Text

Informs That Separate Rept on St Lucie Relief Valve Issue Will Be Issued on 951020
	ML20137P533
Person / Time
Site:	Saint Lucie
Issue date:	10/10/1995
From:	Linda Watson; NRC
To:	; NRC
Shared Package
ML20137P228	List: ML20137P226; ML20137P236; ML20137P391; ML20137P440; ML20137P509; ML20137P533; ML20137P565; ML20137P589; ML20137P595; ML20137P609; ML20137P624; ML20137P631; ML20137P647; ML20137P675; ML20137P686; ML20137P700; ML20137P714; ML20137P719;
References
	FOIA-96-485 NUDOCS 9704090266
	Download: ML20137P533 (103)
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Linda J. Watson (LJW2)-

To:

ATB, BXU

.Date:

Tuesday, October 10, 1995 10:42 am Subjects. St Lucie relief vlv K.

Landis indicated that he was going to re-exit the St. Lucie relief viv issue as an apparent violation on 10/11 and would proceed to set up enf. conf.

He will be issuing separate report on this issue approx. 10/20.

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ANNE T. BOLAND REGIONAL CONTACT DATE OF REQUEST SEPTEMBER 29,1995 REGION 11 LICENSEE FLORIDA POWER AND UGHT COMPANY FACfLITY/ LOCATION ST. LUCIE / JENSEN BEACH, FLORIDA UNIT 1

UCENSE/ DOCKET NO(S).

DPR-67, 50-335 LAST DAY OF INSPECTION SEPTEMBER 16,1995 Of REPORT NO.

NONE DATE OF Of REPORT N/A

SUMMARY

OF FACTS OF CASE (ANNUAL REPORT FORMAT':OR EATS ENTRY) (MAXIMUM OF 300 CHARACTERS)

FAILURE TO TAKE PROMPT CORRECTIVE ACTION RESULTING IN THE FAILURE OF A UNIT 1 REUEF VALVE TO RESEAT WITHOUT OPERATOR INTERVENTION, THE EVENT RESULTED IN APPROXIMATELY 4000 GALLONS OF REACTOR COOLANT ACCUMULATING IN THE UNIT 1 PIPE TUNNEL.

BRIEF

SUMMARY

OF INSPECTION FINDINGS (IF NOT SUFFICIENTLY DESCRIBED ABOVE) -

ON AUGUST 10,1995, WHILE PLACING THE UNIT 1 SHUTDOWN COOLING SYSTEM IN SERVICE, THE A LPSl HEADER THERMAL REUEF, UFTED

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RESULTING IN THE LOSS OF APPOX. 3500-4000 GALLONS OF COOLANT INTO THE UNIT 1 PIPE TUNNEL. THE ROOT CAUSE OF THE j

PROBLEM WAS A DESIGN ISSUED ASSOCIATED WITH HIGH BLOWDOWN VALUES; HOWEVER, THE LICENSEE FAILED TO EVALUATE AND CORRECT ANOMALOUS RELIEF VALVE BEHAVIOR ON FEBRUARY 20, MARCH 2, AND MARCH 10,1995, WHICH MAY HAVE PREVENT THE AUGUST EVENT. OVER 100 VALVES WERE EVALUATED FOR THIS CONDUCTION AND 15 REQUIRED REPAIR AND/OR REPLACEMENT. NO SAFETY SYSTEMS WERE DETERMINED TO BE INOPERABLE AS A RESULT OF THIS PHENOMENON.

REASON FOR POTENTIAL ESCALATED ACTION SUPPLEMENT l.C.2.B, A SYSTEM DESIGNED TO PREVENT OR MITIGATE A SERIOUS SAFETY EVENT BEING DEGRADED TO THE EXTENT THAT A DETAILED EVALUATION IS REQUIRED TO DETERMINE ITS OPERABluTY.

DELEG ATED CASE YES X

NO MED INST PHYSICIAN NUC PHARM RADIOG 1RRAD WELL LOGGERS ACADEMIC GAUGE MOISTURE DENSITY OTHER TYPE:

CITE SIMILAR CASE: EA NO.

SUPPLEMENT 1.C.2.B SHOULD OE ATTEND ENF CONF X

YES NO NONDELEGATED CASE X

YES NO X

NONDELEGABLE TYPE 01 REPORT / WILLFUL COMPLEX / NOVEL DISCRETION COMM APPROVAL 01 INTEREST SL 1 OR 2 OTHER REASON:

IS THERE A BASIS TO CLOSE ENFORCEMENT CONFERENCE? NO IF YES, EXPLAIN:

ENFOQCEMENT CONFERENCE TO BE SCHEDULED.

EA # ASSIGNED BY OE 6

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ES ASSIGNED

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e t 1 ESCALATED ENFORCEMENT PANEL QUESTIONNAIRE j INFORMATION REQUIRED TO BE AVAILABLE FOR ENFORCENENT PANEL j. PREPARED BY: R. Prevatte 1 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, 1 as well as the trancmittal memo to the Office of Enforcement explaining and l ' justifying the Regiovs proposed escalated enforcement action.) i 1. Facility: E _i.gqig Unit (s): 1 f -Docket Nos: 10-111 i' License Nos: &QZ Inspection Dates: July 30 - September 16. 1995 ~ Lead Inspector: Richard L. Prevatte 2. Check appropriate boxes: ~ [X] A Notice of Violation'(without "boilerplate") which includes the recommended severity level for the violation is enclosed. [] This Notice has been reviewed by the Branch Chief or Division Director and each violation includes the appropriate level of specificity as to how and when the requirement :was violated. -j [] Copies of applicable Technical Specifications or license conditions cited in the Notice are enclosed. l 3. Identify the reference to the Enforcement Policy Supplement (s) that best fits the violation (s) (e.g., Supplement I.C.2) 1.C.2.B I -4. What is the apparent raot cause of the violation or problem? Enaineerina evaluation and orioritization of potential eautoment oroblem was not timely. ~ -5. State the message that should be given to the licensee (and industry) i i ~ through this enforcement action. Imorove prioritiration and timeliness of response to olant problems. 6.- Factual information related to the following civil penalty escalation or l +

.. ~ h 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? i Was. it; self-disclosing? Was it identified as a result of a: generic notification?). Licensee identified anomalous behavior of safety related thermal relief valves on February 20. March 2. and March 10. 1995, but did not take action until a failure also occurred on Auaust 10. 1995 and NRC auestioned corrective action. e b. CORRECTIVE ACTION: Although we expect to learn more information regarding corrective action.at the enforcement conference, describe preliminary information obtained during the inspection 3 'and exit interview. 6 See item A., t What were the.immediate corrective actions.taken upon' discovery of i the violation, the development and implementation of long-term corrective action and the timeliness of corrective actions? i -hitial problem was under enaineer na review for several months. After cuestionina by NRC. the prob' em was thorouahly researched and corrected. i What was the degree of licensee initiative to address the violation and the adequacy of root cause analysis? .j hitial - not timely. l Final - aood investiaation and broadened scope led to review of over 100 relief valves. 1 t 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): NCV 94-25-01. Inadeauate desian control of NADH suction relief valves. VIO 94-11-01. Inadeauate corrective action for MOV which stalled durina surveillance. i IE swina bus would not strio on undervoltaae due to '1 VIO 94-12-01. wirina oroblem 94-08-01. Inadeouate corrective action on waterhammer event. a

~. J Inocerable snubbers and SRV PORV tailoines s 94-08-02. Failure to document above non-conformance. 94-06-02. Inadeounte desian control on Unit 2 charaina cumo seauence. 94-06-01. Failure to report DG failure, visntify 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: Ena. Sucoort 1 - 1 4 d. PRIOR OPP 0Ki. UNITY 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? Problem known but not oursued. e.- MULTIPLE OCCURRENCES: Were there multiple examples of the violation identified during this inspection? If there were, -identify the number of examples and briefly describe each one. No. f. DURATION: How long did the violation exist? 1 Problem has existed on thermal relief valves since initial installation. t 1 4 0 h 3 i t 4 -m.

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? W SDDITIONAL COMMENTS / NOTES: t 5) Shutdown Cooling Relief Valve Lift t A.

Background

On February 28, while placing the 1A SDC train in service,- the licensee experienced a lift of 1A LPSI pump suction relief valve V-3483 (see IR 95-04). The valve did not reseat, and the loss of RCS inventory was abated by closing LPSI hot leg suction isolation i valves V-3480 and V-3481, which isolated the valve - t from RCS pressure. The root cause of the lift was determined to be water hammer, which resulted from passing relatively hot RCS fluid through.the suction line at high velocity as the LPSI pump was started. j As corrective action, the~1icensee revised OP 1-0410022, " Shutdown Cooling," to change the methodology of starting the LPSI pump to the following: e Shut LPSI pump discharge isolation and LPSI header-' isolation valves e Start the LPSI pump e Immediately open the LPSI pump isolation valve 3 e Throttle open two LPSI header isolations to 150 gpm per header e Run for 15 minutes l e Start the second pump 4 e Throttle open the remaining LPSI header isolation valves to 150 gpm per header e Wait 5 minutes e Incrementally open header isolation valves to obtain full flow. The licensee reasoned that this methodology would result in a slow increase in flow, allowing controlled system heatup and minimizing the potential for water hammer. B. LPSI Discharge Isolation Valve Lift On August 10, while placing the Unit 1 SDC system in service to support a cooldown required due to inoperable PORVs (see IR 335/95-16), V-3439, the A LPSI header thermal relief, lifted resulting in a loss of approximately 3500-4000 gallons of RCS coolant in-the Unit 1 Pipe tunnel. The following timeline was developed from operator interviews, logs and d instrumentation data: 0018 A LPSI pump start (ANPS, NWE, Logs) Pressurizer level begins to drop (strip chart data). F . - - - ~

.4 0025 ANPS directs SNPO to tour pipe tunnel'due to minor reduction in pressurizer level (ANPS). No increases in HUT, RWT, etc noted (ANPS) i 3 SNPO reports no unusdal conditions in pipe t tunnel 0105 B LPSI. pump start (ANPS, NWE, Log) Pressurizer level recovers and oscillates (strip chart) 0140 Cooldown flow established (ANPS, NWE) 0210 Fire watch calls control room, reports water issuing from watertight door isolating pipe tunnel from RAB (ANPS, NWE) i 0215 SDC secured (ANPS, NWE) I Pressurizer level increases and stabilizes (strip chart) 0226 Floor drain isolation valves (FCV 25-1 through

7) noted to be closed on control panel (ANPS,

~ NWE) i Drain valves subsequently opened (ANPS, NWE) i Flooding in RAB ONOP entered (ANPS) Water levels in pipe tunnel weren't dropping due to clogged floor drains (NWE) 0345 Water in pipe tunnel pumped by maintenance personnel to floor drains in RAB (ANPS) Operators cycle various isolation valves looking. for leak 0611 1A LPSI pump started with NWE observing in pipe tunnel (ANPS) 4 0612 NWE identifies V-3439 as giassing water (ANPS) l The licensee concluded that the cause of the relief valve lift was a pressure surge while LPSI pumps were operating in a low-flow condition. The combination of RCS pressure (a maximum of 267 psia at the. time) and LPSI pump discharge head at essentially no flow (approximately 182 psid) combined with possible perturbations (when starting the pump) was considered enough to challenge the relief valve setpoint (485-515). This condition existed from the time the 1A LPSI pump discharge isolation valve was opened until i operators initiated flow through the LPSI header isolation valves. V-3439 was designed to provide a 10 percent blowdown, which, if applied to the lower acceptable lift ~ setpoint of the valve (485 psig), would require a 48.5 psia reduction in pressure to allow reseat. Given these aarameters, should V-3439 open, RCS pressure would lave to drop to 436.5 psia to allow valve resent (assuming only a 10 percent blowdown). The volume of the RCS and pressurizer would preclude such a reseat J until significant volumes of coolant were lost. l The volume of coolant lost during the event was a m m m

