Text

Partially Deleted Reply Re St Lucie Operator
	ML20137Y479
Person / Time
Site:	Saint Lucie
Issue date:	03/18/1996
From:	Michael B; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	Linda Watson; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML20137Y003	List: ML20137Y001; ML20137Y022; ML20137Y042; ML20137Y052; ML20137Y479; ML20137Y498; ML20137Y503; ML20137Y507; ML20137Y514; ML20137Y518; ML20137Y540; ML20137Y544; ML20137Y552; ML20137Y556; ML20137Y576; ML20137Y594; ML20137Y609;
References
	FOIA-96-485 NUDOCS 9704230097
	Download: ML20137Y479 (73)
	v • d • e

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I From: Beverly Michaele O i To: ATP2(LJW2) L(A>473o% f2 i Date: 3/18/96 9:44am j

Subject:

st. Lucie operator -Reply ;

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EAW . 101,153 03-11-96 04:15p ENFINAL . 16,928 03-25-96 03:46p FINALSIG.SDE 40,801 04-01-96 03:06p NAMEADDR.OPR 1,010 03-21-96 04:28D.

RABREIF . 96,672 03-11-96 03:42p REPORT . 142,495 03-11-96 03:23p l

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i 15.2.4 4 CHD(ICAL AND VOLLHE CONTROL SYSTDi MALFUNCTION - BORON DILUTION !

IVENT

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15.2.4.1 Identification of Causes -

The chemical and volume control system (CVCS ) descri bed in Section 9.3.4 regulates both the chemistry and the quantity of coolant in the reactor coolant system. Changing the baron conceheration in the reactor coolant system is a part of normal plant operation, compensating for long-term

~ reactivity effects, such as fuel burnup, xenon buildup and decay, and plant !

startup and. cooldown.

For refueling operations, borated water is supplied from the refueling water tank, which assures adequate shutdown margin. An inadve rtent boron dilution in any operational mode adds positive reactivity,

produces power and possibly temperature increases, and, in Modes 1 and 2 (startup and power, operations) can cause an approach to both the DN3R and CTM limits.

Boron dilution is conducted under strict administrative procedures which specify permissible limits on the rate and magnitude of any required change in i baron concentration. 3oron concentration in the reactor coolant system can be 4 2

decreased either by controlled addition of unborated makeup water with a corresponding removal of reactor coolant (f eed and bleed) or by using the deborating ion exchanger. The de borr. ting ion exchanger is normally used f or boron removal when the boron concencration is low ((ppm) and the f eed-and-bleed method becomes inef ficient . A boronometer is located in a line upstream of the deborating and purification ion exchangers in the CVCS. This ins t rume nt provides a continuous measure of boron concentration and high-low boron concentration alarms.

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During normal operation, concentrated boric acid sof r-ion is mixed with demineralized makeup water to the concentration reqair;d for proper plant 3

operation and is automatically introduced into tha volyse control tank in response to a low water level signal from the volume evccrol. To effect boro n dilution, the makeup controller mode selector switch must. be set t o " Dilute" and the demineralized water batch quantity selector set to the desired

[;'

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quantity. When the specific amount has been injected, the demineralizer vater control valve is shut automatically.

Dilution of the reactor coolant can be terminated by isolation of the makeup wate r system , by stopping either the cakeup water pumps or the cha rging pumps, or by closing the charging isolation valves. A charging pump must be running in addition to a makeup watar pump for boron dilution to take place.

The CVCS is equipped with the f ollowing indications and alarm functions, which will inform the reactor operator when a change in boron concentration in the a

reactor coolant system may be occurring:

a) Boronometer high and low alarms and concentration indication b) Volume control tank level indication and high and low alarus i

15.2.4-1

J i c) Makeue flow indication and alarms d) Volume control tank isolation. _

Changes in boron concentration while the reactor is on automatic control at full power are compensated for by repositioning the CEA's. Ebwever, t o assist

, the reactor operator in maintaining an adequate shutdown margin, CEA insertion

_ below a position that would provide a minimum of one percent shutdoun margin-(assuming one stuck CEA) is accompanied by control room alarms. Because of h the procedures involved and the numerous alarms and indications available to ,

the operator, the probability of a sustained or erroneous dilution is very low. (

15 2.4.2 Analysis of Effects and Consequences 15.2.4.2.1 Method of Analysis 11e rise required to achieve criticality from a suberitical condition due to boron dilution is based on the initial and critical boron concentrations, the baron reactivity worth, and the rate of dilution. Reactivity increase ratesE due to boron dilution are based on the boron worth and the dilution race.

Cases have been analyzed f or all six o perational modes, i.e. , power operation, startup, hot stand by , hot shutdown, cold shutdown. and refueling. " In each case, it is assumed that the boron dilution results from pumping unborsted deminerali:ed water into the reactor coolant system a t the maximum possible rate of 132 gpm (3 x 44 g pm per charging pump) and that the boron concentrations are uniform at all time s.

The boron dilution race is calculated by CESEC f or all cases except dilution during refueling. CESEC described in Section 15.1.4-1 divides the reactor coolant system into 15 control volumes with the continuity equation beine ' ..

satisfied by all nodes. The charging rate of non-borated water and the NAR$e~^ {

content of the system are inputs to CESEC. The maximum d.ilution rate h

(10.5 pps/ minute) occurs at the initiation of the transient. For dilution during refueling the reactor coolant system is assumed to be one control volume with the boron concentration calculated by: the time rate of changa o,f.

boron equals flow in times the beron concentration sninus flow out times baron concentration.

The uniformity of the boron concentration can be assured f or the dif ferent modes of operation as follows:

a) During refueling Prior to cooldown, the reactor coolant system boron concentration is increased to a minimum of 1720 ppa. The boron is mixed by the reactor coolant system pumps. Because the boron is chemically dissolved in the reactor coolant , it will not precipitat e . The only possible means of obtaining a nonuniform solution is by the addition of domineralized water via the charging pumps. However, because the maximum water An additional boron dilution event would be via the Iodine Removal System (NaOH spray additive). This event is not governing, however.

See Reference 42.

15.2.4-2 l I

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CORRECTIVE ACTIONS !

PORV TESTING IMMEDIATE ACTIONS TAKEN l

l UPGRADED POST MAINTENANCE TEST PROCEDURE M-0037, -

l "PORY MAINTENANCE" TO VERIFY VALVE OPERATION IN ADDITION TO SEAT LEAKAGE TEST TESTED BOTH PORVs ABILITY TO LIFT PRIOR TO )

INSTALLATION 1

l UPGRADED INSERVICE TEST PROCEDURE AP 0010125A, DATA SHEET #24, " VALVE TESTING" IN ADDITION TO ACOUSTIC MONITORS' RESPONSE, VERIFY RCS PRESSURE DECREASE >5 PSIG TO CONFIRM MAIN DISC OPENING ,

l OTHER CONFIRMING SYSTEM PARAMETERS ARE RECORDED AND EVALUATED:

QUENCH TANK TEMPERATURE 4 QUENCH TANK LEVEL QUENCH TANK PRESSURE

PORV TAILPIPE TEMPERATURE l

l 950352 22 - 9/21/95

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CORRECTIVE ACTIONS

! PORV TESTING !

, (continued) i EXPECTED ACTIONS TO PREVENT RECURRENCE COMPLETION REVISE THE UNIT 2 PORY POST MAINTENANCE 11/15/95 TEST PROCEDURE TO VERIFY THAT THE MAIN DISC ACTUATES REVIEW POST MAINTENANCE TESTING OF OTHER SAFETY RELATED EQUIPMENT TO ENSURE THE TESTING ADEQUATELY DEMONSTRATES COMPONENT OPERABILITY:

l CONSOLIDATE TEST GROUPS UNDER COMPLETE A SINGLE MANAGER REPORTING TO 1

THE OPERATIONS MANAGER !

1 REVIEW UNIT 2 OUTAGE SCOPE POST 11/9/95 MAINTENANCE TEST PROCEDURES TO l ENSURE CRITICAL COMPONENT FUNCTIONS ARE ADDRESSED REVISE PROCESS FOR POST MAINTENANCE 1/1/96 ,

TESTING TO IMPROVE COORDINATION i AMONG OUTAGE MANAGEMENT, l

OPERATIONS AND MAINTENANCE 1 l

REVIEW UNIT 1 OUTAGE SCOPE POST 4/96 MAINTENANCE TEST PROCEDURES TO ENSURE CRITICAL COMPONENT FUNCTIONS ARE ADDRESSED l

950352 - 23 9/21/95

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l EVALUATION OF SAFETY SIGNIFICANCE l

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! FUNCTIONS OF PORVs i

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POWER AND SHUTDOWN OPERATION ASSESSED ;

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l PROBABILISTIC SAFETY ASSESSMENT 1

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950352 24 9/21/95

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EVALUATION OF SAFETY SIGNIFICANCE (continued)

FUNCTIONS OF PORVs o POWER OPERATION ,

I PORVs PREVENT LIFTING CODE SAFETIES DURING OPERATING TRANSIENTS AND ARE NOT RELIED UPON FOR ANY SAFETY RELATED OPERATING FUNCTION EMERGENCY OPERATING PROCEDURES (EOPs) USE PORVs AS CONTINGENCY FOR BEYOND DESIGN BASIS EVENTS THAT INVOLVE MULTIPLE SINGLE FAILURES SUCH AS COMPLETE LOSS OF SECONDARY HEAT REMOVAL l

o SHUTDOWN OPERATION PORVs ARE REQUIRED FOR LTOP MITIGATION AND ARE INCLUDED IN THE TECHNICAL SPECIFICATIONS:

MODE 4: RCS COLD LEG s 304 F

MODE 5 & 6: VESSEL HEAD ON AND RCS NOT VENTED 950352 25 - 9/21/95

k is EVALUATION OF SAFETY SIGNIFICANCE (continued) ,

POWER OPERATION ASSESSMENT.

VARIOUS METHODS FOR DEPRESSURIZATION PRESSURIZER SPRAY SYSTEM .

