ML19325C567

Advises That NRC Determined That Series of Audits Necessary to Verify Proper Implementation of B&W Owners Group Safety & Performance Improvement Program Recommendations at Each B&W Designed Facility.Audit Planned for Wk of 900402 at TMI-1

ML19325C567
Person / Time
Site:	Crane
Issue date:	10/05/1989
From:	Hernan R Office of Nuclear Reactor Regulation
To:	Hukill H GENERAL PUBLIC UTILITIES CORP.
References
TAC-68206, NUDOCS 8910170072
Download: ML19325C567 (25)

Text

{{#Wiki_filter:(t October S, 1989 -Docket'No. 50-289

Mr. _ Henry D. Hukill, Vice President and Director - TMI-1 GPU Nuclear Corporation P; 0. Box-480 Middletown, Pennsylvania 17037

SUBJECT:

SAFETY AND' PERFORMANCE IMPROVEMENT PROGRAM (SPIP) IMPLEMENTATION AUDIT AT THREE MILE ISLAND UNIT 1 (TMI-1) (TAC N0. 68206)

REFERENCE:

Letter from Dennis M. Crutchfield, NRC, to Walter S. Wilgus, BWOG, dated May 4, 1988, " Status of the evaluations of previous NRC requirements, reconnendations and concerns applicable to B&W designed plants." As you are aware, the NRC staff has determined that a series of audits is necessary _to verify the proper implementation of the. Babcock & Wilcox Owners Group (BWOG) Safety and Performance Improvement Program (SPIP) recommendations y at each B&W designed facility. This audit series consists of a SPIP programmatic audit (which has been completed already), a SPIP recommendation implementation l; . audit, and a followup audit, if necessary. The implementation audit for THI-1 K is planned for the week of April 2, 1990. In conjunction with this audit the audit team will also evaluate your disposition of previous NRC issues applicable to B&W designed plants (See the Enclosure to this letter). .It is anticipated that the audit will commence with an entrance meeting at 9:00 am on-April 2,'1990 and conclude with an exit meeting on April 5, 1990. We plan to conduct the audit at the TMI site in Middletown, PA. I will firm up specifics with your staff as the week of the audit approaches, Sincerely, /s/ Ronald W. Hernan, Senior Project Manager Project Directorate I-4 Division of Reactor Projacts - I,11 Office of Nuclear Reactor Regulation

Enclosure:

.As stated. cc w/enciosure: ~ See next page DISTRIBUTION e M M C'PTTE7 ' NRC & Local PDRs Plant File S. Varga (14E4) B. Boger (14A2)

5. Norris.

R. Hernan 0GC E. Jordan (MNBB3302) .B. Grimes (9A2)' C. Harbuck (13D18) G.Hsii(8E23) ACRS(10) .LA:PDI-4 PM:PDI-4 k PF 4 s SH O W RHernin: m JSt .10////89 10/ /89 10/ b Of 0FFICIAL RECORD COPY. Document Name: TAC 68206 8910170072 891005 l PDR ADOCK 05000289 .J 9%0

't- 'T 15 J w4 7[ NUCLEAR REGULATORY COMMISSION ~ l ~ O :. 3 = WASHINGTON, D. C, 20566 ~ October 5.-1989 .. gg. Docket No. 50-289-1 I' 'a i .ME. Henry D. Hukill.Vice President W and Director - TMI 1 .'GPU Nuclear Corporation "j 'P.'.0.' Box 480 ~ Middletown, Pennsylvania-17057 j p

Dear Mr. Nukill:

j L Sl!BJECT:, SAFETY AN0 FERFORMANCE IMPROVEMENT PROGRAM (SPIP) IMPLEMENTATION 1 H ~ AUDIT AT THREE MILE ISLAND UNIT 1 (TMI-1) (TAC N0. 68206) 1

REFERENCE:

Letter from Dennis M. Crutchfield, NRC, to Walter S. Wilgus, BWOG, dated May 4, 1988,~" Status of the evaluations of previous NRC I requirements,'reconnendations and concerns applicable to B&W L designed plants."- / i L As you are aware, the NRC staff has determined that a series of taudits is j necessary to verify the proper implementation of the Babcock & Wilcox Owners - 1 c Group:(BWOG)' Safety and Performance Improvement Program (SPIP) reconnendations - l l-u l1 at each B&W designed facility. This audit series consists of a SPIP. programmatic - 1 audit..(which has been completed already), a SPIP reconnendation implementation audit, and a followup. audit,'if necessary..The implementation audit for THI-1 is planned;for' the week of April 2,1990. In conjunction with this audit the audit team will also evaluate your disposition of previous NRC issues applicable i r ) B&W designed. plants (See the Enclosure to this letter). .It:is. anticipated-that the audit will connence with an entrance reeting at 9:00 am E on April 2, 1990 ard conclude with an exit meeting on April 5, WSO. We plan Lw to; conduct the audit at the THI. site in Middletown, PA.. I will firm up j ~ specifics with your staff as the week of the audit approaches.- L Sincerely, k @ta % tde / p. Ronald W. Hernan, Senior Project Manager Project Directorate I-4' L Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Enclosure:

As stated 1

ce w/ enclosure:

l- .See next page i l-

n Ihi n. 4 ?! 1,.. =.. Mr.- Henry D. Hukill Three Mile Island Nuclear Station, GPU Nuclear Corporation -Unit No.-1 r Cc: 'G. Broughton-Francis I. Your.g NkuSMCohNtion bbO[RgidentInspector(TMI-1) Post Office Box 480 : Post Office Box 311-Middletown, Pennsylvania 17057 Middletown, Pennsylvania 17057 Richard J. McGoey. 4

Manager, PWR Licensing-Regn. Administrator, Region I GPU Nuclear Cceporation U.S. Nuclear Regulatory Coraission 100 Interpace Parkway 475 Allendale Road

{ Parsippany, New Jersey 07054 King of Prussia, Pennsylvs nia 19406 C. W. Sgth . Robert B. Borsum. .TMI-1 Licensing Manager Babcock & Wilcox GPU Nuclear Corporation Nuclear Power Generation Division Post Office Box 480 Suite 525 Middletown, Pennsylvania 17057 1700 Rockville Pike Rockville, Marylard 20852 Ernest' L. Blake, Jr., Esq. Governor's 0.ffice of State Planning Shaw, Pittman, Potts & Troworidga .and Development 2300 N Street, N.W. ATTN: Coordinator, Pennsyjvania Washington, D.C. 20037 State Clearinghouse Post Office Box 1323 r p Harrisburg, Pennsylvania 17120 ' Sally S. Klein, Chairman Thomas M. Gerusky, Director Dauphin County Connissioner Bureau of Radiation Protection i; Dauphin County Courthouse- . Pennsylvania Department of L Front and Market Street-Environmental Resources Harrisburg, Pennsylvania 17120 Post Office Box 2063 Harrisburg, Pennsylvania - 17120 j-Kenneth E. Witmer, Chairman L~. . Board of Supervisors of Londonderry Township 25 Roslyn Road .-Eilzabethtown, PA 17022 Mr., Henry D. Hukill, Vice President and Director - TMI-1 GPU Nuclear Corporation P. O. Box 480~ Middletown, Pennsylvania 17057