L i estimated by the inspector, based upon floor layouts-as displayed on drawings. Given water depths reported l - by the NWE (up to approximately 14" in some areas), l i-the inspector estimated that approximately 3500 - gallons were lost. - The CVCS makeup integrator, measuring volume added to the VCT in maintaining pres'serizer level on setpoint, indicated that 4000 gallons were added _to.the VCT. - The licensee concluded that the closed floor drain isolation valves, HCV-25-1 through 7 (a set of 7 ganged valves) were the result of valve stroke testing in preparation for Hurricane Erin. During testing conducted by control room operators, some of the valves had failed to stroke properly. As a result, the valves were left closed for troubleshooting and were never reopened. OP 1-0010123, Rev 99, " Administrative Control of Valves, Locks, and Switches," required, in step 8.1.6, that "All valve or switch position deviations or lock openings shall be documented in Appendix C, Valve Switch Deviation Log..." The inspector reviewed archived Appendix C logs completed in July and August and control room open Appendix C logs and found no evidence that HCV-25-1 through 7 were logged as being out of position. The failure to enter the valves' closed status into the valve deviation log is an example of a violation (VIO 335/95-15-01, " Failure to Follow Procedures," Example 4). STAR 950917 was initiated to develop a PM for verifying that floor drains were unclogged. The licensee prepared an evaluation of the effects of the subject setpoint/ blowdown values on plant operation. JPN-PSL-SENP-95-101, Rev 1, " Assessment of the Effects on Plant Operation of Lifting the LPSI Pump Discharge Header Thermal Relief Valve," concluded that the subject condition would not have a significant effect on safe plant operation during normal, shutdown, and design basis accident conditions. In reaching this conclusion, the c evaluation noted the following: i e As flowrate through the relief valve (at lift setpoint pressure) was approximately 40 gpm, the loss of inventory was within charging system capacity (44 gpm per pump). l e During the injection phase of an accident, the LPSI pumps would draw suction from the RWT, thus pressure developed by the' pump would not compound a high pressure suction source and the relief valve's lift setpoint would.not be challenged. f ..m... .. m

a; 4 4 j o-The relief valve in question discharged to a floor _ drain which directed flow to the . safeguards room sump. The. sump was designed to be pumped down in level to the EDT automatically i-when offsite power is available. Thus, with offsite power available, no flooding hazard t would exist. Under conditions with no offsite power available, the water level in the. safeguards room (after the sump. overfilled) i -would not rise to the level of the HPSI pump j motors until approximately 7 hours after the lift. Before this time elapsed, the licensee reasoned that sump. high level alarms would alert operators to the event, allowing operator intervention prior to the loss of the HPSI pump. e The licensee noted that, while SDC was assumed to be placed in service during postulated small break LOCAs, ESDEs, and SGTRs (when RCS pressure may have been high enough to have led to a relief valve lift), the FSAR analysis demonstrated that fuel damage (and thus the a release of significant amounts of radioactive . material to the RCS) was involved only because of extremely conservative assumptions. The i evaluation went on to state that "A review of FSAR analysis of small break LOCAs, ESDEs and SGTRs demonstrates that these accidents will not result in fuel damage if assumptions that reflect the actual operating history of the plant are applied. Therefore, the radiological consequences of these FSAR accidents will not be L increased and the FSAR offsite doses remain bounding." 1' i The inspector took exception to the licensee's conclusion. The subject passage was included in 1 Section 4 of the evaluation, " Analysis of Effects of 3 i Lifting V3439," in a section entitled " Increases in Radiological Consequences 'of Design Basis Accidents." ) The inspector found that, in choosing to neglect 3 design basis assumptions in their analysis of the event- (specifically, a return to power and fuel J failure resulting from the most reactive rod failing j to insert), the licensee did not evaluate the increases in the radiological. consequences of design ] basis accidents. Rather, the licensee evaluated the i radiological consequences of a less significant set of i accidents and. concluded, without providing ] quantitative results, that the radiological consequences of design basis accidents bounded the i i noted relief valve lift. While the inspector agreed i ~ with the licensee's position that the circumstances assumed in design basis accidents were,

d probablistically, of low likelihood, the inspector pointed out that the assumptions were the approved licensing basis of the plant and, as such, provided J the appropriate common ground upon which to evaluate the event's significance. The inspector brought this to the attention of the licensee, who stated that they would consider the issue. At the close of the inspection period, the licensee had not presented a final position on the issue. As a result, this issue will be tracked as an unresolved item (URI 95-15-04, " Adequacy of Engineering Evaluation Regarding Unit 1 Loss of Inventory via V-3439"). On August 12, the inspector requested data on-approximately 25 relief valves on both units which communicated with the RCS in some way. The requested data included lift and blowdown setpoints, tolerances, relief capacity, and normal operating 'ressures experienced by the valves. Shortly 4r ir requesting the information, the licensee informeu the inspector that a team had been formed to evaluate all safety-related relief valve data. The team included members from Engineering, Maintenance, Operations, Tech Staff, and Licensing. The team's review was documented in JPN-SPSL-95-0334, "St. 1.ucie Units 1.and 2 Design Review of Safety Related Relief Valves," transmitted to the site by letter dated August 30. The inspector found the methodology of the study to be sound, considering worst case combinations of system operating pressures, relief valve setpoint, and blowdown. Relief valves were evaluated for their margin to lift and blowdown margin (the difference between reseat pressure and normal system operating pressure). The document reported that, of 114 relief valves reviewed, 8 valves on Unit I and 5 valves on Unit 2 required further review due to unacceptable margins of lift or blowdown. Corrective Actions were specified for the following valves: Unit 1 valves e V2324, V2325, and V2326 - Charging Pump Discharge Relief Valves - MEP 107-195M was issued to reduce the design superimposed backpressure from 165 psig to 115 psig. ~ e V2345 - Letdown Relief Valve - PC/M 108-195 issued to reduce letdown backpressure to 430 i psig and to reduce the valve's blowdown from 25 percent to 15 percent. e V3412 - HPSI 1B Discharge Header Relief Valve - d

] .t w 4 1 i 'EP 115-95 was-issued to increase the design

i setpoint from 1735 psig to 1750 psig and to reduce blowdown from 25 percent.to 10 percent.

o V3417 - HPSI Pump 1A Discharge High Pressure-Header Relief. Valve -design setpoint increased from 2400 psig to 2485 psig and blowdown reduced 2 from 25, percent to 15 percent. I e V3468 and V3483 - SDC Suction Relief Valves - STAR 950430 was issued to evaluate new setpoints j 'and blowdown values.. Unit 2 Valves i i e . V2345 - Letdown: Relief Valve - At the close of the inspection period, an EP was being prepared-to implement actions similar to those implemented on Unit 1 for this valve. e <V3412 - HPSI 28. Discharge High Pressure Header i Relief Valve - At-the close of the inspection period, an EP was being prepared to reduce blowdown from 25 percent to 10 percent. j e V3417 - HPSI Pump 2A Discharge High Pressure Header Relief Valve - At the close of the inspection period, an EP was being prepared to increase the valve's setpoint from 2400 psig,to 2485 psig and to reduce blowdown from 25 percent E to 10 percent. e V3439 and V3507 - Low Pressure A and B Discharge Relief Valves - At the close of the inspection period, an EP was being prepared to increase the i p valve's setpoint from 500 psig to 535 psig. t As a result of the licensee's investigation, and through discussions with vendors, the licensee determined that some relief valves had been provided with unacceptably high blowdown values. This was, apparently, due to procedural problems at the vendor's test facility. At the close of the inspection period, the vendor (Crosby) was considering the 10 CFR 21 ramifications of the issue. The licensee documented l the conditions on STAR 951024. The inspector reviewed the STAR and noted that it had not been identified as j an "N" STAR (indicating a nonconfonaing conditior) The inspector brought this to the attention of QC, and ~ the condition'was corrected. The licensee identified-i the affected relief valves and segregated them -appropriately. The inspector reviewed the licensee's STAR database >h, W 6 l.

i 1 for events similar to the subject event and found the l ~ following: ' i o ' STAR 2-950167, initiated February 20, documented t 4 the lifting of SDC heat exchanger CCW relief valve SR-14350 when stroking CCW "N" header isolation valves closed. Once open, the relief. l valve had to be isolated.(by closing an upstream . valve in the process line) to bring about a reseat., l T ~ e STAR 0-950234,_ initiated March 2, documented the fact that relief valves had lifted and that-blowdown values placed the reseat pressure of i = the valves in the' operating. ranges of the systems they protected. i - e STAR 1-950269, initiated March 10, documented relief valve lifts on the Unit 1 CVCS letdown line during evolutions which should not have challenged the valve's setpoint. e STAR 0-950917, initiated August 18, documented the subject SDC relief valve lift. ] In addition to the STARS referenced above, IR 95-05-01 1 4 discussed work performed on the Unit 2 CVCS system to prevent letdown line relief valve lifts. The IR also described the' failure of the relief valve to resent (once lifted) due to a blowdown value which placed the t reseat pressure below the system's normal operating i pressure. The inspector reviewed the status of the STARS listed above and found them all to be open. The inspector discussed the timeliness of the resolutions to the subject STARS with the licensee. The licensee stated that their focus had been on the methodologies for setting blowdown values on the valves in question, rather than on the appropriateness of the setpoints 1 themselves. The licensee offered STAR 950234 as being illustrative of this point. The proposed corrective actions included: e Completion of SRV test benches, which would ^ allow the licensee to better set and test SRVs for lift setpoint and accumulation. It was noted that the bench had.only limited blowdown test capability. -e Performing an engineering design basis review of all safety related SRVs to validate or correct setpoints and issue a design document that summarizes the design data. ? .[' ~ c .m m

Enhancing journeyman training on SRVs. e While the inspector found the licensee's proposed activities prudent, it was noted that nothing precluded engineering from addressing the 'setpoint issue earlier in the process. The licensee stated that the STAR was addressed in stepwise fashion and. that the maintenance-related-items were addressed prior to forwarding the STAR to engineering. The inspector found that the licensee's corrective actions for the subject event were comprehensive and sound. However, the inspector concluded that the actions could have reasonably been expected to be performed.in a much more timely fashion. The subject phenomenon was identified as early as February,1994, and repeated itself on no less than 3 separate systems, and on both units, prior to the most recent event. Once focused on the issue, an engineering product.of high quality was developed, and corrective actions initiated, in approximately 2 weeks and identified valves requiring attention in a comprehensive action. 10 CFR 50, Appendix B required that, for conditions adverse to quality, prompt corrective action be taken to prevent recurrence. The licensee's failure to take prompt corrective action to the February / March events is a violation (VIO 335/95-15-02, " Failure to Take Prompt Corrective + Actions for Repeated Relief Valve Lifts") f V