ATMOSPHERIC DUMP VALVES STEAM BYPASS CONTROL SYSTEM

SOME PRESSURIZED WATER REACTORS DO NOT HAVE -

PORVs TECH SPECS ALLOW BLO'CK VALVE CLOSURE >

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PORVs ARE NOT REQUIRED FOR UFSAR CHAPTER 15 ;

ACCIDENT MITIGATION l

CODE SAFETY RELIEF VALVES PROVIDE PROTECTION i

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4s e EVALUATION OF SAFETY SIGNIFICANCE ,

(continued) i PROBABILISTIC SAFETY ASSESSMENT (PSA) o PORVs MODELED IN BASE PSA i

o EFFECT OF PORVs ON CDF PSA WITH PORVs: 2.1 X 10~5/ YEAR PSA WITHOUT PORVs: 7.6 X 10 5/ YEAR (i.e., LOSS OF SECONDARY HEAT REMOVAL) l o CDF REMAINS LESS THAN NRC SAFETY GOAL OF 10 d/ YEAR

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ADDITIONAL CREDIT FOR OPERATOR ACTIONS TO RESTORE FEEDWATER TO S/Gs WOULD FURTHER REDUCE l

i CDF l

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EVALUATION OF SAFETY SIGNIFIr \NCE I

(continued) l i

SHUTDOWN OPERATION ASSESSMENT l

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i l REVIEWED UNIT 1 OPERATING HISTORY l

HAVE NEVER EXPOSED THE REACTOR VESSEL /RCS TO OVER PRESSURE EVENT (LTOP RANGE) )

1 REVIEWED THE ANALYZED LTOP TRANSIENTS i

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CONSIDERED MASS AND ENERGY ADDITION TRANSIENTS l

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MASS ADDITION TRANSIENTS ARE MOST LIMITING i

HPSI AND/OR CHARGING PUMPS 1

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l 950352 9/21/95

SHUTDOWN OPERATIONS ASSESSMENT EFFECTS ON REACTOR VESSEL ,

REACTOR VESSEL ASSUMPTIONS

- ISOTHERMAL EVENT g M A - 1/4 T FLAW (2.2" X 13")

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- LOCATION LIMITING WELD (LOWER AXIAL

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- RT ndt CORRECTED FOR IRRADIATION

- BOUNDING CRACK INITATION CURVE -Kic

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(ASME APP. A)

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a (PRES 95035246-RO) 950352 9.21/95 l

l-t i EVALUATION OF SAFETY SIGNIFICANCE i

(continued)

SHUTDOWN OPERATION ASSESSMENT (continued)

LTOP TRANSIENTS ANALYTICAL RESULTS ;

A88 E RCS TEMPERATURE MAXIMUM gOURC OF AL OM LE

( F) I }

PRESSURIZATION (PSIA) s200 HPSI PUMP s 1300 s 1510

> 200 ^ 5 2575' s 2750 3 CH PU S l

I CODE SAFETY RELIEF PRESSURE (2500 PSIA) PLUS ACCUMULATION (75 PSIA) o IMPOSED STRESSES ARE LESS THAN REQUIRED FOR CRACK INITIATION "

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950352 30 9/21/95 l

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EVALUATION OF SAFETY SIGNIFICANCE (continued) )

l CONCLUSIONS NO EFFECT ON CHAPTER 15 SAFETY ANALYSES CDF REMAINS LESS THAN NRC SAFETY GOAL

UNDER LTOP, IMPOSED STRESSES LESS THAN THOSE REQUIRED FOR CRACK INITIATION NO UNDUE RISK TO PUBLIC HEALTH AND SAFETY ;

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l 950352 31 - 9/21/95 j l

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, ENFORCEMENT AND INVESTIGATION l

I COORDINATION STAFF i

REGION II i

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l REFERENQIPPACKAGE l

EA 95-180 I

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SEVERITY LEVEL III, $50,000 CP ;

FLORIDA POWER AND LIGHT COMPANY ;

ST. LUCIE NUCLEAR PLANT i

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REFERENCE. DOCUMENT CHECKUST REGION ll ENFORCEMENT AND INVESTIGATIONS COORDINATION STAFF ENCLOSURES i EA 95-180 FLORIDA POWER AND UGHT COMPANY

- ST. LUCIE NUCLEAR PLANT

[1] ' NRC Inspection Reports and other documentation of the case:

a. NRC Inspection Report 50-335/95-16 AND 50-389/95-16 l [2] Licensee reports (Submitted at the Conference):

I

a. Engineering Evaluation - Evaluation of PORV Unavailability on Plant Operations

b. LER 95-238, PORVs Inoperable Due to Personnel Error, l August 22,1995

! c. Topical Quality Assurance Report

[3] Applicable license conditions:

Technical Specification 3.4.13 Technical Specification 4.0.5 Article IWV-3000, Test Requirements i

[4] Applicable licensee procedures or extracts

a. General Maintenance Procedure No.1-M-0037

b. Second Ten-Year Inservice inspection Interval inservice Testing Program 1

[_] Copy of discrepant licensee documentation referred to in citations such as NCR, inspection record, or test results

[5] Enforcement Panel Questionnaire /PEC Briefin0 Paper

[8] Licensee PEC Presentation Materials

[] Referenced ORDERS or Confirmation of Action Letters (CALs)

! [] Other miscellaneous documents:

! ENFORCEMENT COORDINATOR: LINDA WATSON (404) 331-4192- - -- - -

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REFERENCE DOCUMENT CHECKLIST REGION 11 ENFORCEMENT AND lNVESTIGATIONS COORDINATION STAFF

! ENCLOSURES l EA 95-180 l FLORIDA POWER AND UGHT COMPANY ST. LUCIE NUCLEAR PLANT l [1] NRC Inspection Reports and other documentation of the case: ,

s. NRC inspection Report 50-335/95-16 AND 50-389/95-16 l [2] Licensee reports (Submitted at the Conference):

l l a. Engineering Evaluation - Evaluation of PORV Unavailability on Plant Operations

! b. LER 95-238, PORVs inoperable Due to Personnel Error, 1 August 22,1995

c. Topical Quality Assurance Report l [3] Applicable license conditions:

Technical Specification 3.4.13 Technical Specification 4.0.5 Article IWV-3000, Test Requirements l [4] Applicable licensee procedures or extracts l

a. General Maintenance Procedure No.1-M-0037 l b. Second Ten-Year Inservice inspection Interval inservice Test'ng Program j [_] Copy of discrepant licensee documentation referred to in citations such as

( NCR, inspection record, or test results 1

[5] Enforcement Panel Questionnaire /PEC Briefing Paper

[6] Licensee PEC Presentation Materials

[] Referenced ORDERS or Confirmation of Action Letters (CALs) l [] Other miscellaneous documents:

i

._ENgRCEMQCOOMNATOR:,_UNDA WATSON (404) 3314192 l

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REFERENCE PACKAGE Ril l EA NO: 95-180 i l

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SEP 26 '95 69:39R1 FPL -' UK.IE r,1X4 , P.2 l REVISION 5

( 01/6/95 i

FLORIDA POWER and LIGHT COMPANY l NUCLEAR ENERGY SERVICES 700 Universe Boulevard Juno Beach, Florida 33408 l

SECOND TEN-YEAR INSERVICE INSPECTION INTERVAL INSERVICE TESTING PROGRAM l TOR PUMPS AND VALVES

'N ST. LUCIE NUCLEAR PO UNITNC.g w,9 NRC DOCKE . 50-135 DOCUMENT NUMBER' '

' DML1 kruttion s 1

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ST. LUCIE P W IENS AND ROVALS:

PREPARED D DATE: if FI ~ IST ENGINEER e

APPROVED BY: <

DATE:

l P M sst s CODE SUPERVISOR '

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hk 5- REVISICN 5 01/6/95 INSERVICF, TEST::NG (IST) PROGRAM PLAN

> ST. LUCIE UNIT 1 1

1.0 INTRODUCTION

i Revision 5 of the St. Lucie Unit 1 ASME Inservice Inspection (IST) Program will be in effect through the end of the second i 120-month (10 year) interval unless revised and reissued for reasons other than the routine update required at the start i of the third interval per 10 CFR (g). The second

inspection interval is defined as foli neKLDA

Februsry 11, 1988 3, ruary 10, 1998 i

This document outlines the ram for St. Lucie Plant,

! Unit 1, based on the re j f the AsME Boiler and Pressure Vessel Code (t tion XI, 1983 Edition 1 through suancer 1983 Add =

ces in this document to "IWP" or "INV" cor a ons IMP and IWV, respectively, of t S n XI, 1983 Edition, unless otherwise noted ,

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The IST pro rpo the requirements of ASME/ ANSI j CH-1987, P Testing of a for Inservice Performance l l r at Pressure Relief Devices" for j i the to Nsafe relief valves. The use of i ASMB/ "987, P requirements as an alternative to j

ASME XI, 1983 Edition, Subsection INV-3510 ret was approved by the NRC Safety Evaluation of St.

! Lut :

it 1 Inservice Testing Program Relief Requests dated Ses r 27, 1994.

i The intervice testing requirements identified in this Plan were prepared to verify the operational readiness of pumps l and valves which have a specific function in sitigating the 4

d consequences of an accident or in bringing the reactor to a safe shutdown.

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In this regard, the general requirements of Paragraphs IMP-1100 and IWV-1100 foss the following basic scope document as

} it applies to ISI Class 1, 2, and 3. Specifically ocuponents

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I REFERENCE PACKAGE Ril EA NO: 95-180 Ril/EICS NO: 95-E 026 ENCLOSURE .NO: .-.-

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, NRC CLOSED PREDECISIONAL I 1

ENFORCEMENT CONFERENCE 1 1

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ST. LUCIE NUCLEAR PLANT 1 4

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SEPTEMBER 25,1995 i l

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NRC CLOSED PREDECISIONAL ENFORCEMENT CONFERENCE ST. LUCIE NUCLEAR PLANT SEPTEMBER 25,1995 IAH TITLE 1 Predecisional Enforcement Conference Agenda

2 Expected Attendees, Meeting Announcement 3 Opening Remarks and Introductions i 4 NRC Er.forcement Policy 5 Summary of the issues l 6 Statement of Concerns / Apparent Violations -

7 Inspection Report No. 50-335/398/95-16 l 1

8 Enforcement Pre-Panel Questionnaire 9 50.72 Report, LER 95-242 i i

10 Closing Remarks: l l

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PREDECISIONAL ENFORCEMENT CONFERENCE AGENDA ST. LUCIE SEPTEMBER 25,1995, AT 10:00 A.M.

NRC REGION ll OFFICE, ATLANTA, GEORGIA l

I. OPENING REMARKS AND INTRODUCTIONS )

S. Ebneter, Regional Administrator !

l

11. NRC ENFORCEMENT POLICY ;

B. Uryc, Director ;

Enforcement and Investigation Coordination Staff l l

Ill.

