ML18089A166

Responds to Re ATWS Events & Reactor Trip Breaker Malfunctions.Language Used in IE Bulletins Revised to Ensure Specific & General Responses

ML18089A166
Person / Time
Site:	Salem
Issue date:	05/16/1983
From:	Palladino N NRC COMMISSION (OCM)
To:	Markey E HOUSE OF REP., INTERIOR & INSULAR AFFAIRS
Shared Package
ML18089A167	List: ML18089A166 ML18089A168
References
NUDOCS 8306060388
Download: ML18089A166 (31)

Text

U.NITED STATES e

NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 May 16, 1983 The Honorable Edward J. Markey, Chairman Subcommittee on Oversight and Investigations Committee on Interior and Insular Affairs United States House of Representatives Washington, D.C.

20515

Dear Mr. Chairman:

I This is in response to your April 7, 1983 letter to me expressing your concerns regarding the anticipated transient without scram events that occurred at the

~plant on February 22 and 25, 1983.

In this regard, your letter pose~-fiVe questions involving reactor trip breaker {RTB) malfunctions.

This letter addresses the five specific questions and the more general concerns that are expressed in your letter. The five specific questions posed in your letter are addressed in Enclosure 1 and its attachments.

Your letter highlights the fact that all RTB failures have not been reported to the NRC.

You then point out that the Licensee Event Report (LER) system cannot do its job of analyzing operational data and preventing accidents based on pre-cursor information unless the NRC is provided with accurate data in a timely fashion.

We share your concerns in this regard and we have been assessing the matter. Toward this end, we are reviewing the reporting requirements of opera-ting facilities and we are also endeavoring to determine whether additional RTB malfunctions have occurred that have not been reported to the NRC.

Additionally, you bring to our attention that although the Southern California Edison response to IE Bulletin 83-01 was technically correct, it did not mention the RTB failures experienced at San Onofre Units 2 and 3.

We too are concerned with this situation, and therefore, we are revising the language used in bulletins to ensure that licensee responses not only address the specific action items de-lineated in bulletins but that they also respond to the more general.concerns of the bulletin. The actual testing of the General Electric RTBs by Southern Cali-fornia Edison appears to be responsive to the general concerns expressed in IE Bulletin 83-01.

The test failures were separately reported to the NRC as a LER in a timely manner and these reported failures led directly to the issuance of "IE Bulletin 83-04.

