ML20127A414

Insp Repts 50-206/85-13,50-361/85-12 & 50-362/85-11 on 850304-08 & 0401-05 & 15-19.Violation Noted:Maint Orders Issued to Replace Studs W/O Performing Design Analysis, Constituting Improper Design Change

ML20127A414
Person / Time
Site:	San Onofre
Issue date:	06/06/1985
From:	Johnson P, Narbut P, Willett D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20127A403	List: ML20127A401 ML20127A409 ML20127A414
References
RTR-NUREG-0737, RTR-NUREG-737 50-206-85-13, 50-361-85-12, 50-362-85-11, IEB-79-27, IEB-80-06, IEB-80-6, IEB-82-02, IEB-82-2, IEB-83-06, IEB-83-6, IEB-84-02, IEB-84-2, NUDOCS 8506210134
Download: ML20127A414 (13)

See also: IR 05000206/1985013

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U. S. NUCLEAR REGULATORY COMMISSION

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REGION V

Report Nos. 50-206/85-13,50-361/85-12,-50-362/85-11

Docket Nos. 50-206, 50-361, 50-362

License Nos. ~DPR-13, NPF-10, NPF-15

Licensee: Southern California Edison Company

P. O. Box 800

2244 Walnut Grove Avenue

Rosemead, California 91770

Facility Name: San Onofre Units 1, 2 and 3

Inspection at: San Clemente, California

Inspection conducted: March 4-8, 1985 and April 1-5, 15-19, 1985

Inspector: ) %lJt

P. P. Narbut, Project Inspector

5/31/8 5

Date Signed

M)& J. Willett,

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eactor Inspector

5 $/~~ S$

Date Signed

Approved By: M [

P. H. nson, Chief Date Signed

Reacto Projects Section No. 3

Summary:

Inspection during period of March 4-8 and April 1-5, 15-19, 1985 (Report Nos.

50-206/85-13, 50-361/85-12 and 50-362/85-11)

Areas Inspected: Unannounced inspection by regional inspectors of unresolved

and followup items, and licensee actions on IE Bulletins and TMI Action Items.

The inspection involved 86 inspection hours by two inspectors.

Results: One violation was identified regarding an unauthorized design change

(paragraph 4).

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8506210134 850607

PDR ADOCK 05000206

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.1. -Individuals' Contacted

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a.  : Southern' California Edison Company (SCE)

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. Kr.*J.:G._Haynes, Station' Manager

J . J*R. W. Kreiger, Deputy Station Manager

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' *D. B. Schone, Site QA Manager

' *P. ' A'. Croy, Manager, Compliance

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.(*D. E. Brown,. Supervisor, Maintenance Codes.and Welding

. . *R. F. Penn, St'ation Codes Engineer

• D. Sheridan,-Supervisor Plant Maintenance, Unit 1

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' *H.'. Newton, Manager Maintenance Engineering

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• T.7 A.~ . Mackey,. Supervisor Compliance

• C..A. Kergis,: Compliance Engineer '

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F. Briggs,'NSSS,' Mechanical Supervisor

'*J. Reilly, Technical Manager

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_~ s 't' .M. Wharton, Supervising Engineer, NSSS Support ,

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v'J. Boardman, ISI Engineer

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W.~ Savage,. General Foreman, Maintenance Planning

/ *W. M. Lazear, QA Supervisor- , ,. .

V.-Gow, Lead QA Engineer, Codes

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4 , jL. Rice,, Supervisor,, Site Supply 7 ,

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• w Denotes those. personnel in' attendance at the exit.. interview on April:

19, 1985. -

Licensee Actions on Bulletins and-Previously Identified Inspection Items

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.(Closed) IE Bulletin 79-27,: Loss of Non-Class IE Instrumentation

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-  :-and Control Power System Bus During Power Operation', Unit 1

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Thebulletinrequiredlic[enseesttoreviewpowerlsuppliestosafety

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and non' safety related instrumentation'and control' systems which

could affect the ability,to achieve cold shutdown,3 to" describe any

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resulting design modifications.and*to p'r'epire J

emergencyLprocedures

for such an event. The' licensee'ssactions.for thisibulletin were

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examined in Inspection' Reports'50-206/80-16, 80,-19,~80-28 and 80-31.