ENCLOSURE g 1 } $TATUS OF-EVALUATIONS OF PREVIOUS NRC REQUIREMENTS REcomENDATIONS,

ANDCONCERNSAPPLICABLETOSABC0CK&WILC0XDISIGNEDPLANTS The' staff has perfossed computer searches using the labcock & Wilcox (t&W) docket

numbers and key words from specific subjects-addressed in the Sabcock & Wilcox Owners Group (BWDG) " Safety and Performance Improvement Program" (SPIP)' report, N

.BAW-1919, to identify the documents the NRC staff believes should have been reviewed during the SPIP review. :In addition, the staff has identified safety-- 'related issues that should have been & valuated on a routine basis by the S&W utilities. Large, stand alone program efforts such as the TMI action plan-items,- anticipated tranaients without scram (Ahts), fire protection, and equip-ment quslification, are not incleded because the staff believes they have been adequately _ covered. In addition, previous concerns related to the intepfated control system /non-nuclear instrumentation These were addressed in Appendix E af the $5(ER, issued Merch 1988.ICS/NNI) syst uded. The' documents identified in the computer search that contain reconnendations and identify concern 6 are primarily NURIG reports, NRC orders, ar.d Inspection and Enforcement bulletins,' circulars, and information notices. This enclosure ec,n-tains the issues derived frcm these documents. In addition, Nuclear Safety li Analysis Center - 3 (March 1980) recommendations are included where appropriate. This information has been civided into the eight categories listed below.

1) reactor coolant syst*m and emergency core cooling systems l

L

2) safety-related electt? cal systems and/or components 3) instrument auxiliar y

4) decay heat removal y systems system l

(5) valves (including safety / relief) L (6) main feedwater system l (7) auxiliary / emergency feedwater system L (8) administrative controls l Ior'simplificationandtoavoidduplication,eachissuethathadbeenidentified' was placed in what appeared to be the most applicable category (i.e., one entry). .The results of this search are presented in the table below, which provider a . number column that categorizes the issue for easy reference, a source column that identifies the referer.:e document, and an issue column that describes the Mquirements, recommendations, or concerns contained in the source document. o I The staff requests that' the BWDG fr conjunction with the individual utilities l where required, provide for each entry a response that describes the action 'taken to address each issue and identifies the docuneit or method by which the issue was addressed. Each entry should be referenced by category, number, and, if' applicable, item. 9 1 l-

. ~ cfj-y ,.f l Y? .c r '^ g { jD iCATEGORY le : REACTOR COOLANT SYSTEM AND EMERGENCY.C0RE COOLING $YST t l [ No. ~ Source; -!ssue-(1-1 NUREG-0667~ This rep"rt' recommends the following: N. Recommendation 4 i W -(Rec.) 2.2-(9) . Following reactor trip, pressurizer level _should re-i

• a Mayl1980 sain on scale and pressurizer pressure should remain-above the high-pressure injection (NPI) actuation set M_

point.- ~ These objectives should bw met independent of all + manual operator actions (e.6., control of feedwater, 1 letdown isolation, startup.of a makeup pump). Rec.L2.2 (11)- Plant modtfications should be ~ande to. reduce or elini, J-nate manual immediate actions from ear.rgency proedures. 1 NRC Office of-' This notice concerns ~ failure of the riactor coolant puing Inspection (RCP) shaft at Crystal River. The initial indications.- 'and Enforce-were motor frame vibration and RCP.thrcst bearing high l ment'Informa- . temperature; the operators manually tripped the RCP. e h' tion Notice- . Further symptoms related to such an event are listed '(IEIN) 86-19 below. L Mar. 2J, 1986 o D A shaft fracture was found at a nonfunctional groove below thermal barrier, h L Ultrasonic' testing (UT) of three other RCPs showed cracks at same locations. 1 y All four capscrews joining the shaft to impeller had i L broken as a result of intergranular stress corrosion -l cracking (IGSCC). L UT at Davis-Besse found. cracks in on RCP and prob-able cracks.in three other RCPs. ] E - Similar capscrew failures to those above were found on J the RCPs at the Palisades plant because the pt3 scribed proloading could not be achieved because of rou@- threads. 1-3 IE Bulletin This bulletin addresses the potential failure of multi-J IE8) 84-03 pie pumps of the eeer ncy core cooling systes (E CS) J Oct. 8, 1986 as a result of the s: le failure of the air-operated j Fr valve in the minimum-f ow recirculation Ifne. L.- Itaa 1 Promptly determine whether or not the fac.'ety has a c sintile failure vulnerability N the minimum flow rec' eculation line of any ECC$ pumps that could cause the failure of more than one ICCS train. l - L l 2 l E 1 )

,7 %, .r

No.

Source Issus t Item 2

If the problem exists:

Fromptly instruct all operating shifts of the problem and measures to recogni:e and mitigate the problem. J Promptly develop and implement corrective actions to t. o bring the facility into compliance with General Design Criterion (GDC) 35. i CATEGORY.'2:. SAFETY-RELATED ELECTRICAL SYSTEMS AND/0R COMP 0NENTS a No. Source-Issue 2-1 IE8 80 06 This bulletin addresses engineered safety feature (ESF) g L; Mar. 13, 1980 reset controls and recommends that the licensee: ) Item-1 Review the drawings for systems serving safety-related functions at the schematic level to determine whether L' or not all' associated safety-related equipment remains in its emergency mode if the ESF actuation signal is reset. Item 2 ' Verify that the actual installed instrumentation and controls at the facilicy are consistent wit 5 the draw-ings reviewed by conducting a-test to demonstrate that L all equipment remains in.its emergency mode if the actuating signal-is removed and/or the various isolating er actuation signals are reset manually. l Item 3 If any safety related equipment does not remain in its i L emergency mode when an ESF signal is reset, describe the proposed system modific& tion, design change, or other corrercive action planned to resolve the problem. 2-2 NUREG-0667 Thir, report indicates that: Rec. 2.2 (4) l May 1980 Steam line break detection and mitigation systems should be modified as necessary to eliminate adverse int.oractions with the auxiliary feedwater (AFW) system. Steam line break detection and mitigation systems should be re-evaluated and modified so that they are [ capable ~ of df fferentiating between an actual steam i line break and undercooling/ overcooling caused by L feedwater transients. p 2-3 IE8 80-16 This t != tin covers miscpplication of pressure trans-June 27, 1980 sitter;. < indicates that the licensee: 3 1' - - ~..,.... - -

