prior Implementation to Crystal River Requirement Required by Transient (2/26/80)
Impact on the Crystal River Transient
- 11. Review and modify procedures for removing 4/16/79 In place.
Potentially positive luyact in that,this requirement safety-related systems from service (and (Exception Double verifica-assured availability of engineered safety features some restoring to service) for maintenance and to 3/31/80 tion at Crystal of which called upon during the transient.
testing to assure operability status is granted)
Alver.
vertfled and known.79-05a (Item 10)
- 12. Assure all operating and maintenance 4/16/79 Yes Potentially positive impact; same as item III.I.
personnel are aware of the seriousness and consequences of the simultaneous blocking of both AfW trains and the other actions taken during the early phases of the IMI-2 accident.79-054 (Item II)
- 13. Develop procedures and train operators 5/8/79 Training accom-Potential positive impact; natural circulation on melleds.of establishing natural circu-plished; procedures established and vertfled during the. transient by 11 tion. Include: means of monitoring in place, utill Ing these procedures.
cfficiency by available lastrumentation; cssure RCS is at least 50 F subcooled; Precautions for pressurizer level Indi-cetion, pressure contrul. F-T Ilmits; and procedures in the event of LOFW while in natural circulation.79-058(Item 11)
- 14. Modify design and procedures which reduce 4/23/79 Modification of Difficult to judge impact in that PORV failed cpen, the Ilkeilleod of automatic PORV ilfting design and pro-PORY setpoint changed from 2255 to 2450 psig (08-1 during anticipated transients. Lower high-cedures was setpoint 2400) Illgh pressure trip setpoint changed prsssure reactor trip setpoint.79-058 complete.
from 2355 to 2300 psig.
(Itan 3)
- 15. Provide a manual trip (procalures and 3/8/79 Hanual trip No impact, since reactor tripped automatically on pressure translents:
provided.
high RCS pressure (2300 psi).
training) for the following high
\\
~
1, 4
em
III. SOURCE: IE SULLETINS RELATING TO THI-2 Complete Prior Impirwntation to Crystal River Requirement Required by Transient (2/25/80)
Impact on the Crysta) River Transient
- c. Loss of Main feedwater (Note - industry representatives feel that reactor b.
Turbine Trip trips on turbine trip asul low pressurlier level are
- c. Mala 5 team Isolation Valve Closure notdesirable.)
d.
Low Steam Generator Level and f.
Low Pressurizer Level D. Provide design review and schedule for 5/23/79 Design review completed No lapact on this transient, however, requirement lapleecatation of a safety-grade reactor and schedule submitted, may not go far enough to assure a positive effect during trip upon:
anticipated transients. (Note: Key element is signal selection.)
c.
Loss of feedwater b.
Turbine Trip c.
Significant Reduction in Steam GeneratorLevel79-058(Ites5)
)7. Propose changes to Technical Specift-5/23/79 No impact since technical specification changes were not citions which must be audified as a (Exception in place, result of lawlementing IE Bulletin to 3/31/80 items.79-058 (Ites 7) granted)
D. Perform and, submit a report of LOCA 8/28/79 Report of LOCA Positive lapact.
analysis for a range of small break analysis subaltted siz:s and a range of time lapses asul documented in between reactor trip and RCP trip.
Determine primary coolant tengerature anil identify any area wiscre, a ary m
coolant temperature is greater whan 2200 F.79-05C (Short-tern Item 3) 09, Based upon the analyses done in require-8/28/79 New guidelines Impact unknown; further analyses should provide relevant meat 18 above, develop new guidelines for developed.
Infonsation upon which procedures for transient alliga-oper' tor action for both LOCA and non-tion could be based.
LOCA events that take into account the cffect of RCP trip.79-05C(Short-ters item 3).
[
O
III. SOURCE: IE BULLEllHS RELATINr. TO TMI-2 Complete Prior Implementation to Crystal River Regulrusent Required by Transient (2/26/80)
Impact on the Crystal River Transient
& Based upon guidelines developed in require-9/12/79
. flew guldclines Positlyq lap 4Ctl no small Lrc4L during this transient, ment 19 above, revise amergency procedures developed.
howeverf new guidelines did Proylde operators with and train all operators.79-05C (Short-effective procedures for contro]Ilng Lists transient, term Itas 4)
- 1. Provide analyses and develop guidelines 8/28/79
flew proeddures for Procedures had a positive impact in that they assured and procedures for inadequate core cooling.
harmiling inadequate adesguate core cooling. (Notes Reactor coolant pump Define RCP restart criteria.79-05C (Short-core cooltre in restart criteria not yet classified.) -
term Itan 5) place. (notes Updated version of requirement listed initasIII.19.)