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Theremaining.actionw'as'toverify'thatthe/ licensee lhadprepared

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emergency procedures for'the-loss of the'125v.DC busses prior to the "

JUnit 1. return to power in;1981. Unit!I returned.to" power on June-

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17, 1981.

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-This Inspection

'The.' inspector examined the current' emergency procedure'for the. loss

^of the-125v DC buses. The current procedure is' Abnormal Operating

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Instruction S01-2.6-4,, Revision 0, dated NovemberL17, 1983. The

l procedure contains information, as required by the bulletin, to ,

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enable the operator to recognize the symptoms of a -loss of the buses

and describes the required operator: actions. The inspector verified

that the procedure had in fact .been . issued prior to the 1981 return

to power. The procedure atLthat time was S01-1.7-3,. issued May 4,

1981, and contained ,the same ,information which was reorganized and

later. issued as th'e superseding (and. current) procedure S01-2.6-4. 1

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Based on the licensee's actions this bulletin is considered closed.

, b. (0 pen) IE Bulletin 80-06, Engineered Safety Features (ESF) Reset

Controls

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The licensee actions regarding the subject bulletin were examined in

report 50-206/81-29. All aspects were closed except valves MOV-

1100B, C.and D which did not respond properly after the ESF signal

was reset. The-licensee had established administrative (procedural)

controls for an interim control measure. A letter from SCE to NRC

dated February 27, 1984 defined the commitment to rework the control

circuitry for the valves during the Cycle 9 refueling in 1986.

This item remains open pending licensee action on valves MOV

1100-B,C, and D.

c. (Closed)-IE Bulletin 82-02, Degradation of Threaded Fasteners

Background

The bulletin addressed corrosion and cracking of reactor coolant

pressure boundary fasteners and imposed inspection requirements

beyond those already addressed in inservice inspection (ISI)

requirements. The licensee's actions and responses to the bulletin

were addressed in inspection report 50-206/84-11 which closed

certain items. The remaining licensee action was to inspect

specified fasteners when otherwise removed for maintenance or repair.

and report the results to the NRC.

The licensee had requested certain exemptions from the Bulletin

requirements as follows:

Results of fastener inspections will be included in ISI

inspection reports available for NRC review at the site, rather

than reported separately to the NRC as required by the bulletin

(reference SCE to NRC letter dated April 3, 1985).

Nondestruction examinations will only be done if a fastener is

to be reused and if-it has significant degradation, otherwise

only a visual inspection will be performed, unless an

evaluation of the visual examination otherwise dictates

(reference SCE letter dated April 3,1985).

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This Inspection. _

The inspector examined the procedures in place to perform the

bulletin required inspections. The procedures ' applicable are:

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'S0123 XVII-1.0, Revision 0, dated October 3, 1984, " Inservice

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Inspection Program Implementation"'

S0123-I-7.19, Revision 9, date'd August 24,.t1983, " Monitoring of

Threaded Fasteners (Bolts / Studs) in the Reactor Coolant

Pressure Boundary (RCPB)" ,

Theprocedureswerefou$dtoadequatelydefinethebulletinrequired

actions for fastener inspection.

The inspector examined the licensee's inspection data for bolting

from the following components.

Pressurizer Relief Valve bolts for valves 533 RV, 532 RV

Steam Generator manway bol'ts for steam generators A, B and C,

hot and cold leg sides

RHR valves MOV 833, MOV 813, M0V 814

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The bolt inspection results were satisfactory for the pressurizer

code safety valves and the steam generator manways. The inspector

verified that data were recorded and actions were taken in

accordance with procedure requirements.