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my' m o 9 6 No. Tsobece! " Issue i 1 e' [, Item 1D ~ Detemine if the facility has installed or plans to. linetall' Rosemount;Inc. ' Model 1151 or 1152 pressure i -transmitters with outpe', codes "A" or "D" in any " safety-related application.. 4 o'

Jtem'2' LIf it is determined that the facility.has the trans-1

- eitters described in item 1 above in sny safety-related i application, determine whether they can be exposed to 'j -input pressures that could result in anomalous. output signals during normal operation, anticipated transients : -l or design-bas's accident,s. If affected transmitters y can be: exposed to input pressures that could result in anomalous. output si0n 1s, perform a' worst-case analysis to determinne whether the anomalous signals could result in violating any design basis assumption. The safety-related appUcation shall include control, protective, or-indication functions. If any safety-related appli-- cation does not confom to these requirements,' address the basis for continued plant operation until the -oroblem can be resolved and provide-an er,alysis of all po'tential adverse system effects-that could occur as a result of the postulated pressure transmitter salopera-t tion" described in enclosure 1 of the bulletin. In each { ~ - instance,-the analysis should include the effects of L the postulated transmitter saloperation as it relates to indication, control, and protective functions. 4 ^ The' analysis'shall address-both incorrect automatic - system operation and incorrect operator actions caused by the erroneous. indications. Address conformance to Institute of Electrical and Electronics Engineers (IEEE) Standard 279, Section 4.20, in-the analysis. The analy- .l sis should include the-specific infors.ation required by the bulletin.- ' Item 3 Submit a complete description-of all corrective actions - required as a result of the-analyses and evaluations, along_with the schedule for accomplishing the corrective. m i actions. , 2-4 IEIN 86-10 . This notice alerts licensees to the cor.cern of safety L-Feb. 13, 1986 - 1>arameter display system (SPDS) operability. It was found that two out of.five plants had declared the SPDS to be operational-when the $PDS was not operational. 1, The major deficiencies that were found in the SPDS sur-3 voy are listed below. lack of availability because of gross system malfunctions l display of unreliable or invalid data and alares ) l 4 I i +- a v -..w- --4-n- ---a ,+----.,..---.n- ~ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

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r-w., .4 h- ,g . No. - Source 1 Issue U : ' problems ptance by operators becaus, of reliability poor accc + l- +.. management; failure to integrate SPDS into-operational y environment inadequate: documentation and failure to control system m = E testing and modifications p, .~ 6 low system response to some cperator commands.. r. 2-5 IEB 79 25-1This bulletin requires licensees to determine if West-i Nov.'2, 1979, inghouse model BFD/NSFD relays of the type identified k. ' Item 1 in the bulletin are used at their facilities. 'If they are used,-identify the safetyrrelated systems involved, i' the function of the relays, and plans'for test and/or ~ replacement programs. l1 Item 2 -Establish a prograr to ensure performance of affected p relays. This program should include periodic testing l. and/or replacement, the basis for. test intervals, h development of approved procedures for. testing and/or L -replacement. and documentation of relay failures found' lo

during testing.

L ' Item 3 A written report should te submitted to address the actions taken under items 1 and 2 above. [t: )Apr.;17.1979 - type AK-2 breakers are used at their facilities an 2-61 1EB 79-09 This bulletin requires licensees to determine if GE 3 4 p L Item I if used, identify the' safety system involved ano the plans for developing a preventive maintenance program, u, Item.2 The preventive maintenance program should incluse a preventive maintenance schedule, the necessary qualifica-tions for persor.nel to perfore the maintenance, and the status of the= recommended corrective' actions described in GE Service Altrt Letter No.175. Item 3 A written report should be submitted to address the actions taken under items 1 and 2 above. 7 IE Circular This circular urnes Ifeensees to review the specific OEC) 81-12 hans presented on the Description of Circumstances" l July.22, 1981 Section of this IEC as they relate to circuit breakers. Also review the procedure for surveillance testing of n circuit breakers to ensure that the procedure provides for independent tr. 'ng of each trip function. If the procedure does no E <e provisions for independent test-b ing of each trip fk nion, then modifications should be made to include such features, t l 5 s 6- ~w s r,,na-,, ,a,,.. -.mm,,,-.~w

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A: qQ [ G E No. ' Source' Issue y 90 1 L28' IEC 81 14' LThis! circular urges 1.icensees to review operating ys Nov.,5; 1981< experience with main steam isolation valves-(MSIVs) to-u

identify probless related to the failure of the valves 1

1 Lto close'end to equipment degradation that prevents a-i M.n ' valve from closing and would require other than rout.ine y p%, maintenance to correct. r q ~ Evaluate the corrective actions identified in the main-tenance records to ensure that the action! were adequate to' solve the root-cause problems; if not adequate, 4 develop plans for-further corrective action. I 1 k If: air quality control is suspected of contributing to. problems With the MSIVs, review the air system to ensure-1 that measures have been or will be taken to prevent future air evality degradation and consider the installa-tion of monitors and/or alares to provide warning of air li quality deterioration. j !f system binding is suspected of contributing to prob-lems with the MSIVs, review maintenance procedures to 1 ?' ensure that they include precautions to be taken against - dettirental effects (such as those caused by inappropri-ate lubricants) and that the procedures include tests j that demonstrate the valves will perform under operating [ - conditions before being placed in service, f 2-9 IEIN'82-35 This notice identifies a concern that stop check valves L - Aug. 25, 1982 manufactured by Velan would fail to pass flow because of } maintenance and design problems. L 2-10 IEIN 82-48 Thic notice identifies potentially significant defici- 'l . Dec.:3, 1082 encies with Agastat.CR 0095 relny sockets. In 1979 General Electric tested 2200 of these relcy sockets and found a significant number exhibited contact retention d problems'and potential electrical connection problems. These relay sockets were redesigned and modified by the manufacturer and all relay sockets shipped after March 1979 were of the new type. Another means of correcting the problem in the old relay sockets is to install cardboard insulator strips behind the relay sockets. L . 2-11' IEIN 82-50 This notice idertifies a potentially sinnificant prob-Dec. 20. 1982 les pertaining to aisapp11 cation of sol'd state ac undervoltage relays type ITE-27, series 2118 ar.d 211L. manufactured by Brown Bovery Electric Inc. These R relays were found to be used on Class 1E switch 0 ear L that requires a source of dc control power for proper e operation and these relays were used in some plants to h monitor ac %s undervoltage conditions.