SUMMARY

OF THE ISSUES S. Ebneter, Regional Administrator i IV. STATEMENT OF CONCERNS / APPARENT VIOLATIONS E. Merschoff, Director Division of Reactor Projects V. LICENSEE PRESENTATION

W. Goldberg, President !

St. Lucie Nuclear Plant VI. BREAK / NRC CAUCUS :

VII. NRC FOLLOWUP QUESTIONS Vlil. CLOSING REMARKS S. Ebneter, Regional Administrator l

EXPECTED ATTENDEES I l

L

~ Licensee

~ J. Goldberg, President, Nuclear Division l

D. Sager, Vice President, St. Lucie Site l W. Bohlke, Vice President, Engineering L. Bladow, Nuclear Assurance Manager D. Denver, Site Engineering Manager l iL. Rogers, Systems and Component Engineering Manager t

J. Marchese, Maintenance Manager

~J. West, Operations Manger I

i NEC l

Stew Ebneter, Regional Administrator, Region.Il (Ril) ;

Ellis Merschoff, Director,. Division of Reactor Projects (DRP), Rll Al Gibson, Director, Division of Reactor Safety (DRS), Ril Bruno Uryc, Director, Enforcement and Investigation Coordination Staff-

-(EICS), Ril l Charles Casto, Chief, Engineering Branch, DRS, Ril j Kerry Landis, Chief,-Reactor. Projects Branch 2, DRP, Ril i

' Linda Watson, Senior Enforcement Specialist, EICS, Ril Carolyn Evans, Regional Counsel, Ril Richard Prevatte, Smior Resident inspector, St. Lucie, DRP, Ril L Robert' Schin, Project Engineer, Reactor Projects Section 2B, DRP, Ril Edwin Lea, Project Engineer, Reactor Projects Section 2B, DRP, Rll George Hopper, Operator Licensing Examiner, DRS, Ril 1

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l OPENING REMARKS AND INTRODUCTIONS (S. Ebneter)

Good morning. I am Stew Ebneter, Regional Administrator for the ,

Nuclear Regulatory Commission's Region 11 Office. This morning we

will conduct a predecisional enforcement conference between the NRC and St. Lucie which is CLOSED to public observation.

1 The agenda for the conference is shown in the viewgraph. Following i my brief opening remarks, Mr. Bruno Uryc, the Director of the Region ll l

Enforcement Staff, will discuss the Agency's Enforcement Policy. I 1 will then provide introductory remarks concerning my perspective on the events to be addressed today. Mr. Ellis Merschoff, Dir~ector of the Division of Reactor Projects, will then discuss the apparent violations.

You will then be given an opportunity to respond to the apparent violations. In this regard, I wish to reiterate to you that the decision to i

hold this conference does not mean that the NRC has determined that l l

violations have occurred or that enforcement action will be taken. This conference is an important step in arriving at that decision.

1

Following your presentation, I plan to take about a 10-rninute break so I that the NRC can briefly review what it has heard and determine if we i

have follow-up questions. Lastly, I will provide concluding remarks.

l At this point, I would like to have the NRC staff introduce themselves !

and then ask you to introduce your participants.

[lNTRODUCTIONS] i l

i Thank you.

l Mr. Uryc will now discuss the Agency's Enforcement Policy. l 1

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l NRC ENFORCEMENT POLICY l (B. Uryc)

NRC Enforcement Policy and Procedure I After an apparent violation is identified, it is assessed in accordance ,

i-

with the Commission's Enforcement Policy, which was recently revised and became effective on June 30,1995. The Enforcement Policy has been published as NUREG-1600.

The assessment of an apparent violation involves categorizing the ,

apparent violation into one of four severity levels based on safety and regulatory significance. For cases where there is a potential for escalated enforcement action, that is, where the severity level of the apparent violation is categorized at Severity Level 1, ll, or lil, a predecisional enforcam'ent conference is held.

There are three primary enforcement sanctions available to the NRC i 1

and they are Notices of Violation, civil penalties, and orders. Notices of Violation and civil penalties are issued based on identified violations. 1 Orders may be issued for violations, or, in the absence of a violation, l because of a significant public health or safety issue.

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~ .m This predecisional enforcement conference is essentially the last step of the inspection or investigation process before the staff makes its final enforcement decision.

The purpose of this conference is not to negotiate a sanction. Our purpose here today is to obtain information that will assist us in i

determining the appropriate enforcement action, such as: (1) a '

common understanding of the facts, root causes and missed i

opportunities associated with the violations, (2) a common understanding of corrective action taken or planned, and (3) a common understanding of the significance of issues and the need for lasting comprehensive action.

The apparent violations discussed at this conference are subject to further review and they may be subject to change prior to any resulting enforcement action, it is important to note that the decision to conduct this conference does not mean that NRC has determined that a violation has occurred or that enforcement action will be taken.

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. m I should also note at this time that statement of views or the expression of opinion made by the NRC staff at this conference, or the

/ack thereof, are not intended to represent final determinations or beliefs.

1 Following the conference, the Regional Administrator in conjunction with the NRC Office of Enforcement and other NRC Headquarters offices will reach an enforcement decision. This process should take i about four weeks to accomplish.

Predecisional enforcement conferences are normally closed to the public as is this conference. However, the Commission implemented a trial program in July 1992 to allow certain enforcement conferences to be open for public observation. [ July 10,1992 - Federal Register]

This trial program was recently extended for additional evaluation.

Finally, if the final enforcement action involves a proposed civil penalty or an order, the NRC will issue a press release 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the ,

enforcement action is issued.

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SUMMARY

OF THE ISSUES '

(S. Ebneter) {

Issues: St. Lucie Power Operated Relief Valves Inoperable ;

Power Operated Relief Valves V-1404 and V-1402 were i reassembled incorrectly and did not receive adequate post-maintenance testing. The PORVs were reinstalled in the Reactor Coolant System on November 5,199 , without adequate surveillance testing sufficient to provide reasonable assurance ,

that the valves would perform their intended function while in )!

service.

.MW ik j Defect:

ig\df '

As a result of the inadequate reassembly, post-maintenance ]

I testing, and surveillance testing the PORVs were inoperable from {

the. time they were installed in the RCS during the 1994 refueling ;

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outage until they were removed and reworked in August of 1995. l

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Plant data indicated that the valves could not have performed ]

their intended safety function. Plant data also indicated several i

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instances in which an operable PORV was required by TS, but l

i l was not available.

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f t Consequences:

Inoperable equipment was installed in the Reactor Coolant System I

which'was unable to perform intended safety function.

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-l STATEMENT OF CONCERNS / APPARENT VIOLATIONS (E. Merschoff) ;

F This is a Predecisional Enforcement Conference to discuss three apparent violations associated with PORV maintenance and operability. l The first and second apparent violations involve the adequacy' of the i procedure used to perform post-maintenance testing and the !

procedure used to perform surveillance testing. The third apparent violation addresses the operability of the PORV as required by TS.

We are concerned with those activities which resulted in the PORVs' being rendered inoperable. Encompassed in our concerns cre the facts that post-maintenance and surveillance testing should be of sufficient scope, and acceptance criteria of sufficient technical rigor, to ensure l component operability. We are also concerned with the consequences .

of operating outside TS limits, due to inoperable PORVs.

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l Our findings are documented in NRC Inspection Report 50- )

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335/389/95-16, which were transmitted to you on September 8, !

1995. At this conference, we are affording you the opportunity to i

i provide information relative to:

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--- Any errors the inspection reports i

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I --- The severity of the violations I

--- Any escalation or mitigation considerations i

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--- Any other application of the Enforcement Policy relevant to this issue. 1

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- m ISSUES TO BE DISCUSSED

1. 10 CFR 50, Appendix B, Criterion XI required, in part, that a test program be established to ensure that all testing required to demonstrate that components will perform satisfactorily in service be performed and that the program include proof tests prior to installation. FPL Topical Quality Assurance Report TOR 11.0, revision 4, " Test Control," stated, in part, that a test program shall be established to assure that testing required to.

demonstrate that structures, systems and components will perform satisfactorily in service and that the program shall include proof tests prior to installation.

In November,1994, valve maintenance was performed under a work package, which directed the rebuilding of Power Operated Relief Valves V-1404 and V-1402 per licensee procedure 1-M0037, Revision 6, " Power Operated Valve Relief Valve Maintenance. The post-maintenance testing was limited to a bubble test for seat leakage prior to reinstallation. The procedure contained a note explaining that lift set point testing was not required, as the valve was lifted based upon solenoid valve input.

The procedure did not require a verification that the valve would change state under pressure prict to installation.

NOTE: The apparent violations discussed in this predecisional enforcement conference are subject to further review and are subject to change prior to any resulting enforcement decision.

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ISSUES TO BE DISCUSSED

/

2. 10 CFR 50, Appendix B, Criterion XI required, in part, that a test program be established to ensure that adequate test '

instrumentation is available and used. FPL Topical Quality

_ Assurance Report TOR 11.0, revision 4, " Test Centrol," stated, in part, that a test program shall be established to assure that ;

testing required to demonstrate that structures, systems and components will perform satisfactorily in service is performed and '

that the program shall include operational tests. TOR 11.0 further states that test procedures shall incorporate requirements -

and acceptance limits in the applicable design and procurement documentation. ,

On November 25,1994, and on February 27,1995, operational ,

surveillance testing, performed under Administrative Procedure 1- l' 0010125A, revision 39, Data Sheet 24, did not employ adequate test instrumentation to detect the inoperability of both valves and j did not employ test acceptance limits derived from the valves' design documentation. Specifically, the use of acoustic data, as ;

opposed to system pressure reduction derived from valve l I capacity, to indicate valve position was insufficient to discern the difference between bypass flow through the PORV pilot valves l

.and actual changes in main valve position.

NOTE:- The apparent violations discussed in this predecisional enforcement conference are subject to further review and i are subject to change prior to any resulting enforcement i decision. 7 1'

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T ISSUES TO BE DISCUSSED i

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3. Technical Specification 3.4.13' requires, in part, that two Power Operated Relief Valves be operable in " Mode 4 when the temperature of any RCS cold leg is less than or equal to 304 F, l Mode 5 and Mode 6 when the head is on the reactor vessel; and I the RCS is not vented through a greater than 1.75 square inch' vent." TS 3.4.13 AS (c) required that, "with two inoperable l l- PORVs, at least one PORV be returned to an operable status or l that the RCS be completely depressurized and vented through a

minimum 1.75 square inch opening.within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ." ,

From November 22 through 27,1994, and from February 27. ,

through March 6,1995, St. Lucie Unit 1 was in conditions l requiring operable Power Operated Relief Valves but no operable i relief valves were in service. The inoperability of the Power i Operated Relief. Valves resulted from a combination of personnel i

error during maintenance and inadequate post-maintenance and i surveillance testing. ;

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NOTE: The apparent violations discussed in this predecisional enforcement conference are subject to further review and are subject to change prior to any resulting enforcement decision. -

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CLOSING REMARKS (S. Ebneter) !

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{ in closing this predecisional enforcement conference, I remind the Licensee of two things:-

First, the apparent violations discussed at this predecisional enforcement

, conference are subject to further review and may be subject to change prior to any resulting enforcement action.

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Second, the statements of views or expressions of opinion made by NRC employees at this predecisional enforcement conference, or the lack thereof, are not. intended to represent final agency determinations or beliefs.

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ENFORCEMENT PRE-PANEL -

ST. LUCIE w :. .

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8/23/95 -

2:00 cm F ,.