We appreciate your interest in this matter and trust that the information con-veyed in this letter and its enclosure is responsive to your inquiry.

~~~

Nunzio J. Palladino

Enclosure:

Response to Rep. Markey's questions cc: Rep. Ron Marlenee

(. 8306060388 830516

!

• PDR COMMS NRCC CORRESPONDENCE PDR

QUESTION 1.

ANSWER.

Please provide a list of breaker failures not previously reported to the NRC.

We presently know of fifteen RTB malfunctions that had not been previously reported to the NRC.

These malfunctions are described by licensees in their responses to Action Item 4.d. of IE Bulletin 83-04.

These malfunctions and the licensees' reasons for not having previously reported them to the NRC are listed in Table 1, Summary of RTB Malfunctions not Previously Reported to the NRC (Attachment 1).

It should be noted that several of the listed malfunctions did not involve the safety concern that precipitated the bulletin, namely, failure of the undervoltage device to trip the RTB.

Erclosure 1 i

J

QUESTION ~

ANSWER.

Is it possible that there have been additional reactor trip breaker failures not reported to the NRC than those identified in response to question 1?

It is likely that other such failures have occurred.

The basis for this assumption includes the fact that failures that occur: (a) following main-tenance but prior to returning equipment to service; (b) before issuance of an operating license; or (c) while in modes not requiring the breakers to be operable, (e.g., cold shutdown -0r refueling) may not be reportable.

In addi-tion, interpretations of Technical Specifications requirements vary between plants with respect to component failure reportability.

Thus, all RTB fail-ures may not be reportable.

However, any RTB failure of an automatic or manual trip function that occurs while the plant is in a mode requiring the breakers to be operable is clearly reportable.

QUESTION 3.

ANSWER.

Specify what, if any, reporting requirements applied to the reporting of the incidents identified in response to ques-tion 1 at the respective sites where they occurred (taking into account when they took place).

The specific reporting requirements for operating plants using standard technical specifications (STS) are stipulated in Section 6.9, "Administrative Controls - Reporting Requirements 11 of.the plant's technical specifications.

These STS requirements are based on the recommendations contained in Regulatory Guide (RG) 1.16, 11 Reporting of Operating Information - Appendix A Technical Specifications" (Attachment 2).

The technical specifications for plants not using STS, (i.e., those licensed prior to 1975) have been amended on a case by case basis to include the major recommendations in RG 1.16; however, because of different limiting conditions for operation among plants, there are some differ-ences in the reporting of these events.

The bases for some of these differences are described in Attachment 3, 11 Reporting Requirements.

11 Enclosure l

QUESTION 4.

ANSWER.

Does the NRC believe that its reportable requirements were violated by the failures to report any of these incidents?

Is any enforcement action being contemplated?

Since NRC followup of the licensee's responses to the IE Bulletins 83-01 and 83-04 is not completed, determinations of potential violations of regulatory requirements remain open.

Table 1, (Attachment 1) identifies the licensee's stated reasons for not previously reporting certain failures.

With regard to the Oconee event, it should be noted that although the licensee did not submit a LER, Region II was aware of the malfunction.

In fact, the malfunction was described in Inspection Report No. 50-269/81-02, 50-270/81-02, 50-287/81-02 dated February 24, 1981.

Region II reviewed this matter and concluded that failure to report was due to a misinterpretation of the plant's technical specifications. Since appropriate corrective actions were taken by the licensee, it was concluded at that time that further enforcement actions were not warranted.

Enclosure

QUESTION 5.

ANSWER.

Does the NRC believe that the history of test failures calls into question either the reliability of 25 cycle pre-startup testing or the reliability of this component.

The present testing requirements for using Westinghouse breakers with under-voltage trip devices as reactor trip breakers were recently (subsequent to the Salem ATWS event) developed jointly by Westinghouse and the users as requested by and concurred in by the NRC.

These tests to demonstrate operability are much more stringent than those in place prior to the Salem events of February 22 and 25, 1983.

The acceptance testing re~uirements are such that the manufacturer must perform a 25 cycle test with no failures permitted on each breaker prior to shipment.

Additionally, the licensee must perform a 10 cycle test with no failures permitted on each breaker prior to placing it in service or subsequent to performing maintenance.

These tests should provide reasonable assurance that each breaker will perform reliably while in service.

Recent experience indicates that the undervoltage trip attachments may have an unacceptable reliability for RTB applications.

The reliability of this method of tripping the breaker is currently being reviewed.

In the interim, enhanced*

survei 11 ance, ma i*ntenance and operating procedures provide reasonable assurance that the breakers will operate.

. TABLE 1

SUMMARY

OF RTB MALFUNCTIONS NOT PREVIOUSLY REPORTED TO THE NRC (Based ?n responses to Action Item 4.d. of IE Bulletin No. 83-4)

Plant

• Arkansas-1 (AK Breakers)

• Arkansas-1 (AK Breakers)

• Arkansas-2 (AK Breakers)

• Arkansas-2 (AK Breakers)

• Arkansas-2 (AK Breakers)

Calvert.

Cliffs-1 (AK Breakers)

Calvert Cl iffs-2 (AK Breakers)

Calvert Cliffs-2 (AK Breakers)

Calvert Cliffs Davis-Besse (AK-Breakers)

Date of Event 8/23/82 10/6/82 12/14/79 9/11/81 11/12/82 3/1/78 12/20/82 12/20/82 12/20/82 8/7 /77 Licensee's Description "B" RTB failed to trip on demand "B" RTB failed.to trip on demand from the non-safety related shunt coil during test.

A new UV trip coil installed in a spare RTB.

Reason for replacement not known; actual malfunction not described Breaker would not stay closed.

New UV trip mechanism installed on a RTB during replacement of of a blown control power fuse.

Reason for replacement not known; actual malfunction not described.

TCB-2 failed to trip during preventive maintenance.

TCB-7 failed to trip during preventive maintenance.

TCB-1 operated sluggishly during preventive maintenance TCB-4 operated sluggishly during preventive maintenance "Trip Confirm" light failed to clear after the breaker was closed.

(Breaker 11 8 11

)

Licensee's Reason for Not Previously Reporting All control rods inserted/

reactor subcritical at the time of the occurrence.

Safety related portion of breaker functioned properly.

Event occurred during re-furbishment of a spare breaker that was not in service at the time.

Breaker Technical Specifica-tion requirements not vio-lated, i.e., breaker tripped.

All control rods inserted/

reactor subcritical at the time of occurrence.

Occurred during refue 1 i ng while plant was shutdown; Occurred during refueling while plant was shutdown.

Occurred during refueling while plant was shutdown.

Occurred during refueling while plant was shutdown.

Not stated, but assumed to based on the fact that the breaker was operable.

be

• Licensee states that the ANO~l and AN0-2 Technical Specifications only require reporting of RTB failures to trip when the reactor protection system is required to be operable.

11.ttachment 1 I I /

Plant Davis-Besse (AK-Breakers)

Davis-Besse (AK:..Breakers)

McGuire Unit 2 (OS-Breakers)

Oconne Unit 3 (AK-Breakers)

Three Mile Island-1 (AK-Breaker)

~ TABLE 1 (Continued)

~

Date *of Event 9/28/79 9/1/82 Early 1983 12/17 /78 Licensee's Description Trip Confirm light failed to clear after the breaker was closed.

(Breaker 11A 11

)

Trip confirm failed to clear after the breaker was c 1 o s e d *

( B r.e a k er 11 B " )

During preoperational testing, one RTB failed to trip on a UV demand.

The breaker failed to trip on five occa-sions while testing, after which the breaker was reworked with no apparent problems.

However, on 3/18/83, this breaker failed three times out of 125 cycles while being tested per the requirements of I.E.Bulletin 83-04.

During a step in the startup procedures test, one RTB did not open.*

• 11/19/76 One RTB failed to trip during "Post Maintenance Testing.

1

~

Licensee 1s Reason for Not Previously Reporting Not stated, but assumed to be based on the fact that the breaker was operable.

Not stated, but assumed to be based on the fact that the breaker was operable.

Not specifically stated, but presumed to be based on the fact that the plant was still in a preoperational state.

At the time of the event, it was not considred reportable; however, such an event is now reportable pursuant to existing Technical Specifi-cations at the Oconee facility.

Failure was due to binding~

which was attributed to effects of transporting and installing the breaker subsequent to maintenance.

Malfunction was detected prior to returning RTB to servic~.

Attachment l.

U.S. NUCLEAR REGULATORY COMMISSION d

• OMB 3150-0011 Approve

• Revision 4 August 1975 REGULATORY GUIDE OFFICE OF STANDARDS DEVELOPMENT REGULATORY GUIDE 1.16 REPORTING OF OPERATING INFORMATION-APPENDIX A TECHNICAL SPECIFICATIONS A. INTRODUCTION Section 50.36, "Technical Specifications," of 10 CFR Part 50, "Licensing of Production and Utilization Facilities," requires that each applicant for a license authorizing operation of a nuclear power plant include 74 tomic Energy Commis-in its application proposed technical specifications.

ub shed. Revision 2 of Regula-These technical specifications, as issued by the NRC, are '

sion reflected results of a.staff incorporated into the facility license and are conditions infonnation needed to permit of the license. Technical specification.s are now included Commission of safety-related activities as two appendices to the license: Appendix A technical erating phase of plant life. Significant specifications relate to health and safety, and Appendix 1sion 2 were:

B techniCal specifications relate to environmental im-porting requirements were updated to reflect pact.1 Each of these appendices includes a sectiono in reports required by Appendix A technical

~epor~ing requiremen~s. ~e reporting progra1:11 descri ci 1cations. In general~ these changes involved:,

m this regulatory guide mvolves the reportm u

a.

a change m frequency of submittal of ments of Appendix A technical specificatio o y. In

. routine operating reports; some cases, this program may need to be su

b.

elimination of the first~year operating or modified because of unique plant design report; other factors. The need for a supplemental or

c.

fonnalization of reporting of operating program will be determined on a case-by-case basis.

infonnation on a monthly frequency;

.Reporting of informati cerning radioactive

d.

deletion of certain items of infonnation no discharges, radiological en

• monitoring, and

• longer required to be submitted on a routine basis; noruadiologfo.t en:e e co md environ-

e.

clumge. In the format and imm*di"'l' of mental impact is

• n atory Guide 4.8,.

reporting required for certain.types of abnonnal occur-

"Environmen. tal Te "cal eci 1cations for Nuclear rences (now called reportabl'e occurrences); and Power Plants.~'

f.

improved guidance concerning defmitions In additi to g requirements necessary and categories of significance of abnormal occurrences.

for compliance technical specifications~ specific re-.

2.. Appendices were added to provide the desired porting requirem are included in Part 50, as well as format for radiation exposure reports and monthly in other Parts *of 1de 10, Chapter I, Code of Federal operating reuorts.

Regulations. A compilation of all repoi:ting requirements

3.

A lis!iq.g of reports other than those. required applicable to the various types of NRC licensees, includ-by Appendix A technical specifications* was eliminated.

ing identification of the proper NRC addressee 'or ad (See Introduction above.)

dressees and designation of the number of copies reqliir-ed, is included in Regulatory Guide 10.l, Compilation 1 A few facilities. *have a single appendix that

• combined aspect of Appendices A and B.

USN RC REGULATORY GUIDES contains the Regulatory Guides are issued to de&cribe and make avail*ble to the public

• methods acceptable to tho NRC staff of implementing specific puts of th*

Commission's regulations, to delineate techniques used by the stmff in evalu-ating specific probloms or postulated accidents. or to provide guidance.to appli-cants. Regulatory Guides are not substitutes for ragulation1. and compliance with them is not required. Methods and solutions different from tho10 aet out in the guides will be acceptable if they provide a basis.for the finding!' requisite to the issusnce or continuance of a p~rmit or license by tho Commission.

Comments and suggestions for improveme~ts in these guid9S, er; encouraged at all times, and guides will be revised, as apprOpriate, to accommodate com-ments and to reflect ne.w information or experience. However. commant* on this guide, if received within about two months after its issuanc~. will be JNr*

ticularly useful in evaluating the need for an early revision.

2The Atomic Energy Commission was abolished by the Energy Reorganization Act of 1974, which also created the Nuclear Regulatory Commission and gave it the licenSing and related regulatory fu'.nctions of the AEC.

Com.ments should be sent to the Secretory of the c'ommiuion. U.S. Nuclear RegUlatory Commiuion. ~ashington, D.C. 20556, Attention: Docketin; and Service Section.

The guidos ore isauod in the following ten bro*d divisions:

1. Power Reactors

2. ReBe*rch *nd Test Reactors J. Fuela and Materi*ls Faciliti~*

4. Environmontal and Siting s; Mtaterials and Plant Protection

_i. Products

1. Transponation

• S.. Occupational Hemlth

9. Antitrust RovleW

10. Genorol

~ Copies of publia.h*d ~uides m*y be obt*ined bv written request i~dlcating the divisioms dnired to the U.S. Nuclnar Regulatory Commiuion, Washington. D.C.

~*Attention: Director. Office of Standards Development.

Comments were invited within 60 days of publica-tion of Revision 2 for use in conjunction with early revision of the guide. As a result of comments received on the guide and additional staff review, the staff developed Revision 3. Significant changes in Revision 3 were:

1.

The startup report was revised to be more specific as to the test results to be reported.

2.

The annual report section was revised to (1) further quantify the term "reduction in power," (2) provide further guidance on reporting of occupational radiation exposures, and (3) revise the information to be submitted on fuel performance.

3.

The abnormal occurrence report section was revised to (1) provide for prompt notification by telephone and confirmation of such notification by telegraph, mailgram, or facsimile transmission of the types of abnormal occurrences listed under Section 2.a; (2) be more specific on the types of abnormal occur-rences reported, (3) delete radiological effluent releases from Appendix A technical specification reporting requirements, ( 4) provide for reporting of the types of abnormal occurrences listed under Section 2.b within 30 days of occurrence of the event, and (5) make Section 2.c of Revision 2 of the guide a separate section (Section 4).

In previous revisions of Regul.3.tory Guide 1.16, the term "abnormal occurrence" was used to designate any unscheduled or unanticipated operational event reported to the Commission. Included in these reported events were (1) events that could or did have significance from the standpoint of public health or safety and (2) events reported to NRC for performance evaluation and trend determinations. In Section 208 of the Energy Reorgani-zation Act of 1974 (Pub. L.93-438), an "abnonnal occurrence" is defined for the purposes of the reporting requirements of the Act as *an unscheduled incident or event which the Commission determines is significant from the standpoint of public health or safety. In order to be consistent with this definition, the events desig-nated in previous revisions of this guide as "abnormal occurrences" are designated "reportable occurrences" in Revision 4. Any "reportable occurrences" that are.

determined by the Commission to be significant from the standpoint of public health or safety will be further designated "abnormal occurrences."

C. REGULATORY POSITION In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following program for reporting of operating information provides an acceptable basis to the NRC staff for meeting the reporting requirements of Appendix* A teclmical specifi-cations. Reports submitted in accordance with this guide should be addressed to the Director of the appropriate NRC Regional Office unless otherwise noted.

• lines indicate substantive changes froni previous issue.

1.

Routine Reports

a.

Startup Report.

A summary report of plant startup and power escalation testing should be submitted following (1) receipt of an operating license, (2) amendment to the license *involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

The report should address each of the tests identified in the FSAR and should in general include a description of the measured values of the operating conditions or characteristics obtained during the test program a..'1.d. a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation should also be de-scribed. Additional specific details may be included in license conditions based on the applicant's commitment to applicable regulatory guides and should be included in this report.

Startup reports should be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or com-mencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest.

• If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test pro-gram, and resumption or commencement of commercial power operation), supple"menl:ary reports should be submitted at least every three months until all three events have been completed.

b.

Annual Operating Report. 3 Routine operating reports covering the opera-tion of the unit during the previous calendar year should be submitted prior to March 1 of each year. The initial report should be submitted prior to March I of the year following initial criticality.

The primary purpose of annual operating reports is to permit annual evaluation by the NRC staff of operat-ing and maintenance experience throughout the nuclear power industry. The annual operating reports r.11<1.de by licensees should provide a comprehensive summary of the operating experience gained during the year, even though some repetition of previously reported informa~

tion may be involved. References in the annual operating report to previously submitted reports should be dear.

Each annual operating report should include:

(1) A narrative summary of operating experi-ence during the report period relating to safe operation of the facility, including safety-related maintenance not covered in item l.b.(2Xe) below.

3 A sini;lc submittal may be made for a multiple unit station. The submittal should combine those sections that arc common to all units at the st.ation.

1.16-2

(2) For each outage or forced reduction in power4 of over 20 percent of design power level where the reduction extends for more than four hours:

(a) the proximate cause and the system and major component involved (if the outage or forced reduction in power involved equipment malfunction);

(b) a brief discussion of (or reference to reports of) any reportable occurrences pertaining to the outage or power reduction; (c) corrective action taken to reduce the probability of recurrence, if appropriate; (d) operating time lost as a result of the outage or power reduction (for scheduled or forced out-ages, 5 use the generator-off-line hours; for forced re-ductions in power, use the approximate duration of op-eration at reduced power);

(e) a description of major safety-related corrective maintenance performed during the outage or power reduction, including the system and component involved and identification of the critical path activity dictating the length of the outage or power reduction; and (f) a report of any single release of radio-1 activity or single radiation exposure specifically associ-ated with the outage which accounts for more than 10 percent of the allowable annual values.

