ML20116M959
| ML20116M959 | |
| Person / Time | |
|---|---|
| Issue date: | 07/31/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-BR-0125, NUREG-BR-0125-V01-N5, NUREG-BR-125, NUREG-BR-125-V1-N5, NUDOCS 9608210189 | |
| Download: ML20116M959 (8) | |
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OFFICE OF NUCLE AR RE ACTOR REGULATION U.S. NUCLEAR mREGULATORY COMMISSION E
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NEWSLETTER
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Learning from Operating Experience by Thomas E. Murley In carrying out our responsibilities for ensuring the safe damage given that the precursor event occurred in the manner it g
i operation of nuclear power plants, we should pause from time did. It can, therefore, be considered a measure of the residual totime and ask ourselves what the operatingexperience of the protection against severe core damage that was available during i
112 licensed plants is telling us. It is not enough to assume that the actual event.
simply carrying out our normal reviews and inspections will ensure safety. We must also be alert for indications that there Within NRR I have asked the Risk Application Branch (RAB) to may be unexpected sequences or trends that are overlooked in analyze the results of the ASP Program for insights important for l
our safety reviews. When such trends are found, we should our mission of protecting public health and safety. Keepingin mind carefully analyze the operating experience data and, if war-the care that must be used in treating random fluctuations in the ranted, take action to ensure continued safe oxration. One data for low-frequency events, RAB has reported that it is, none-example that comes to mind is when several mstances of loss theless, possible to draw some important conclusions. These are:
of decay-heat-removal capability during mid-k>op operation of pressurized-water reactors (PWRs) indicated that poten.
(1)
There has been a significant decrease in the inferred tially serious accidents could occur even when a plant was in a mean core damage frequency since the TMI-2 acci-i cold shutdown condition. In this example,we issued Generic dent in 1979 (at least an order of magnitude de-Letter 88-17 ("Imss of Decay lleat Removal") to all PWR crease).
l licensees asking them to inform NRC of the actions they had l
taken to limit the potential for loss of shutdown-decay-heat-(2)
Eighty percent of the most significant precursors in l
removal capability.
1987 involved loss of electrical power.
1 One of the most important NRC programs for systematically (3)
Approximately one-half of the precursor events in examining operational experience is the Accident Sequence 1987 involved human error to some degree.
Precursor (ASP) program that has been sponsored by AEOD and carried out at the Oak Ridge National Laboratory since (4)
Precursor experience tends to validate the PRA 1979 Accident sequence precursors are actual initiating functional and system event trees (that is, there are events or equipment failurcs that, when coupled with other fewer surprises from operating events each year in postulat ed events, could result in inadequate core cooling and, terms of unanticipated sequences).
l ultimately, a severe core damage accident. A few thousand Licensee Event Reports are screened cach year, and, from what we note in them, several precursors are selected for detailed reviewif thcy meet certain significanm criteria. Methmis What' sin this issue?
developed originally for Probabilistic Risk Assessments (PRAs)
I, are used to estimate the conditional probability of potential See Page 2 severe core damage associated with each precursor. This m
i probabilityis an estimat e of the chance of potential severe core r ;.?'
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complicated sequence ofevents thbt starts many months in advance.
Thefollowing article details the way emergency exercises areplanned
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IN THIS ISSUE and caluatedin NRWon W.
Emergency preparedness exercises are scheduled well in advance, Learning from Operat, g Expen,ence at meetings between licensees, the State, counties, the NRC, and m
the Federal Emergency Management Agency (FEMA). The by Thomas E. Murley....
1 Emergency Preparedness Exercise Evaluation exercise schedule must be crafted so that exercises do not overlap (straining agency inspection resources); fall on Federal, State or by James E. Foster -
2 local holidays; or otherwise cause conflicts. Exercises may be Peach Bottom Public Meetings tentatively scheduled up to 5 years in advance. Once an exercise is by James Linville 4
Shutdown Decay Heat Removal scheduled the dates are difficult to change, because licensees and offsite agencies expend considerable resources in exercise plan-by Warren C. Lyon and Robert C. Jones -
4 ning and coordination.