The results for the RHR valves were unsatisfactory. The licensee

had performed a walkdown for corrosion as part of the return to

service effort for Unit 1. That walkdown identified corroded

fasteners on the RHR valves. The bulletin's fastener inspection

requirements were properly invoked by the licensee maintenance

personnel in conjunction with their decision to replace the

fasteners.

As will be discussed separately in paragraph 4 of this report, the

selection of replacement fastener material was not proper; a lower

strength material was used without proper engineering review and

approval.

In regards to bulletin actions however, the fastener inspection data

for the corroded RHR valves also appeared to be improperly taken.

Specifically, the maintenance orders required inspection of the

fasteners (in accordance with the procedure) by visual examination

and then a surface examination by NDE. This was required "even

though existing fasteners may be replaced with new fasteners." The

purpose of the bulletin was to examine corroded fasteners for

evidence of stress corrosion cracking by performing surface

examinations.

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,g lTheLprocedural step signoffs by the maintenance' mechanics (on the-

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maintenance' work orders))were marked."N/A" in the steps for

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y decontamination of the bolts,. examination of'the bolts, and

forwarding-of the results'to' engineering for evaluation. The steps-

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were annotated that the

step was N/A'd because the bolts could not

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'be decontaminatedLand were too corroded for inspection. -The.

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-inspector considered .that, technically, the rationale for not

, e performing the . inspection was poor. The licensee has facilities for

performing' inspections of contaminated-items. Additionally,-the

corroded condition of the bolts made.them good examples of the kinds

, .of~ conditions the bulletin inspections were directed towards,..rather:

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than a reason not_to perform an inspe'ction.

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The maintenance. orders involved.were:

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'MO 306819 for MOV 813 .

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1* MO 307501 for MOV. 814 '

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MO 84042175001 andsMO 307503 for'MOV.833c4 1

Thecompletedmaintenanceobder.(with.tbeinappropriatelyN/A'd

steps) was reviewed and' approved by the procedurally required

organizations which should have flagged ~the; inappropriate decision.

's .These organizations inc1'uded QA-and Engine'ering'. The inspector

"+ considered the issuance of a' Notice of Violation -for'this matter but

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,. concluded a violation?was not involv,ed for the following= reasons:

• - . Lack of Safety Significance: The/ fail'ure to inspect bolts did- 1

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not meet procedure requirements'atf,the, time, but'do meet

current requirements ~('einspect

r for. reuse ~only). Secondly, '

. there has not been a history of stress corrosion cracking of >

this material. The bolts were. examined by.the licensee'(as a

. catch-up-effort during this' inspection) and none had stress ,

corrosion ~ cracking problems. .

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Previous Licensee Actions: As'a result'of other' violations the

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licensee had since taken. aggressive actions regarding the."N/A"

of procedural steps. These actions were'not.in place at the-

time of the Solt inspections in 1984.

Conclusion:

Based on the licensee's sctions and the revised requirements

' reflected-in the licensee's April 3, 1985 letter.to the NRC which

reduce the data gathering and reporting requirements, this. bulletin

is considered closed.

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d. l(Closed) IE Bulletin 83-06, Nonconforming Materials Supplied by

g }lC Tube-Line Corporation

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.Backgroundi

The-licensee's actions for this bulletin were previously examined in

inspection. reports 50-206/84-23 and 84-30. At that time the . .

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inspector'had found additional installed Tubeline material which'had >

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not been identified by the licensee's search. The remaining actions

were for the licensee to submit a revised response to the bulletin,

to include ~ actions to prevent recurrence of an incomplete search and

to ensure previous actions taken regarding generic communications

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(on material problems) were properly investigated'and evaluated.