- 2. Propose and sulmolt a design which will llot yet implemented, Impact unknown at this time, however, potentially cssure automatic tripping of the RCPs Operators instructed negative impact since there was no vaall break but under all circumstances in which this to perform trip tripping the reactor coolant pumps terminated forced action may be required.79-05C (Long-
- manually, core" cooling.
term Itas 1) 2 F
4 e
g g
.* 4
a IV. SOURCE: GENERIC EVALUATIDH OF SMALL BREAK LOSS-OF-COOLAlfi ACCIDENT
- BEllAVIOR IN BW DESIGNED 177-fA OPERAilHG PLANT (NUREG-0$65.1/80)
Complete Prior Implementation to Crystal River Requirement Required by Transient (2/26/80)
Impact on the Crystal River Transient 1.
Provide a systen'to cause pressurizer block 1/1/81 No yet installed.
flo impact since not installed; NRC staff reevaluating valves to close at RCS pressures below the under NRC staff its position on the need for this requirement.
PORV epen set point.
review.
2.
Revise trip and valve setpoints to assure 1/1/81 Action complete.
This requirement had no effect since PORY failed that the PORV will open in less than open.
5 percent of all anticipated over-pressure transients.
3.
Report failures of PORVs and safety valves 4/1/80 Fallures presently No impact; and NRC staff has not completed analysis to reciose (promptly) and challenges to being reported, of such failures.
FORVs and safety valves (annually) to the NRC.
4.
Submit a report discussing safety valve 1/1/81 Action not completed.
Reporting only and no lupact.
failure rate in B&W plants.
5.
Revise and de r E analysis methods used 1/1/82 Same as item IV.4.
No impact.
for small brist 8 9.
6.
Suin3t plant-spr&,T p talculations, using 1*/1/82 Osse as iten IV.4.
No impact.
NRC models, for --tli ht.'=ks to show compilance with c ; im C AG.
7.
Perform analyses to detewSaa If core flood 7/1/80 Same as item IV.4.
No impact, tank lajection modelt dra conservative.
8 Install anomatic reactor coolant pump 1/1/81 Same as iten IV.4.
No impact, trip act'ivated by small break LOCA.
g.
Reviewandupgradereliabilityandrehndancy' Action Plan Same as item IV.4.
No impact.
of non-safey grade equipment used in sitigating small break LOCAs.
- Notre-although NUREG-0565 his been distributed and in some cases acted upon by B&W owners, it has not been. formally issued as a set of requirements i.e., no cover letter to B&W owners has been sent.
e 9
0
IV. SOURCE: GENERIC EVALUATION Of SHALL BREAK LOSS-8f-COOLANT ACCIDENT BEllAVIOR IN B&W DE5lGNED 177-FA OPERATING PLANT (NUREG-0565.1/80)
Complete Prior Implementation to Crystal Alver Requirerent Required by Transient (2/26/80)
Impact on the Crystal River Transient
- 03. Providt $8LOCA slaulator training for all 1/1/81
- BW operators have Positive effect.
operators, already been required to complete 1NI-2 simulator training.
- 11. Verify esperleentally, various modes of 1/1/82 Future implementation. No impact. (Note: Duke Power. Florida Power and BW feel this sinnald be s'one in a scale nudel experiment ain! M two-phase natural circulation.
inacosacrcla. plant.)
- 12. I; stall additional control room lastru.
1/1/81 Same as ites IV.11.
No impact.
eentation to vertfy natural circulation.
- 13. P;rform analysis of plant response to a 6/1/80; Same as item IV.11.
No tapact.
small break'which is isolated, causing RC5 repressurization and subsequent stuck-open PORV.
- 14. Perform analyses of plant response to a 5/1/80 Same as ites IV.11.
No impact, small break in the pressuriser spray line tith a stuck-open spray line isolation v:1ve.
- 15. Evainte the effects of water slugs in 5/1/80 Same as ites IV.11.
Ilo impact.
1:: piping cased by llPI and core flood tank flows.
- 16. Provide pretest predictions for LOFT test Schedule no 5ame as ites IV.11.
No lepact.
L3-6(RCPsrunning).
finalized.
- 17. Provide both techalcal justification for 5/1/80 Same as ites IV.11. ~
leo tapact.
omitting rad olytic decomposition of 1;;jected ECC water as a source of non-condensible gast and confinnatory Infor-1 mation to verify the predicted condensation he:t transfer gradient.
i 1
1 l'
i
V.