Inservice Inspection Programs for Nuclear Power Plant Components A licensee is requested to send draft exercise objectives and a by George Johnson -
5 Substandard Vendor Products scenario outline to the Regional office approximately 75 days i
before the exercise date. The NRC and FEMA then evaluate these by E. William Brach.
6 Interfacing Systems LOCA submittals to determine if all the exercise objectives can be met.
by Sammy Diab, Kazimieras Campe, and Richard The ability to meet certain minimum emergency criteria must be demonstrated each year, while the ability to meet some others Barrett 7
(such as post-accident sampling) can be demonstrated less fre-quently. The ability to meet all parts of the plan must be demon-NEWSLETTER CONTACT:
sirated every 5 years.
Valeria Wilson, NRR 492-1208 At this stage the draft scenario often lacks specific details; nonethe-I less it is redewed for realism, technical adequacy, degree of challenge, opportunity to demonstrate acceptable performance in areas previously found deficient, and to demonstrate the listed objectives. The technical adequacy may be reviewed by the (5)
Precursor experience does, however, point out Resident inspector, a consultant, the Regional Emergency Prepar-two weaknesses in PRA methodology:
edness staff, or a combination of these, depending on the plant involved and the complexity of the scenario. Technical reviews
-inadequate modeling of human errorsofcom-address the realism of the scenario, the expected response of plant mission and systems and operators, and include an analysis of the source term, postulated releaws, and offsite radiation doses.
. inadequate modeling of systems interactions.
The NRC's exercise evaluations are team inspections. The inspec-(6)
Precursor data show that older plants have a tion Team Leader will select the areas to be observed, and the team j
smaller precursor rate than newer plants. These member (s) who wili perform the observations. Areas observed are data appear tobe in contrast to the fact that,in selected on the basis of the importance of the area, past inspections general, PRAs predict a lower core damage fre-and exercise performance.
quency for newer plants than for older plants.
Team members may be assigned on the basis of their general expertise in the area assigned (post accident sampling system expertise, for example). If one or more consultants are to be Emergency Preparedriess involved, they are selected based on their area of expertise (opera-Exercise Evaluation
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h by James E. Foster, DRSS, Region ll1 The Team Leader then establishes the general logistics for the evaluation team, including travel dates, arrival times, hotel selec-(
The NRC requires an emergency plan at all plants, and the tion, an ngements for site-specific training for team members, effectiveness of the plan must be demonstrated annually in an licensec entrance and exit mterviews, time and iocation of the team exercise. Although some NRCstaffers may be under the impres-meeting, and attendance at public meetings.
sion that the NRC evaluation of an emergency exercise is easily accomplished with a short visit to the site during the exercise, what During the entrance interview, the Team Leader introduces the actually hapnens during the exercise is reallyjust the midpoint in a team members and the areas they ate ta evaluate, obtains copics of 2
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the most recent revision of the exercise scenario, discusses Open It is not unusual for the Team Leader to be contacted by the local Items from past inspections, obtains current facility status, and media at the end of the exercise. However, the Team leader confirms schedules for the exit interviews and public meeting. cannot discuss any of the specifics of the evaluations until the exit l
interview takes place.
l Licensee personnel will provide site-specific training for team members who are not currently familiar with the site and/or who The exit interview will be held with senior licersee personnel and do not have current site badges. Team members who will enter members of the plant management staff on the day after the r:diological areas of the plant must either have current plant exercise. During this meeting, thc Team leader will discuss overall
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badges, receive radiological training from the licensee, or be and specific exercise performance, and will tent atively characterize escorted.
the findings of the team as Exercise Weaknesses, Unresolved
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Items, Open items, Improvement Items, or Comments. Licensee As a general rule,the NRC team attends at least a portion of the personnel may be asked to provide comments or clarification on licensee's exercise controller training sessions. 'Ihe Team leader specific items.
often introduces the team to the controllers and provides gen-eralinformation on the role of the NRC observer / evaluator.