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This Inspection

The inspector reviewed the licensee's revised response to the

bulletin. The revised response was dated December 4,1984. The

licensee identified the reason for the initial incomplete search as

having failed to review the material provided through Bechtel field

purchase orders. This search was a general one (for material

searches) during the period of September 1980 through September

1984, when no Bechtel field organization was in place at SONGS 1 to

review field purchases. During this time the Bechtel search mode

was limited to home office purchase reviews. It should also be

noted that the specified scope of the bulletin was ASME piping, and

the Tubeline fittings found at San Onofre 1 were in systems not

constructed to the ASME Code.

To prevent the recurrence of incomplete searches, the licensee has

initiated significant action in the form of a program called COPE

(Control of Problem Equipment) which is planned to be totally in

place by June 1985. The program was generated in response to the

inspector's findings regarding this bulletin and Bulletin 84-02

(paragraph 2.e) regarding HFA relays. The completion of the COPE

program will be examined as part of the followup of that Bulletin.

Additionally, SCE instructed Bechtel to perform a reinvestigation of

material searches conducted during the period that a Bechtel field

procurement group was not in place. SCE audited those Bechtel

remedial actions and found them satisfactory. Additionally, Bechtel

identified, in a letter dated January 11, 1985 (McClusky to Nunn),

how they would revise and proceduralize future material searches to

ensure completeness.

Subject to the followup of the licensee's new COPE program to be

followed up as part of Bulletin 84-02, this bulletin (83-06) is

considered closed.

e. (0 pen) IE Bulletin 84-02, Failure of GE HFA Relays

Background:

The licensee actions for this bulletin were examined and described

in inspection reports 50-206/84-30 and 84-33. The inspector had

determined that the licensee's material search had not been

comprehensive; the inspector discovered the subject relays in the

warehouse. The remaining items regarding the licensee's actions

were verification of the COPE program (Control of Problem

Equipment), and verification of the procurement Quality Assurance

surveillances of the adequacy of licensee actions for previous

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material searches conducted for IE Bulletins and Information

Notices.

This. Inspection

'The inspector reviewed site order S0123-CP-1, Revision 0, dated

February.27, 1985, which outlined the program to be established and

defined responsibilities.

The inspector met-with responsible personnel and reviewed the

schedule, which shows-full implementation by June 1985.

-This bulletin remains,open pending' implementation of the COPE

program and ' review of the procurement QA verification actions.

3. Licensee Actions on Previously Identified Inspection Items

-a. (0 pen) Followup Item 50-206/80-11-03, No cold-to-hot

relief pressure correlation factor for the Pressurizer Code Safety

Valves setpoint testing

Background:

The inspector had previously. observed in the 1980 inspection that

the licensee performed required periodic set pressure tests of the

Unit.I' Pressurizer Code Safety valves in a cold condition. Cold

testing of safety valves is a generally acceptable method of

verifyingfset pressure provided that the expected difference in the

pressure at which the valve actuates (in the hot versus cold

condition)'is known and accounted for.

This requirement, for.a cold-to-hot correlation factor, is derived

from the Technical Specification requirement to perform periodic

relief pressure tests.of the subject valves in accordance with the

, requirements of ASME Section XI. The ASME Section XI requirements,

in turn,' reference ASME PTC 25.3 for the' performance of safety and

L relief valve testing. The. introduction of PTC.25.3,~Section.0,;

cautions that""...if the~ temperature'of the medium used to test the

valve differs substantially- from th'e te'mperature to which the valve

i, subjected while in service, the, opening and closing

prassures. . .will be different from the test pressure. . .it ~is

naressary to develop ap'propriate corrections for the' valve under

. ' test to account for thesefdifferences.",'i.;

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The open status of this item was discussed at 'a regional inspector's

exit interview on December: 21,'1984.

This Inspection

The inspector reviewed the current versions of the maintenance and

test. procedures applicable to the pressurizer code safety valves

(PRV 532 and 533) and confirmed the licensee had not yet included a

cold-to-hot correction factor in'their set point test procedure.

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y The current set point e'st procedure is S01-1-2.3, Revision 2, dated

October 14, 1984. The valve is tested cold with a nitrogen medium.