A RRJDGE 6 : (301) 415 - 7605 s _

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SENIOR DRP MANAGEMENT REPRESENTATIVE - ELLIS MERSCEOFF SECTION CHIEF - KERRY LANDIS

[EICS REPRESENTATIVE - LINDA WATSON REGIONAL COUNSEL - BRAD FEWELL SENIOR PROGRAM _ OFFICE REPRESENTK?IVE - JAN NORRIS DRS MANAGEMENT REPRESENTATIVE - AL GIBSON_ l l

DRS REPRESENTATIVE -

CHUCK CASTO i 1

l DRS REPRESENTATIVE - TON PEEBLES l

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i l . ESCALATED ENFORCEMENT

PANEL QUESTIONNAIRE l INFORMATION REQUIRED TO BE AVAILABLE FOR ENFORCEMENT PRE-PANEL i

. PREPARED BY: Mark S. Miller i

i NOTE: The Section Chief is responsible for preparation of this questionnaire and

its distribution to attendees prior to an Enforcement Panel. (This information

.will be used by EICS to prepare the enforcement letter and Notice, as well as the

transmittal memo to the Office of Enforcement explaining and justifying the l Region's proposed escalated enforcement action.)

1. Facility: St. Lucie Unit: 1

Docket Nos: 50-335 License Nos: DPR-67 i

Inspection Dates: July 2 - Auaust 28. 1995 Lead Inspector: Richard L. Prevatte i

j -2. Check appropriate boxes:

, [X] A Notice of Violation (without "boilerplate") which includes the j- recommended severity level for the violation is enclosed.

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! [] This Notice has been reviewed by the Branch Chief or Division Director and each violation includes the appropriate level of l specificity as to how and when the requirement was violated.

[] Copies of applicable Technical Specifications or license conditions j cited in the Notice are enclosed.

3. Identify the reference to the Enforcement Policy Supplement (s) that best !

fits the violation (s) (e.g., Supplement I.C.2) !

i I B.l* 1.C. 2. (b f '

l I.C.6' l i As stated in new enforcement policy 1

[ 4. What is the apparent root cause of the _ violation or problem?

) Personnel error in maintenance combined with a failure to perform adeauate

! post-maintenance testina and a failure to provide adeouate acceptance

! criteria for surveillance testina of PORVs.

5. State the message that should be given to the licensee (and industry) through this enforcement action. i l

Post-maintenance and surveillance testino should be of sufficient scope. l and acceptance criteria of sufficient technical ricor, to ensure component ;

operability.

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6. Factual information related to the following civil penalty escalation or mitigation factors (see attached matrix and {

10 CFR Part 2, Appendix C, Section VI.B.2.): '

a. IDENTIFICATION: (Who identified the violation? What were the facts ,

and circumstances related to the discovery of the violation? Was it i self-disclosing? Was it identified as a result of a generic !

notification?) l l

The condition was identified by control room operators when it was noted that operation of the PORVs durina surveillance testino did not result in expected chances in RCS and auench tank parameters.

The identification of the misalioned main disk ouide was identified by the licensee when the valves were removed from the system for inspection. Failure to Derform adeauate oost-maintenance and surveillance testina was identified by the licensee in course of a ricorous root catise effort,

b. CORRECTIVE ACTION: Although we expect to learn more information regarding corrective action at the enforcement conference, describe !

preliminary information obtained during the inspection and exit ;

interview.

The discrepant conditions were corrected by the licensee. See attachment.

What were the immediate corrective actions taken upon discovery of the violacion, the development and implementation of long-term corrective action and the timeliness of corrective actions?

The PORVs were declared inoperable. the unit was placed in a condition not reauirina their operability and then was cooled down and depressurized. Immediate corrective actions were approDriate and the subject failures were investicated aaaressively. Short term corrective actions included properly assemblina and testina the subject PORVs. addina a OC hold ooint to the PORV maintenance Drocedure to insure oroDer disc auide installation and to reauire a bench lift test under air oressure. Additionally, the licensee chanced the acceptance criteria in the PORV surveillance testina procedure reouirina documentation of RCS pressure chanae. PORV tail _oice temperature chance, and auench tank pressure, level and temperature chanaes.

What was the degree of licensee initiative to address the violation and the adequacy of root cause analysis?

Root cause determination was well-coordinated, timely, and comprehensive.

c. LICENSEE PERFORMANCE: This factor takes into account the last two years or the period within the last two inspections, whichever is longer.

List past violations that may be related to the current violation (include specific requirement cited and the date issued):

No violations involvino the adeauacy of post-maintenance testina have been issued in the last two years.

Identify the applicable SALP category, the rating for this category and the overall rating for the last two SALP periods, as well as any trend indicated:

The subiect functional area is Maintenance and Surveillance, which was most recently rated SALP 1 (January 94). The previous SALP Deriod, the area was rated a SALP 1.

d. PRIOR OPPORTUNITY TO IDENTIFY: Were there opportunities for the licensee to discover the violation sooner such as through normal surveillances, audits, QA activities, specific NRC or industry notification, or reports by employees?

Proper post-maintenance testina would have been effective in identifyino the inoperable status of the PORVs. Additionally. '

adeouate surveillance testina would have detected the inoperability.

The orior opportunities to identifiv the subiect conditions is at

_the heart of this enforcement action,

e. MULTIPLE OCCURRENCES: Were there multiple examples of the violation identified during this inspection? If there were, identify t' number of examples and briefly describe each one.

Multiple occurrences have not been identified, except to the extent that the noted test failures applied to two PORVs.

f. DURATION: How long did the violation exist?

Since November, 1994.

ADDITIONAL COMMENTS / NOTES: See attached description of the sub.iect events. Note 1

that, in addition to violations relatina to cost-maintenance and surveillance testina adecuacy, a violation of TS 3.4.13, reoardina PORV operability for LTOP, j existed. The information in this document is current as of Auaust 22, 1995.

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ESCALATION AND MITICATION FACTORS (57 FR 5791, February 18, 1992) i IDENTIFICATION rmerCTIVE LICENSEE PRIOR 80LTIPLE DURATION ACTION PERFORMANCE OPPORT12flTY TO m erNCES IDENTIFY

+/ 50% +/- 50% +/ 100% + 100% + 100% + 100%

Licensee Timeliness of Current Licensee should Multiple used for identified (M) corrective violation is an have identified examles of significant (To be applied action (M) isolated violation violation reguls y even if (Did NRC have failure that is sooner as a identified message to licensee could to intervene to inconsistent result of prior during have licensee. (E) accog lish' with licensee's opportunities inspection identified the satisfactory good such as audits (only for SL 1, l violation short-term or performance (M) (E) 11 or !!!

sooner) remedial action (Ell violations) (E)

NRC identified Prom tly Violation is Opportunities OTHER CONSIDERATIONS (E) developed reflective of available to i schedule for licensee's poor discover 1. Legal aspects and potentfal long term or declining violation such litigation risks corrective performance (E) as through action (M) prior 2. Negligence, careless dis

notification regard, willfulness and (E) management involvement

-; Self- Degree of Prior Ease of earlier disclosing 3. Economic, personal or licensee performance and discovery (E) corporate gain (M 25% if initiative (M) effectiveness there was [To develop of previous initiative to 4. Any other regulatory frame.-

corrective corrective work factors that need to be~

identify root actions and action for conaidered: pending action-cause) root cause] similar with regard to licensingi violations coussission meeting, or press Licensee Adequacy of the conference.

SALP - Period of time identified as root cause Consider: between a result of analysis for

5. What is the intended message SALP 1 - (M) violation and for the licensee and the-generic the violation SALP 2 - (0) notification incbstry?

notification (M) SALP 3 - (E) received by (M) L icensee (E) ......--.- NOTES --------.

4 comrehensive Prior similarity corrective enforcement between the action to history violation and i

prevent including notification occurrence of escalated and (E) similar non escalated j violation (M) enforcement Isenediate Levs! of corrective managesaent action not review the taken to notification restore safety received (E) and compliance (E) mm -

SAFETT SIGNIFICANCE: In determining the safety significance of a violation in conjunction with the enforcement proces t the evaluation should conalder the technical safety significance of the violation as well as the reptatory significance. Consideration should be given to the matter as a whole in light i

of the circunstances surrounding the violation. There may be cases in which the technical safety significarce of the matter is low while the process control f ailure(s) may be significant, and, therefore, the severity level determination should be based more on the process control f ailure(s) than on the technical safety issue. The following factors should also be considered: 1) Did the violation actually or potentially impact p@lle health and safety? 2) What was the root cause of the violation?

3) Is the violation an isolated incident or is it indicative of a programmatic breakdown? 4) Was management aware of or involved in the violation? 5) Did the violation involve willfulness?

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f. Prooosed Violation A 10 CFR 50, Appendix B, Criterion XI states, in part, that a test program shall be established to ensure that all testing required to demonstrate that components will perform satisfactorily in service and that the program shall include proof tests prior to I

installation and operational testing. The criterion further states that test procedures shall

, include provisions to assurbk that adequate test instmmentation is available and used. j

~ FPL Topical Quality Assurance Report TQR 11.0, revision 4, " Test Control," states, in 1 part, that a test program shall be established to assure that testing required to demonstrate that stmetures, systems and components will perform satisfactorily in service and that the i

program shall include proof tests prior to installation, operational tests, and retest after repair. TQR 11.0 further states that test procedures shall incorporate requirements and acceptance limits in the applicable design and procurement documentation.

9 Contrary to the above, on November 5,1994, and on November 6,1994, Power U

Operated Relief Valves V-1404 and V-1402, respectively, were installed in the Unit 1 Reactor Coolant System, placed into operation on November 22, and relied upon to be

! operable for approximately nine months without adequate post-maintenance and i ~surviellance testing sufficient to provide reasonable assurance that the valves would l perfonn satisfactorily in service. As a result of these failures, both Unit 1 PORVs were i inoperable from the time of their reinstallation after maintenance until August 11,1995,

) when the reactor coolant system was depressurized and vented. Examples are:

d

1. No post-maintenance bench test was performed to ensure that the valves' main discs would change state in a pressurized environment.

j 2. On November 25, 1994, and on February 27, 1995, operational surveillance j testing,_ performed under Administrative Procedure 1-0010125A, revision 39c j

Data Sheet 24, did not employ adequate test instrumentation to detect t -

j moperability of both valves and did not employ test acceptance limits derived i

from the the valves' design documentation. Specifically, the use of acoustic data, as opposed to system pressure reduction derived from valve capacity, to indim* ;

valve position was insufficient to discern the difference between bypass flow i

} through the PORV pilot valves and actual changes in main valve position.

'Ihis is a Severity I.evel II violation (supplement I).

j Proposed Violation B >

Technimi Specification 3.4.13 requires, in part, that two Power Operated Relief Valves L be operable in " Mode 4 when the temperature of any RCS cold leg is less than or equal

to 304*F, Mode 5 and Mode 6 when the head is on the reactor vessel; and the RCS is not vented through a gitater than 1.75 square inch vent."

k Contrary to the above, from November 22 through 27, 1994, and from February 27 through March 6,1995, St. Lucie Unit I was in conditions requiring operable Power

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! Operated Relief Valves but no operable mieif valves in service. The inoperability of the

~ Power Operated Relief Valves resulted from a combination of personnel error during maintenance and inadequate post-maintenance and surveillance testing.