SOURCE: LETTER OF 9/13/79 FROM D.G. EISEmilui TO ALL OplRATING NUCLEAR l*DidER PLANIS Complete prior Implementation to Crystal River W irasent Required by Transient (2/26/80)
Impact on the Crystal River Transient
- 1. lipgrah emergency plans to conform with 1/1/80 New plans have been No impact. because state and county implementation of Regulttiry Guide 1.101.
submitted but Crystal emergency plans not required during this transient.
River is operating (Note: closest residence 3 miles from Crystal River.)
under entsting plans.
- 2. Establish an Emergency Operations Center for federal. state and local officials.
c.
Designate location and alternate location Mid-1980 Interim emergency Impact unknown.
and provide communications to the plant, center estabitsbed.
b.
lipgrade Emergency Operations Center in 1/1/81 Not yet completed.
conjunctlen with in-plant technical support center. (See1.34) 3, leprov3 offsite monitoring capability.
Mid-1980 Implementation not
. s No impact. but implementation would have provided
- complete, assurance that no off-site releases has taken place.
- 4. Assure tdequacy of state / local plans.
- c. Against current criteria.
Mid-1980 Completed.
Same as item
.1.
b.
Against upgraded criteria.
1/1/81 Not completed.
No impact.
O. Conduct test exercises.
a.
T:st Ilcensees new emergency plans.
Mid-1980 Not completed.
b.
Test states new emergency plans.
Mid-1980 Not completed, c.
Jol:t test exerc'lse of new emergency
~
1/1/81 Not completed, plans (federal. state, local. licensee).
I9
l l
e e
VI. SOURCE: LETTERS OF SEPTENDER 28, 1979 AND JANUARY 9, 1900 FROM M.W. REID (NNN) TO 8&W LICENSEES Complete Prior Implementation to Crystal Alver Regul. w nt Required by Transient (2/26/80)
Impact on the Crystal River Transient 1.
By analysis and/or esperiment, address the 6/1/80
' Work has been completed. Results may suggest need for w o design changes.
sechanical ef fects of laduced slug flow, No lepact.
which may occur during natural circulation /
reflux belling transitions. on steam generator tubes (11/21/79).
2.
Evaluate tapact of RCP seal damage and leakage 6/1/80 Evaluation in progress. Positive effect, caused operating staff to evaluate due to loss of seal coollag on loss of off-seal condition prior to restart of reactor coolant site power (11/21/79).
pumps.
3.
Ev:Imate PORV and safety valve lift frequency 1/1/81 Future analyses.
No tapact.
and increase la reactor trip frequency based on revised PORV and high pressure reactor tripsetpoints(9/28/79).
4.
Perf:ra an analhls of potential velding 1/1/81 Future analyses.
No impact.
1:3 the RCS during anticipated transients (1/9/80).
e
.8 e
e.
m 44 b.
e
~
APPENDIX B Chronology and Comments on the Saturazion Meter Installation at Crystal River The following comments are presented to substantiate the inference that some haste was involved in the installation of the required saturation meter at Crystal River. The schedule below indicates the key dates.
The requirements for the meter to be installed by January 1,1980 was delineated in Mr. Eisenhut's letter of September 13, 1979.
Florida Power actions then were:
Purchase Order Issued:
10/30/79 P. O. Acknowledged by B&W:
12/4/79 Equipment Received:
1/24/80 Quality Assurance Documents Received 3/27/80 It is obvious that despite best efforts the January 1,1980 date could not be met and a waiver was granted by NRC until February 15, 1980 to complete the installation.
The best effort apparently required some improvisation at Baily Meter and/or by the technicians at Crystal River.
A detailed examination of photographs of the buffer amplifier card that created the transient revealed that it was not identical with the other similar, and supposedly technical, cards in the rack. The following differences were noted:
1.
The new buffer printed circuit card base was a punched-out production type card having the same dimensions as the other cards.
But the printing process appeared to be different. The previously obt: 6 d cards were tinned and connected by what appears to be a production type wave soldering equivalent process.
The new buffer card appears to be manufactured by a different process, possibly being hand made in the laboratory.
2.
The circuit was slightly different.
On the new card, a 200,000-ohm resistor was omitted and previously jumpered wires were now printed connections.
The printed wire routine was somewhat different.
3.
The principal active element in the circuit (probably an integrated circuit operational amplifier) had its leads soldered to the board in a different and obviously poorer manner.
It appeared as though the original circuit element had been removed and replaced by a new one by someone with less skill than the original solderer.
4.
No QA stamps were in evidence on the new card, whereas the prior cards were clearly marked by inspectors' stamps.
It is doubtful that the poor soldering on the new card would have passed inspection.
Note also that QA papers did not arrive at Crystal River until a considerable time af ter the equipment (and after the incident).
All of the above items may indicate some pressure and haste to get the saturation meter installed, but in all f airness, did not contribute directly to this incident.
.