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W L+J K g jo. ' Source ~ Issue 1 N 'ITE-27, Series 211R,- relays should be used since they do not drop out on loss of de power and do not resu1* +1 in' an ir. advert 6nt isolation of Class 1E.switchgear. .l 2-12 21EIN 82 This notice identifies potentially significant problems Dec.- 27, 1982-with a certain batch of Westinghouse NSFD relays that s appear to have a higher-than-expected failure rate. L The notice contcins a copy of a Westinghouse technical !Y -letter that discusses the problem and provides inspec-tion and test methods for verifying operability of the relays as well as suggested corrective action. 2-13 IEB 83-04 This bulletin requires PWR licensees with other than Mar.=11, 1983 WDB-type breakers in the reactor protection system. ~ Item 1 TRPS)to:- Perform surveillance tests-of the undervoltage trip function that are independent of tests of the shunt L trip function. Review their maintenance program to ensure conformance with manufacturer's recommendation. Ensure that the appropriate emergency' operating pro-cedures for the event of failure-to-trip p.nd other operating events are reviewed with each operator. e Provide a written report containin results of the j. Labove actions, a description of al RPS breaker mal-p functions that have not been reported previously. and verification that procurement, testing, and saintenance activities related to the RPS breaker and c 4 undervoltage devices are treated as safety related. 4 2,-1.4 IEIN 83-76 This notice suggests that utilities using General Elec-L Nov. 2, 1983 tric type AK-2-25 breakers with undervoltage trip l devices, visually inspect each underv11tage armature to - ensure it is in its proper position after each opera-tion (i.e., the fully down position and not the aid position). ~2 IEB 64-02 This bulletin requires licensees to: Develop plans and Mar. 12. 1984 schedules for replacing nylon or Lexan coil spool-type 2 tem 1 NFA relays that are used in energized safety-related applications and nylon coil spool-type HFA relays that are used in normally de energized applicationst Item 2 Ir. the interin, before the relays are replaced, develop 1 and implement surveillance plans that include sonthly e } functional tests and visual insper.tions. 7 L, -. - ~. ...m.....

p4 I [; No; 'Sourc~e ' Issue ~ Item'3 Provide the basis for continuint operation until.the j nomally energized relays are replaced. - Item.4L Provide a written report describiig the above actions' .q and including complation schedules.- 2-161 IEIN M-20 .This notice identifies a probles pertaining to the - Mar. 21,: 19M service life of relays in safety related systems. It spectfically identifies the earlier-then-anticipated, = j end-of-service life failures of Agsstat G-P series s relays and Sylvania GTE ac relays.- The notice suggests-j; that utilities review their safety-related systees-.to determine if these relays have been installed er are being held as spare parts. Preventive maintenance pro-grams should recognize ~the. application-dependent t. L. (energized /de-energized) service life of relays and the i current surveillance interval should be compared with k'

the service life of.the relays as used in the system to D

detemine if it is acceptable. The notice indicated that. these problems are similar, to those discussed in 1EB M -02 and the general concorrs associated with HFA p relay failures discussed in that bulletin apply, 2-17. -1EIN M 37 This notice provides the following guidance to eliminate May 10, 1984-problems encountered in the use of lifted leads and jumpers during-maintenance and surveillance-testing: p Install :..ermanent test hardware. Include additional procedural checks of system config-uration during surveillance and maintenance testing. lp Review procedures to* ensure instructions'for surveil-lance and maintenance clearly specify the reconnec-tion of any lifted leads and the removal of any jumpers. E Use at-least two qualified operators to independently v verify proper system configuration before safety-

4 relatea equipment' is returned to service.

Perfsm functional tasts to verify proper system con-figuration is restored before safety-related equipment 1 is returned to service. Review w'th operators and maintenance personnel spe-cific instances of errors involving lifted leads or $mpers and the safety impact ref such errors. 2-18..3EIN 85-58 This notice identifies a potentially significant prob-t ~' July 17,1985 tem pertaining to the failure of General Electric (GE) Supplement 1 type AK-2-25 reactor trip breakers (RTBs) that are r Nov. 19, 1985 installed in facilities designed by B&W and Combustion i 8 l e y. m

E., DM! ~ j -I jga = No -Source issue-6.V -Engineerin!iled to trip open when its undervoltage(CE 'y ;e the RTBs.f i 0 attachment was actuated. The RT8s at. Rancho Seco that-1 were sent to GE for refurbishment, were only visually-inspected when returned. The utility had developed procedures to perfom checks of. critical parameters of-breakers' based on S&W guidance. Supplement 1 to this-t notice identifies additional GE type AK-2-25 breaker failures resulting in slow closure times that occurred at Calvert Cliffs and Oconee. - These failures were re-lated to laminated sections of the armature that slipped p y down causing ~ contact between the faminationr,and pole %e face at Calvert Cliffs.. At Oconee a new undervoltage device in 4.he RTB reduced clearance,s between the arma-w ture~ and heads of-mounting bncket studs:and could have caused contact. This notice calls attention to GE's fervice Advice Letter No. 300, which outlines corrective actions. k 19 IEIN 85-93. This notice alerts utilities that the electric circuit l t l' Dec. 6, 1985-breaker closing function of Westinghouse (W) type DS circuit breakers would not operate if the spring re-lease latch-levers were broken. W issued Technical Bulletin No. NS10-TB 85-17 advising utilities of this potential an1 function. The W bulletin identifies the following corrective actions? (1) advise personnel 1 W - breaker may-be closed manually; (2) user should evalu-t ' ate ~ function to determine if it affects safety; and (3) inspect latch lever during' normal scheduled maintenance / inspection of type DS circuit breakers. 2-20' IEIN 87-08 This notice alerts utilities-to potentially defective L - Feb..4, 1987 de motors that were manufactured by H. F. Porter (now Peerless-Winsmith) between December 1984 and December L l' 1985 and installed in Limitorque actor operators. 2'-21u IEIN 87-12 Tr.is notice alerts utilities to potential problems with Feb. 13, 1987 GE type AKF-2-25 circuit breakers failing to fully open. These circuit breakers 1.re susceptible to failures as a result of; binding within the breaker-can mechanism unless M oper maintenance procedures-are developed and followed by trained individuals. The notice contains ' the followitM1 maink aan:e information that was provided by CE to be e ncorporated in utility programs for type AFK-2-25 breakers: (1) saintenance/ inspection interval:, and complete overhaul should be-every 22 months or each refueling outage; (2) only specified lubricants should be used; (3) only qualified properly trained personnel should perform maintenance; and (4) breakers that have not yet been convertec; to a specified lubricant should be-cycled. + 9 , b yrx--. 4 e,