Affll( ) \\ l '4 9 4 EVALUATION OF SAFETY SIGNIFICANCE EDEI 4 sedsmed 4000 sanoes remetw soolms misseed imo Unit i mas maowims bebg of LPSI dashesgo therunal relief VM39 on S/1095 duria$ abusedews oseams y - 1 6 e, e rep N Evelmaud espots hr shermate operating conditions (normal & sondes0: i s. 71ooding onashey related eqwpmens i b. Less o(RCSinventory c. Insrommes in udieiosical consequenas I 1. Suscepthie designbasis accidents j e. Considered operating hissory of U:st 1 Credble but senservenhe " backward noolf is sensistems with regulmory guidenes and abervehies. s EEE3H t 71ooding desseted prior to loss of"4CCS penps ) I I l Lees of M hvassory (=40 gran) withist makeup capabusty of %4 and/ct 19$1 pumps. No incmasas in FS AR radhkycal consequences if operenes tuseory e(@e plantis-ammalemed. remotor trip. b. ssLOCh Essonnauy so bel falkn since very and break and imemory loss wnhan capabihty of HPSl and/or chargny 4 1 0 -vv-- r

W 26-1995 06:00AM,,. F+ Lucie Restc'ent Offica 407 461 4622 P.02 1 No additional ibal fhilures caused by event; limited ofBite releases c. SGTREs: (unag leakage to RAB) since no c,peks with primary-to-t secondary leakage and DEQ 131 at T/S limits. d. In adefon to the 'above event evainations, a reenstic does cahm assuming sonne that fhilure shows no increase in PSAR radbbgka! consequences. LER filed describing / analyzing event and corrective actions 1 e i 4 1 I 4 i e w-unt a. m

Oe-28-1995 OG8 ?O@t C* Luc 1e Resadmt Office

407 461 4622 P.03 r

I i 1 \\. 4 EVALUATION OF SAFETT SIGNIFICANCE i (additional supporting information for engineering evaluation) APPROAQ ,An evaluation was performed to determine safety significance following the release of reactor coolant into the RAB from ' the lifting of V3439 during shutdown cooling operation on 8/10/95. The evaluation involved. a realistic backward look at the potential consequances of FSAR design basis accidents assuming that V1439 was part of the accident scenario. Considering the operating history of.the plant, the evaluation concluded that consequences of such accidents would be bounded by the offsite doses currently in the FSAR. The evaluation focused on design basis accidents where RCS conditions necessary to lift the relief valve may be present. Excess steam demand events (ESDEs), omall break loss of coolant accidents (SBLOCAs) and steam generator tube-rupture events (SGTREs) were identified as the types of accidents that could 3 involve partial depressurization of the RCS, and therefore, may involve shutdown cooling operation at a sufficiently high RCS j pressure to cause V3439 to lift. ESDEs l ESDEs are characterized by rapid and significant cooling of the RCS due the faulted S/G. In these events, the most reactive control rod is normally assu;aed to be stuck out of the core. The positive reactivity addition from the stuck rod and the cooler RCS temperatures causes the reactor to experience an overpower transient which results in fuel exceeding minimum DNBR. St Lucie i Plant has never had a failure of its control rods to insert upon reactor trip. With all rods inserted during an ESDE, the reactor l would neither return to power, nor would it experience power distribution anomalies, and very limited fuel failure would be l_ expected. The lifting of V3439 during shutdown cooling operation following such an event would not involve consequences exceeding i FSAR offsite doses. i l l l l SBLOCAs The NAR analyzed SBLOCA includes depressurization of the RCS, high pressure safety injection at 1275 psia, dumping of the safety injection tanks at 215 psia and low pressure safety injection at i 200 psia. Shutdown cooling operation would occur at a very low RCS pressure after securing high pressure safety injection, and therefore, would not involve the lifting of V3439. Only small SBLOCAs, involving partial depressurization of the RCS, provide the l 1 B e.we-. n -e-m e, ~_a- --~c

-l 09-26-M95 Ge:00R1 St Luc 1e Reaaownt Office

407 461 4622 F.04 i

conditions necessary to lift V3439. For small SBLocAs, either high safety injection or charging will provide sufficient i i pressure The makeup, and fuel failure, if any, would be extremely limited. lifting of V3439 during shutdown cooling operation following such an event would not involve consequences exceeding FSAR offsite j doses. 4 50TREs Since the FSAR SGTRE involves shutdcwn cooling operation following a controlled reduction in RCS pressure, it could result in lifting of V3439. The FSAR analysis of this event provides offsite deses j resulting from plant operation at the Technical Specification 2 allowed maxianna DEQ 131 concentration coupled with a maximum i gym primary to secondary leak. No additional fuel failures occur as a result of 'this event. St Lucie Plant has neither operated with primary to secondary leakage nor with DEQ 131 concentrations at the l Technical Specificatien limit. The lifting of V3439 during shutdown cooling operation following such an event would not' involve consequences exceeding FSAR offsite doses. a T ^ i i o I 2 I j 4

_..m ) 09 @ 1995 @ 09R1,, e+ 4 ucle Res dent OffIco 407 461 @22 P.05 j LER: ASSES 8 MENT OF GAFETY CONSEQUENCES i h l $OURCES OFINFORMATION i ] 10 CPR 80.73(b)(3)-The LER shall contain' 'An assessmart of the safety consequences j and impiloation of the event. This casessmerd must include the availability of other eyelems or components that could have performed the same function as the components j and systems that failed during the event? l Statemente of Consideration for 10 CFR 50.73 (b)(3) (48FR33867, dated July 28, 1983)- Paragraph (b)(3) requires that the LER include a summary assessment of the 1 l artisand potential salsty consequences and implications of the event. This assessment may be based on the conditims existing et the time of the event. The evaluation must be carried out to the extent necessary to fully assess the safety consequences and safety j margins associated with the event. An assessment of the event under attemative conditions must be included if the incident would have been more severe (e.g., the plant would have been in a condition not analyzed in the Safety Analysis Report) under l reasonable and aedible altamative conditions, such as power level or operating modo.... i i NUREG 1022, Rev.1 Second Draft, " Event Report!ng Guidelines 10 CFR 50.71 and i i 50J3, pgs.114 and 115 "Give a summary assessment of the actual and potential safety consequences and implications of the event, including the basis for submitting the report. Evaluate the event to the extent necessary to fully assess the safety consequences and safety margins associated with the event. include an assessment cf the event under altemotive conditions if the incident would have l been more severe (e.g., the plant would have been in a condition not analyzed in its latest l 8AR) under reasonable and credible attemative conditions, such as a different operating l mode. For example, if an event occurred while the plant was et low power and the same i event could have occurred at full power, which would have resulted in considerably more serious consequences, this siternative condition should be assessed and the j consequenoss roported. [ Reasonable and credible attemative conditions may include normal plant operating conditions, potodial accident conditions, or additional component failures, depending on the event. Normal allemative operating conditions and off normal conditions expected to j occur during the life of the plant should tre considered. The intent of this section is to obtain the result of he considerations that are typical in the conduct of routine operations, such as event reviews, r>ot to requke extraordinary studies." ,,,wr .w 4 - -, +. ..-...nn n, a

& 20-1995 08r 09r41.,

Luc 1e Resid:nt Off1ce
407 461 4622 P.06 L

Othar Vendons of NUREG 1022, Sept.1983; Supp.1, Feb.1984; Supp. 2, Sept.1985; l Rev.1, Sept. itM -The other versions of NUREG 1022 are similar in wording, however, I several notable differences are prov!ded below-. i NUREG 1022 Rev.1, Septamtser 1991 "A conclusion about the actual or implied effect on public health and safety of the event may be included as part of the assessment, but is not required." 4 NUREG 1022, S@plement 2, September 1985 "A discussion of the safety i consequenose and imp 6 cations is required in at least a few sentences or a paragraph that ' is clearty identifiable as a safety essessment. This discussion should indicate: (a) all of the safety consequences of the event including an assessment of the consequences had it been poselble for the event to occur under a more severe set of initial conditions, or (b) if there were no safety consequences or implications, it should explicitly state why there were none." NUREG 1022, Supplement 1, February 1984 - " Question 12.5 Does the term "reasonabis and crodele" conditions swally refer to normal plant operating conditions or j to potertial accident conditions? In addition, do we have to consider additional wrpew.t failures as " reasonable and credible" alternative conditions? Answer: "Rossonable and credible" altsmative conditions may include either normal plant J operating conditions, additional cenw.ee; failures, or potential sooident conditices depending on the event. Each licensee is required to assess its operating experience. in order to determine the safety significance and implications of operating events, consideration will normally be given to the implications of the event under normal altemotive operating conditions such as reactor power conditions expected to actually occur during the life of the olent. The intent of this section is to obtain the results of such i

09-29 '956 08 10AM e* Luc e Resaomt OHiC2 407 461 4622 P.07 l PAGE i cAIEtrLartant COWER azzaT l* l calculation Nos mat-irsw-es-oo1 Titles. '=*4*ata af eks offsita nsam can :ca.ne.. frenn t .b a. et t_ par en--a-, amit.# Valve. v3ast DRAFT l l l l l i 1 i O INITIAL ISSUE 1 Lw...+r-=vn "I Fubw wmp = we womu nyp y. { mesrvernman JFN Form 82A, Rev. 7/90 _____}

. _.. ~.. - -. ... ~..... -.. 09-28 '995 EiB110r41 '

- Lucie Ree: dent Offics
407 461 4622 P.00 i

FAEE 11 LIM or arraerTTu paans I m g m __y m ty.vu.es. oat - _ RET. _ a 38T FASS SSCTICW any pagg gggggg ggy o L LL O 0 LLL 0 1 0 2 0 3 4 0 0 5 n 6 ~ i f f t i i JPN Form $323, Rey, 779g,

09-E-1995 00819941 ~ e+ Luc 1e h atc:nt Offico .307 461 4622 P 09 1

Phos 111 j

1 TABLE OF G2ETENTS e C&LC9LATZoWI alUMSER

at-1F3W-GE.OO1 REY.