This is a Severity level II violation (Supplement I),

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i St. Lucie Unit 1 PORV Inoperability l Ooerational Events l On August 9, the licensee performed ASME Section XI strob testing on V-1402 and V-1404 (Unit 1 PORVs) per AP 1-0010125A, revision 39, "Surveilhnee Data Sheets," Data Sheet 24.

The test was performed with RCS pressure controlled at 257 to 268 psig. The methodology of l- the test involved placing the PORV. control switches in " override," (which ensured that the valves would not open) removing High Pressurizer Pressure bistables from the RPS cabinets (which would send an "open" signal to the PORVs which would be blocked by the status of the control switches), and then, for each PORV, placing the control switch in " normal," which would send the open signal to the PORV. The stroke time for each PORV was to be measured from the time the control switch was taken to " normal" to the time that acoustic monitors indicated that the subject valve had opened Once a valve stroke time had been obtained, the subject valve's control switch was to be retumed to override to close the valve.

The results of the subject testing indicated that the valves did not stroke open. No acoustic signal was received in the control room. The licensee then returned the valves to service while questions of acoustic monitoring calibration and threshold levels were considered. The test was reperformed approximately one hour later with temporary acoustic monitors and the resulting acoustic signals indicated that both valves stroked in under one second. LTOP was placed back in service, but Operations personnel began to question the validity of the test results, as no changes were noted in either RCS or Quench Tank parameters. While evaluations were bein;;

conducted, the unit was taken to Mode 4. At 7:03 p.m. on August 9, the valves were retested and found to be inoperable based, in part, on observations of RCS and Quench Tank parameters.

Each was declared out of service and the licensee entered TS 3.4.13 Action (c), which required depressurization of the RCS and venting through a 1.75 square inch or greater opening within

'24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

- At 9:37 p.m., operators were directed by management to perform a cooldown of the RCS.

When placing the SDC system in service, the SDC discharge relief valve lifted and would not rescat without securing the SDC pumps to reduce SDC system pressure at the relief valve. This j issue will be discussed in a separate enforcement action; however, the inopembility of the SDC j system (due to the relief valve issue) precluded the licensee from cooling down and depressurizing Unit 1. Consequently, the licensee entered TS 3.0.3 at 10:45 a.m. on August 10 and began a heatup to greater than 304T, a plant condition for which TS 3.4.13 did not i apply. 305T was achieved at 11:53 a.m.

'Ihe SDC system was returned to service on August 11. A cooldown was corr.menced at 6:25 i a.m. the same day. The licensee made plans to re-enter TS 3.4.13 AS (c) during the cooldown, l and to create the required vent path by removing the bonnet of PCV-1100F, one of two pressurizer spray valves, which would create the required vent path to RCS cold leg IBl. 'Ihe subject AS was entered at 7:15 a.m. on August 11 and exited at 8:40 p.m. the same day, when the system was vented. l PORV Operation

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h He PORV design in question is a Dresser Industries Model 31533VX-30. The valve is a 2.5"

inlet by 4" outlet pilot operated valve with a relief capacity of 153,000 lbm/hr. The internals

of the valves are displayed in Figure 1.

i The valve, as installed in Unit 1, is actuated by a solenoid valve, which acts on the pilot valve ;

operating lever to open the pilot valve. The open pilot valve creates a vent path from below the main disc of the main valve, through holes machined in the main disc guide, to the quench tank. .

1. The reduction in pressure below the main valve main disc allows the disc to move open (down, I j in Figure 1) under the force of system pressure acting on the main disc.

( When pressure has been reduced below the applicable rescat pressure (depending upon PORV mode - normd LTOP low range, or LTOP high range), the solenoid valve is dagkod,

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which closes the pilot valve and isolates the vent path from below the main valve disc to the quench tank. Once the vent path is isolated, pressure builds up below the main valve disc as i

system pressure is admitted to the space through an orifice in the main valve retainer plug.

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j. When pressure has been built up below the main valve disc in this manner, the disc is moved i into a closed position under pressure aided by spring force.

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, Diarnostic Mahitenance i

The subject PORVs were removed on August 12 and placed on a test bench for lift tests to be 4

conducted under air pressure. Both valves were tested at a number of pressures within the

{ LTOP range and were found to be inoperable. Disassembly and inspection revealed that the j main disc guide was installed upside down, with the holes (required to vent the space below the j main disc) located at the upper extreme of the main disc cavity such that proper venting below i the main valve disc could not take place.

As a function of diagnosing the root cause, the licensee reversed the main disc guide orientations j 4

(to the proper orientation) and retested the valves under air pressure. Both valves tested satisfactorily. The licensee'also sent a spare valve to Wylie Laboroatories for testing under water and steam pressure, as these condidons could not be established at the site. The spare

- PORV was tested under water and steam with the main disc guide misoriented (the as-found condition of the Unit 1 PORVs) at pressures ranging from LTOP pressures to NOP ranges. De

! PORV failed to open under any condition with the main disc guide misoriented. . Additionally, <

l- it was found that:

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Under water pressure at 335 and 450 psig,10-15 psig was developed at the i discharge of the pilot valve, indiming that some leakage around the main disc guide was possible, but not enough to provide venting sufficient to open the PORV.

e Under steam pressure from 50 psig to 450 psig in 50 psig increments, 20-60 psig I was developed at the pilot valve discharge.

o Under steam pressure at 2400 psig,1500-1800 psig was developed at the pilot - .

valve discharge. '

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The pressures and media flow detected at the pilot valve discharge indicated that acoustic data may be received during PORV testing without being indicative of a PORV changing state.

Maintenance '

The subject PORVs were last reworked in November,1994, as part of the Unit I refueling .

outage. The rework was conducted by employees of Funnanite, which were used by the licensee for outage-nlated valve work. The PORVs were each worked by the same two workers. The work package which directed the rebuild invoked the licensee's procedure 1-M-0037, revision 6, " Power Operated Valve Relief Valve Maintenance." The licensee determined that step 9.8 " reassembly of Main Valve," step 7, which directs the installation of the main disc guide, did not include a QC hold point to verify proper installation. It was noted that this is the only component which can be installed improperly and result in undetected inoperability. The procedure was revised to include a QC Hold Point prior to the valves' reassembly.

The inspector questioned the licensee as to post-maintenance testing requirements as applied to the PORVs. The licensee stated that post-maintenance testing was limited to a bubble test for seat leakage prior to reinstallation. The inspector noted that 1-M-0037 only required the noted bubble test as post-maintenance testing and, in fact, contained a note explaining that lift set point testing was not required, as the valve was lifted based upon solenoid valve input. The procedure r did not require a proof that the valve would change state under presseure prior to installation, l but did include a check for main disc mechanical freedom.

In discussing post-maintenance testing with the licensee, it was stated that, while no documented lift test existed, air lifts were typically performed as a function of preparing for seat leakage :

tests. It was explained that, upon initial reassembly, the PORVs mrely, if ever, satisfied seat l leakage criteria due to relative misalignment between the main valve disc and its seat. As a result, the licensee stated that lifts under air pressuit were performed as a matter of course to allow the main disc to orient itself properly against its seat. The inspector noted that the governing procedure included a note to this affect, but no evidence existed to indicate that lifts had occurred on the test cench.

The licensee stated that, in discussions with the Furmanite Supervisor who oversaw the rebuilding of the PORVs during the 1994 outage, the Supervisor stated that he recalled at least 6 lifts under air pressure per PORV in attempts to obtam satisfactory seat leakage tests. No documentation existed to validate the claim. The licensee stated that, when testing of the spare PORV at Wylie was complete, and the PORV was returned to the site, Furmanite was going to be allowed to rebuild the valve and demonstrate that lifts could be achieved with the main disc guide installed backwards. The inspector diw=~i the plausibility of such lifts with the valve ,

vendor representative on site, who stated that, in principle, such liAs were possible it sufficient l

gaps existed between the main disc guide and the gasket below the guide. The results of the test i are pending.

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'Ihe inspector concluded that post-maintenance testing, described in 1-M-0037, was inad--'a to verify that maintenance had been satisfactorily performed on the PORVs. As described below, surveillanea testmg was performed on the PORVs during unit heatup and repressunzation

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I l j following the Unit 1 outage. However, the inspector concluded that insuffi 1 performed on the PORVs, prior to installation, to obtain a reasonable assura

would perform surveillance test. satisfactorily in the LTOP conditions which would exist prioj

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i The inspector discussed the issue of post-maintenance testing with Opemtions

confinned that Operations had accepted the subject PORVs from maintenance w

assumption installation. that they had been properly tested and, as such,' considenu them i

t Maintenance personnel, as in-situ surveillance testing w i test to I&C of perform the valve overhaul. The inspector reviewed PWOs 63/8104 and 63/8105 ,

which directed PORV solenoid valve inspection and testing on V-1402 and V-1404,

respectively. Included in the task description of each PWO was a requirem Dept to perform valve stroke time verification test per QI 11-4/AP l-0010125A D .

! referenced procedure and data sheet were the same as that described below fo

time testing. The inspector concluded that the Maintenance and Operations D under completely different impressions of the status of the valves.

i

i. ~ Surveillance Testine -

J l

The inspector questioned the licensee as to whether any in-sian testing had bee the PORVs since their installation during the 1994 outage. The licensee stated thatl been performed; one on November 25, 1994, with RCS pressure at 245 psia, and one on February 27, 1995, with RCS pressure at 1750 psia. i Both tests were documented as !

satisfactory. The satisfactory results were, by procedure, based upon acoustic to system parameter changes (e.g. RCS pressure, quench tank conditions). As stated above, resuhs from testing at Wylie indicated that sufficient flow could be developed thr around a misinstalled main disc guide (and then out an open pilot valve) to provide acous;;

without actual main valve movement.

The inspector concluded that the acceptance criteria provided for verifying PO in OP l-0010125A November 25,1994, was insufficient to demonstrate valve operability in that tests pe PORVs. and February 27, 1995, did nct detect the inoperability of the subject PORV Onemhility

'Ihe iapr reviewed the licensee's activities with regard to root cause determination e for th subject PORV conditions. In particular, the ingetor noted the following:

e Bench testing of both PORVs, once removed from the system and prior to individual valve dimembly, indicated that the valves would not lift under air pressure at any process pressure from the LTOP range to the NOP range.

e Disassembly of each PORV resulted in the discovery ofincorrectly installed main valve disc guides.

1 Upon correction of main valve disc guide orientation alone (i.e. no other piece part changes or replacement) for each PORV, bench testing under air pressure resulted in satisfactory lifts.

Wylie I.aboratory testing of a spare PORV, under water and steam, under pressure conditions ranging from below LTOP setpoints to above NOP, indicated that no lift was possible with the main valve & c guide installed backwards.

! As a result, the inspector concluded that the PORVs were inoperable from the time they were l installed in the RCS during the 1994 refueling outage until they were removed and reworked in August,1995.