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><, m g ), M 3 '2*22 !!!N 87 24' This notice-alerts utilities to potential problems.in f.. June 4, 1987' -volving inverter losses. The notice refers to an NRC a case study report AE00/C605 -en operational' experience R . involving losses o,f electrical inverters in whi h three failure mechanisms were identified. The notice suggests-Lthat utilities consider (1) monitoring temperature and/or humidity internal to inverter enclosures and. input:end output voltages of the. inverter unit during steady-state and transient conditions and (2) reviewing maintenance and testing procedures and practices. lL f' n p, . CATEGORY 3: INSTRUMENT AIR SYSTEMS No. -Source Issue 3-11 IEIN 81-38 This notice informs. licensees about an NRC staffL Dec. 17, 1981 review of a number of. problems and instant.es related to contamination of air systems in operating plants. -The review indicated that air-operated components and p systems will occasionally become inoperable because of y contamination with oil, water, desiccant, rust, or 1 other corrosion product.- The notice described the L following actions,-which are known to minimize air E system problems: I frequently monitor the dew point of the instrument l E air periodically check the desiccant cartridges to deter-mine if they need. regenerating or replacing L p periodically blow down lines to remove cil, moisture, El and crud in the instrument.ir system 1 - periodically inspect filters downstream of.the desiccant cartridges to ascertain that the desiccant has not-been pulverized to' the point that it is g escaping from the cartridge and possibly ciogging the i filters li avoid using service air as a backup to the instrument l air system when alternative backups are available frequently monitor the instrument air system to ensure that it has not been contaminated with oil, moisture, or crud when service s'.r has been used as a backup to the instrument air system l l l: ) 10 -. -. ~..

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3 -y4 L CATEGORY 4:? DECAY l HEAT REMOVAL (DHR)- [' ! No. Source Issue H 4-1 IEIN 80-20 This notice describes an eve'nt'at Davis Besse that May 9,-1980' occurred while in a refueling mode and that resulted in ,a loss of DHR capability for approximately fle hours. The following three factors contributed to this loss. (1) inadequate procedures and/or administrative con-j trols,--(2) extensive maintenance activities; and (3) the~ two out of. four safety features actuation system (SFAS) 11ogic. It was suggested that licensees evaluate the~ susceptibility of their plants to lose OHR' capability ' f by these causes, 4 2-IEIN 86-39 This notife alerts licensees to a potential common-mode May 20,-1956 failure of multiple residual heat removali(RNR) pump L actors and pump internals. An event occurred in which l-the high temperature of a lower guide bearing went un-l^ noticed because of several other, alarms -3 days later J the motor caught fire.. It was found that lower pump impeller wear rings on multiple RHR pumps' had separated - from the impeller as a result of.IGSCC.. Motor guide bearing failures are significant because of-the potential L failure of pump internals. Other potential causes of-J internal damage include inadequate flow and' lubrication. = 4-3 IEC 81-10 This-circular advises licensees to: July 2, 1981

Review their. operating procedures for plant'cooldown,-

emergency, and abnormal operations as they relate to natural circulation to ensure sufficient information is available for operators to recognize symptoms of-reactor coolant system (RCS) voiding and take appro-priate actions to recover from a voided condition. Inform each licensed operator of the infonmation dis-y cussed in the circular. a Cor. sider including this information in operator train- + ing and retraining classes, 4-4 GL 87-12 The subject of this letter 1s-the loss of RHR capability p h; July 9, 1987 while the BCS is partially filled. The GL transmits l huREG-1269, e report on the Diablo Canyon loss ef RHR L system incident of April 10, 1987. It requests licensees to proviet a description of plant operations with a par-tially filled RCS. The description should address such topics as the initial conditions, instrumentation and alarms, available pumps, containment closure capability, procedures, training, additional resources, requirements - for other modes, any changes to be made and the schedule for making them. l-l 11 ^#--# 6 '-me -w-+----- =c ---e-- -=e .e-r --me=. -e-.- = =w e --.--------------------------------------- ------------e-

7 __ 4 ke CATEGORY 5: VALVES (INCWDING SAFETY / RELIEF VALVES) n 'No; Source-Issue i , 5 IEC 79 22 This circular advises licensees to conduct a review to ~ Nov.-_16, 1979-determine if periodic surveillance _of power-operated _. relief valves (PORVs) is necessary to ensure that the ? ~p0RVs will perform as intended, c ~5 IEIN 80-41l This: notice describes the failure of a Velen check y Nov. 6,l1980. -valve in the decay heat removal system at Davis-tesse. 6 The valve disk and are separated from the valve body and were lodged under the valve cover plate. The bolts and locking mechanism that holds the am to the valve body were alssing. J 1 L 5-3 IE8 81-02 This bulletin requires licensees to ascertain whether j L, Apr. 8, 1981 any PEMD motor-operated gate valves have been installed Supplement. or are s.aintained as spares for installation in safety-u. Aug.'18, 1981 related systems where they are required to close against Item 1 differential pressure, g If the affected valves have been installed, licensees should take corrective action and evaluate the effect on system operability.1f the valves fail to close. If -affected valves are spares, licensees should replace or modify the valves before installation. f Licensees should affected valves, provide a written report listing the service, and maximum differential pressure required to close and describing the safety consequences if the valves fail to close and the cor-rective actions taken or planned along with a schedule for completing these actions. E-4 IEIN 81-35 This notice addresses Metropolitan Edison's report of Dec. 2, 1981-loose valve internals in the hi h pressure injection pump discharge crane 3-inch,15g0 pound tilt check valves that resulted from corrosion of the' seat holddown devices.. As a result of these findings, a continuing-inspection program for TMI-1 was developed and imple-8 mented which led to the discovery that some tilt check valves could not prevent back flow because the hinge pin and ring s u t retention devices failed. Many valve fabr6 cation inconsistencies also were discovered that any have initiated or contributed to the failures. 0 '5-5 IEIN 82-20 This notice alerts iicensees to a potentially signifi-June 28, 1982 cent problem pertaining to internal rJamage to swing check valves of the same or sisilar design and service L as those manufactured by Alloy Steel Products Company and Pacific Company. 12 h c