O j BIG 119E Z12LE PM cover sheet i List of Sifestive Pages 11

== Table of contents 111 1.0 Purpose / Scope 1 2.0 References 1 3.0 Methodology 1 4.0 Assumptions / Bases 2 S.O calculation 3 6.0 Resulta 5 1 ATTAdEDORIT MO. TITLE MUMBER OF PAGES i a Determination of avaiable curies 1 of "'I Deq. d d 4 o f i 4 A JPN Form #82c, new. 1/90

~... - -. e-A -1536 06:1.41 st Lucie Res tornt OH ace 1 407 461 4622 P.;0 N W E0. Det-ir.2m-e s-noi RST. O SERET No. 1 REF. I Puepose/seope The purpose of this calculation is to estimate the Thyacid Does (CDB) at the Emelusion Area soundary (EAB) from the valve leakage during a postulated accident. The 543 samves as a reference distance for comparison to other accident aamiyees and or Part 100 limite. The Thyroid Does is of conoorn because the leaked fluid wenld.be from the containment sump and, therfore, contain very listle noble pas. II amferomeos 1. St. Lucie Unit 1 FSAR Amendment 14. i 2. Reg-Guide 1.4, Rev 2, " Assumptions used for Evaluating the Potential Radiological coneespaeaces of a Loss of coolant Accident for Pressurised water Rosatore" i 3. 3FA-400, October 1991, " Manus 1 of Protective Action Guides sad Protective actions for Nuclear Zaeidente" 4. NUR30/CR - 1465

Accident source Terms for Light-Water Wuclear Power Plents"; Draft for comment

{ 1 111 Methodelegy The Thyroid Dose at the EAR will be determined using PSAR styled calculations and assumptions. Then, more realistic fastere and aseemptions will be applied. This will establish the boundaries of the potential doses. The process involves detesmining the enount of iodine released from the leakage to the Reactor Auxillary neilding (RAS), them to the environment, estimating the ' downwind' concentrations and finally seoversion to a dese. The remaLnder of the process will be applying, i sequentially, realistic factors and assumptions, and demonstrating their effect on the doses. i JP98 FCest 83 Rw. 3/

06-Ac. #J @ 1.m 5t u.,c ie Res tornt Lit t ice . =07 aoi oil F. ;; mtJarrawrgg 3g, pat.ir.ru.es. cot RST. A sESpr HO. 2 mar. IT Aaeumptions/Sases 58W 1 1. Accident type : samil Break Loca (SBLOCA) p' This accident type is moeded to maintain primary pressure sufficiently high to cause the relief valve to open upen initiation of shut-down cooling (80C). 2. The SB %$v1 a 1% cladding failure 3. Primary systes make-up estring the 3314cA is from NFSI at a flow of 300 gpe; NPs! flow is constaat until entering SDC, at which time ESPI is shut off. 4. Enter sec at 12, or 18, or 24 hours after start of 331mcn. Upon l initiating 50C, NFSI is secured, plant switches to recirculation mode and LPSI pump starts. i 5. Relief valve leak rate is 40 gpas; duration of 8 hours, then the valve ressats - leak stops. Volume of susap water leaked = 19200 gallons. f 6. 4000 gallons of the leaked water is ' trapped' in the pipe tunnoi and secociated drains; the remainder (15200 gallons) leaks to the RAS floor. l 7. Radiciodines do not Cacay or deposit once they've been released from the plants this is a conservative simplification. l i i i ) I

- 4 '#M JM reest 83 Aev. 3/t

- ~.. Dir-26-J f.; 44:51W1 Et L.ucie Res sesnt OH ice ' 407 461 4622 P.12 o l gagm EskEI M 30.- ear ip.sm.ta.nas 337. n SEEEE ED. 2 1 EEP. a Y celeslation A) curies of '88I Dog available From cap for release (14 say; (see Attachment 1): 9.4 3+4 9 12 hr. 9.04 3+4 4 is kr 8.72 5+4 0 24 hr. 2 as 1.4 p.esumes that 50 4 beoems airborner therefore, 504 is in the sump water - the curies available to be leaked are e t 4.7 34 4 9 12 hr. 4.52 5+4 8 18 hr 4.36 3+4 6 24 hr. B) volume of water in sump and primary system : Initial acs volume (65000 gallone) + usF2 injected (300 ype for 12,18,24 hr) water volume (thousands of ga12cas) 281 0 12 hr 389 9 18 hr 432 9 26 he c) Coseentration, ci

I Dog / allon j

0.167 9 12 hr 0.116 e 18 hr 0.088 e 24 hr D) Curies leaked to ras floor (conc, ci/ gal

15200 gal) 2.54 3+3 ci 0 12 hr 1.76 3+3 ci 0 18 hr 1.33 B+3 C1 0 24 hr F) Fraction of leaked Iodine that becomes airborne (curies leaked a partition factor); partition facter = 0.1 from the seTR analysis.

This becemos the curies released to the environment. 254 ci 0 12 hr 176 CL 6 18 hr 133 Ci e 24 hr

0) Downwind eoneentration (release rate, C1/sec

X/9, sec/m3).

X/9 at the SA5 = 8.55 E-5 ses/m' hoouse release duration = 1 hourt exposure duration equals release duration, so the time will null in the next step. (remester that the time, in hours, is hours poet trip when the release began) s 6.033-6 ci/m' 912 hr 4.185-6 Ci/m' S 18 hr 3.163-4 ci/m 4 24 hr I) Convert concentration to Dose (ci/ms

1.3E+6 Rem /hr per C1/m )

8 7.84 Rom 9 12 hr 5.43 Ram 9 18 hr 4.11 Ram e 24 hr At this point, following an rama styled analysis, the thyroid does eenooguences raange from 7.8 to 4.1 Rom for the event. The following page of eatenlations apply the more realistic assumptions and factors. i NAS-4. JFil gyg yggg gy g, p ;

w ' *m1 ^7: ocie hes aens.r,' ice ~~ ~~ ~ - ~ '-~ ~- ~ ^ ~ -~ ~~ 62-2

85, ~, _ _.

.g;; .,0,,,g ] .~ 4 catarraertur go. PSLelF h 9BeOO1 33y. o 3333T 30. A Y caleslaties eastiewed of RG 1 4, in that some (50%) of the Zodines

2) tfeing the concept released to the contairument atmosphere plate-out er eendonoe este surfaces, and applying the same plate-out fraotica to this seenarie reduces the iodines released, and the resultant dose, to see half that calculated in step N.

The caege would then be 3.9 to 2.06 men for the event. The FEAR discussion of BSF leakage indicates that for water J) temperatures less than 212 *F, a partition f actor of 0.01 Le more i appropriate. The water, having passed through the shutdeem heet i likely be less than 212 'F. This leads to a i emekanges would most ten-fold decrease in the iodines released, and the receitant dese, to one half that calculated in step 3. The range would then be 0.7s to 0.41 men for the event. 4 R) NUR34-1465 diseusses the sequestration of iodines by chemically i binding with cesium. Although this would increase the aetivity in the sump water a water leaked to the RAB floor, the reduced ambility of the iodine leads to a lesser (1/108) release than that estestated i in step M. Holding the partition f actor constant the offeet of Ce! formation would be 954 todines in water

5% available for partitioning 0.0475 not fraction of iodines available for partitioning.

This Lo, about l 1/102 of the 0.50 fraction of iodines avaliable for partitieming aesweed in 30 1.4. 2 I l l 1 i N i. ) teh8-4.JW g g yg,,,g3 g,,, 37

~ ~.?_i..:. a.,,.

c. _ :.. A ; 2 6.;,~, i s '

~~^. i~,

2. 1 T.~.,

~ f caussarras so.

==r.-irm-s a-na t asy. o susse no. _ m Bar. TI Esemits A) The doses calculated following the FSAR styled factors and assesytions range free 7.8 Res, for a release occuring 13 hours after reactor trip, to 4.1 Rom for a release occuring 24 hours after reester trip. 'more realistie' assumptions and

5) The doses osiculated using the0.04 Rea, for a release occuring 12 hours after fasters range from resetor trip, to 0.02 Rea for a release coeuring 24 hours after remotor trip.

doses calculated following the FSAR styled factors and C) The ase g ions are a small fraction of Part 100 limits. I I i . MM-4.JPW JM FostM 83 Aey, 3/

C:r-i:.:3 w. A. $. '.e. e -e r.. aer.t ^^. i -. :. - _ '~ ] ort.;e ememermegas so, set tysm.es-nei anv. o SEEST No. l j REF. AT2aOEMENT 1 DeterminatLon of Available curies of *I Deq Factor (Wf) gag Weight 9 T=0 EFh=400 1,3 mue1L6e T1/2, Mr cap Ci DCFe 2-131 1.93 B+02 8.86 E+06 1.30 E+06 1.00 E+00 1-133 2.26 E+00 1.82 3+O6 7.70 E+03 5.92 E-03 2-133 2.03 3+01 7.83 E+06 2.20 E+05 1,69 E-01 1-134 S.67 E-01 2.10 5+06 1.30 E+03 1.00 E-03 1-135 6.68 3+00 4.34 E+06 3.80 E+04 2.92 E-02 (1) Wf a DCF Iodine-zw + DCF Iodine-131 j 12 Bours post trip a Gap c1 Osp x Wf I-131 0.49 3+06 S.49 E+06 4 I-132 4.59 3+04 2.72 E+02 I-133 5.20 2+06 S.40 E+05 DBQ 2-131 I-134 1.43 E+02 1.43 E-01 From it Oap 2-135 1 25 E+06 3.65 E+04 9.40 s+06 e.01 = 9.40 s+04 18 pours post trip Sap ci Gap a vf 2-131 S.31 E+06 S.31 I+06 I-132 7.29 E+03 4.32 E+01 2-133 4.23 E+06 7.17 E+05 D3G I-131 I-134 1.18 5+00 1.18 E-03 From 1% Cap I-135 6.70 3+0S 1.96 3+04 9.04 E+06

.01 = 9.04 E+04 1

24 Ecure poet trip Gap cisap z Nf I-131 S.13 B+06 S.13 E+06 I-132 1.16 s+03 6.85 B+00 1-133 3.45 B+06 S.84 E+05 DEQ 2-131 1-134 9 75 E-03 9.75 E-06 From 16 Osp 1-135 3.40 3+05 1.05 E+04 8.72 E+06

.01 = 8.72 E+04 l

JN FCIDI 83 Rev. W9 Mh8-4 *IN TOTAL P.15 1

Proposed Violation B 10 CFR 50,-Appendix B, Criterion XVI, " Corrective Actions," requires, in i part,-that measures be established to assure that conditions adverse to quality are promptly identified and. corrected. t Contrary to the above, prompt corrective action was not taken in the case of St.'Lucie Action Requests which reported anomalous relief valve behavior and which were initiated on February 20, March 2, and March 10, 1995. The failure'to take prompt corrective action for these conditions led to a repetition of the anomalous behavior on August 10, 1995, when Unit I relief valve V-3439 li'ted and failed to-reseat without operator intervention. The subject ev.. ' resulted in approximately 4000 gallons of reactor coolant accumulating in the Unit I pipe tunnel. This is a Severity Level III violation (Supplement I). P f I ) I

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V Notice of Violation 10 CFR 50.54(m)(2)(iii) requires', in part, that, for a nuclear power unit in an operational mode other than cold shutdown or refueling,'a person holding a senior reactor. operator license shall be present in the control room at all times. 10 CFR 50.54(m)(2)(iii) requires, in part, that, for a nuclear power unit in an operational mode other than cold shutdown or refueling, a licensed operator or senior operator shall be present at the controls at all times. Contrary to the above: 1. On January 22, 1996, a boron dilution event occurred at St. Lucie Unit I which demonstrated that the senior reactor operator in the control room was not aware of the dilution in progress and, therefore, was incapable of providing the oversight function required by the subject rule. 2. On January 22, 1996, a boron dilution event occurred at St. Lucie Unit I which demonstrated that the licensed operator at the controls was not aware of the dilution in progress and, therefore, could not and did not continuously monitor plant instrumentation or properly manipulate controls associated with the dilution to prevent the reactor from exceeding 100% power, thus failing to satisfy the requirements of the subject rule. This is a Severity Level III Violation (Supplement I) ,p Of '% e i 1 4 h