If FEMA evaluated State and local participation in the exercise, FEMA will sponsor a public meeting one or two days after the Exercises can begin at any hour of the day, and be from one to exercise. At this meeting the preliminary exercise evaluations of three days long. During the exercise, the Team Leader usually both FEMA and the NRC willbe providedl(FEMA personnelwill l
moves from area to area, generally startingin the Control Room be contacted before the meeting to coordinate presentations.)
l cnd moving to other areas as they are activated in response to the Team leaders from NRC and FEMA will present their findings scenario. This gives the Team Leader an overview of the and answer questions from media representatives and the public.
l licensee's actions. During a typical exercise, the Team Leader l
will observe actions in the Control Room, Technical Support When he/she returns to the Regional Office, the Team Leader Center, Operations Support Center, Emergency Operations gathers, the full-text input of the NRC team members, organizes Ftcility, and possibly the Joint Public Information Center. If them into the appropriate format, and begins the process of resources have allowed, each area will have an assigned evalu-producing a draft inspection report. Management resiews the stor who observes the entire exercise, and when the Team report for accuracy and consistency with guidance and policy. The Leader visits each area, he/she will discuss the evaluator's report is also checked for format and grammar. The goal is to observations. Otherwise, the Team Leader's observations will cumplete the report 20 days after the last day of the inspection.
be the sole evaluation.
Most exercise evaluation reports are completed well within this time. Exercise Weaknesses and Open items are entered on the 1
l Each team member evaluates the licensee's actions on the basis Open items List (OIL) maintained by the Region.. If an Open item I
of the inspection procedure and his/her knowledge of NRC from a previous report is shown to be resolved during the exercise, l
emergency response guidance and philosophy. Detailed notes that item is' closed in the inspection report and the OIL If the Open are taken, and a chronology of actions developed.
Items resulting from the exercise can be closed during a routine 1
inspection, they will be reviewed during the next scheduled routine At the end of the exercise, the Team Leader or assigned team inspection. Often, the nature of an item resulting from an exercise member attends one or more of the licensee's self-critiques, is such that it can only be closed during a subsequent drill or l
where licensee personnel discuss the findings of the licensee's exercise.
observers. The team does this to evaluate the licensee's capabil-
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ity to identify deficiencies in the facility's emergency response The final product, the inspection report, may make it appear that program. At this meeting, the Team leader or an assigned a minimum of effort was involved in the inspection because the i
mspector may address the exercise players.
total sequence of events is not outlined (this is consistent with other l
NRC inspection reports). However, the real story is that, as with l
After the exercise, the evaluation team meets (generally in the other NRC products, a number of dedicated professionals spent i
evening) to discuss general findings in each of the areas ob-considerable time shaping the end product.
l served. Discussing findings with other observers often clarifies the exact sequence of events and provides details of noted l
deficiencies. Then the Team Leader begins the overall exercise evaluation, including identification of exercise deficiencies and their relative significance.
At this point, the team members are asked to develop" bullets" i
-- short descriptive sentences or paragraphs summarizing their evaluation of assigned area (s) of observation. The Team Leader will use these " bullets" as the basis for the exit interview and any public meeting on the inspection.
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4, Pcach Bottom Public Meetings Shutdown Decay Heat Removal by James Linville, DRP, Region I by Warren C. Lyon and Robert C. Jones, DEST After the NRC issued an Order to shut down the Peach Bottom Atomic Power Station (PBAPS) on March 31,1987, the Agency The loss of decay heat removal (DH R) capability during shutdown i
received many informal requests for public hearings. These operations is a concern at pressurized water reactors (PWRs), and c me from members of the public, and from local and State Generic Letter 88-17 is almed at resoMng this concern. Over the officials up to the Governors of Maryland and Pennsylvania. The years many losses of DHR have occurred during PWR shutdown.