{3) A tabulation on an annual basis of the number of station, utility, and other personnel (in-cluding contractors) receiving exposures greater than 100 mrem/yr and their associated man-rem exposure according to work and job functions,6 e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe mainte-nance), waste processing, and refueling. The dose assign-ments to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements.

Small exposures totalling less than 20 percent of the individual totaJ dose need not be accounted for. In the aggregate, at least 80 percent of the total whole body dose received from external sources should be assigned to specific major work functions. See Appendix A to this guide for a standard format for providing this information.

( 4) Indications of failed fuel resulting from irradiated fuel examinations, including eddy current 4The term "forced reduction in power" as used in this guide and as normally defined in the electric power industry means the occurrence of a component failure or other condition that requires that the load on the unit be reduced for corrective action immeruately or up to and including the very next weekend.

Note that routine preventive maintenance, surveillance, and cah"bration activities requiring power reductions are not covered by this section.

5The term "forced outage" as used in this guide and as normally defined in the electric power industry means the occurrence of a component failure or other condition that requires that the unit be removed from service for corrective action immediately or up to and including the very next weekend.

6This tabulation supplements the requirements of § 20.407 of IO CFR Part 20.

tests, ultrasonic tests, or visual examinations completed d!Jring the report period.

c.

Monthly Operating Report.

Routine reports of operating 'statistics and shutdown experience should be submitted on a monthly basi_s. The report formats set forth in Appendices B, C, and D to this guide should be completed in accordance with the instructions provided. The completed forms should be submitted by the tenth of the. month following the calendar month covered by the report to the Director, Office of Management Information and Program Control, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the appropriate NRC Regional Office.

2.

Reportable Occurrences Guidance concerning reportable occurrences that should be reported in different time frames is provided below. Supplemental reports may be required to fully describe final resolution of the occurrence. In cases of corrected or supplemental reports, a licensee event report should be completed and reference should be made to the original report date.

a.

Prompt Notification With Written Followup.

The types of events listed below should be reported as expeditiously as possible, but within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirmed by telegraph, mail-gram, or facsimile transmission to the Director of the appropriate NRC Regional Office, or his designee, no later than the first working day following the event, with a written followup report within two weeks. A copy of the confirmation and the written followup report should also be sent to the Director, Office of Management I Information and Program Control, USNRC. The written followup report should include, as a minimum, a completed copy of the licensee event report form (see Appendix E to this guide) used for entering data into the NRC's computer-based file of infonnation concerning licensee events. (Instructions for completing these licensee event report forms7 are issued individually to each licensee.) Information provided on the licensee event report form should be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.

(1) Failure of the reactor protection ~ystem or other systems subject to limiting safety-system settin~

to initiate the required protective function by the time a monitored parameter reaches the setpoint specified as the limiting safety-system setting in the technical specifi-7Instruction Manual, Licensee Event Report File, Office of I Management Information and Program Control, U.S. Nuclear Re~latory Commission, Washington, D.C. 20555.

1.16-3

• .'r

cations or failure to complete the required protective function. The following are examples:8

-(a) Reactor pressure exceeds limiting safety-system setting value without automatic µip.

(b) Inability to trip and insert sufficient control rods to achieve the technical specification shut-down margin.

(c) Failure of the reactor protective system to complete the required protective action once initi-ated.

Note: Instrument drift discovered as a result oftesting need not be reported under this item but.may be report-able under items 2.a(5),2.a(6), or 2.b(l) below.

(2) Operation of the unit or affected systems when any parameter or operation subject fo a limiting condition for operation is less conservative than the least conservative aspect of the limiting condition for opera-tion established in the technical specifications. The fol-lowing are examples:

(a) Shutdown not begun within the speci-fied time when unidentified reactor coolant leakage ex-ceeds the technical specifications limit.

(b) Failure of a system other than the systems subject to limiting safety-system settings (see 2.a(l) *above) to actuate, or actuation of such a system at a monitored parameter value less conservative than that listed in the technical specifications for the system.

(c) Operation with unacceptable contain-ment leak rate type B or C test results.

(d) System cooldown at a rate exceeding the technical specifications limit.

Note: If specified action is taken when a system is found to be operating between the most conservative and the least conservative aspects* of a limiting* condi,tion for operation listed in. the technical specifications, the limiting condition for operation is not considered to have been violated and need not be reported under this item, but it may be *reportable under item 2.b(2) below.

(3) Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment. The following are examples:

(a) Through-wall failure of piping or com-.

ponents of the reactor coolant pressure boundary.

(b) Steam generator tube thinning in excess of acceptance limits in Regulatory Guide 1.83,

• "Inservice hspection of Pressurized Water Reactor Steam Generator.Tubes."

(c) Welding or material defects greater than those allowable by applicable codes.

Note: Leakage of valve packing or gaskets within the

• limits fcir.1dentified leakage set forth in te-chitical specifi-cations need not be reported under this item.

8Examples are* Intended to be illustrative only.

( 4) Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady-state conditions during power operation greater than or equal to 1 % il.k/k; a calculated reactivity balance J indicating a shutdown margin less conservative than specified in the technical specifications; short-term react-ivity increases that correspond to a reactor period of less than 5 seconds or, if subcritical, a_i:i unplanned reactivity insertion of more than 0.5% il.k/k; or occurrence of any I unplanned criticality.

• (5) Failure or malfunction of one or more com-ponents which prevents or could prevent, by itself, the fulfillment of the functional reqtiirements of system(s) used to cope with accidents analyzed in the SAR. The following are examples:

(a) Clogged fuel line(s) resulting in failure to supply fuel to the emergency generators.

(b) Multiple ins~rument drift resulting in loss of protective function.

(c) HPCI failure to start or failure to con-tinue running once initiated.

( 6) Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfill-ment of the functional requirements of systems required to cope with accidents analyzed in the SAR. The follow-i.iig are examples:

(a) Failure _to restore a safety system to operability following test or maintenance.

(b) Improper procedure ieading to in-correct valve lineup which resulted in closure of one manual valve in each of two redundant safety injection subsystems. and would have prevented injection on demand.

Note: For items 2.a(5) and 2.a(6) reduced redundancy that does not result in loss of system function need not be reported under this section but may be reportable under items 2.b(2) and 2.b(3) below.

(7) Conditions arising from natural or man-made events that, as a direct result of the event, require plll!J.t shutdown, operation of safety systems, or other protective measures required by technical-specifications.

The following are e}Wllples:

(a) Threatened civil disturbances requiring plant shutdown.

(b) Damage to the facility caused by fire, flood, earthquake, or other similar occurrences.

(8) Errors discovered in the transient or accident analyses or in the methods used for such analyses as described iri the safety analysis report or in the bases for the technical specifications that have or could have perriiitted reactor operation in a manner less conservative than assumed in the analyses. The following are examples:

(a) Loss of condenser vacuum resulting in reactor pressure and flux transients that peak at values higher than analyzed.

1.16-4

(b) Reactivity insertion delay times by reactor protection system longer than those used in the technical specification bases.

(9) Performance of structures, systems, or com-ponents that requires remedial action or corrective measures to prevent operation in a manner less conserva-tive than that assumed in the accident analyses in the safety analysis report or technical specifications bases; or discovery during plant life of conditions not specifically considered in the safety analysis report or technicai" spe-cifications that require remedial action or corrective measures to prevent the existence or development of an unsafe condition. The following are examples:

(a) Axial' flux ratios less conservative than those for which correlations with overpower AT were based on core bumup projections.

(b) Failure of a safety injection pump to deliver the flow rates assumed in the FSAR.

(c) Degradation of hydraulic shock sup-pressors to the extent that they could not perform their required safety function.