Imnact on Unit 2 1

l Unit 2 is not susceptible to the same failure, as Unit 2 employs Garrett / Crosby PORVs, which are of distinctly different design. Additionally, the Unit 2 PORVs provide direct main valve L !

position indication, provided by a indexing rod attached to the main valve disc which activates !

l a reed switch. The inspector reviewed AP 2-0010125A, revision 43, " Surveillance Data l Sheets," Data Sheet 24, which directed surveillance testing for Unit 2 PORVs. The inspector found that the procedure directed that stroke time be based upon indicated valve position change, as opposed to acoustic data. [ FIND OUT IF UNIT.2 PROCEDURE HAS BEEN REVVED TOO)

Imoact on Unit 1 l !

The St. Lucie Unit I design employs two PORVs, which provide overpressure protection both during normal operation and for LTOP concerns. Additionally, the PORVs are employed in '

EOPs for once through cooling in the event of a loss of other core heat removal options. During power operations, PORVs are designed to open only in the event of a high pressure reactor trip, j and are sized to allow the unit to suffer a loss of load trip from full power without lifting a l pressurizer code safety valve. Accident analyses do not credit PORV operation. During low-mode conditions, the PORVs operate on one of two selectable LTOP setpoints, depending upon cold leg temperature and whether a heatup or cooldown is in progress.

I Following the Unit I refueling, the unit was filled solid on November 22. The RCS was L pressurized and in a condition requiring LTOP from November 22 through November 27. The unit was subsequently at NOP until a Short Notice Outage (SNO) in February,1995. During the SNO, the unit was in conditions requiring LTOP protection from February 27 through March

6. Notably, on March 4,1994, Unit 1 experienced a loss of shutdown cooling event with the unit in a solid water condition. The condition was corrected by operators, but not before RCS pressure had exceeded the LTOP anticipatory alarm setpoint. No LTOP lift of PORVs was demanded or experienced (peak pressure was 343 psia, LTOP setpoint at the time was 350 psia).

On July 11,1994, Unit 1 suffered a high pressure trip (see IR 95-14) which, according to the

licensee at the time of the trip, included a lifting of both PORVs. The conclusion was supported I

at the time by the inherent design of the system, the fact that acoustic data indicated that the w

. . _ - - . - . - - - . - - . . - . - - - - . . - - . ~ . - _ - . - . . - - . - . - . - -

7 i .

l PORVs lifted, and noted inemases in Quench Tank temperature. The licensee is now doubtful i f that the PORVs lifted during the trip, based upon a review of data (which suggested that i 4 pressure drifted above the PORV serpoint, as opposed to plateauing) and of analyses which l

showed that the post-trip loss of heat source acts, in conjuncdon with steam milefs to limit i pressure incmases.

, J i

l As regards once through cooling functions of the PORVs, the St. Lucie IPE includes PORV use !

j in early post-accident heat removal for reactor trips, loss of pressure control events, loss of l offsite power events, main steam line break accidents, and steam genemtor tube ruptures.

j The licensee performed a PRA analysis which quantified the change in CDF for common-mode

] failures in PORVs. 'Ihe licensee d etermined that the Unit 1 CDF had incmased by an j approximate factor of 3 for the periot of inoperability of the subject PORVs.

With regard to LTOP concerns,' the licensee analyzed the linpact of a loss of LTOP PORV function for the energy and mass addition events in the original LTOP design basis. The licensee determined that, based upon current levels of Unit I fluence, the maiximum allowable

, vessel stress would not be exceeded for any of the previously-analyzed LTOP events. Pmssure relief by pressurizer code safety valves, or shutdown cooling relief valves (depending upon the

) event considered), were found to be sufficient to limit peak pressures to below maximum j allowable values.

I i 3

l l i i e

I t

i Figure 1 1

J PanialSection View ofPower OperatedReuef Valve (PORV)

OURET

\ l

\ j ,

x MAIN DISC g

MAJN DISC

,' I!g MAJN DISC '

GUIDE u P N_ ; ( ,

MJ s ,

4 .. ,,,,, - fs A .. -

. N ,I 5

. M.

i, :l:

s ,I..._

RETAHER nuo

! ! \N GASKET fee a m uunone GUlOE s (pepc or.e leho ) -, ,,

~

.hr-- --dL

= a -

N - -

/

9%P r} *i omes.fau me a mai mn a.weam wi

4

. 6 i

l l

l i

REFERENCE PACKAGE Ril EA NO: 95-180 Ril/ElCS NO: 95-E-026 j ENCLOSURE NO:

b

!/ _ ..

i

! l j ll,4 Ppt!CA8!LITY t

, SURVEILLANCE REQUIREMENTS l J 4.0.1 Surveillance Requirements shall be applicable during the OPERATIONAL s 'iODES or other conditions specified for individual Limiting Conditions for.

i Operation unless otherwise stated in an individual Surveillance Requirement.

4.0.2 Each Surveillance Requirement shall' be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25%

j of the specified surveillance interval.

1 'k

t.0.3 Failure to perform a Surveillance Requirement within the allowed i .urveillance interval, defined by Specification 4.0.2, shall constitute t

noncompliance with the OPERA 8ILITY requirements for a Limiting Condition

for Operation. The time limits of the ACTION requirements are applicable -

! at the time it is identified that a Surveillance Raquirement has not been i performed. The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .to j permit the completion of the surveillance when the allowable outage time j limits of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Surveillance l Requirements do not have to be performed on inoperable equipment.

4.0.4 Entry into an OPERATIONAL MODE or other specified applicability a condition shall not be made unless the Surveillance Requirement (s) associated i cith the Limiting Condition for Operation have been performed within the

'tated surveillance interval or as otherwise specified. This provision 1

aall not prevent passage through or to OPERATIONAL M00ES as required to comply with ACTION requirements.

4.0.5 Surveillance Requirements for inservice inspection and testing of ASME i Code Class 1, 2 and 3 components shall be applicable as .follows:

i

a. Inservice inspection of ASME Code class 1, 2 and 3 components and inservice testing ASME Code Class 1, 2 and 3 poses and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50. Section 50.S5a(g), except where specific writ'en .

relief has been granted by the Cossaission pursuant to 10 CFR 50,

' Section50.55a(g)(6)(1).

3 b .' Surveillance intervals specified in Section XI of the ASME Soiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:

2

hw -

ST. LUCIE - UNIT I 3/4 0-2 Amendment No. 25.48.f8.pg,108

APPLICABILITY SURVEILLANCE REOUIREMENTS (Continued) 4.0.5 (Continued)

ASME Boiler and Pressure Vessel Code and applicable Required frequencies Addenda terminology for for performing inservice inservice inspection and inspection and testing testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 caths At least once per 92 days Semiannually or every 6 months At least once per 184 days Yearly or annually At least once per 366 days !

c. The provisionc of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities.

d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements. l

e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

ST. LUCIE - UNIT 1 3/4 0-3 Amendment No. g,118

l l

I l

EACT00 COOLANT tv5Try POWED OPEcaTED DELfEF valves LIMITING CONDITfrN fit. ODEpATION 4

3.4.13 Two power operated relief valves (PORVs) snall be OPERABLE. with their setpoints selected t the low temperature mode of operation as follows: l

a. A setpoint of less than or equal to 350 psia shall be selected:

1. During cooldown wnen the temperature of any RCS cold leg is less than or equal to 215'F and

2. During heatus and isothermal conditions when the temperature of any RC! cold leg ts less than or equal to 193*F.

I

, b. A setootnt of less than or equal to 530 psia shall be selectM:

' 1. Durtn; caeldown when the temperature of any RCS cald leg is greater tian 215*F and less than or equal to 281*F. a

2. During heatus and isothermal conditions when the temocrature of any RC5 cold leg is greater than or equal to 193*F ano less tsan or equal to 304*F.

ApptfCABILITY: !

M00f 4 when the temoerature of any RCS cold leg ts less than l or toual to 304*F, ' # L 5, and MODE 6 when the heaa is on the reactor vessel; and the RCS is not verted through greater than a 1.75 square inch vent. l l

ACTION:

a. With one 90.*V inoperable in MODE 4. restore the inoperable PORV to OPERABLE status within 7 days; or depressurtre and vent the RCS througn greater than a 1.75 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b. With one PORY inoperable in MODES 5 or 6. either (1) restore the D inoperable PORY to OPERA 8LE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or (2) complete i

i depressurization and venting of the RCS through greater than a 1.75 !

serare inch went within a total of 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months , i c~ 7 With both PORVs inoperable, restore at least one PORY to operable

' status or complete depressurtzation and venting of the RCS througn greater thar a 1.75 square inch vent within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d. With the Ra vented sir ACTIONS 4, b. or c. vertfy the vent pathway at least once per 31 days when the pathway is provided by a valve (s) inat is locked, sesled. or otherwise secured in the open position; otherwise, verify the vent pathway every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

e.

In the evert either the PORVs or the RCS vent (s) are used to attigate an .45 pressure transient, a Special Report shall be prepared and subettted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating tan transient, the effect of the PORVs or RC5 vent (s) on

- the traasten , and any corrective action necessary to prevent recurrence.

f. The provisions of specificatten 3.c.4 are not appitcable. --*- '

sumtrit1 AMEF RFoulDEMENTS 4.4.13 Each PORY sha*1 be demonstrated OPERABLE by:

a. Verifying l

heers; andthejPORY isolation valve is open at least once per 72 s

b. Performance d a CHAlelEL FUNCTION TEST, but excluding valve '

8 operation. J! least once per 31 days; and b

c. Performance of a CHAletti. CALIBRATION at least once per 18 months.

,. m. ... ,....

g ST. LUCIE - UNIT 1 3/4 4-59 Amendment 46.9hfe4.132 l

-~

f %%

i i

l APPLICABILITY l- SURVEILLANCE REQUIREMENTS i 4.0.1 Surveillance Requirements shall be applicable during the OPERATIONAL i MODES or other conditions specified for individual Limiting Conditions for

Operation unless otherwise stated in an individual Surveillance Requirement.

4.0.2 Each Surveillance Requirement shall be performed within the specified I surveillance interval with a maximum allowable extension not to exceed 255 l of the specified surveillance interval. l l

l l 4.0.3 Failure to perform a Surveillance Requirement within the allowed

. surveillance interval, defined by Specification 4.0.2, shall constitute j" noncompliance with the OPERA 8ILITY requirements for a Limiting Condition

for Operation. The time Ifmits of the ACTION requircunts are applicable i at the time it is identified that a Surveillance Requirement has not been '

! pe rfo rmed. The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .to j permit the completion of the surveillance when the allowable outage time i limits of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Surveillance l Requirements do not have to be performed on inoperable equipment.