--} n W.. y - q [F g ~ Source Issue y 5-6 !EIN-83 57L This notice alerts licensees to a potential problem: Aug.- 31,'1987' with ASCO three-way solenoid-operated pilot valves,. ' type NP-8316 in 3/8+ and 1/2-inch national pipe thread ' sizes. The manufacturer's installation instruction bulletin, issued in 1978, provides incorrect assembly . instructions for certain parts of this valve. o IE!N 84-33 -This notice alerts licensees to a potential problem-5-7' 1 E Apr. 20,;1984 with main steam safety valves that have failed because L of cotter pin failures. i 1 5-8 IEIN 84-66 This notice identifies events where turbine-driven AFW O-Aug.-17, 1984 pumps were unavailable because the. steam supply was ? E, -isolated (trip and throttle valve was not latched). The-NRC recompends that licensees review these events and [: consider the following preventive actions: i i. s design change to provide positive control room indi-l cation of a trip velve "Istched" condition i E regular adjustment and testing of the limit' switches to ensure operability . local-verification of position after resetting trip valve O visual verification daily or once per shift to see [ that the valve is not tripped t local mechanical valve position indication installed L and permanent tags attached to the valve providing instructions for operation h on-shift training in operation of the trip valve for all personnel who are required to operate the valve improved housekeeping to prevent fouling external valve linkages warning sign installed near the trip lever 5-9 -!EIN 84-48 This notice alerts licensee: to a potential deficiency June 18, 1984 in the design, app'lication, or maintenance of Rockwell Supplement 1 International globe valves that have resulted in two 4 Nov. 16, 1984 types of failures: (1) the stem separating from the disk and (2) the disk being backed off its disk nut. The manufacturer attributes these failures to the high cavitation loads the valve disk experiences when used in severe throttling conditions and recommends that these valves be replaced with smaller size valves so that the valve disk will be in a more fully open position. 13 1 a. 4 "4 Sr-m>M-e-ww-=- re+-ww+ew--w -swwaw-u--twersesu---m--vese---u--e-Te wmr-w-we

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N# 'No. So' urce - !ssue' ., e - R*, 5-105 1EIN 85 This notice identifies a potentially significant problem Apr.530, 1985 related to Parker-Hannifur Corporation check valves supplied _by Anchor / Darling Valve Company. These valves may degrade the capability for closing main steam iso- ^' 1ation valves (MSIVs) or feedwater isolation valves or. inhibit other safety functions. An event at Syron Unit I resulted in the failure of two MSIVs-to close. on'an isolation signal because the instrument air check valves. failed to seat in response to gradually decreasing air pressure. 11 IEIN 85-59 This notice identifies a potentia 11y'significant prob-July 17, 1985-les related to stress corrosion failurei, of valve stems i and shafts that are not routinely examined. Four -l instances are described where cracks were found in 410 stainless steel valve stems. Each instance involved a different plant and valve manufacturer, and the cracks 4 were discovered after failure or disassembly. ' l L LL IEIN 85 84 This notice discusses the possible failure of MSIVs to K Oct. 30, 1985 close under low-or no-steam flow conditions and the ^ testing of these valves with non-safety-related motive-g j, power (instrument air) in place.. l determine the need for a test program to establish . reliability L 5 13 IEB 85 -The common-mode failures of motor-operated valves with Nov. 15, 1985 improper switch settings during plant transients led the NRC to request licensees to: Deielop and implement a program'to ensure that valve operator switches are selected,' set, and maintained properly'for MOVs in high pressure coolant injection / core spray and emergency feedwater systems beactor core isolation cooling (RCIC) system for BWRs) that are required to be tested for operational readiness in accordance with 10 CFR 50.55a(g). This should include the following components: (1) Review and document the design basis for the opera-tion of each valve. This documentation should include the anximum differential pressure expected during opening and closing'the valve for normal and 1 abnormal events to the extent that these valve oper-ations and events are included in the existing, approved design basis (e.g. mented in pertinent licensee, the design-basis docu-submittals such as FSAR analyses and fully approved operating and emer-gency procedures). When determining the maximum 14 \\. L 1 i

_ _ _. ~.. _ _ % mf i y +. k 1. -( differential pressure, those' single equipment fail-- J ("M ,ures and inadvertent' equipment operations-(such as J. inadvertent valve closures or openings) that are. . ithin the plant design basis should be assumed. w '(2) Using the.results of ites (1) above, establish the correct switch settings. 'This shall-include a pro-g' gram to review and revise. 'es-necessary.. the methods for selecting and setting all switches-(i.e., torque bypass, position limit, overload) for.each: valve . operation (opening and closing).- y p' If.the 1icensee determines that 's valve is inoper-J abit, the licensee.shall also make an appropriate: K justificationforcontinuedoperationinaccordance p with the applicable technical specification.c i l:

(3)

Individual-valve setting-shall be chanted,-as-l' appropriate, to~those established in item (2), p above. Whether the valve setting is changed or not, E the valve will be demonstrated to be operable by testing the valve at the maximum differential pres-1 V., sure determined in item (1) above with the excep-- tion that testing MOVs unfar conditions simulating. a break in the line containing the valve.is not-required. Otherwise. justification should be pro-vided for any cases w,here testing with the maximum !.[w differential pressure cannot practicably be per-i J formed. This jur,tification should include the. l= alternate to max'imum differential pressure testing. K which will be used to verify the correct settings, n i Each-valve shall be stroke tested, to the extent l practical; to verify that the settings defined in L o item (2) above have been properly ~ implemented.even L. if testing with differential pressure cannot be ~ performed. (4) Prepare or revise procedures to ensure that correct switch settings are determined and maintained throughout the life of the plant. Ensure that r applicable industry recommendations are considered in the preparations of these procedures. 5-14 IEIN 86-05 This notice discusses incorrect factory-set rina settinas Jan. 31, 1986 (not allowing full disk travel and hence, relfe" capacity) T for main steam safety valves (M55Vs) and for PWR primary - system safety valves that say not be known because full-flow tests are not performed or required. I I 15 7 -.-,.--4.~,,,. __~,,..._._-....m._..~

l {. } 4 m i ' No souce,- ,y,,,,- ,.p: .NUREG-1195-This report ~ discusses.the need;for a maintenance program. H H 5-15; : - t z. i k Febh 1986 for manual-isolation valves to ensure ~ continued > y" j .Section:10.1 ' operability. : R , Item 3 m

5-16 IEIN P4-29 This, noticediseusses the importance of fully understand-i 1

m Apr. 25' 1986-standing the effects of changes to MOV switch settings. L L For instance. the readjustment of a switch that was on., same shaft as the torque bypass switch to the closed: i l, position in response to Its 85 03 caused an excessive. s L' 'cooldown rate because the 5/D HX isolation valves were D shown to be closed when-they were really up to 16 percent l open. Evgn though maintenance and operations personnel were aware of th"s situation,' such settings could ad-versely affect other plant equipment. 4'