1 )

Background

On' July 11, 1983, the NRC published in Federal Register (48 FR 3k611) the final rule 10 CFR - 50.54(m)(2)(111) which amended NRC regulations to require licensees of nuclear power units to provide a minimum number of licensed operators and senior operators on shift at all times to respond to normal and emergency conditions. 10 CFR 50.54(m)(2)(iii) states "When a nuglear power unit is in an operational mode:other than cold shutdown or refueling, as defined by the unit's technical specifications, each licensee shall have a person holding a senior operator license for the nuclear power unit in the control room at all times.. In addition to this senior operator, for each fueled nuclear power unit, a licensed operator or senior operator shall be present at the controls at all times." In the Statements of Consideration: The requirement for.a senior operator's conti':ua;:s presence in the control room would assure that: ' (1) _A person is available who can provide the oversight function of the supervisor so that the probability of correctly detecting abnormal events early enough to mitigate potential adverse consequences might be increased: (2) the senior operator in the control room is aware of plant conditions prior to and resulting from an abnormal event so that the senior operator will be able to use extra experience, training and knowledge to act promptly to mitigate that event; and (3) the reactor operator is able to direct attention to performing the immediate actions necessary to mitigate that event rather than having to brief the senior operator about.the background of that event if that person were absent from the control room. The presence of a senior operator...will also increase the probability of correctly detecting abnormal events and human error early enough to mitigate. potential consequences of any accident. For the Licensed Operator: The requirement that an operator be at the controls..will assure that plant instrumentation is continuously monitored and that controls are properly manipulated.

s. ENFORCEMENT ACTION WORKSHEET i (ST LUCIE OVERDILUTION EVENT] ) 4 PREPARED BY: R. Schin DATE: February 5, 1996 This Notice has been reviewed by the Branch Chief or Divis Director and-each violation includes the appropriate level of s. f as to how and when the requirement was violated. A v 51gnature 1 Facility: St. Lucie Unit (s): 1 Docket Nos: 50-335 License Nos: DPR-67 Inspection Report No: 50-335,389/96-01 Inspection Dates: January 26-30, 1996 Lead Inspector: R. Schin 1. Brief Gummary of Inspection Findings: Concern with operator attentiveness related to a reactivity addition event, and related operator violations of procedures: Operators failed to stop dilution when the proper amount had been a. added. l b. There was inadequate watch turnover for the operator at the controls during dilution. Operators failed to follow the Conduct of Operations procedure in c. performing the dilution procedure (lack of strict / verbatim 1 compliance). d. Operators failed to adequately report the event to licensee management. l Also, operators exceeded the steady state licensed power limit of 2700 megawatts thermal (100% power). In addition, the licensee nade a change to the procedures as described in the SAR without a 10 CFR 50.59 safety evaluation. See the attached draft NOV, General Description of Event, Detailed Sequence of Events, Summary of Draft Preliminary Inspection Findings, Control Room Diagram, CVCS Charging System Diagram, Procedures, and FSAR. PROPOSED ENFORCEMENT ACTION - NOT FOR PUBUC DISCLOSURE WITHOUT THE APPROVAL OF THE DIRECTOR. OE 2 t w .+ m

f ... i ~ f r : a 2 i .2.' Analysis of Root Cause:.. ] Operator inattentiveness to reactivity addition. I ji 3.- Basis for Severity Level (Safety Significance):= V Inattentiveness to duty on the part of licensee personnel, while 1 I.C.3 adding reactivity to the reactor, and j 4 I.C.7 A breakdown in the control of licensed activities involving a number of violations that are'related that collectively represent a significant lack of attention or carelessness toward licensed responsibilities.. 4. Identify Previous Escalated Action Within 2 Years or 2 Inspections? !x EA 95-180 (EEI 95-16-01); LTOP'inoperability due to PORV failure l Event date 8/9/95 5. Identification Credit? No L Identified through an event. The licensee initiated an In-House Event Report and gave.a copy to the NRC resident inspector promptly after the event. The event occurred at approximately 0220 on January 22, 1996. l Missed opportunities: a. In response to SOER 94-02, dated September 1994, which described a I similar Turkey Point overdilution event and several inadvertent dilution events at other utilities, the licensee reviewed the St. I' Lucie operating procedures related to dilution and concluded that no changes were needed. This was a missed opportunity to strengthen operating procedures to prevent the 1/22/96 overdilution event. b. The Unit 2 dilution procedure had been changed in December 1995, but not the Unit 1 procedure, to more accurately describe dilution the way the operators had performed it for years (in manual and direct to the charging pumps). During the event, manual dilution could not be accomplished by using the Unit 1 procedure in compliance with the Conduct of Operations (strict / verbatim compliance). 6. Corrective Action Credit? Yes The licensee initiated an In-House Event Report summarizing the event and began distribution of that report'within about four hours after the event. The licensee also immediately removed the reactor operator who had initiated the event from licensed duties, promptly issued a Night . Order and conducted training on the event with operators on each shift; revised the Unit 1 procedure for dilution so that manual dilution could be performed by strict compliance to the procedure steps; revised the . Conduct of Operations procedure to require the RO to get prior approval PROPOSED ENFORCEMENT ACTION - NOT FOR PUBLIC DISCLOSURE WITHOUT THE APPROVAL OF THE DIRECTOR, OE F t + r + =r 4.

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i 3 3 { from the SRO for dilution /boration, the SRO to directly supervise [ dilution /boration, no RO or SRO turnover during dilution /boration, and RTGB walkdown prior to RO or SRO short term relief; and initiated ~further review of the event. l Weaknesses in the licensee's corrective actions included: + a.- Potential VIO of 10 CFR 50.59: The revised procedure (after the event) did not support the FSAR Chapter 15 accident analysis assumptions on how dilution was performed. The revised procedure described dilution in manual (with no automatic shutoff) and directly to the suction of the charging pumps. The FSAR assumed i s e h ~ dilution in automatic (with an automatic shutoff) an y I 4 gi, \\ before going to the suction of the charging pumps and result in a lower rate of reactivity addition). The licensee had not performed a safety analysis of this difference and had not revised f) the procedure and/or FSAR to make them agree. b. The revised procedure for manual dilution (after the event) did not require the operator at the controls to remain by the dilution controls and to closely monitor the dilution during a manual i dilution with no automatic shutoff. The licensee initial investigation of the event was not thorough c. j-in that it concluded that maximum reactor power was 100.2%. Subsequent review by the NRC and licensee found that maximum reactor power was approximately 101.18%. l 7. Candidate For Discretion?-[See attached list) Yes - potential i t escalation, 4 During the last year, the licensee's performance in Operations has declined from SALP 1 to SALP 2 (predecisional). There have been several operator violations of procedures that are, in part, related to the' e L current violation: ~ 1) VIO 335/94-22-02, " Improper Modification of Control Room Logs", November 25, 1994 4 i 2) NCV 335/95-07-01, " Failure to Follow Shutdown Cooling Operating Procedures", April 19, 1995 3) VIO 335/95-15-01, " Failure to Follow Procedures and Block MSIS Actuation", October 16, 1995 4) VIO 335/95-15-02, " Failure to Follow Procedures during RCP Seal restaging", October 16, 1995 4 5) VIO 335/95-15-03, " Failure to Follow Procedure and Document abnormal valve position in the Valve Switch Deviation Log", October 16, 1995 PROPOSED ENFORCEMENT ACTION NOT FOR PUBLIC DISCLOSURE l WITHOUT THE APPROVAL OF THE DIRECTOR, OE i r -. e. .-r- ,r- ---r-v,

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~~ 4 6) VIO 335/95-15-04, " Failure to Follow Procedures during Alignment of Shutdown Cooling System", October 16, 1995 7) VIO 38v/95-18-01, " Failure to follow Procedures and Maintain Current and Valid Log Entries in the Rack Key Log and Valve Switch Deviation Log", November 27, 1995 8) VIO 389/95-21-02, " Failure to Follow the Equipment clearance Order Procedure and Require Independent Verification of a TS Related. ] Component", December 8, 1995 1 All of the above VIO/NCVs involved licensed operators with a licensee corrective action comitment to strict adherence to procedures. 8. Is A Predecisional Enforcement Conference Necessary? Yes Why: There is substantial interest in this event and in the NRC message to the licensee and to the industry. The message for this enforcement action should be that operators must treat Dilution /Boration as seriously as control rod manipulations. Also, that unusual operations events must be transmitted promptly to management. If yes, should OE or OGC attend? Yes Should conference be closed? No 9. Non-Routine Issues / Additional Information: 10. This Action is Consistent With the Following Action (or Enforcement Guidance) Previously Issued: I.C.3 Basis for Inconsistency With Previously Issued Actions (Guidance) 11. Regulatory Message: The message for this enforcement action should be that operators must treat Dilution /Boration as seriously as control rod manipulations. Also, that unusual operations events must be transmitted promptly to management. 12. Recommended Enforcement Action: SLIII with CP 13. This Case Meets the Critaria for a Delegated Case. No 14. Should This Action Be Sent to OE For Full Review? No, informal review. 15. Regional Counsel Review To be determined at a later date. No Legal Objection Dated: PROPOSED ENFORCEMENT ACTION - NOT FOR PUBLIC DISCLOSURE WITHOUT THE APPROVAL OF THE DIRECTOR, OE

p 1 5 4 16. Exempt from Timeliness:- No 1 Basis for Exemption: -l Enforcement Coordinator: DATE: I l l 1 i 1 l l I l I J r I-i l 1 ) I ) l l l l l PROPOSED ENFORCEMENT ACTION - NOT FOR PUBLIC D!SCLOSURE ) WITHOUT THE APPROVAL OF THE DIRECTOR, Ou 1 l i t

6 f ENFORCEMENT ACTION WORKSHEET - ISSUES TO CONSIDER FOR DISCRETION' L O Problems categorized at Severity. Level I or II. p O . Case involves overexposure or release of radiological material in excess of NRC requirements. O Case involves particular1y. poor licensee performance. '1 l 0 Case (may) involve willfulness. Information should be included to address whether or not the region has had discussions with 01 regarding I the case, whether or not the matter has been formally referred to 01, ) 2 and whether or not 01 intends to initiate an investigation. A description, as. applicable, of the facts and circumstances that address l-l the aspects of negligence, careless disregard, wi11 fulness, and/or l i management involvement should also be included. S Current violation is directly repetitive of an earlier violation (in i part). O Excessive duration of a problem resulted in a substantial increase in risk. O Licensee made a conscious decision to be in noncompliance in order to obtain an economic benefit. 4 O Cases involves the loss of a source. (Note whether the licensee self-identified and reported the loss to the NRC.) i O Licensee's sustained performance has been particularly good. O Discretion J.nid be exercised by escalating or mitigating to ensure that the proposed civil penalty reflects the NRC's concern regarding the violation at issue and that it conveys the appropriate message to the licensee. Explain. l PROPOSED ENFORCEMENT ACTION + NOT FOR PUBUC DISCLOSURE WVTHOUT THE APPROVAL OF THE DIRECTOR. OE 9 . ~.