4.0.4 Entry into an OPERATIONAL MODE or other specified applicability condition shall not be made unless the Surveillance Requirement (s) associated 1 l with the Limiting Condition for Operation have been performed within the i stated surveillance interval or as otherwise specified. This provision l' shall not prevent passage through or to OPERATIONAL M00E5 as required to

comply with ACTION requirements. -

4.0.5 Surveillance Requirements for inservice inspection and testing of ASME i l Code Class 1, 2 and 3 components shall be applicable as .follows:

l a. Inservice inspection of ASME Code Class 1. 2 and 3 components and

! inservice testing ASME Code Class 1, 2 and 3 pumme and valves

! shall be performed in accordance with Section XI'of the ASME j Soiler and Pres'sure Vessel Code and applicable Addenda as required

! by 10 CFR 50. Section 50.55a(g). except where specific written .

! reitef has been granted by the Commission pursuant to 10 CFR 50 Section 50.55a(g)(6)(1).

l

~

b. Surveillance intervals specified in Section XI of the ASME to11er and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASIE leiler and Pressure Vessel Code and appifcable Addenda shall be appitcable as follows in these Technical Specifications:

O b*

' ' ~ ~

% ~

1 TABLE 5.5-5 PRESSURIZER POWER-OPEllATED STt1EF VALVI PARAMETERS Tag Numbers V-1404, v-1402 )

- 1 Design Pressure, pots 2485 j

Des 15n Temperature. P 675 Saturated Staan, Fluid 0.1% (wt) Beric Acid (

Capacity, ib/hr 153.000 TYPE Solenoid Operated Voltage 125 vde 2 1/2".- 2500 lb by I

! Siza 4" - 300 lb Ser. Pressura, psig 2385

' Body Material ASTM-A-182 F-316 l l Norsia Material ASTH-A-182 F-316 (Sta111ted) kd [

r i

Dise conal 718

Code A Code for Pumps g Valvas for Nuclear J v.,U.

.. Q, C , cp er Class I, How, Igg

[p l

(y raft j

i l

l i

5 5-33 0,.. . - - ,o. .- 2 , , w, . . . . r. . . = 2,n,- = . .m . m ece,. ,_nr

TIMD030 COMPONENT / ASSOCIATE DATA 10/24/95 12:33 Help Data Print Roar

map Exit Options

Facility = PSL Unit. 01 Comp. Status: OK As Of: 07/23/90 Component = V1404 Associate =

01, REACTOR COOLANT SYSTEM NPRDS: Y (Y/N)

Syctem .

Name : POWER OPERATED RELIEF VALVE (PORV) FOR PRESSURIZER TO QUENCH TANK 4 Comp. Type .* Sub-Type = Tech Spec : _ (Y/N)

Safety Channel: IST Reqd. : Y (Y/N)

Train :

RNP Reqd. : Y (Y/N) iLoc. Code : iiCB/80/S-31/E-36 - ~

Loc. Desc. :

EQ Related: ety Cis:

' Comp. Group = _

A Critical Comp.: _ Q Group BQ D c Pac : : 034B -- e vinas :

S21smic : I__ Startup Sys. _

Orig. P.O. : CE-1900(9903305) Work Group : _ Q Basis : _.

' Reg. 1.97 : Reg. Type: BQ Remar)cs :

EQ Comp. Tag : Rev.:

R3g. Catg. : EQ Tab :

..........................................................................=....

Installed: Eng. Ver :

Mfr Code : _

Mod 31 No. : _ P.O. No. :

UTCi: S ock Cd: _

S$ rial No : ,

F9 TO VIEW ADDITIONAL COMPONENT DATA. F10 TO F10.

ACCESS CONFIGURATION X-REF.

Perform Msg F12= Cancel F1.Halp F4. Prompt F5= Search F6= Refresh I

i I

l

m CO*d W 101 10/24/vs ;e:sa TIMD030 COMPONENT / ASSOCIATE DATA Help Data Print Readtrap Exit Options Comp. Status: OK As Of: 07/23/90 Unit. 01

= PSL l Facility .

= V1405 Associate _

Component NPRDS: Y (Y/N)

System Name .: 01_ REACTOR. Sub-Type COOLANT ._ SYSTEMMOTOR Tecn Spec :OPERATED IS Comp. Type =* _

Safety Channel: IST Reqd. : Y (Y/N)

Train : RWP Reqd. : Y (Y/N)

Loc. Code : RCS/79/S-29/E-29 -- m Loc. Desc. : Safety Clos Cortp , Group = EQ Related: ,,_

Q Group : A,_

Critical Comp.: _

EQ Doc Pac : _

Star:up Sys. : 034B e L.r.as s .

Saismic : I_ Work Group : Q Basis : _

Orig. P.O. : CE- 1900 (9002051) :

ROg. 1.97 : Reg. Type: _,

SQ Remarks __

Rev.:

BQ Comp. Tag : _

Reg. Catg. : EQ Tab

....................................................................=..........

Eng. Ver : _

Installed: _

Mfr Code : _

P.C. No. :

Mod 31 No. : LTC: __ stock Cd: _

Sorial No :

F10 TO ACCESS CONFIGURATION X-REF.

F9 TO VIEW ADDITIONAL COMPONENTF6. DATA.

Refresh rio.Perf om Msg F12. Cancel l F1.Halp F4. Prompt FS. Search i

l l

. - ._ _ _ _ _ _ _ ~ _ . . - . _ _ _ _ . . - _ . - _ _ . . _ _ . _ _ . - . _ . _ _ _ . . . _ _

m 6 4 6.0 ADMINISTRATIVE CONTROLS l

i j 6.1 RESPONSIBILITY

! 6.1.1 The Plant 'ianager shall be responsible for overall unit operation and i shall delegate in writing the succession to this responsibility during his j absence.

j 6.1.2 The Shift Supervisor, or during his absence from the control room a ,

designated individual, shall be _ responsible for the control room command l J function. A management directive to this effect, signed by the President -

Nuclear Division, shall be reissued to all station personnel on an annual basis.

6.2 ORGANIZATION ONSITE AND OFFSITE ORGANIZATION

l

! 6.2.1 An onsite and an offsite organitation'shall be established for unit I

operation and corporate management. This onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear j power plant. '

) a. Lines of authority, responsibility and communication shall be established ,

and defined from the highest management levels through intermediate l i levels to and including all operating organization positions. Those relationships shall be documented and updated, as appropriate, in the form of organizational charts. These organizational charts will be I documented in the Topical Quality Assurance Report and updated in accordance: with 10 CFR 50.54(a)(3).

b. The President - Nuclear Division shall be responsible for overall

! plant nuclear safety. This individual shall take any measures needed

. to ensure acceptable performance of the staff in operating, maintaining, i and providing technical support in the plant so that continued ' nuclear

safety is assured.

I

c. The Plant Manager shall be responsible for overall unit safe operation and shall have control over those onsite resourdes necessary for' safe l

! operation and maintenance of the plant.

j d. Although the individuals who train the operating staff and those who carry out the quality assurance functions may report to the appropriate manager onsite, they shall have sufficient organizational freedom to be independent from operating pressures.

(

) e. Although health physics individuals may report to any appropriate manager j onsite, for matters relating to radiological health and safety of

employees and the public, the health physics manager shall have direct i

access to that onsite individual having responsibility for overell unit i management. Health physics personnel shall have the authority to cease 2 any work activity when worker safety is jeopardized or in the event of j unnecessary personnel radiation exposures.

ST. LUCIE - UNIT 1 6-1 Amendment No. 15,$p, N e l

j. 137,

_ _ - . _ .- . . . ~_ _..m_ -.. _ . _ _ _ __ . m_.m _ . _ _ . _. , _.

m '

1 DEaCTOR COOLANT SYSTEM a* POWER OPEDATED DEL!EF vatVES LIMITING CONDITION FOR OPERATION i

3.4.13 Two power operated relief valves (PORVs) shall be OPERABLE with their

setpoints selected to the low temperature mode of operation as follows

a. A setpoint of less than or equal to 350 psia shall be selected:

1.- During cooldown when the temperature of any RCS cold leg is less than or equal to 215'F and

2. During heatup and isothermal conditions when the temperature of a

i any RCS cold leg is less than or equal to 193*F.

b. A setpoint of less than or equal to 530 psia shall be selected:

i 1.

i During cooldown when the temperature of any RCS cold leg is greater than 215'F and less than or equal to 281*F.

1 2. During heatup and isothermal conditions when the temperature of l

i any RCS cold leg is greater than or equal to 193*F and less than s or equal to 304*F.

1 APptfCABILITY MODE 4 when the temperature of any RCS cold leg is less than a

!. or equal to 304*F, MODE 5, and MODE 6 when the head is on the reactor vessel;

, and the RCS is not vented through greater than a 1.75 square inch vent.

, ACTION:

i

a. With one PORY inoperable in MODE 4, restore the inoperable PORV to

+ OPERA 8LE status- within 7 days; or depressurita and vent the RCS through greater than a 1.75 square inch vent within the next 8 4

hours.

b. With one PORV inoperable in MODES 5 or 6. either (1) restore the Inoperable PORY to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or (2) complete depressurization and venting of the RCS through greater than a 1.75 square inch vent within a total of 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months .

. c. With both PORVs inoperable, restore at least one PORY to operable i

- status or complete depressurization and venting of the RCS through greater than a 1.75 square inch vent within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d. With the RCS vented per ACTIONS 4, b, or c. verify the vent pathway 1 1

at least once per 31 days when the pathway is provided by a valve (s) that is locked, sealed, or othemise secured in the open positier.

otherwise, verify the vent pathway every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4 e.

In the event either the PORVs or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be ,

prepared and submitted to the Commission pursuant to Specification i

6.9.2 within 30 days. The report shall describe the circumstances i

initiating the transient, the effect of the PORVs or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence.

4

f. The previsions of Specificatten 3.0.4 are not applicable.

tuewffttANff RfnufREMENTE 4.4.13 Each PORV shall be demonstrated OPERABLE by:

1

a. Verifying the PORY isolation valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and

b. Performance of a CHANNEL FUNCTION TEST, but excluding valve operation, at least once per 31 days; and

c. Performance of a CHANNEL CALIBRATION at least once per 18 months.