NRC committed to hold public meetings in the counties that are These are especially likely to occur when the reactor coolant within the emergency planning zone (EPZ) to receive public system (RCS) inventory is reduced to facilitate maintenance, comments on the licensee's plans to correct the problems that leaving only a small margin between satisfactory DHR operation led to the issuance of the Shutdown Order, and DHR loss. Before 1987, the industry and NRC had issued many publications warning of the need to prevent such events Since September 1987, the NRC has held nine meetings to (e.g., NUREG/CR-2799, Information Notice 86-101, and NSAC-receive and respond to these public comments. Three sets of 52),yet these events continued to occur. On April 10,1987, Diablo l
meetings were held between September 1987 and March 1989. Canyon Unit 21ost DHR while operating with the RCS level near One set was held after thelicensee submittedits Commitment to the centerline of the hot leg piping (commonly called "mid-loop Excellence Plan, one set was held after the submittal of the operation"). Boiling started within about half an hour, and DHR revised Plan for Restart of PB APS, and one set was held after the was restored one-and a-half hours later. After this incident, the completion of the NRC's Integrated Assessment Team inspec-NRC sent an Augmented Inspection Team (AIT) to the site. The tion. A totalof about 150 people spoke at these meetings. Speak-team found deficiencies in procedures, hardware, and training.
crs included employees of the licensee who favored restart, some Later,it was confirmed that these deficiencies were generic to all l
k> cal residents in favor of and some against restart, and other PWRs.
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interested people who spoke against nuclear power in general l
and against restart. State and local officials attended all the Industry and NRC followup to the Diablo Canyon event enhanced meetings and co-hosted the earlier ones.
insight into DHR operation during reduced RCS inventory opera-tion. Some of the lessons learned included the following:
Specific comments and questions and NRC's answers from the e:rlier meetings were documented in the Peach Bottom Restart 1.
Phenomena were identified that influencedlevelinstru-S:fety Evaluation Report, dated October 19,1988. Many of the mentation, level control, loss of DHR, plant response to speakers attended and commented at several of the meetings, loss of DHR, and mitigation of events.
rnd many comments were similar. No new issues relevant to the Peach Bottom restart were raised. Each of the meetings was 2.
Controlof RCS level is difficult in a mid-k>op condition.
transcribed by a court reporter, and copies of the transcripts The range of RCSlevels that will allowadequate DHR j
were placed in k> cal public libraries.
operation and not dump RCS water into containment is smaller than previously thought.
The purpose of the final round of public meetings (February 28 and March 1,1989) was to give the public a presentation 3.
Scenarios were identified for Westinghouse-and Combus-regarding the NRC staff's conclusion that Peach Bottom was tion Engineering-designed plants where loss of DHR rerdy for restart, as well as an opportunity to comment on that could result in core damage within an hour. Earlier it had conclusion. NRC personnel at these meetings included: Re-been estimated that it would be four hours before the top gional management, Resident Inspectors, Headquarters project of the core was uncoveied, m.nagement, Headquarters specialists, and Regional State liai-son personnel.
4.
Scenarios were identified where waterinjected intothe RCS would not reach the core.
These meetings allowed the public to provide input into the Perch Bottom restart readiness review through an informal 5.
Plant response to loss of DHR had not been analyzed.
process that consumed fewer NRC resources than a formal herringwould have. Moreover, the process demonstrated to the Such items as these raised concerns that there might not be public our interest in their views and the ability of NRC to explain effective core damage avoidance strategies for some loss of DHR its position on the important issues associated with a highly events. Furthermore,because the containment building was often controversial shut down.
open to simplify maintenance operations dnring shutdown opera-tions with the RCS in a mid-knop conditiot it was feared that an I
unmitigated offsite release could occur if loss of DHR event caused core damage.
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Probabilistic risk assessment (PRA) usults based upon a pre-Inservice inspection Programs Di:blo Canyon understandmg of plant behavior show that the j
risk of a core damage accident during nonpower operation is for Nuclear Power Plant thout 105 Mid-loop operation causes 85% of the risk. The PRAs show that operator and other personnel interactions Components during event mitigation are greater than during power opera-t,on. The large uncertainty as a result of human mteractions bY Georbe Johnson' EMTB i
increases uncertainty in the PRA results.