( d) Failure of magnetic trip mechanisms on a safety-related circuit breaker to provide trip on instantaneous overcurrent as indicated on the manufac-turer's time-current characteristic curve.

( e) Failure of a safety /relief valve to close after pressure has reduced below the required reseat valve.

(f) Thermal shock to the reactor coolant system resulting from inadvertent safety injection actua-tion.

Note: This iteni is intended to provide for reporting of potentially generic problems.

b.

Thirty-Day Written Reports.

The reportable occurrences discussed below should be the subject of written reports to the Director of the appropriate NRC Regional Office within 30 days of occurrence of the event. A copy of the written report I

should also be sent to the Director, Office of Manage-ment Information and Program Control. The written report should include, as a minimum, a completed copy of the licensee event report form (see Appendix E to this guide) used for entering data into the NRC's computer-based file of information concerning licensee events. (Instructions for completing these licensee event report forms7 are issued individually to each licensee.)

Information provided on the licensee event report form should be supplemented, as needed, by additional narra-tive material to provide complete explanation of the cir-c~mstances surrounding the event.

(1) Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those established by the technical specifications but which do not prevent the fulfillment of the functional requirements of affected systems. The following are *examples:

(a) One of the four scram dump volume level switches failed to operate during surveillance test.'

(b) One of four reactor low-pressure switches operated at 885 psig instead of l.SSS value of 900 psig.

(c) During test, one out of four under-voltage relays failed to perform its function of tripping a reactor trip breaker.

- (2) Conditions leading to operation in a de-graded mode permitted by a limiting condition for operation, or plant shutdown required by a limiting condition for operation. The following are examples:

(af Core spray pump breaker tripped after 20 minutes during test. Trip unit was found to be defective, declared inoperable, and repaired.

(b) Safety injection pump failed to start following system initiation. Required surveillance on redundant components was successfully completed.

(c) One of the two centrifugal charging pumps became inoperable because of a faulty bearing.

Redundant pump operability was confirmed.

Note: Routine surveillance testing, instrument c~bra tion, or preventive maintenance which require system configurations as described in items 2.b(l) and 2.b(2) need not be reported except where test results them-selves reveal a degraded mode as described above.

(3) Observed inadequacies in the implementa-tion of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems. The following are examples:

(a) One of the three diesel generators tripped from high temperature because cooling water valves were lined up incorrectly.

(b) Isolation valve for a low-pressure trip switch was found closed with system pressure locked in.

Trip of switch would not occur at low pressure.

Improper return to operation following maintenance was the cause.

( c) Failure to perform surveillance tests at the required frequency.

( 4) Abnormal degradation of systems other than those specified' in item 2.a(3) above designed to contain radioactive material resulting from the fission process. For example, a through-wall leak in a liquid waste storage tank.

Note: Sealed sources or calibration sources are not included under this item. Leakage of valve packing or gaskets within the limits for identified leakage set forth in technical specifications need not be reported under this item.

1.16-5

3.

Unique Reporting Requirements The above reporting program will in general satisfy the reporting requirements necessary for compliance with Appendix A technical specifications. This program may need to be supplemented or modified because of unique plant design features or other factors. The need for a supplemental or modified program will be deter-mined on a case-by-case basis and so designated in individual operating licenses.

4:

Events of Potential Public Interest The types of events listed below are freqµently of high public interest. While some o.f the events may not be reportable by regulation or defined in other parts of this guide, the Director of the appropriate NRC Regional Office, or his designee, should be informed of such events by telephone as soon as possible after the event has been discovered.

a.

An event that causes damage to property or equipment when such damage affects the power pro-duction capability of the facility.

b.

Radiation ex;posure to licensee personnel or members of the public in excess of applicable exposure limits set forth in 10 CFR Part 20.

c.

Natural or man-made conditions that may require action which need not be reported under item 2.a(7) above.

d.

Discovery of significant radiological event off-site occurring during transport of material for which the licensee was either shipper or consignee.

e.

Unscheduled shutdowns expected to last for more than one week, regardless of cause.

f.

Unusual releases of radioactive material from the site boundary not reportable under other require-ments.

g.

Failure of or damage to safety-related equip-ment which need not be reported under item 2.a above, if the time for repair is likely to exceed the time allowed by the technical specifications.

D. IMPLEMENTATION The purpose of this section is to provide informa-tion to applicants and licensees regarding the NRC staffs plans for utilizing this regulatory guide.

Except in those cases in which the applicant proposes an acceptable alternative method, the reporting program described herein is being used by the NRC staff in order to standardize the reporting requirements section of Appendix A technical specifications of all operating licenses.

For licensees holding operating licenses without Appendix B environmental technical specifications, it may be necessary to include those reports identified in Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from* light-Water-Cooled Nuclear Power Plants," and Regulatory Guide 4.1, "Programs for Monitoring Radio-activity in the Environs of Nuclear Power Plants," in the technical specifications under the unique reporting requirements section of the technical specifications.

1.16-6

APPENDIX A STANDARD FORMAT FOR REPORTING NUMBER OF PERSONNEL AND.MAN-REM BY WORK AND JOB FUNCTION Number of Personnel(> 100 mrem)

Total Man-Rem Contract Workers Contract Workers Work & Job Function Station Employees Utility Employees and Others Station Employees Utility Employees and Others Reactor Operations &. Swveillance Maintenance Personnel Operating Personnel Health Physics Personnel Supervisory Personnel Engineering Personnel Routine Maintenance Maintenance Personnel Operating Personnel Health Physics Personnel Supervisory Personnel Engineering Personnel Inservice Inspection Maintenance Personnel Operating Personnel Health Physics Personnel Supervisory Personnel Engineering Personnel Special Maintenance Maintenance Personnel Operating Personnel Health Physics Personnel Supervisory Personnel Engineering Personnel Waste Processing Maintenance Personnel Operating Personnel Health Physics Personnel Supervisory Personnel Engineering Personnel Refueling Maintenance Personnel Operating Personnel Health Physics Personnel Supervisory Personnel Engineering Personnel TOTAL Maintenance Personnel Operating Personnel Health Physics Personnel Supervisory Personnel Engineering Personnel Grand Total

APPENDIX B AVERAGE DAILY UNIT POWER LEVEL MONTH _________ _

DAY AVERAGE DAILY POWER LEVEL (MWe-Net) 1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 DOCKETNO. ~----

UNIT _____ _

DATE ------

COMPLETED BY TELEPHONE -----

DAY AVERAGE DAILY POWER LEVEL (MWe-Net) 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 INSTRUCTIONS On this form, list tlie average daily unit power level in MWe-Net for each day in the reporting month. Compute to the nearest whole megawatt.

These figures will be used to plot a graph. for each reporting month. Note that when maximum dependable capacity is used for the net electrical rating of the-unit, there may be occasions when the daily average power level exceeds the 100% line (or the restricted power level line). In such cases, the average daily unit power output sheet sho~ld be footnoted to explain the apparent anomaly.

1.16-81

_J

APPENDIX c*

OPERATING DATA REPORT OPERATING STATUS DOCKET NO.~-------

UNIT ----------

DATE ----------

COMPLETED BY ---------

TELEPHONE---------

1. REPORTING PERIOD:

GROSS HOURS IN REPORTING PERIOD: ________

2. CURRENTLY AUTHORIZED POWER LEVEL (MWt):

MAX. DEPEND. CAPACITY (MWe-Netl: _____

DESIGN ELECTRICAL RATING (MWil-Netl:_* ------

3. POWER LEVEL TO WHICH RESTRICTED (IF ANY) (MWe-Net): --------------

4. REASONS FOR RESTRICTION (IF ANY);

THIS MONTH YR TO DATE CUMULATIVE

5. NUMBER OF HOURS REACTOR WAS CRITICAL ***************-----

6. REACTOR RESERVE SHUTDOWN HOURS................... -----

7. HOURS GENERATOR ON LINE.......*.................. -----

B. UNIT RESERVE SHUTDOWN HOURS...................**. -----

9. GROSSTHERMALENERGYGENERATED(MWH) *............ *-----

10. GROSS ELECTRICAL ENERGY GENERATED (MWH)............. -----

11. NET ELECTRICAL ENERGY GENERATED (MWH)......*.....*. -----

12. REACTOR SERVICE FACTOR.*.......*..*..........*..* -----

13. REACTOR AVAILABILITY FACTOR..................,.... -----

14. UNIT SERVICE FACTOR *.*...................*....*.. -----

15. UNIT AVAILABILITY FACTOR.**....**.........**..*.** -----

16. UNIT CAPACITY FACTOR (Using MDC)....*................ -----

17. UNIT CAPACITY FACTOR (Using Design MWe)................. -----

18. UNIT FORCED OUTAGE RATE..**.*............**..**.. -----

19. SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, AND DURATION OF EACH):

20.'" IF SHUT DOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP: _______ _

21. UNITS IN TEST STATUS (PRIOR TO COMMERCIAL OPERATION):

FORECAST ACHIEVED INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION 1.16-9