5-17 IEIN 86-56 This notice. lists more causes of MSSV malfunctions i

July 10, 1986-- :as detailed below 4 .1 ' majority of problems related to:not actuating / resenting ..at. set point second largest group of. problems related to failure to L open and failure to close properly - l 4-u several, cases showed the actual lift point to be sub- + J stantially higher than desired set point in one case, the lift ' point of 11 of 16 valves was ' excessive, indicating a multiple-common-mode problem. o several cases where the resent value was-substantially-H below set point cases indicated that there were more problems with leaking valves than with properly functioning valves i N '5 - IEIN 86-92 This notice discusses pressurizer code safety valves: Nov. 4, 1986 one found 350 psi above set point because of a hole I.- in the bellows causing.boren contamination and L corrosion v one found leaking too much to test as a result of' bad = steam cutting i one found lifted prematurely, causing systen depres-l surization to 1800 psig before ressating, as a result L of cocked spring and improper adjustment of ring i settings by maintenance personnel 16 1' E ..a - - = -

W

Noc

' Source- !ssue 3 ~ ~. one-found actuating spuriously as'a result of the + f'_ wrong test equipment being used to set the lift-e pressure.- w L - 20 other events involving 32 valves-that had drift:in 4 the set point large group with seat leakage 5-19 1EIN 86-93' This notice discusses the importance of 'ly understand ; L Nov. 3, 1986-ing effects of changes to MOV switch settings. generic correlation available several years ago.The that

torque varies linearly from 40 to 100 percent with torque switch settings of.1 to 5 is not valid-for many actuators.

In some cases, at a setting of 1 torque varies from 11 to'55 percent.:Asaresult,anIndividua1' calibration e' curve or bench test is required.- Arbitrarily raising the-torque switch setting to maximum can cause damage since thermal overloads are often removed. Analysis r showed that two valves.in the normal charging line that have to close during ECCS actuation would not be able to close against the differential pressure if the generic . correlation was used to establish switch set points. -In I addition, improperly set thermal overloads can render the HPCI systems incperable. CATEGORY: 6: MAIN FEE 0 WATER SYSTEM (MFW) - No. Source Issue 6-11 NUREG-0667-' This report suggests that B&W licensees should perform

Ma~y 1980 sensitivity studies of possible modifications that Rec ~2.2 (10).

could reduce the response of the once-through steam gen-L erator (OTSG) to secondary coolant flow perturbations. Both active'and passive measures should be investigated to mitigate overcooling and undercooling events. 6-2 GL 81-28 This letter transmits the NRC/AE00 report related to July 31,1981 steam generator overfill. Licensees are requested to: .(formerly o .GL 81-16) determine which scenarios'are credible at their plant i determine possible consequences of steam generator overfill / include this inforsation in an overall operator train-ing program Recommendations of the AE00 report: 17 Ll' .='.

(y; s i 'No! ! source

Issue a

1 1.,, (1)c 0verfill should be considered an unresolved safety ~ issue (USI) because of the= lack of safety grade. i ' equipment:to prevent or sitigate everfill and the j potential; severity,of consequences. -(2) ~ Overfill should be treated as either a separate USI

j

'or made a part of.an existing US! because considera-1 m tion of combined blowdown of primary and secondary. systems resulting from a steam generator tube rup- -ture (SGTR) is not included in the present U51,. j which also ass mes low probability and allows credit-1

for operator actions, q

i' (3) An sedit should be conducted to determine if reactor. operators are aware of the potential-seriousness of j overfill situaticas.- If the audit shows subjects a that are not covered in training programs, interim actions should be "..tiated. ..l L CATEGORY 7: AUXILIARY / EMERGENCY FEEDWATER SYSTEM (AFW) .J v.. No.

Source-Issue

[. ~ ' 7-14 NUREG-0667. - This report discusses the transient response of B&W-l - May 1980 designed reactors and contains several recommendations, L D l Rec. 2.2 (1)~ including those listed below, r h The AFW system on operating B&W plants should be classi-l, fied as an ESF system and upgraded as necessary to meet safety grade requirements. L Rec. 2.2 (2) AFW should be automatically initiated and cor.trelled by ESF (safety grade)Lthat are independent of non-nuclear-instrumentation / integrated control system-(NNI/ICS) and other non safety systems. 1 -l The selection of signals used to initiate AFW flow should be re-evaluated to permit automatic initiation in a more timely manner to preclude steam generator dryout. q The level in steam generators should be automatically e controlled by the AFW to prevent overcooling and to y, terminate flow before overfilling. 1 Rec. 2.2 (3) Installation of a diverse-driven AFW pump should be t expedited at Davis-Besse. J i 18 ) i s,.# c. --,-w.. -,,--,m-,g-f..-e.,n,., ,ny w - -,,,,,.,_, -. -., -,.. ,,..,,m

cy + a ye .D' .No. Source Issue-

e. ~

RecC2.2(7)~ Provide the flexibility to substitute combinations'of' ~ -in-core.thermocouples for-loop resistance temperature W . detectors (devices) (RTDs) used for input'to subcooling- ' meter and the capability of continuous or trending display of,in-core thermocouples. Ree 2.2 (8) Provide safety grade containment high-radiation signal to initiate containment vent and purge isolation. 'o Rec. 2.2 (21) The.need to introduce AFW through the top sparger during a anticipated transients should be re-evaluated by licen-1 sees; consider the reduced depressurization response if + AFW could)e introduced through the main feedwater (MFW) nozzle and could enter the tube region from the bottom. of the unit. 7-2 IEIN 80-23 This notice describes an event at Arkansas Nuclear One c ' May 23, 1980 following a loss'of offsite power and a reactor trip. I Emergency feedwater (EFW) pumps, which started and pro-vided feedwater to the steam generators, list suction which forced hot water through the startup and flowdown j as a result'of flashing in the main feedwater train L h 'demineralizers to the' EFW pump suction where 'it flashed - to steam and caused pump cavitation. Action to prevent reoccurrence included revising the>EFW system operating procedure and plant startup procedure to require shut-ting-EFW suction valve from startup and flowdown domin-eralizers during plant startup after the steam generator:

1 feedwater source has been shifted to main feedwater pump. 3-GL 81-28 See item 6-2 above.