Cleirly, loss of DHR during shutdown operation was an impor. The Code of Federal Regulations in 10 CFR 50.55a requires the t nt safetyissue requiring industry attention.The NRC issued components of a boiling or pressurized water reactor nuclear Generic letter (GL) 88-17, which requested corrective actions Power plant that are important to the plant's safe operation and tt til operating PWRs. The Director of the Office of Nuclear essential for the protection of public health and safety be subjected Re:ctor Regulation sent letters to the Chief Executive Officer of to an insenice inspection (ISI) program throughout the senice life each utility to reinforce the importance of this issue. In recog. of the facility. The rules and requirements for the insenice eition of the operator's role in preventing and mitigating loss of inspection program are those given in Section XI of the American DHR events, he also sent GL 88-17 to each licensed operator. Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. These rules and requirements are applied to those The unique aspects of nonpower operation, including mainte. components classified as ASME Code Class 1,2, and 3, classifica-ntnce needs, required a solution that was flexible yet consistent tions that signify a relative degree of the importance-to-safety with the prime objective of reducing risk to the public health and aspect of the components. ASME Code Class 1 has the greatest safety. To achieve this objective, we met with representatives of degree of importance, and ASME Code Class 3 the least.
the PWR owners' groups and visited plants during periods of nonpower operation. This provided better insight into the The rules and requirements of Section XI of the ASME Code problems and potential solutions.
identify (1) the areas subject to examination, (2) examination methods and procedures, (3) personnel qualifications, (4) fre-GL 88-17 requires licensees to respond to sets of recommenda. quency of examinations,(5) record keeping and reporting require-tions in two areas: expeditious actions and program enhance. ments, (6) procedures for evaluation of examination results, (7) ments. Expeditious actions are those that emphasize mitigating disposition of results of evaluations, and (8) repair and replace-
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em offsite release by developing procedures to ensure contain. ment requirements. Each licensee submits an insenice inspection ment closure before core uncovery. Also recommended are (1) program plan to the NRC before the beginning of cach 10-year immediate, practical actions to reduce the likelihood of loss of interval of plant senice life. The plan must contain the information DHR and (2) short-term improvements in instrumentation and necessary to determine that (1) it is based on the correct edition procedures to provide improved operator response to loss of and addenda of Section XI,(2) the examination sample size and DHR. These actions are to be completed before a unit operates selection are in accordance with Section XI requirements, (3) the in a reduced-inventory condition. Program enhancement rec. correct examinations are being applied, and (4) any required ommendations emphasize improved understanding,better pro. augmented examinations or license conditions relative to inservice cedures, better use of existing equipment, and better instrumen. inspection are included. The plan is reviewed and evaluated by the tition. Recognizing the lead time needed for analysis, design, NRR staff, which will find that it is in compliance with 10 CFR tnd installation, the Generic Letter allows an implementation 50.55a and acceptable for implementation at the facility, condition-schedulc of roughly 18 to24 months.The program enhancement ally acceptable with certain items requiring resolution, or unac-actions address the root cause of loss of DHR and provide ceptable. The staff then documents its findings in a Safety Evalu-improved operator response if a loss of DHR occurs. These ation Report, actions represent a defense-in-depth philosophy of accident prevention, accident mitigation, and containment.
In addition to requiring the components of a nuclear power plant to be examined throughout its senice life,10 CFR 50.55a also In summary, substantial work has been done to understand loss requires that the inservice inspection program be updated to meet of DHR and plant response to such a loss. These investigations the requirements oflater editions and addenda of Section XI of the l
identified needed improvcinents in procedures, hardware, and ASME Code at every 10-year interval beginning with the date of l
tr ining to both prevent and mitigate such events. The discus. commercial operation of the facility. In cases where the licensee sions held with industry during the formulation of GL 88-17 finds it impractical to implement an updated requirement, the helped to significantly reduce the risk associated with shutdown licensee must request relief from the specific requirement and to decay heat removal while minimizing the effects on plant opera. provide information to the NRC in support of the request. The tions. Many of the recommendations reflect good operational staff reviews and evaluates this information, and,if the necessary prictices and control of shutdown activities. We are encourag. findings can be made, the requested relief from the requirement ing licensees to consider the appropriateness of these recom. may be granted. In cases where added assurance of structural mendations for all modes of shutdown decay heat removal.