INSTRUCTIONS FOR COMPLETING OPERATING DATA REPORT Tilis report should be furnished each month by licensees. The name and telephone number of the preparer should be provided in the designated spaces.

The instructions below are provided to assist licensees in reporting the data consistently. The number of the instruction corresponds to the item number of the report form.

1. Reporting Period. Designate the month for which the data are presented. The Gross Hours are normally from 0001 of the first day through 2400 of the last day of the calendar month, with appropriate adjustments for any month in which a change from standard to daylight-saving time (or vice versa) is made. The only two shorter reporting periods are (1) the one in which the initial electrical generation occurs and (2) the one in which the reactor is shut down for decommissioning. In the former, the gross hours, ~xpressed to the nearest tenth of an hour, are those from the time of initial power generation to 2400 of the last day of the calendar month. In the latter case, the gross hours, expressed to the nearest tenth of an hour, are those from 0001 of the calendar month to the specific time.of final shutdown.

2. The Authorized Power Level is the maximum thermal power, expressed L."1 megawatts, currently author.zed by the Nuclear Regulatory Commission.

The net Maximum Dependable Capacity is the gross electrical output as measured at the output terminals of the turbine-generator during the most restrictive seasonal conditions less the normal station service loads.

The net Design Electrical Rating is the nominal net electrical output of the unit specified by the utility and used for the purpose of plant design.

3. Note that this item is applicable only if restric-tions on the power level are in effect. Short-term (less than one month) limitations on power level need not be presented in this item, since one of the important purposes of the item is to determine if, and at what power level, a restricted power level line should be drawn on the chart of average daily reactor power.

Since this information is used to develop figures on capacity lost due to restrictions and because most users of the "Operating Plant Status Report" are primarily interested in energy actually fed to the distribution system, it is requested that this figure be expressed in MWe-Net in spite of the fact that the figure must be derived from MWt or percent power.

4. Reasons for Restriction (if Any). If item 3 is used, item 4 explains why. Brief narrative is acceptable.

Cite references as appropriate. Indicate whether restric-tions are self-imposed or are regulatory requirements. Be as specific as possible within space limitations. Piants in startup and power ascension test phase should be identified here.

5. Show the total number of hours the reactor was critical during the gross hours of the reporting period.

6. Reactor Reserve Shutdown Hours. The total number of hours during the gross hours of reporting period that the reactor was removed from service for administrative or other reasons but was available for operation.

7. Hours Generator On Line. Also called Service Hours. The total number of hours during the gross hours of the reporting period that the unit operated with breakers closed to the station bus. These hours, plus those listed in Appendix D' for the generator outage hours, should equal the gross hours in the reporting period.

8. Unit Reserve Shutdown Hours. The total number of hours during the gross hours of the reporting period that the unit was removed from service for economic or similar reasons but was available for operation.

9. Gross Thermal Energy Generated. The thermal output of the nuclear steam supply system during the gross hours of the reporting period, expressed in megawatt hours.

10. Gross Electrical Energy Generated. The electrical output of the unit measured at the output terminals of the turbine-generator during the gross hours of the reporting period, expressed in megawatt hours.

11. Net Electrical Energy Generated. The gross elec-trical output of the unit measured at the output terminals of the turbine-generator minus the normal station service loads during the gross hours of the reporting period, expressed in megawatt hours. Negative quantities should not be used. If there is no net positive value for the period, enter zero.

12-18. For units still in the startup and power ascension test phase, items 12-18 should not be com-puted. Instead, enter N/ A in the current month column.

These seven factors should be computed starting at the time the unit is declared to be in commercial operation.

The cumulative figures in the second and third columns should be based on commercial operation as a starting date. However, units already in commercial operation, for which cumulative figures have been based on different starting dates, need not recalculate the cumula-tive figures.

1.16-11

12. Reactor SerVice Factor. Compute by dividing hours reactor was critical (item 5) by the gross hours in the reporting.period (item I). Express as percent to the nearest tenth of a percent. During months when the unit is shut down for the entire period because of nonreactor problems, enter "Not Applicable" and explain in the Summary of Appendix D. Do not include reserve shutdown hours in the calculation.

13. Reactor Availability Factor. Compute by divid-ing the reactor available hours (items 5 plus 6) by the gross hours in the reporting period (item 1 ). Express as percent to the nearest tenth of a percent.

14. Unit Service Factor. Col!lpute by dividing hours the generator was on line (item 7) by the gross hours in the reporting period (item I). Express as percent to the nearest tenth of a percent. Do not include reserve shutdown hours in the calculation.

15. Unit Availability Factor. Compute by dividing the unit available hours (item 7 phis item 8) by the gross hours in the reporting period (item 1 ). Express as percent to the nearest tenth of a percent.

16. Unit Capacity Factor (Using MDC). Compute by dividing net electrical energy generated (item 11) by the product of maximum dependable capacity (item 2) times the gross hours in the reporting period (item 1 ).

Express as percent to the nearest tenth of a percent.

17. Uiiit Capacity Factor (Using Design Electrical Rating). Compute as in item 16, substituting design electrical rating for maximum depen~able capacity.

e.

18. Unit Forced Outage Rate. Compute by dividing the total forced outage hours (from the table in Appendix D) by the sum of hours generator on line (item 7) plus total forced outage hours (from the table in Appendix D). Express as percent to the nearest tenth of a percent.

19. Shutdowns Scheduled to Begin in Next 6 Months. Include type (refueling, maintenance, other),

proposed date of start of shutdown, and proposed length of shutdown. It is recognized that shutdowns may be scheduled between reports and that this item may not be all inclusive. Be as accurate as possible as of the date the report is prepared.

20. Self-explanatory.

21. Self-explanatory. Note, however, that this infor-mation is requested for all units in startup and power ascension test status and is not required for units already in commercial operation.

Test Status is defined as that period following initial criticality during which the unit is tested at successively higher outputs, culminating with operation at full power for a sustained period and completion of warranty runs.

Following this phase, the unit is generally considered by the utility to be available for commercial. operation.

Date of Commercial Operation is defined as the date that the. unit was declared by the utility owner to be available for the regular production of electricity, usually related to the satisfactory completion of qualifi-cation tests as specified in the purchase contract and to the accounting policies and practices of the utility.

1.16-12

'?'

w NO.

DATE

SUMMARY

TYPE F: FORCED S: SCHEDULED APPENDIX D UNIT SHUTDOWNS.AND POWER REDUCTIONS REPORT MONTH -------

METHOD OF SHUTTING DOWN DURATION THE REACTOR OR (HOURS)

REASON (1)

REDUCING POWER (2),

(1)

DOCKET NO.------

UNIT NAME

OATE COMPLETED BY TELEPHONE ------

CORRECTIVE ACTIONS/COMMENTS REASON A; EQUIPMENT FAILURE (EXPLAIN)

B: MAINT. OR TEST C: REFUELING D: REGULATORY RESTRICTION E: OPERATOR.TRAINING AND LICENSE EXAMINATION F: ADMINISTRATIVE G: OPERATIONAL ERROR (EXPLAIN)

• H: OTHER (EXPLAIN)

(2)

METHOD 1: MANUAL 2: MANUAL SCRAM 3: AUTOMATIC SCRAM 4: OTHER (EXPLAIN)

UNIT SHUTDOWNS AND POWER REDUCTIONS INSTRUCTIONS This report should describe all plant shutdowns dur-ing the report period. In addition, it should be the source of explanation of significant dips in average power levels (Appendix B). Each significant reduction in power level (greater than 20% reduction in average daily power level for the preceding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) should be noted, even though the unit may not have been shut down completely.1 For such reductions in power level, the duration should be listed as zero, the method of reduc-tion should be listed as 4 (Other), and the Comments column should explain. The Comments column should be used to provide any needed explanation not ade-quately described by the coded columns. Please do not add to the list of codes or legends now furnished. Simi-larly, do not add additional columns.

Number. This column should indicate the sequential number assigned to each shutdown or significant reduc-tion in power for that calendar year. When a shutdown or significant power reduction begins in one report period and ends in another, an entry should be made for both report periods to be sure all shutdowns or signifi-cant power reductions are reported. Until a unit has achieved its first power generation, no number should be assigned to each entry.

Date. This column should indicate the date of the start of each shutdown or significant power reduction. Report as year, month, and day. August 14, 1975 would be reported as 750814. When a shutdown or significant power reduction begins in one report period and ends in another, an entry should be made for both report periods to be sure all shutdowns or significant power reductions are reported.