. ~. ll' ) 7 REFERENCE DOCUMENT CHECKLIST 1 ()- NRC Inspection Report or other documentation of the case: NRC Inspection Report Nos.: () Licensee reports: () Applicable Tech Specs along with bases: i [x] Applicable license conditions [x] Applicable licensee procedures or extracts [] Copy of discrepant licensee documentation referred to in citations such as NCR, inspection record, or test results i [x] Extracts of pertinent FSAR or Updated FSAR sections for citations involving 10 CFR 50.59 or systems operability () Referenced ORDERS or Confirmation of Action Letters ( )- Current SALP report summary and applicable report sections i () Other miscellaneous documents (List): 4 d PROPOSED ENFORCEMENT ACTION - NOT FOR PUBLIC DISCLOSURE WITHOUT THE APPROVAL OF THE DIRECTOR, OE

.~ 8 4 fBGP0 SED VIOLATI0ff j A. Technical Specification (TS) 6.0.1.a required that written procedures be established, implemented, and maintained covering the activities recommended in Appendix A of Regulatory Guide 1.33, Rev 2, February 1978. Appendix A includes operating procedures for the chemical and volume contrci system and administrative procedures for relief turnover, procedural adherence, and authorities and responsibilities for safe-operation. Operating Procedure No. 1-0250020, Boron Concentration Control - Normal Control, Rev 35,' step 8.5.14 required that operators monitor the water flow totaliter and close valve V2525 after the desired volume was added during a boron concentration dilution using the direct path to the o charging pump suction. Administrative Procedure No. 0010120, Conduct of Operations, Rev 79, ' Appendix D, Crew Relief / Shift Turnover, required that, for short ters i watchstander relief, a turnover be conducted including: general watchstation status, off-normal conditions, and tests in progress. Administrative Procedure No. 0010120, Appendix M, Procedural Compliance and Implementation, required that controlled procedures be implemented 3 and complied with in accordance with the instructions provided in QI 5-PR/PSL-1. Procedure QI 5-PR/PSL-1, Preparation, Revision, Review /Apprrwal of Procedures, Rev 67, Section 5.13.2, stated that all 2 procedures shall be strictly adhered to and identified that. Operating i Procedure 1-0250020 was not considered " skill of the trade" and was not to be performed from memory without referring to the procedure. l Administrative Procedure No. 0010120,' Appendix E, Notification of Operations Supervisor /FPL Management, required prompt verbal ~ notification of the Operations Supervisor for unplanned reactivity changes. i Contrary to the above: 1. On January 22, 1996, at approximately 2:30 a.m., Unit 1 operators failed to close valve V2525 after the desired volume was added during a boron concentration dilution using the direct path to the charging pump. Operators had desired to add between 25 and 40 gallons of primary makeup water, but failed to stop the dilution until approximately 400 gallons were added. During this time, the i temporary rel kf operator at the controls was unaware that a boron concentration dilution was in progress, which resulted in an unmonitored reactivity addition. The SRO and other operators in-the control room were also unaware that a reactivity addition was in progress. 2. On January 22, 1996, at approximately 2:30 a.m., the Unit 1 operator at the controls conducted a short term watchstander relief with an inadequate turnover in that it failed to inchde general watchstation status and conditions including that a boron PROPOSED ENFORCEMENT ACTION NOT FOR PUBLIC DISCLOSURE 1 WITHOUT THE APPROVAL OF THE DIRECTOR. OE I = - L

9 concentration dilution was in progress..As a result, the' relief operator at the controls was unaware that a boron concentration dilution was in progress and failed to adequately monitor and control the dilution. 3. On January 22, 1996, at approximately 2:30 a.m., operators performed Operating Procedure 1-0250020 from memory, without referring to the procedure, and without strictly adhering to the >rocedure. At the time, the procedure was written such that the >oron concentration dilution that was performed could not have been performed by strictly adhering to the procedure. 4. On January 22, 1996, between 2:30 a.m. and 7:20 a.m., operators f&iled to give prompt verbal notification to the Operations Supervisor for unplanned reactivity changes that had occurred. B. -The Facility Operating License f

t. Lucie Unit I a orizes the licensee to operate the faci y at a steady state actor core power i

level not in excess on 27 megawatts thermal ). TS 1.25 defines rated thermal power t a total reactor c heat transfer rate to the reactor coolant of 00 Mt. TS 1.33 d nes thermal power to be the total reactor t transfer rate to reactor coolant. Contrary o the above, on Jan y 22, 1996, between approximately 2:20 ard 3-a.m., the reactor re thermal power level limit of 2700 W (1 ) was exceeded, du o operator inattentiveness. 100% react wer was exceeded f approximately 70 minutes. Also, 101% re or power was exceede or approximately 4 minutes and a peak re or power i of approximat 101.18% was reached. C. 10 CFR 50.59 allows the licensee to make changes to the procedures as described in the Safety Analysis Report (SAR), witt'out prior Commission l approval, unless.the change involves, in part, an.tnreviewed safety question. A proposed change shall be deemed to involve an unreviewed safety question if, in part, the probability of occurrence of an accident important to safety previously evaluated in the SAR may be increased. The licensee shall maintain records of changes in procedures made pursuant to this section, to the extent that they constitute changes in procedures as described in the SAR. These records must include a written safety evaluation which provides a basis for the determination that the change does not involve an unreviewed safety question. Contrary to the above, on January 23, 1996, the licensee made a change in Unit 1 procedures as described in the SAR and the records for that change did not include a written safety evaluation. Temporary Change 1-96-017 to procedure 1-0250020, Boron Concentration Control - Normal Operation, Rev. 35, added instructions for dilution in manual and directly to the suction of the charging pumps. However, the SAR, paragraph 15.2.4.1, states that boron dilution is conducted under strict administrative procedures which limit the rate and magnitude of any required change in boron concentration. Further, the SAR states that boron dilution must be conducted in automatic (such that when the PROPOSED ENFORCEMENT ACTION - NOT FOR PUBUC DISCLOSURE WITHOUT THE APPROVAL OF THE DIRECTOR, OE + _u___.__________.__.____.______

e 10-i

specific amount has been injected, the demineralized water control valve P

is shut automatically)- and describes introduction into the volume a . The SAR concludes that, in part, because of the control tank (VCT)d, the probability of a sustained or. erroneous procedures involve dilution is very low. The licensee implemented Temporary Change 1, l 017 on January 23, 1096, without a written safety evaluation. ? I 1 General Descriotion of the Event At approximately 0225 on January 22,1996,. the Unit I control board Reactor d f ' Controls Operator (RCO) began a manual dilution to the RCS by aligning primary i makeup water (demineralized water) directly to the. suction of the IB Charging Pump.. Moments after beginning the dilution, the board RCO responded to a secondary plant annunciator and then saw the desk RCO return from the kitchen. He requested that the desk RCO relieve his so that he could prepare his' lunch. During the turnover,-there was no discussion of the dilution in progress. Following the turnover, the relief operator at the controls and the Nuclear Plant Supervisor (NPS),.who was at the desk RCO station, were not aware that a I dilution was in progress. The original board RCO returned between 5-10 minutes later and immediately recognized his error. He informed the other RCO of the overdilution, which was overheard by the NPS, and stopped the dilution. 1he NPS directed the ANPS take charge and.begin a manual boration. Unit I entered 2-hour TS LCO Action Statement 3.2.5 for T, greater than 549'F. The t maximum T, obtained was 549.9'F and the maximum reactor power was 101.18%. T, was above the TS limit of 549'F for approximately 50 minutes and reactor power 5 was above 100% for approximately 70 minutes. The TS LCO Action Statement for l T, was not exceeded and the guidance of the Jordan memorandum on maximum - reactor power was not exceeded. The operators did not verbally notify plant management or the NRC of this event. Detailed Seouence of Events (Note that the times for the sequence of events are approximate and only relevant events are mentioned) 1/21/96 11:00 p.m. Incoming mid shift assumed Unit I responsibility with the Unit at 100% power, 870 MWe, Tavg at 575 degrees F, Thot at 600 degrees F, Tcold at 548.9 degrees F. RCS Boron concentration at 376 ppm, Xe worth at -2722 pcm, all CEAs fully withdrawn and manual, and no Technical Specification action statements in effect. Major evolution planned for the shift was to place the waste gas system in service. Further, there was an annunciator alarm E-9 associated with circulating water pump-lube water supply strainer delta P high that was intermittently coming in due to a failed pressure switch. 11:45 p.m. Board RCO reset to zero the primary water (to VCT or charging pump) flow totalizer in preparation for inventory balance (RCS PROPOSED ENFORCEMENT ACTION NOT FOR PUBLIC DISCLOSURE WITHOUT THE APPROVAL OF THE DIRECTOR. OE ,~.,

11 l' leak rate calculation) 11:00 p.m.- 2:00 a.m The board RCO recalled performing at least two dilutions of approximately 35 gallons each between 11:00 p.m. and 2:20 a.m. without resetting the totalizer. l 1/22/96 xx:xx a.m NPS arrived in Unit I control room to gather data for morning i report meeting and sat near desk behind control boards. STA was also present near NPS xx:xx a.m. ANPS turned over control room senior reactor operator l responsibility to NPS and proceeded to the kitchen to prepare breakfast I xx:xx a.m. Desk RCO left control room to go to the kitchen i 2:20 a.m. Normal cantinued fuel burnup resulted in indicated Tc of 548.7 degrees F on RTGB-104 (digital meter). At this point the board RCO decided to restore Tc to maximum allowable program value of 549.0 degrees F. xx:xx a.m. Desk RCO arrived in the control room with his meal 2:25 a.m. The board operator began a manual dilution by aligning primary water to the suction of the charging pumps by opening FCV-2210X and A0V-2525. The flow rate was approximately 44 gpm. 2:26 a.m. Annunciator E-9 associated with circulating water lube water supply strainer high delta P was received. The board RCO walked to the panel and acknowledged the annunciator. 2:27 a.m. After acknowledging the annunciator, the board operator decided to proceed to the kitchen to prepare his meal. The board operator conveyed this to the desk operetor and requested that he take over the board operator responsibilities. However, he did not mention the ongoing dilution. The desk operator got up and proceed to the board in the vicinity of panel 103. The original board operator proceeded to the kitchen and started preparing his meal on a skillet that had been kept warm. At this time the NPS and the STA were in the control room at the desk area. The NWE had been in and out of the control room throughout the shift. The relief operator at the controls, NPS, STA, and NWE were not aware of the ongoing dilution. 2:35 a.m. The original board operator returned from the kitchen with his meal. Upon approaching the board, he realized that he had left the control room with an ongoing manual dilution. He exclaimed that he had overdiluted and immediately began securing the dilution. The desk operator questioned how much water was added and the board operator noted from the totalizer that approximately PROPOSED ENFORCEMENT ACTION NOT FOR PUBUC DISCLOSURE WITHOUT THE APPROVAL OF THE DIRECTOR, OE