reliabilityis deemed necessary, the NRC may require the licensec 5
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l to follow an augmented inservice inspection program for the quality of vendor products,but they are not specifically desig l
systems and components for which this determination is made. detect counterfeit or fraudulent activities. NRC expec procedures and actions by licensees and their representatives con-i Inservice inspections are donc primarily during plant refueling forming to the Appendix B criteria will detect substandard outages. These required periodic examinations have resulted in poor quality products, and traditionally they have done the timely detection of flaws that could have progressed to ever, as the recent past has shown, such procedures a component failures, such as those detected in steam generator may not necessarily detect counterfeit or fraudulent' shcIl girth welds, feedwater elbows, recirculation risers, and products.
nozzle-to-safe end welds. The results of the examinations and/
or tests scheduled to be performed during an outage are summa-The NRC relics on the integrity of its licensees and their su rizedintheInserviceInspectionSummaryReportandsubmitted to ensure the implementation of the commitments made to thej to the NRC within 90 days of completion of the inspection. Any NRC with respect to the quality and functional capability of powe unteceptable indication detected during an examination of a reactor systems. To be able to do so is important for ba j
component must be corrected by repair or replacement, or it plant systems whose failure can challenge safety syste may be accepted for continued service on the basis of further cate the response to an accident, as well as for those syste covered by the quality assurance requirements of 10 CFR 50 evaluation.
Appendix B. Although, to date, no major safety problems have i
A well-prepared inservice inspection program that is in compli-been identified as a result of the instances of counterI ence with the requirements of Section XI of the ASM E Code, the fraudulently marketed products, the NRC is concerned tha regulations, and any augmented requirements and that is prop-checked, this problem could have significant impact on re i
erly implemented provides (1) assurance of the continued struc-safety.
tural reliability of the facility's components,(2) assurance of the j
integrity of its pressure-retaining boundary, and (3) assurance of the protection of public health and safety.
Causer ofthe Problem An obvious question arises: Why is the nuclear industry detecting so many more cases of misrepresented vendor products than it Substandard Vendor Products detected in past years? There are four possible answers to the question.
by E. William Brach, VIB First,the shrinking U.S. domestic nuclear market has caused many manufacturers toleave the nuc! car market or to reduce the product lines offered under nuclear quality assurance production stan-The Problem dards. Consequently, NRC licensees have, on some occasions, Duringthe past severalyears the NRC and the nuclear industry been forced to search out new vendors who m have identified numerous instances in which vendor products with or appreciate the need for strict conformance with nucl failed to meet appropriate standards.The causes of these sub-quality requirements.
standard products included lack of material control, missed in-Second, a large fraction of safety-related equ spection holdpoints, testing equipment that was out of calibra-tion,usc of unqualifiedtechnicians, inadequate engineering and is procured by intermediate suppliers and upgra safety-related applications. The intermediate suppliers may not lack of design control, have all the correct engineering, design, and material drawings and However,in 1988 and 1989 an increasing number of substandard specifications for the item being upgraded and they vendor products have resulted from apparently counterfeit or to verify whether the item has been altered sinc manufacture, fraudulently marketed products. Examples of apparently counter-feit vendor products include fasteners (nuts, bolts, and screws), Third, obviously, there is an economic incent piping material (such as fittings, flanges, lugs, and plate materi-als), pumps, valves, valve replacement parts, and electrical differencebetweennuclear andcommercialprices, equipment, such as molded-case circuit breakers and metal-clad prices for refurbished and newcomponents. Th create an incentive for misrepresenting the product.
circuit breakers.
Appendix B to 10 CFR 50 establishes the quality assurance Fourth, NRC and the nuclear industry may be m criteria for safety-related structures, systems, and components misrepresented products for nuclear service are for nuclear power plants. These requirements cover design, place. Thus the increase in the number of casec is th procurement, receipt inspection and testing, and construction of our sensitivity toward identifying counterfeit or end operation of nuclear power plant structures, systems, and products.
components. The critcria are ge nerally structured to confirm the 6
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- NRCActions these cases are under investigation. Prosecution of two vendors identified by the NRC resulted in guilty pleas and NRC's plan to address the counterfeit and fraudulently mar-verdicts invoMng corporate and individual penalties.