4 ST. LUCIE - UNIT 1 3/4 4-5g Amendment 40,eM44.132

i 1

ARTICLE IWV-2000 DEFINITIONS IWV 2100 ACTIVE AND PASSIVE (c) Category C- valves which are self-actuating in

\ALVES response to some system characteristic, such as pressure (relief valves) or flow direction (check valves)

(a) active ralres - valves which are required to (d) Category D - valves which are actuated by an change position to accomplish a specific function energy source capable of only one operation, such as (b) passive ralres - valves which are not required to rupture disks or explosively actuated valves change position to accomplish a specific function IWV.2300 OTHER DEFINITIONS IWV 2200 CATEGORIES OF VALVES (a) exercising - the demonstration based on direct Valves within the scope of this Subsection shall be or indirect visual or other positive indication that the placed in one or more of the following categories. moving parts of a valve function satisfactorily When more than one distinguishing category charac- (b) inservice life - the period of time from installa-teristic is applicable, all reqairements of each of the tion and acceptance until retired from service individual categories are applicable, although duplica- (c) inservice rest - a special test procedure for tion or repetition of common testmg requirements is obtaining information through measurement or obser-not necessary. vation to determine the operational readiness of a (a) Category A - valves for which seat leakage is valve limited to a specific maximum amount in the closed (d) maintenance - routine valve servicing or work position of fulfillment of their function on a valve undertaken to correct or prevent an (b) Category B - valves for which seat leakage in abncrmal or unsatisfactory condition the closed position is inconsequential for fulfillment of (c) operational readiness - the capability of a valve their function to fulfillits function 218

1 e ARTICLE IWV-3000 i 1

TEST REQUIREMENTS I IWV 3IDO PRESERVICE TESTS IWV-3400 INSERVICE TESTS -

Each valve, after installation and prior to service, CATEGORY A AND B VALVES i i

- shall be tested as required by this Subsection. These IWV 3410 VALVE EXERCISING TEST tests shall be conducted under conditions similar to those to be experienced during subsequent inservice 3411 Test Frequency tests. Safety and relief valves which will be removed Category A and B valves shall be exercised at and bench tested during subsequent inservie; tests least once every 3 months, except as provided by need not be installed prior to the preservice test. 1WV 3412(a),IWV 3415, and IWV 3416.

l IWV 3412 Exercisies Procedure j (a) Valves shall be exercised to the position re- l quired to fulfill their function unless such operation is !

IWV 3200 VALVE REPLACEMENT, not practical during plant operation. If only limited REPAIR, AND MAINTENANCE operation is practical during plant operation, the valve When a valve or its control system has been shall be part-stroke exercised during plant operation replaced or repaired or has undergone maintenance t and full-stroke exercised during cold shutdowns.

that could affect its performance, and prior to the time Valves that cannot be exercised during plant operation it is returned to service, it shall be tested to demon- shall be specifically identified by the Owner and shall stra'e that the performance parameters, which could be full-stroke exercised during cold shutdowns. Full-be affected by the replacement, repair, or maintenance, stroke exercising during cold shutdowns for all valves are within acceptable limits. not full-stroke exercised during plant operation shall be on a frequency determined by the intervals between shutdowns as follows:

For intervals of 3 months or longer - exercise during each shutdown.

For intervals of less than 3 months - full-stroke exercise is not required unless 3 montS have passed IWV 3300 VALVE POSITION INDICATOR sine &t shutdown exercise.

VERIFICATION (b) The necessary valve disk movement shall be Valves with' remote position indicators shall be determined by exercising the valve while observing an observed at least once every 2 years to verify that valve appropriate indictar, which signals the required operation is accurately indicated. change of disk position, or observing indirect evidence (such as changes in system pressure, flow rate, level, or temperature), which reflect stem or disk position.

%amples of ===*===a that coukt afect valve puformance IWV-3413 Power Operated Valves parameurs am adjustment of suen pechng; mnoval of the ha==st stan assembly, or actuator; and disconnocuon .if hydraunc [4) The limiting value of full-Stroke time of each or eisetru:al hoes. power operated valve shall be specified by the Owner.

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i l I%%3413-IWV 3423 SECTION XI - DIVISION 1 1983 F4 tion full-stroke time is that time interval from initiation of corrective action shall be initiated immediately. If the l the actuating signal to the end of the actuating cycle, condition is not, or cannot be, corrected within 24 hr, l fb) The stroke time of all power operated valves the valve shall be declared inoperative. When correc-shall be measured to the nearest second, for stroke tive action is required as a result of tests made during times 10 see or less, or 10% of the specified limiting cold shutdown, the condition shall be corrected before stroke time for full-stroke times longer than 10 sec, startup. A retest showing acceptable operation shall be whenever such a valve is full-stroke tested. run following any required corrective action before the valve is retumed to service. '

l IWV-3414 Valves ir, Reguist Use I Valves that operate in the course of plant operation i

at a frequency which would satisfy the exercising IWV 3420 VALVE LEAK RATE TEST !

requirements of this Subsection need not be addition-ally exercised, provided the observations otherwise I% V-3421 Scope required for testing are made and analyzed during Category A valves shall be leak tested, except that such operation, and are recorded in the plant record at valves which function in the course of plant operation intervals no greater than specified in IWV-3411. in a manner that demonstrates functionally adequate seat tightness need not be leak tested. In such cases, i

the valve record shall provide the basis for the IWV-3415 Fall-Safe Valves conclusion that operational observations constitute When practical, valves with fail-safe actuators shall be tested by observing the operation of the valves upon loss of actuator power. If these valves cannot be tested IWV-3422 Frequency once every 3 months, they shall be tested during each Tests shall be conducted at least once every 2 years.

cold shutdown; in case of frequent cold shutdowns, these valves need not be tested more often than once every 3 months. IWV-3423 Differential Test Pressure Valve seat leakage tests shall be made with the pressure differentialin the same direction as when the !

IWV 3416 Valves in Systems Out of Service valve is performing its function, with the following l exceptions.

For a valve in a system declared inoperable or not I required to be operable, the exercising test schedule (a be-type valves may be tested with pressure i need not be followed. Within 30 days prior to return of under the seat.

j (b) Butterfly valves may be tested m_ either direc-the system to operable status, the valves shall be 1 exercised and the schedule resumed in accordance tion, provided their seat construction is designed for with requirements of this Article, sea ing against pressure on either side.

(c) Gate valves with two-piece disks may be tested by pressurizing them between the seats.

** ** cept check valves) may be tested in IWV-3417 Corrective Action . . .

either direction if the function differential pressure is (a) If, for power operated valves, an increase in 15 psi or less, stroke time of 25% or more from the previous test for (e) Leakage tests involving pressure differentials 1 v:1ves with full-stroke times greater than 10 sec, or lower than function pressure differentials are permit-50% or more for valves with full-stroke times less ted in those types of valves in which service pressure than or equal to 10 see is observed, test frequency shall will tend to diminish the overall leakage channel be increased to once each month until corrective opening, as by pressing the disk into or onto the seat l I tetion is taken, at which time the original test with greater force. Gate valves, check valves, and I frequency shall be resumed. In any case, any abnor- globe-type valves, having function pressure differential mality or erratic action shall be reported. applied over the seat, are examples of valve applica-l (b) If a valve fails to exhibit the required change of tians satisfying this requirement. When leakage tests l valve stem or disk position or exceeds its specified are made in such cases using pressures lower than limiting value of full-stroke time by this testing, then function maximum pressure differential, the observed I

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7 NRC CLOSED PREDECISIONAL ENFORCEMENT CONFERENCE ST LUCIE NUCLEAR PLANT MARCH 8,1996 IAR TITLE 1 Predecisional Enforcement Conference Agenda 2 Expected Attendees, Meeting Announcement 3 Opening Remarks and introductions 4 NRC Enforcement Policy i

5 Summary of the issues 6 Statement of Concerns / Apparent Violations 7 Inspection Report No. 50-335,389/96-03 8 , ,

Unit 1 Control Room Arrangement, CVCS Charging System Flow Diagram, Enforcement Pre-Panel Questionnaire 9 Licensee Procedure OP 1-0250020, Boron Concentration Control - Normal Operation; and TC 1-96-017 to OP 1-0250020 of 1/23/96 10 Licensee Procedure Ol5-PR/PSL-1, Preparation, Revision, Review / Approval of Procedures 11 Licensee Procedure AP 0010120, Conduct of Operations; and TC 0-96-014 to AP 0010120 of 1/29/96 mo

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i 12 Memo from E. Jordan on Licensed Power Level j of 8/22/80 :

1 13 St. Lucie Unit 1 FSAR l 1

1 l- 14 Closing Remarks 4

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'PREDECISIONAL ENFORCEMENT CONFERENCE AGENDA j ST LUCIE MARCH 8,1996, AT 10:30 A.M.

NRC REGION 11 OFFICE, ATLANTA, GEORGIA l i

1. OPENING REMARKS AND INTRODUCTIONS S. Ebneter, Regional Administrator ll. NRC ENFORCEMENT POLICY B. Uryc, Director ;

Enforcement and Investigation Coordination Staff . l

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SUMMARY

OF THE ISSUES S. Ebneter, Regional Administrator !

IV. STATEMENT OF CONCERNS / APPARENT VIOLATIONS A. Gibson, Director Division of Reactor Safety V. LICENSEE PRESENTATION T. Plunkett, President - Nuclear Division Florida Power & Light Company l

VI. BREAK / NRC CAUCUS Vll. NRC FOLLOWUP QUESTIONS l Vill. CLOSING HEMARKS :

S. Ebneter, Regional Administrator i j

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4 EXPECTED ATTENDEES i

I LLqensee l

l Tom Plunkett, President - Nuclear Division, FPL i

Bill Bohlke, Vice President, St. Lucie Nuclear Plant j Jim Scarola, Plant General Manager, St. Lucie j Dan Denver, Engineering Manager, St. Lucie j Ed Weinkam, Licensing Manager, St. Lucie

! Peter Honeysett, Nuclear Plant Supervisor, St. Lucie l Frank Cone, Reactor Controls Operator, St. Lucie l Hank Holzmacher,' Reactor Controls' Operator, St. Lucie !

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4 MEC Stew Ebneter, Regional Administrator, Region ll (Rll) i - Luis Reyes, Deputy Regional Administrator, Ril

) Al Gibson, Director, Division of Reactor. Safety (DRS), Ril Ellis Merschoff, Director, Division of Reactor Projects (DRP), Ril 4

Gene Imbro, Director, Project Directorate 11-2, NRR l James Beall, Enforcement Coordinator, Office of Enforcement (OE)-

) Johns Jaudon, Deputy Director, DRS, Rll l Jon Johnson, Deputy Director, DRP, Rll j

[ Bruno Uryc, Director, Enforcement and Investigation Coordination Staff j i (EICS), Ril

. Charles Casto, Chief, Engineering Branch, DRS, Ril !

! Tom Peebles, Chief, Operations Branch, DRS, Ril !

Kerry Landis, Reactor Projects Branch 3, DRP, Ril Jan Norris, Project Manager, NRR Linda Watson, Senior Enforcement Specialist, EICS, Rll Carolyn Evans, Regional Counsel, Ril Mark Miller, Senior Resident inspector, St. Lucie, DRP, Ril Robert Schin, Reactor Inspector, Engineering Branch, DRS, Ril Edwin Lea, Project Engineer, Reactor Projects Branch 3, DRP, Ril

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s From: Anne Bolandj C,d l To: ATP1.JWY J1}st/z;' RL ;

l Date: 7/24/9611:15amV '

Subject:

END DATE FOR ST. LUCIE INSPECTION WHAT IS THE END DATE FOR THE ST. LUCIE INSPECTION WHICH IDENTIFIED i i

THE 50.59 VIOLATIONS? ..

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ML20137Y479

Contents

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Subject:

1.0 INTRODUCTION

SUMMARY

SUMMARY

SUMMARY

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