12 t 400 gallons was added. 2:35 a.m. Soon after, annunciator M-16 associated with RCP controlled bleedoff pressure high was received. At this point the Tc was noted by the desk operator to be 549.6 degrees F. Entry into two-hour action statement associated with Technical Specification 3.2.5, DNB paramenters.was recognized and later logged. 2:36 a.m. The desk operator directed the board operator to initiate boration to restore Tc to program. The NWE calculated the amount of borated water to be added to the RCS. The NPS asked the desk operator. to notify the unit ANPS to come to the control room. i x:xx a.m. ' ANPS walked into the control room. 2:41 a.m.- Tc reached the highest noted value of 549.9 degrees F. MWe reached 875 and indicated reactor power was approximately 101.2% x:xx a.m. Operator secured boration. 3:14 a.m. Tc noted below 549.0 degrees F. Technical Specification action statement was exited. x:xx a.m. STA initiated an In-House Event Report and notified HPES personnel [ by telephone. 5:45 a.m.- 6:00 a.m. Shift turnover occurred. It appears that the dilution event was not discussed with the oncoming shift. 6:25 a.m. In-House Event Report was E-mailed to standard distribution, which included plant management, by the STA. 6:30'a.m. The Operations Manager toured the control room but was not informed of the over dilution event. i 7:20 a.m. The Operations Manager read the control room logs (in his office by computer) and questioned the log entry associated with the ) overdilution event. l 7:30 a.m. Licensee initiated a detailed investigation associated with the event. 7:45 a.m. Senior Plant management was notified of the event during the morning meeting. 10:00 a.m. NRC resident inspector was given the event report that was initiated associated with the event. l PROPOSED ENFORCEMENT ACTION - NOT FOR PUBUC DISCLOSURE l WITHOUT THE APPROVAL OF THE DIRECTOR, OE L

13 'ST. LUCIE ONSITE EVENT FOLLOWUP INSPECTION OVERDILUTION EVENT of 1/22/96 (Exit was at 10:00 a.m. on 1/30/96) Inspectors: R. Schin, S. Sandin,,B. Desai Summary of draft creliminary findinas: 1. Magnitude of power and temperature excursion a. Reactor power Peak reactor power was approximately 101.18% 100% power was exceeded for approximately 70 minutes 101% power was exceeded for approximately 4 minutes The event was within the accident analysis The guidelines of the Jordan memo were not exceeded b. Cold leg temperature Peak Tc was approximately 549.9 degrees F T5 limit of 549 was exceeded for approximately 50 minutes TS 2-hr. action statement was properly entered and was not exceeded 2. Concern with operator attentiveness - Potential / Apparent VIO of procedures (Enforcement panel form completed on this issue): a. Operators failed to stop dilution when the proper amount had been added. f b. There was inadequate watch turnover for the operator at the controls during dilution. c. Operators failed to follow the Conduct of Operations procedure in performing the dilution procedure. d. Operators failed to adequately report the event to licensee management. 3. Concern with control room command and control - Weakness a. The SRO in the control room was not aware of the dilution in progress. b. The board operator did not inform the SR0 of dilution - this was a general practice at the site and not required by procedures. c. The watchstander board was not maintained (on Saturday). d. The SR0 in the control room was allowed to be in the ANPS office for unlimited time, out of sight of control room activities and out.of hearing range of almost all control room activities except annunciator alarms (not applicable.during this event). PROPOSED ENFORCEMENT ACTION - NOT FOR PUBUC DISCLOSURE WITHOUT THE APPROVAL OF THE DIRECTOR, OE

(_ Lr 4. Weaknesses'in-procedures a. The Unit 2 dilution procedure had been changed, but not the Unit 1 procedure, to more accurately describe dilution the way the operators had performed.it for years (in manual and direct to the charging pumps). During the event, manual dilution could not be accomplished by using the Unit 1 procedure in compliance with the Conduct of Operations.. b. Procedures and practices for dilution (before and during the event) did not support the FSAR accident analysis assumptions on how dilution was performed. The FSAR assumed dilution in automatic and to the VCT. c. Procedures for dilution (before and during the event) did not require the operator at the controls to remain by the dilution controls and to closely monitor the dilution during a manual dilution with no automatic shutoff. 5. Weaknesses in corrective action a. Potential VIO of 10 CFR 50.59: Revised procedure (after the event) did not support the FSAR Chapter 15 accident analysis assumptions on how dilution was performed. The FSAR assumed dilution in automatic and to the VCT. b. The revised procedure for manual dilution (after the event) did ,not require the operator at the controls to remain by the dilution controls and to closely monitor the dilution during a manual dilution with no automatic shutoff. c. The licensee initial investigation of the event was not thorough in that it concluded that maximum reactor power was 100.2%. Subsequent review by the NRC and licensee found that maximum i reactor power was approximately 101.18%. 6. Weakness in Operational Experience Feedback a. In response to SOER 94-02, dated September 1994, which described a similar Turkey Point overdilution event and several inadvertent dilution events at other utilities, the licensee reviewed the St. Lucie operating procedures related to dilution and concluded that no changes were needed. This was a missed opportunity to strengthen operating procedures to prevent the 1/22/96 overdilution event. 7. Other comments a. There was no clearly noticeable indication of dilution in progress..The dilution clicker was quiet (might not be heard from the desk area) and sounded identical to the nearby clickers that routinely made noise. j PROPOSED ENFORCEMENT ACTION - NOT FOR PUBLIC DISCLOSURE I WITHOUT THE APPROVAL OF THE DIRECTOR. OE -I

- -.. ~ E b. Operators routinely did not log reactivity additions; however, the licensee's Conduct of Operations procedure. stated that operators should log reactivity changes. l 4 LICENSEE DISSENTING C0fMENTS 1. The licensee had dissenting comments on item 5.a. above, the potential 2 violation of 10 CFR 50.59. The inspectors told the licensee at the exit that those dissenting comments would be included in the inspection report, for further review by NRC management. The dissenting comments, i from the engineering manager (Dan Denver) and the licensing manager (Ed Weinkas), included: a. The previous procedure allowed diluting in manual and directly to the suction of the charging pumps, and that had been the practice .for many years. Therefore, the temporary change on 1/23/96 (after 4 the event) did not change the method of dilution, but only i clarified a previously existing procedure and made it conform to " verbatim compliance" rules. The inspectors did n'ot disagree. In l fact, further review, as requested by the inspectors, found that i the first time the dilution procedure was changed to allow opening of valve 2525 (directly to the suction of the charging pumps) was in a change to rev. 2 of the procedure, in 1976, before the operating license was issued. b.' The design of the plant (piping, valves) always was such that , dilution in manual and directly to the suction of the charging pumps was possible. The inspectors did not disagree. c. The accident analysis assumed a worst case dilution event with demineralized water going directly to the suction of the charging pumps and thne charging pumps running. That would be three times the flowrate of this event and therefore that analysis boun<is this g event. The inspectors did not disagree. d. The FSAR Chapter 9 description of the Chemical and Volume Control i System did not prohibit dilution in manual and directly to the j suction of the charging pumps. The inspectors did not disagree. e. The automatic mode of dilution is less safe than the manual mode, in that there is more opportunity for a malfunction that could result in a maximum flowrate approaching the design limit. The inspectors did not comment on that position. f. The procedure change that first allowed dilution directly to the suction of the charging pumps was made before the operating i license was issued, therefore 10 CFR 50.59 did not apply to that change. The inspectors did not comment on that position. g. Since the rperating procedure that was in effect at the time the operating license was issued allowed dilution in manual and directly to the suction of the charging pumps, that method was PROPOSED ENFORCEMENT ACTION NOT FOR PUBUC DISCLOSURE WITHOUT THE APPROVAL OF THE DIRECTOR, OE .--~,

7.. 7 4 IfL 2 included in the. original licensing basis of the plant. The inspectors did not agree with that position. l h. After receiving these licensee' comments, the inspectors' concern remained unchanged: The Temporary Change of 1/23/96 (after the J event) described procedure steps for dilution in manual and directly to the suction of the charging pumps. That procedure was. different from the one described in the FSAR. The licensee's-procedure differed from the FSAR in that it allowed a faster rate of reactivity addition and without an automatic shutoff. The licensee had not performed a safety analysis of this difference and had not revised the procedure and/or FSAR to make them agree. i 1 i 2. The licensee also had a dissenting comment on item 5.c. above, the weakness in the licensee's initial investigation. The dissenting comment, from the Plant Manager (Jim Scarola), was: 4 a. The initial investigation, for the In-House Event Summary, was j done by the STA. Timeliness was more important than quality at that time. Subsequent more thorough review would be performed by l l the licensee. The inspectors acknowledged the licensee's comment. t 1 i L i F d 1 I l 2 PROPOSED ENFORCEMENT ACTION NOT FOR PUBLIC DISCLOSURE i WITHOUT THE APPROVAL OF THE DIRECTOR, OE

s ~ C. ~ n

Proposed Operator.

NOTICE OF VIOLATION \\ I Docket No. 55-1.icense No.0P- -EA(s) TBD During an NRC inspection cc. ducted on January 26-30, 1996, violations of NRC requirements were identified.- In accordance with the " General Statement'of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, the violations . are listed below:. j Technical Specification 6.8.1.a required that written procedures be established, implemented, and maintained covering the activities i recommended in Appendix A of Regulatory Guide 1.33, Rev 2, February 1978. Appendix A includes operating procedures for the chemical and i volume control system and administrative procedures for relief turnover, ' procedural adherence, and authorities and responsibilities for safe operation. !L Operating Procedure No. 1-0250020,-Boron Concentration Control.- Normal l Control, Rev. 35, step 8.5.14.rgquired that operators monitor the water flow totalizer and close valve V2525 after the desired volume was added l duringaboronconcentrationdilutionusigthedirectpathtothe charging pump suction. s Administrative Procedure No. 0010120, Conduct of Operations, Rev 79, Appendix D, Crew Relief / Shift Turnover, required that, for short term i watchstander relief, a turnover be conducted including: general s watchstation status, off-normal conditions, and tests in progress. l Administrative Procedure No. 0010120, Appendix M, Procedural Compliance and Implementation, required that controlled procedures be implemented and complied with'in accordance with the instructions provided in QI 5-PR/PSL-1, Preparation, Revisinn, Review / Approval of Procedures, Rev 67. Procedure QI 5-PR/PSL-1 Section 5.13.2, stated that all procedures shall i be strictly adhered to and specifically identified that Operating Procedure 1-0250020 was not considered " skill of the trade" and was not. to be performed from memory without referring to the procedure. Contrary to the above: i 1. On January 22, 1996, at approximately 2:30 a.m., the Unit 1 operator. failed to close valve V2525 after the desired volume was added during a boron concentration dilution using the direct path 'to the charging pump. The operator had desired to add between 25 and 40 gallons of primary makeup water,'but failed to stop the dilution until approxim'ately 400 gallons were added. During this time, the temporary relief operator at the controls was unaware that a boron concentration dilution was-in progress, which l resulted in an unmonitored reactivity addition. The SR0 and other . operators in the control room were also unaware that a reactivity addition was-in progress. i .., +