keted vendor products issue can be separated into four activities:
(1) provide licensees with information to facilitate their correc-Rulemaking. In March 1989, NRC issued an Advance tive actions, (2) share information with other Federal agencies, Notice of Proposed Rulemaking (ANPR) to request public (3) investigate and initiate appropriate prosecutorial actions, and nuclear industry comment on whether existing regula-and (4) determine whether additional NRC rulemaking or other tions should be changed to address counterfeit and fraudu-actions are needed to help correct the problem. Each of these is lent vendor products. The ANPR focuses on procurement discussed below.
and dedication programs for safety-related applications i
and requests public comment on questions pertaining to information to Licensees. During the last year and-a-new requirements. The areas of the ANPR include in-half, the NRC has issued six Bulletins and Bulletin creased receipt testing, establishing traceabilityand unique Supplements to inform the industry of instances of sus-identification requirements for procured equipment and pect vendor products and to request the industry to take material, better audits of vendors, joint utility procure-l specified actions. Also,12 Information Notices and ments, and improved criteria for commercial-grade dedica-Information Notice Supplements were issued to the nudcar tion programs. The ANPR provides for a 120-day public industry to alert licensees to additional instances of comment period that closes in July 1989.
suspect vendor products. The NRC staff has also partici-pated in numerous meetings and conferences with licen-Recognizing that rulemaking generally is a lengthy process, l
sees, industry trade organizations, and national stan-the staffissued Generic Letter 89-02 to bring to the indus-i dards and code organizations to discuss NRC findings, try's attention positive aspects of procurement programs i
activities, and actions.
that appear to have had benefit in reducing the use of counterfeit and fraudulently marketed products. The pri-Information to Other Federal Agencies. In July 1988, mary areas emphasized in this Generic letter are increased l
NRC asked the Office of Management and Budget (OMB) engineering involvement in all aspects of procurement and l
to organize an interagency meeting on the topic of sub-commercial-grade dedication of vendor products, better standard vendor products with expedation that the meeting receipt inspection and testing, and improved vendor audits.
would be the initiator of an interagency network for sharing, informing, and coordinating Federal agency TheNRChasalsoaskedtheNuclearManagement andResources activities. More than 20 Federal agencies attended the Council (NUMARC) to coordinate industry efforts to address l
resulting meeting in August 1988, and OMB reported procurement, dedication,and vendorissues. In mid-1988,NUMARC that the President's Council on Integrity and Efficiency formed anindustryworking group to coordinate and focus nuclear and the Council on Management Improvements would utility efforts and to review existing procurement and dedication coordinate and direct the Federal agency interactions.
programs. The initial efforts of this working group appear prom-ising with respect to improving the procurement process through-The NRC also developed direct staff contacts with other out the industry. The NRCis monitoring the NUMARC efforts Federal agencies including Department of Energy (DOE), and will adjust its own programs as the NUMARC initiatives National Aeronautics and Space Administration (NASA), become effective.
and Department of Defense (DOD), to share informa-tion on counterfeit or fraudulently marketed vendor products as well as on the broader topic of substandard Interfacing Systems LOCA l
vendor products. NRC has provided information on suspect vendor activities to OMB, DOE, DOD, NASA by Sammy Diab, Kazimieras Campe, and and other Federal agencies on numerous occasions dar-mg the past year.
Richard Barrett, RAB Investigation of Wrongdoing. NRC is attempting to ensure that those vendors involved in counterfeit and Intahrtion fraudulent activities are identified and prosecuted. De-pending on the merits of each case, the NRC is pursu-Nuclear plant operating experience of the past few years, in the ing enforcement sanctions in some cases and is referring United States and abroad, indicates that there is a need for others to the Department of Justice for possible prosecu. reassessing the likelihood of an unisolable interfacing systems loss-tion. In 1988 and 1989, the NRC inspection and investi-f-coolant accident (ISLOCA) that bypasses containment. An gative staffs were involved in a large number and a wide unis table ISLOCA outside containment can have severe conse-variety of vender cases invoMng suspected counterfeit quences as a result of the release of radioactivity directlyinto the and fraudulently marketed vendor products; many of environment. An event of this type can occur if a low-pressure l
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system is accidentally subjected to the high pressures associsted Operations (INPO) issued recommendations to licensees for es-with the reactor coolant system.
tablishing preventive maintenance programs for PlVs.