- July 31, 1981 7-4 IEB 85-01 This bulletin relates to steam binding of AFW pumps and Oct. 29, 1985 requests licensees to develop procedures for monitoring Item 1 fluid conditions within the AFW system on a regular basis during times when the system is required to be operable. This monitoring should ensure that fluid temperature at the AFW pump discharge is maintained at about ambient temperature. Monitor < ng of fluid condi-I tiens, if used and the primary basis for precluding L steam binding, is recomunended each shift. )' - Item 2 Develop procedures for recognizing steam binding and for restoring the AFW system to operable status, should steam binding occur. ? Item 3 procedural controls should remain in effect until com- -pletion of hardware modification to substantially reduce F the likelihood of steam binding or until superseded by u, action implemented as a result of resolution of Generic 1 Issue 93. j D ig 1 l\\ j v y ..iy --s.n...,. U .-m '...~*._e......._,.w,ww..-., ..~..e..

,f,

la f ', . No'. ' Source issue: ^ c m. .a C ' 7-li IEIN 84-06 This notice indicates that leakage into the AFW system-iJan.- 25; 1984 from the MFW system constitutes a common mode' that can- ~ lead to.a loss of all AFW capability. as a result of-staan - binding. In addition there is a potential for water hammer damage of AFW pump if relatively cold water dis-4 charges into a region of the piping systes that contains 7 steam. o i m Item 14 It is not clear that overcooling transients such as L i L occurred at Rancho Seco are within the bounds of the FSAR analyses.. 7-5 IE1H $6-14 This noti .turbineskediscussesoverspeedtrip/lockoutofAFW Mar. 10, 1986 nd the various causes listed below. i Supplement l' o Dec. 17, 1986-immediate clearing and reset of overspeed trip signal-caused trip on restart of AFW L leaking steam supply valve caused non-zero initial speed f undrained condensate in long steam supply lines b trip on restart because the governor is designed to- + start with no initial control of oil pressure (Iow pressure) (Because the oil pressure does not-decay y quickly and the provision to dump the oil is local and manual the turbine will not restart until the j -cil pressur,e is reduced.)- L The supplement-to this notice addresses an NRC/AE00 re-port en turbine overspeed. The dominant causes were problems related to governor speed control, trip valve. and overspeed trip mechanism. The report recommends a procedura1' change-to start up the turbine by warning it with a small steam flow before exposing it to full steam -flow. -Both procedural inadequacies and human errors were found to contribute to' improper setting.of governor speed. o

CATEGORY 8: ADMINISTRATIVE CONTROLS No.-

Source Issue , 1 NUREG-0667 This report recommends that mandatory 1-week simulator May 1980 training should be required for all licensed 8&W oper-Ree2.2(15) ators, oriented toward undercooling and-overcooling events, solid waste system operation, 6nd natural circu-lation cooling. Item 7 Evaluate procedures and training for reporting events to the NRC Operations Center. Review the adequacy of 20 D (; ~

w q

t CATEGORY 8: ~ ADMINISTRATIVE CONTROLS f': (; - No.' ' Source Issue shift. staffing for ensuring that knowledgeable indivi-duals will:be available for properly implementing:the H emergency plan during complex and long operational events.

Item 13 Yerify that plant procedures involving " drastic" actions are sufficiently precise and clear to ensure proper ss

implementation, i 2 NUREG-1195 This report states that the anticipated transient oper-Feb. 1986 ating guidelines (AT0G) supplied by the BWDG include an ~ Section 10.1 explicitgrocedureforlossof.~ICSpower. However this Item 4 procedure may not be included in the utilities' eme,rgency operating procedures (E0Ps), as it should be. Item 5 E0Ps direct operators to trip the appropriate pumps to terminate flow if feedwater flow cannot be isolated; however, operators seem reluctant to do so. Item 6 Operator. training and procedures should be adequate to resolve conflict between avoiding the pressurized thermal shock region and regaining pressurizer level. s Item 7 Operators should receive classroom and/or simulator ,u training on overall plant. response to either loss of ICS de power or the restoration of ICS de power. n Item 9 Non-licensed operators may only.be receiving walk-through or talk-throug! training where hands on train-ing may be necessary. Item 10 Radiological. control and emergency preparedness programs and training may not be adequate if events occur which H result in other than minor radiological consequences. a \\ l 8-3 . NUREG-1195 This report covers the adequacy of annunciator procedures Feb.-.1986. sanual concerning implications of ICS alarm'and value to

Section 10.2 operators in recognizing or restoring a loss of ICS de 6

L Item 2 power. Item 5 protective clothing or respiratory protection should be readily available in the event of an incident requiring emergency entry into a contaminated area. Item 6 Following plant modifications, review procedures, other than those directly affected, to determine applicability. j.. ' Item 7 Determine whether the staffing required by Technical Specifications and other regulatery commitments is adequate to mitigate the effects of an overcooling transient such as occurred at Rancho Seco. 21 l o. -..;.,~_ _, ~, -, _ _ ~,. ~..,,... -, _., _. -

m \\ .t y No. Source Issue Ei '8-4 NuclearSafety-fhisreportstatesthatproceduresfororderlyplant Anal;J,N$AC)-3 rsis Cen. shutdown following less of power supply should be pre-s tst pared or reviewed / revised, as necessary. Reacter March'1980 systes coeldown limits, and the basis for these liatte ip Rec. II.C should also be reviewed. \\ Rec. !!.D The industry should further analyse and resolve with l the NRC the current reacter coolant pump trip precedures to be followed during a small-break less of-coolant accident (LDCA). tec. II.C The industry should review the current high pressure injection pump requirements and resolve any procedural isens with the NRC. Procedures that avoid er aintette e,hallenges.to safety valves, risary systes tuellytothecontainmentbutgdingitselfar_andeven-e needed. Rec. II.F Procedures for declaring an emergency should be reviewed to deterstne if respons' bility for monitoring plant con-t ditions, which lead to declaring a specific category of L l> emergency, should be assigned to a specific individual. L Rec. 111.g.1 Data handling and display tystems should be reviewed to detemine their adecuacy. Ree 1.8 Power supply failures and their effects en control sys- ~ tems should be reviewed. ) -{ Rec. I.C. The work practices of instrumentation technicians and q their effects en plant safei,y should be reviewed in plant training sessions. Rec. II.A Written procedures should be established for switching instruments between power supplies in the event of power j l supply failures, including designating the preferred 3 bus for each instrument.' ] 1 - .j Rec. II.g Procedures for steam generator Pupture metrix er its equivalent should be reviewed in conjunction with post-l TMI requirements en steam-driven energency feedwater pumps to detereihe if aggravating effects axist during less of heat sink. i 22 i ,_ __._ -.._.._. ~ -.- - ~ -- - - - - - -}}