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The inspector met with t.he responsible site engineering manager and

his technical' staff to discuss'the item. The staff presented

'information which demonstrated that the issue of: proper testing of

safety and relief valv'es. is an-industry wide subject of interest.

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There have been several initiatives by various organizations aimed

at resolving the issue. ,Two notable efforts are (1) an NRC

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requirement (imposed in the technical specifications) for newer

plants to test safety valves hot,and with the same medium as seen in

- service and (2) an ASME code committee'(OM-1) effort to revise the

code requirements to require essentially the same thing but provide

some' options. The licensee had gathered some data, from TVA and

'EPRI, but the data showed significant scatter. The data do not

apply directly to the Unit 1 Crosby valves but are consistent in

that the hot tested' set point is always lower than the cold tested

setpoint (the co'nservative direction).

The inspector-requested the licensee to provide rationale to

' demonstrate that the change in set point (from cold-to-hot) for the

Unit I safety valves would be in the conservative direction and to

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affirm that'no safety analyses would be adversely affected by the

expected change in set point. The licensee presented general

industry information which demonstrated that safety valves have a

hot set point.from 1 to 180 psi. lower than the cold set point. The '

fact that San Onofre does not have a history of inadvertent relief

valve lifts indicates that the change in set point for a hot valve

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is not an operational problem at normal operating pressures. The

licensee'also presented a memorandum dated A,ril 16, 1985 from the

Westinghouse site representative to Station Technical, SCE. The

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memorandum states:

Westinghouse has no directly applicable data for the Unit 1

valves but the same type of conservative change can'be

expected.

There is no unresolved safety concern with a cold-to-hot set

point drift (in the lower set point direction).

At the exit interview the inspector discussed the findings regarding

this item, specifically that no apparent action toward resolution

had occurred since the item was identified in 1980 and that the

information provided indicated there was not an unresolved safety

question involved. The inspector requested the licensee to identify

the path to resolution of this item. Licensee management stated

that the resolution of this item would be addressed in the context

of a larger initiative, that being the licensee's plans for a

general revision and reformatting of their technical specifications

to be accomplished in conjunction with their SEP actions.

Specifically, the licensee plans to update the FSAR (and technical

specifications) after the completion of the SEP evaluation

(scheduled for 1985) in accordance with 10 CFR 50.71(e)(3)(ii).

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This item remains open'ending

p 'further licensee action.

4. Maintenance Issues

During the examination of IE Bulletin 82-02'regarding degradation of

fasteners, the inspector examined certain aspects of the licensee's

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controls for maintenance and maintenance testing. These aspects included

compliance with ASME Section XI repair and testing requirements, material

control, engineering involvement.in design changes, records' accuracy,

procedure adequacy, torquing requirements, proper test pressure, and

supplier material certifications. Specifically, these aspects were

reviewed as they related to the replacement of corroded carbon steel

fasteners with stainless steel fasteners on RHR valves MOV 833, 813 and

814. All aspects examined were found satisfactory with the following

exceptions:

Improper Design Change

The maintenance orders issued to replace corroded body to bonnet studs on

the RHR valves replaced the original carbon steel fasteners with lower

strength stainless steel fasteners. There were two central problems

identified by the inspector:

No design analysis was performed to assure the lower strength

material was adequate for service.

Licensee personnel failed to recognize the material substitution as

a design change.

The specific details involved are as follows:

The valves involved are RHR MOV-813, 814 and 833. The valves are reactor

coolant pressure boundary components. Valve 813 is the first isolation

valve in the reactor coolant system loop C outlet to the RHR system; and

therefore sees RCS pressure in service.

The associated documentation for the valves is:

Valve Maintenance Work Order ASME Section XI Abstract MERS*

MOV 813 306819 S01-011-83 010-83

814 307501 S01-009-83 007-83~

833 307503 S01-010-83 008-83 Rev. 1

• Maintenance Engineering Repair Specification

The original construction design code was ANSI B31.1, 1964 Edition, but

repair and replacement are to be done to ASME Section XI,1977 ' Edition.