In one recent occurrence, there was a release of primary coolant Recent operating experience indicates that in most of the uniso-outside the primary containment as a result of the use of lablelSIDCAs, human errorwasa major contributortoinitiation improper procedures in operating a check valve in a residual or aggravation of the event. It is interesting to note that,in these heat removal (RilR) system suction line. In another case, cases, huma i actions also were credited for terminating the same c.perator error led to the opening of one of two isolation valves events. The lessons learned from these events include the follow-between the primary and RIIR systems, followed by unsuccess-ing: (1) ISLOCA events may have a higher likelihood than previ-ful attempts to open the second isolation valve. l{ad the operator ously estimated;(2) the contributions of human error and recovery succeeded in opening the second isolation valve, it is likely that actions to the risk associated with ISLOCA events may not be a large LOCA would have occurred outside the containment.
modeled adequately by current PRAs; and (3) the traditional i
1 Event-V (i.e., failure of two check valves in series) is only one of Other examples of related recent experiences include inadver-several ISLOCA scenarios.
tent opening or misalignment of pressure isolation valves (PlVs),
grossleakage through redundant check valves, disabling of a PIV In view of the lessons learned from recent operating experience,it pressure interlock, and the identification of previously unana-is believed that further steps are needed to achieve the NRR goals lyzed ISLOCA paths in the component cooling water (CCW) described ab(we.
system.
ISLOCA Risk Assenment Program Events of this type suggest that the probability of a ISLOCA may not be as low as previously believed on the basis of existing NRR has initiated a joint program with RES, AEOD and the l
probabilistic risk assessments (PRAs). Because some ISLOCAs Regions for bringing the ISLOCA issue to a closure. The principal could lead to significant offsite doses in a relatively short time, elements of this program are:
the potential for their occurrence is a significant safety concern.
Ilence, NRR has initiated an effort to obtain a comprehensive (1) human reliability considerations with respect to the initia-assessment of the likelihood of and consequences associated tion and mitigation aspects of the ISLOCA, with an ISLOCA and to determine,if necessary, the steps that can be taken to reduce the risk to an acceptable level.
(2) determination of the most likely size and location of any breaks in low-pressure systems (stress and thermal hydrau-The NRR effort to resolve the ISLOCA issue has three main lic analyses),
goals: (1) achieving a high level of confidence that the event will not occur for the current generation of reactors; (2) ensuring (3) focused PRA analyses that include human reliability as well that procedures, equipment, and operator training are such that as stress and thermal hydraulic considerations, there are adequate provisions to minimize offsite radiological consequences of an ISLOCA.
(4) a pilot inspection program of a few selected plants, and
Background
(5) accident management and consequence mitigation.
As early as 1975, the ISIDCA, referred to as Event-V, was The focused PRA analyses, the pilot inspection findings, and the identified in WASH-1400. At that time, the event was character-accident management and mitigation assessment will be the three ized in terms of the failure of two check valves in series that form main elements addressed in a safety evaluation report (SER).
l the interface between high-and low-pressure systems. To I
address the Event-V ccmccrns, in 1981 NRR issued orders to The lessons learned from the pilot inspections will forrt the basis more than 30 plants, calling for mandatory surveillance and for an approach to meet the NRR goals for all plants. As appro-limiting conditions for operation (LCOs) for PIVs. Neverthe-priate, the approach willinvolve generic recommendations, plant-less, since that time, there have been a number of PlV failures as specific inspections,or some combination of the two. To the extent l
a result of mechanical causes as well as a result of operator possible, the activities in this program will be coordinated with the ongoing effort with respect to Generic Issue 105,"PlV Intersystem l
crrors.
LOCA."
To alert licensees to the potential for similar failures in their plants, the NRR staff issued Information Notices discussing some of the PIV failures. Also,in a case study, AEOD described a number of the PlV failures and some of the subsequent overpressurizations at boiling water reactors. In 1983, Region I
(
undertook plant inspections for Event-V type of vulnerabilities l
cssociated with PlVs. In 1986, the Institute for Nuclear Power o
n