Type. Use "F" or "S" to indicate either "Forced" or "Scheduled," respectively, for each shutdown or signifi-cant power reduction. Forced shutdowns include those 1 Note that this differs from the Edison Electric Institute (EEi) definitions of "Forcl'Al Partial Outage" and "Scheduled Partial Outage." For these terms, EEI uses a change of 30 MW as the break point. For larger power reactors, 30 MW is too small a change to warrant explanation.

required to be* initiated by no later than the weekend following discovery of an off-normal condition. It is recognized that some judgment is required in categori-zing shutdowns in this way. ln general, a forced shut-down is one that would not have been completed in the absence of the condition for which corrective action was taken.

Duration. Self explanatory. When a shutdown extends beyond the end of a report period, count only the time to the end of the report period and pick up the ensuing down time in the following report periods. Report dura-tion of outages rounded to the nearest tenth of an hour to facilitate summation. The sum of the total outage hours plus the hours-the generator was on line (item 7 of Appendix C) should equal the gross hours in the report-ing period (item 1 of Appendix C).

Reason. Categorize by letter designation in accordance with the table appearing on the report form. If category H must be used, supply brief comments.

Method of Shutting Down the Reactor or Reducing Power. Categorize by number designation in accordance with the table appearing on the report form. If category 4 must be used, supply brief comments.

Corrective Actions/Comments. Use this column to am-plify or explain the reasons for each shutdown or signifi-cant power reduction, with the corrective action taken, if appropriate. The Comments column entries should provide identification of each shutdown or significant power reduction that occurs as a direct result of a re-portable occurrence on which a report has been or will be submitted. (This information may not be immediately evident for all such shutdowns, of course, since further investigation may be required to ascertain Whether or not a reportable occurrence was involved.) When a direct correlation can be made between a given shutdown and a specific reportable *occurrence report, the Comments column entry should state the reportable occurrence report number and date.

Summary. Write a brief summary description (3 to 4 sei1tences) of the highlights of operation of the unit for the reporting month. Include any comments required by item 12 of Appendix C.

1.16-14

APPENDIX E LICENSEE EVENT REPORT CONTROLBLOCK i.__..__.....__,___J.___.__~I LICENSEE NAME 1

6 LICENSE NUMBER 25

@iilll I

1-1 1-1 9

. 14. 1~5_..__.,__...__.___,_~~~__._....._~

REPORT,REPORT 26 LICENSE TYPE I I CATEGORY TYPE

SOURCE DOCKET NUMBER EVENT DATE I l I I I 69

~fON'T~ ~ ~ 6

_1 1~~1~1_-~1~1~1~1_6

_,J EVENT DESCRIPTION EVENT TYPE LLI 30 31 32 RE;PORT DAT'!'

I I I I I I I 74 75 80

@l@.lL-.~~~~~~~~~~~~~~~~~*~~~~~~~~~~~~~----

7 8 9 8(l

@Iii,__~~~~~--~~~~~~~~~~~~~~~~~~~~~~~~~~--'

7 8 g 80

~~

1 1

s~9~.----------------~--------------------_..;a--'o

@Ifil*.,._~~~~~~~~~~~~~--~~~~~~~~~~~~~~~~

7 8 9 80 1~1~'--~~~~~~~~~~~~~~~~~~~~~~~~~~~~~-'

7 8 9.SYSTEM*CAUSE' cor!4P~

CODE CODE COMPONENT CODE pp 1~MLD-u- cr-1 *1 1 r-1 s_uu~fl 7

89 10!

11 12 17*

431 CAUSE DESCRIPTION COMPONENT MANUFACTURER VIOLATION r* 1 i 1 r*

• u 44 47*

. 48 80

@:@].,.__~~~~~~~~~~~--~~~~--~~~~~~~~~~~~-'

~9

~

7 3~9----------------------------------------8 0

ITEi 7

8~9---------------------------------------.,........80 FACILITY STATUS

% POWER IIE1 LJ I I ! I 7

8

.9 10 12 13 l-8T~ CONTENT OTHER STATUS METHOD OF DISCOVERY LJ.

44 45

~

RE~EAiEO OLJLEASE AMOUNT OF ACTIVITY 7

8 9

10 11 44 45 PERSONNEL EXPOSURES r;r:;;t i\\JUMBER TYPE OESl°CRIP.TION DISCOVERY DESCRIPTION 46 80 LOCATION OF RELEASE l..'.J.:'.ll I I J LJ 7 s 9 11 121

-3.,... ------------------------------8-"0 PERSONN.EL INJURIES

[1]3~~ ~o-E_sc_R_1r_T_1o_N _______________________________ _.

7 8 9 11 12 80

.I PROBABLECONSEGUENCES

(!0L__

_J 1

S'Q

• ----~--~~~------~--~--------~-~~---~-~s*o 1_oss OR DAMAGE TO FACILITY

(![3TLJ DESCF!iPi'iC:JN 7 s 9 rn 80 PUBLICITY J]j]

"1 J 8 9 80 ADDITIONAL FACTORS IT@

7 6;~9---------~-~---------~----~-~~~-

80

!IIfil L~. ------

____ _J 7

8 9 80 NAME: _____

PHONE: ____

1.16-15

REPORTING REQUIREMENTS Regulatory Reporting Each licensed nuclear power facility must report certain types of operational events to the NRC.

Reporting requirements are delineated in various parts of Title 10 (

11 Energy 11

) Code of Federal Regulations (lo CFR) Chapter 1 and in each licensee 1 s technical specifications and/or license provisions.

Those of specific interest are in 10 CFR 50.36, 11Technical Specifications.

11 The regula-tory intent has always been to have items of potentially serious safety concern reported, such as the failure of protective devices to function properly.

Regulatory reporting requirements have existed since the beginning of licensing activities, and initially were stated in the individual licenses for nuclear power plants.

In. the early 1970s, Regulatory Guide (RG) 1.16, 11 Reporting of Operating Information, Appendix A Technical Specifications" was developed to provide uniform guidance on the information to be supplied to the regulatory body (the Atomic Energy Commission at the time).

During the early to mid 1970s, the importance of the collection, assessment and feedback of operational experience grew as the nuclear power industry expanded.

As a result, RG 1.16 was revised several times to clarify what information was de~ired. Revision 3 to the RG 1.16, issued in January 1975, not only specified the types of events to report, but also contained many notes and examples for added guidance.

Revision 4 to RG 1.16 was issued for public comment in August 1975; however, the sections pertaining to the type of events to report, and the associated notes and examples, wer~ not

Significantly affected.

As discussed.further in the "License Technical Specifica-tion Requirements" section all plant technical spe~ifications presently contgin reporting requirements based on and essentially identical to the reporting cri-teria in RG.1.16.

There are hov1ever, some differences in the specific guidance for reportable occurrences among plants and also differences in the interpretation of the requirements.

Further, there are differences among plants in other parts of the techncial specifications, such as the exact nature of the limiting condi-tions for operation.

Consequently, reporting varies among licensees. The specific reporting requirements applicable to the reporting of RTB and RTS failures are discussed in the following sections.

License Technical Specification Requirements 10 CFR 50.36, 11 Technical Specifications, 11 specifies items to be included in individual plant technical specifications, including specifying requirements for the licensee to notify the Commission when safety limits are exceeded, or limiting conditions for operation are not met.

In addition, the technical specifications also include requirements for certain reports titled 11 Reporting Requirements."

Included in these requirements is the reporting of certain off-normal events (

11 Reportable Occurrences 11

) involving safety-related matters including co~ponent, system and structure failures.

The technical specifica-tions define two types of Reportable Occurrences:

(a) those requiring prompt notification with written followup within two weeks, and (b) those requiring thirty-day written reports.

The written reports, whether two-week or thirty-day, are known ~s Licensee Event Reports* (LERs);

• , In October 1974, the Office of Nuclear Reactor Regulation (NRR) initiated a

• program to update the reporting requirements specified in the technical

• specifications of power reactor licensees.

Licensees were issued specifications that either referred to RG 1.16 or used words from certain sections of the Guide.

Based on the later revisions to RG l.16, NRR initiated action to again update reporting requirements in the technical specifications.

As a result, all plant technical specifications now contain reporting criteria essentially identical to RG 1.16 regarding the types of events to report.

However, none contain the specific examples to RG 1.16.

In addition, not all plant technical specifica-tions. include the clarifying notes from RG 1.16. A typical set of reporting requirements which do contain the clarifying notes is shown in Appendix A.

Experience has.shown, however, that the benefits of standardizing these reporting requirements have not been fully realized, not only because of differing interpretations of these requirements, but also because other sections of the technical specifications (e.g., LCOs - limiting conditions for operation) to which the reporting requirements refer remained non-standardized among plants.

Thus, since the LCOs may vary among plants, the reporting may also vary.

Furthermore, there may be ambiguity in the technical specifications on whether or not failures are reportable which are found during plant modes when equipment

. is not required to be operational.

Another ambiguity is associated with the definition of component failure and whether sluggish or delayed operation must be reported.

Finally, there is a question whether RTB failures are reportable when no limiting condition for operation exists in the technical specifications.

In summary, the requirements for reporting single apparent random failures of the RTB vary, and may depend upon the specific circumstances.

The guidance in RG 1.16 and the typical technical specifications have been interpreted to require for reporting of Reactor Trip System (RTS) failure, (i.e., failure of two series RTBs to open as in the Salem events) and RTB related failures as follows:

a.

Failures of the reactor to trip (either automatically or manually) are clearly reportable under Appendix A, Paragraph 1.2.a.1.

The reporting requirements are very specific regarding the total system's failure to

• function.

Such an event is also reportable under 10 CFR 50.72, which became a regulatory requirement in February 1980.

To date, the February 22 and 25 failures of Salem Unit 1 to automatically trip upon receipt of a valid RPS signal are the only known events of this type.