The manufacturer was Crane Valve Company, drawing DR 33473 (for 6" valve

833) and drawing DR 33463 (for 8" valves 813 and 814).

The original bolting material specified by Crane was ASTM A-193 Grade B7

Chrome-Moly Carbon steel (which has a minimum tensile strength of

approximately 125 ksi). The replacement material installed by SCE was

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ASTM A 193 Grade B8 Chromium Nickel Stainless steel (which has.a minimum

tensile strength of approximately 75 ksi).

The responsible station engineering manager stated a design

reconciliation analysis had not been performed to verify that the

material substitution was technically adequate. During the inspection,

preliminary analysis was performed in response to the inspector's

findings and the preliminary results showed that the new material was

satisfactory for service, that design margins have possibly been reduced,

but that there was no cause for an immediate safety concern. The

responsible engineering manager so stated ~during the exit interview on

April 19, 1985.

The Code of Federal Regulations, 10 CFR 50, Appendix B, Criterion III,

state in part that:

" Design changes including field changes shall be subject to design

control measures commensurate with those applied to the original design."

The licensee's design control measures for repair and replacement

activities are specified in the SCE Topical Quality Assurance Manual,

Appendix IV "ASME Code Section XI Repair and Replacement Program" which

states in part " Repairs and replacements performed at Unit I will

implement the requirements of ASME Code Section XI, 1977 Edition through

Summer 1978 Addenda...."

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ASME Section XI, IWB 7600 states in part " Materials shal1 comply with the

requirements to which the original component or part was constructed."

IWA 7210 allows for replacement material provided: " Modified or altered

designs are reconciled with the Owner's Specification through the Stress

Analysis report, Design Report or other suitable method whi-h

demonstrates satisfactory use for the specified design and operating

conditions...."

Contrary to the above, carbon steel body to bonnet studs were replaced

with significantly weaker stainless steel studs on RHR valves MOV 813,

814, and 833 without a design reconciliation by a suitable method which

demonstrated satisfactory use for the specified conditions.

This is an apparent violation. (Violation 50-206/85-13-01)

The inspector also-noted a similar carbon to stainless bolt material

substitution was authorized in 1981 for pipe flange bolting. Flange

bolting generally sees a lesser loading than body to bonnet valve bolting

and is generally very conservative in design strength. There was no

evidence at the time of inspection that the material substitution

authorized was, in' fact, implemented. Therefore no violation was

identified. The material substitution was authorized by a change to the

" Piping Design and Material Specification" Number M-18668 Sh 179 Revision

0, Configuration Change Notice 1 dated January 19, 1981.

At the exit interview the licensee representatives committed to determine

if the material substitution authorization was implemented and to verify

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that the substitution authorization is a technically sound design change.

'(Followup item 50-206/85-13-02)

Improper Test Pressure

RHR Valve MOV 813 is a boundary gate valve between'the RCS and RHR

systems. As such, in normal operation it is shut and sees RCS normal

operating pressure rather than the much lower normal operating pressure

of the RHR system on its "high pressure" side and in the valve bonnet

area. Upon' replacement of-studs the ASME Code requires a leakage test at

"not less than the normal operating' pressure associated with 100% rated

reactor power".

The studs in questions see RCS system pressure in service but were leak

tested at the lower RHR system pressure.

This matter is primarily one of test data record validity since the studs

did eventually see RCS system pressure and normal planned QC walkdowns

would have identified any leakage problems. At the exit interview, the

licensee committed to examine methods to improve the detailed

specification of test requirements for system boundary valves. (Followup

Item 50-206/85-13-03.)

5. TMI (NUREG 0737) Activities (Unit 2)

The inspector reviewed the licensee's program for/and implementation of

NUREG 0737 items I.C.1, I.D.2, and II.F.2: Accident and emergency

procedure implementation, safety parameter display system, and

instrumentation for detection of inadequate core cooling, respectively.

a. I.D.2 (0 pen) Safety Parameters Display System (SPDS)

The accident monitoring system includes the Critical Factors

Monitoring System and the Safety Parameters Display System. A

sub-set to this is the Q-SPDS (SPDS for seismically qualified'

parameters). The licensee's implementation of the requirements in

I.D.2 was as follows:

The operability requirements of the SPDS (required by

NUREG-0696) are included within the technical specification

requirements for the Accident Monitoring System

instrumentation. The SPDS is installed and complete except for

that which is supplied by the heated junction thermocouples

(HJT). The licensee stated that the design change package

which incorporates the HJT's will be complete when the head is

installed, the instruments are connected to the SPDS and the

cables are rung-out, tested, and calibrated. This work is

scheduled to be completed just prior to restart after the first

refueling outage.

The inspector observed the use of the SPDS system by license

candidates in the control room, and verified their training and

understanding of the system. Closure of this item will be

verified at a later inspection when the HJT work is complete.

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.. b . II.F.2 (open) Instrumentation for Detection of Inadequate Core Cooling

Generic Letter No. 83-37 identifies those TMI items for which

technical specifications are required. ~ Enclosure 1, item 10,

provides guidance for II.F.2. Enclosure 3 provides model technical

specifications. The Subcooling Margin Monitor (SMM), the Heated

Junction Thermocouple System-and the Core Exit Thermocouples (CET's)

comprise the instrumentation for detection of inadequate core

cooling. The licensee's implementation of the requirements of

II.F.2 was as follows:

San Onofre-2 technical specifications (amendment 31) address

some item II.F.2 instruments in numbers 11, 18 of the accident

monitoring instrumentation tables for limiting conditions of

operation and surveillance requirements. On January 11, 1985,

the C.E. Owners Group submitted (after discussions of

fundamental disagreement) to NRR proposed HJT System Technical

Specifications. The licensee has submitted a Technical

Specification proposal (consistent with the C.E. Owners Group

proposal) to NRR for consideration.

Procedure implementation and training for prompt recognition of

inadequate core cooling using existing instrumentation (per

NUREG-0578) was observed during license examinations on the

plant simulator. Procedures used by operators to recognize

inadequate core cooling, which rely on data from the SMM, HJT,

and CET's .(per NUPIG-0694) will be evaluated when TMI item

I.C.1 is reviewed.

This item will be closed out after:

(1) Completion of the;HJT system work 4

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(2) When NRR has reviewed and amended the~ technic 1 specifications

(3) When a review of selected ' emergency procedures which use SMM,

HJT, CETs, and the SPD System is completed'during

a follow-up

inspection of TMI. item I.C.1.

c. TMI Item I.C.1 (open) Accident and' Emerge'ncy Procedure

Implementation ,

• .

This TMI item requires' licensees to: ,

[

.

y

Perform analyses of transients and accidents (including small

break LOCA's and inadequate core / cooling), and prepare

emergency procedures (symptom based) for multiple and

,

consequential failures.

Revise procedures to address inadequate core cooling,

transients and accidents (per NUREG-0694 and NUREG-0578),

maintaining consistency with the Final Long-Term guidance

contained in NUREG-899 (implementing document for-the emergency

procedures pilot program--TMI item I.C.8).

.- a-. - - -

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The' inspector verified thelexistence, use and training of personnel,

in symptom based single / multiple ~and consequential transient and

accidentprocedures,during}estingoflicensecandidatesonthe

plant simulator.

. '

. .

- ~ Procedures which address inadequate, core cooling will be reviewed

, 'during a follow-up inspection ~when the HJT System and QSPD. System

are; complete-and the technical specifications which address them are

approved by NRR. <

, ,

-6.

.

Exit Interview

The inspector met with the licensee personnel denoted in paragraph 1 on

April 19, 1985. The inspection details and findings as noted in this

report were discussed.

.

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