b.

A failure of the UV trip device, or any other failure which causes one RTB to fail to open an actual operational demand, is considered reportable under Appendix A, Paragraph 1.2.a.9.

~.;

c.

A failure of the RTB to open during surveillance testing is considered reportable under Appendix A~ Paragraph 1.2.b.2 for plants that have limiting conditions for operation imposed on the reactor trip breakers.

For those plants without LCOs on t~e RTBs, reporting could be required under Paragraphs 1.2.a.9, 1.2.b.1 or 1.2.b.3 of Appendix A.

The intent of RG 1.16 seems clear in this regard, and is as interpreted and implemented by most licensees.

For example, the current version of of RG 1.16 contains an example for Appendix A, Paragraph 1.2.b.1, which though portraying an event involving an UV relay not actuating at the proper setpoint to trip the RTB, provides for the reporting of a single random failure in the RPS.

The example reads, "During test, one out of four undervoltage relays failed to perform its function of tripping a reactor trip breaker.

11 In summary, failure of the RTS to trip the reactor is clearly reportable, and the ~andom single failure of an RTB to open in response to an actual or test signal is also considered reportable. However, the specific technical specifi-cation requirements for individual plants govern.

P.ttachment 3

Typical Technical Specification Section on Reporting Requirements for Reportable Occurrences

1.

Reporting Requirements In addition to the applicable.reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the Administrator of the cognizant NRC Regional Office unless otherwise noted.

1.1 Routine Reports*

1.2 Reportable Occurrences Reportable Occurrences, including corrective actions and measures to prevent recurrence, shall be reported to the NRC.

Supplemental reports may be required to fully describe final resolution of an occurrence.

In case of corrected or supplemental reports, reference shall be made to the original report date.

[These reporting requirements apply only to Appendix A (of the license) Technical Specifications.]

a.

Prompt Notification With Written Follow-Up.

The types* of events listed below shall be reported as expeditiously as possible, but within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirmed by telegraph, mail-gram, telecopy or facsimile transmission to the Administrator of the cognizant NRC Regional Office, or his designate no later than the first working day following the event, with a written follow-up report within two weeks.

The written follow-up report shall include material to provide complete explanation, cause of the event, the circumstances surrounding the event, any corrective action, and component failure data.

1.

Failure of the reactor protection system or other systems subject to limiting safety system settings to initiate the required protective function by the time a monitored para-meter reaches the setpoint specified as the limiting safety system setting in the Technical Specifications or failure to complete the required protective function, Note*:

Instrument drift discovered as a result of testing need not be reported under this item but may be reportable under items 1.2.a.5, 1.2.a.6, or 1.2.b.1 below.

• The 11 Nofes'~ shown in this Appendix are not included in all licensees* technical specifications.

/l,ppendi x A

2.

Operation of the unit or affected systems when any parameter or operation subject to a limiting condition is less.con-servative than the least conservative aspect of the limiting condition for operation established in the Technical Speci-fications.

Note:

If specified action is taken when a system is found to be operating between the most conservative and the least conservative aspects of a limiting condition for operation listed in the Technical Specifications, the limiting condition for operation is not considered to have been violated and need not be reported under this item, but it may be reportable under item 1.2.b.2 below.

3.

Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment.

Note:

Leakage of valve packing or gaskets within the limits for identified leakage set forth in the Technical Specifica-tions need not be reported under this item.

4.

Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady state conditions during power operation greater than or equal to 1%

k/k; a calculated reactivity balance indicating a shutdown margin less conservative than specified in the Technical Specifi-cations; short term reactivity increases that correspond to a reactor period of less than 5 seconds or, if sub-critical an unplanned reactivity insertion of more than 0.5%

k/k; or occurrence of any unplanned criticality.

5.

Failure or malfunction of one or more components which prevents or could prevent, by itself, the fulfillment of the functional requirements of system(s) used to cope with accidents analyzed in the FSAR.

6.

Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in the FSAR.

..... 7.

Note:

For items 1.2.a.S and 1.2.a.6 reduced redundancy that does not result in a loss of system function need not be reported under this section but may be reportable under items 1.2.b.2 and 1.2.b.3.

Conditions arising from natural or man-made events that, as a direct result of the event required plant shutdown, Appendix A

e e operation of safety systems, or other protective measures required by Technical Specifications.

8.

Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the FSAR or in the bases for the Technical Specifications that have or could have permitted reactor operation in a manner less conservative -than assumed in the safety analyses.

9.

Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the FSAR or Technical Specifications bases; or discovery during plant life of conditions not specifically considered in the FSAR or Technical Specifica-tions that require remedial action or corrective measures to prevent the existence or development of an unsafe condition.

Note:

This item is intended to provide for reporting of potentially generic problems.

b.

Thirty Day Written Reports.

The reportable occurrences discussed below shall be the subject of written reports to the Administrator of the cognizant NRC Regional Office within thirty days of occur-rence of the event.

The written report shall include narrative material to provide a complete explanation of the cause of the event, circumstances surrounding the event, any corrective action; and component failure data.

1.

Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those established by the Technical Specifications but which do not prevent the fulfillment of the functional requirements of affected systems.

2.

Conditions leading to operation in,a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation.

3.

Note:

Routine surveillance testing, instrument calibration, or preventive maintenance which require system configura-tions as described in items 1.2.b.l and 1.2.b.2 need not be reported except where test results themselves reveal a degraded mode as described above.

Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause red4ction of Appendix A degree of redundancy provided in reactor protection systems or engineered safety feature systems.

4.

Abnormal degradation of systems other than those specified in item 1.2.a.3 designed to contain radioactive material resulting from the fission process.

Note:

Sealed sources or calibration sources are not included under this item.

Leakage of valve packing or gaskets within the limits for identified leakage set forth in Technical Specifications need not be reported under this item.

1.3 Unique Reporting Requirements (Plant Specific)

Appendix A

Variations Noted in Licensee's Bases for Reporting the Reactor Trip Breaker Failures (1)

Reported to NRC by )Information Letters The licensee for Robinson Unit 2 reported two events, occurring on 9/23/81 and on 12/20/82, to the NRC as information letters rather than LERs.

(2)

Reported on Basis of Paragraph 1 The letters forwarding the Davis Besse and Kewaunee LERs referenced only the general heading of 11 Reporting Requirements" (Appendix A, Paragraph 1) as the basis for reporting the events without identifying which specific subparagraphs of Paragraph 1 where applicable.

(3)

Reported on Basis of Paragraph 1.2 The forwarding letters for two of the Zion Unit 1 LERs referenced only the genera 1 introductory paragraph of 11 Reportab1 e Occurrences 11 (Appendix A, Paragraph 1.2) as the basis of reporting without identifying which specific subparagraphs of Paragraph 1.2 were applicable.

(4)

Reported on Basis of Paragraph 1.2.a The forwarding letters for two Arkansas Unit 1 LERs and one Robinson Unit 2 LER referenced only the general introductory paragraph of 11 Prompt Notifi ca-tion with Written Followup" (Appendix A, Paragraph 1.2.a) as the basis of reporting without identifying which specific subparagraph of 1.2.a were applicable.

For the Arkansas events, one occurred during testing and was reported as a 14 day report, while the other occurred on demand during reactor trip and was reported as a 30-day report.

(5)

Reported on Basis of Paragraph 1.2.6 The forwarding letter for one Arkansas Unit 1 LER only referenced the genera 1 introductory paragraph of "Thirty Day v.Jri tten Reports 11 (Appendix A, Paragraph 1.2.b) as the basis of reporting without identifying which specific subparagraph of Paragraph 1.2.b was applicable.

(6)

Reported on Basis of Paragraph 1.2.b.l The forwarding letters for three LERs cited Appendix A, Paragraph 1.2.b.l as the basis for reporting.

Appendix B (7)

Reported on Basis of Paragraph 1.2.b.2 The forwarding letters for 17 LERs cited Appendix A, Paragraph 1.2.b.2 as the basis for reporting.

Of these 17 30-day LERs, 13 involved RTB failure during surveillance testing while the other 4 (the 1/6/83 event at Salem Unit 2 and all three Zion 2 LERs) involved RTB failure on demand during reactor trip.

In regard to the 3/27/77 event at Zion 2, the licensee's forwarding letter states that, "this event was previously classified as non-reportable but, as the result of an internal audit was reclassified as a 30-day report on 6/7/79." Accordingly, the licensee submitted the 1977

.report on 6/7/79.

Appendix B