i 2. On January 22, 1996, at approximately 2:30 a.m., the Unit-1 t operator at the controls conducted a short term watchstander relief with an inadequate turnover in that he failed to include general watchstation status and conditions including that a boron concentration dilution was in progress. As a result, the relief operator at the controls was unaware that a boron concentration dilution was in progress and failed to adequately monitor and control the dilution. 3. On January 22, 1996, at approximately 2:30 a.m., the Unit 1 operator performed Operating Procedure 1-0250020 from memory, without referring to the procedure, and without strictly adhering to the procedure. At the time, the procedure was written such that the boron dilution that was performed could not have been performed by strictly adhering to the procedure. M These violations represent a Severity Level III p 1. r (Supplement ). ' Pursuant to the provisions of 10 CFR 2.201, ************* is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555 with a copy to the Regional Administrator, Region II, aw a qv to the NRC Resident Inspector at the facility that is the subject of this Notice, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a "Repl,y to a Notice of Violation" and should include for each violation: (1) the reason for the violation, or, if 3 contested, the basis for disputing the violation, (2) the corrective steps that have been taken and the results' achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved. Your response may reference or include previous docketed correspondence, if the correspondence adequately addresses the If an adequate reply is not received within the time required response. ( specified in this Notice, an order or a Demand for Information may be issued as to why the license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time. Under the authority of Section 181 of the Act, 42 U.S.C. 2232, this response shall be submitted under oath or affirmation. Because your response will be placed in the NRC Public Document Room (PDR), to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be placed in the PDR without redaction. However, if you find it necessary to include such information, you should clearly indicate the specific information that you desire not to be placed in the PDR, and provide the-legal basis to support your request for withholding the information from the public, t Dated at Atlanta, Georgia this ' day of Februay 1996 t A y

fb f, Page 91 of 174 ~ N; ST. LUCIE PLANT ADMINISTRATIVE PROCEDURE NO. 0010120, REVISION 79 CONDUCT OF OPERATIONS

0*N#[f b

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Page 1 of 20 FLORIDA POWER & LIGHT COMPANY ST. LUCIE UNIT 1 OPERATING PROCEDURE NO. 1-0250020 I p , se' REVISION 35 W f N l { l 8l 1.0 T_l, g.: '. ~. ! ri BORON CONCENTRATION CONTROL - NORMAL OPERATIOhrnec=ur4%dtiuNIDN 2.0 REVIEW AND APPROVAL: Reviewed by Plant Nuclear Safety Committee 5/30 1974 Approved by _ K.N. Harris Plant General Manager 6/31974 Revision 35 Reviewed by Facility Review Group 8/10 & 8/17 193)_ Approved by C. L. Burton Plant General Manager 8/17 1995

3.0 PURPOSE

This procedure establishes a method of operation to supply makeup water to the Reactor Coolant System (RCS), Safety injection System and Refueling Water Tank (RWT) at a desired baron concentration and provides instructions for the following modes of control: 3.1 BORATE 3.2 DILUTE 3.3 MANUAL 3.4 AUTOMATIC 3.5 Shutdown Cooling (SDC) Boron Concentration Control S 1 OPS DATE DOCT PROCEDURE DOCN 1-0250020 SYS COMP COMPLETED ITM 35

Pcge 14 of 20 __ _ f ST. LUCIE UNIT 9 OPERATING PROCEDURE NO. 1-0250020. REVISION 35 BORON CONCENTRATION CONTROL - NORMAL OPERATION ~%.

8.0 INSTRUCTIONS

(continued) .,;,w 8.4 (continued) p Mt 3. Enter the number of gallons to be added into the PMW Batch Integrator $ and set desired flow rate on FRC-2210X (Makeup Water Flow). 4. Start one Primary Water Pump if not running. 5. Place V2512 in the OPEN position. 6. Place Mode Selector switch in DILUTE and observe flow indication of FRC-2210X. 7. Monitor VCT level to ensure tank does not fill up to high level alarm. For extended dilutions, match makeuo flow with charging flow using the PMW makeup flow controller to prevent over-filling the VCT while diverting letdown. 8. Upon completion of dilution, retum V2512 control switch to AUTO or CLOSED position. 9. Retum Mode Selector Switch to AUTO or MANUAL.

10. Ensure that the desired reactivity change occurs.

8.5 Manual Mode of Operation i 1. Determine.the desired volume to be added to the VCT and calculate the i i proper blend ratio using the most recent chemistry boron samples of the ? 1A or 1B BAMT and the RCS. If the chemistry sample for the RCS is not y available then use the boronometer reading. ~

f P ge 15 cf 20 ST. LUCIE UNIT 1 .k OPERATING PROCEDURE NO. 1-0250020,. REVISION 35 BORON CONCENTRATION CONTROL - NORMAL OPERATION

8.0 INSTRUCTIONS

(continued) 8.5 (continued) 6 J NOTE The following formulas can be used to determine volume and blend ratio.?L i Remember to make note of the current totalizer readings. j h Volume to be added = desired VCT level % - actual VCT level % X 33.8 gal / k Blend ratio = BAMT Concentration divided by RCS Concentration minus one BAMT-1 RCS i 2. Ensure Mode Select switch is selected to MANUAL. l! 3. Place FRC-2210Y and FRC-2210X to manual and close FCV-22 FCV-2210X by taking the controller output to zero. I 4. Ensure 1 A or 1B primary water pump is running. 5. Ensure the BAM pump recire valves V2510 and V2511 are open. } 6. Start either the 1 A or 1B BAM pump. i Open the Boric Acid Makeup isolation valve FCV-2161. . ) 7. I Ensure FCV-2210X, Reactor Makeup valve, selector switch is in AUTO. 8. i Ensure FCV-2210Y, Boric Acid valve, selector is in AUTO. 9. J \\ If blending directly to the VCT, then open V2512, Reactor Makeup Water 10. stop valve. If direct path to the charging pump suction is desired, then open valve 11. i V2525, Boron Load Control Valve. M

P ge it of 20 m ST. LUCIE UNIT 1 OPERATING PROCEDURE NO. 1-0250020. REVISION 35 3ORON CONC 3NTRATION CONTROL -.40RMAL OPERATICN

8.0 INSTRUCTIONS

(continued) 8.5 (continued) b CAUTION To preclude lifting the VCT relief valve while using V2525, do not allow the ~ combined PMW and boric acid flowrates to exceed the running charging I ]mp(s) capacity. )

12. Adjust FRC 2210X and FRC-2210Y to the desired flow rates.

NOTE Monitor VCT level for increase. I NOTE The addition of Boric Acid should be completed before the PMW, such that, the total blend volume remaining allows for at least 30 gallons of primary makeup water alone, to flow through the lines and flush out any remaining f i boric acid. l

13. When the desired amount of Boric Acid has been added, place the h

selector switch for FCV-2210Y to CLOSE.

14. When the Boric Acid and water flow totalizers show that the proper I

amounts have been added to the VCT, then close V2512 or V2525, which ever was used. 1 / l

15. Place the running BAM pump switch to AUTO and ensure pump stops.

\\'.

16. Close FCV-2161.

j

17. Close FCV-2210X.

18. Monitor for any abnormal change in temperature. Check Boronometer for undesirable change in Boron Concentration.

l

'g QI 5-PR/PSL-1.,. __.. _ Revision 67-December,1995 Page 92 of 101 l PIGURE 4 TEMPORARY CHANGE REQUEST 4 (Page 1 of 3) 4 A Referenos_intatmangn- (Origineter to complete) I-96' of 7 st.1.ucie unit # PSL 1 TC # l Procedure

Title:

'Goeoa Cow.errama Our#4 - den dtawn,J l Procedure Numbec oP f-e.:Les.: a5 Rev. 3f' i Reason for change: Aub Baxmoven. Gwanact bz. CtK;s oua n a, m bruera i A m Auw w e Tkt Rex. As dmean na is tw sme ss,,c St.oor ta C#.4-sMtr. mao, 24v as. 4 Originator: E haA Phone: K %69 Date:Jan/ 23 /1994 B Procedural Controts: (originetor to comp 6ete) Yes O is the intent of the procedure altered? (Tech. Spec. 6.8.3.A) If yes, a TC is E WhNa A PCR is required. [ jls this Temporary Change for a one-time use? If yes, this T O i [g]p, only. If no, this TC may be used up to 90 days, and the eiWiaiGr of the TC shall submrt a procedure change request incorporating this TC at the same time the TC is approved. hMM l MIN Department Head or Designee / a_ O r l C f is this T.C. for a Q.l.7 If yes, the Quality Manager or designee and the Dept. Head or r designee who is jurisdictionally responsible for the Q.l. shall sign. l l Quality Manager or Designee / / Department Head or Designee / / C Temocrary chanoe contents: (originater to cornpiete) Does this Change: Yes No l O g incorporate cornpiex or extensive changes? If Yes Subcommittee required. subcommrttee initials i C Modify instrument setpoints? O g Delete an independent verification? l C [ Alter a QC holdpoint? j O-p Modify a procedural step which alters a regulatory requirement as identified in the procedure? O g Alter the first execution of a procedure? (Proop, LO!) O g Addition of any chemicals? NOTE If any of the above criteria are marked yes, pnor FRG towew is required. /R6'

QI 5-PR/PSL-1 Revision 6T ~ December,1995 Page 93 of 101 FIGURE 4 TEMPORARY CHANGE REQUEST (Page 2 of 3) I TC #.1-%- ut 7 D 1o crn ssLas screemng Yes No i 1. Does me ehenge mswesent a omnge e me tenay as essenbed m the SAR? L Does the ehenge reswesent a snenge e paseewee as essenbed in tw sAR7 Y A is me enange messouses wah a test er esenment not essenbed in me sAR7 [ 4. Ceum the ehenge asset nuemar sehey in a way not penoissy medussed h me sAR? 5. Does me ehenge reques a change to me Technmal W7 M i If the answerto& the 10 CFR 50.59 screerung questens are no, (Questions 1 + 0, then a samy evd ,is ngsfeguired. sTA towawisionnem $1tdMim / 6l3 94 Dane / j E Dose run snenge: (NPS e Yes No 1. Compnmuss me sesenden cd tama of agupment? L Poteneedy leches presens ~ L Defeat enemmen olyess? i 4. Defoot sneenanmal or einesteel inledeces? 5. Aaer the completon of en evolueen eue e en opener woric amund. W yee m No. 8. aumormenen torn em Plast Genere Manager or She Vlos PreeMard ened te obumed. Den / / Yes No O Prior PMG review secused? M If any of i cnteria are rnarked yes, discuss possible alternatives with the O onginator. I NPS Sl re M( M y Does I / / F PMG mecew: V Plant Generes Manager Approval Dale / / FRG Number TNe enange shed be rewomed (rf pnar PRG renew is not regused) by the Feomy Rewow Group and approved by me Plant Genomi Manager semin 14 days of me sumanasson ease. (reen, spec. s.a.:LC) Raase?so try enGment Generw unnager Dane _.j r i Reason. 1 Rehen m W tt is me :- : - Moorn, destroy 48 Seid oopen and haft 48 outeeguent owWusono uen l

E QI 5-PR/RSLs1. _ __ _. Revision ST 4 December,1995 Page 94 of 101 RGURE 4 i TEMPORARY CHANGE REQUEST l (Page 3 of 3) i j 1 G Tcs (M(,-str 85252g3) (This change shall have prior p i NPS and one member of the piant ensnagement seen.) (Tech. spec. a.a.: ) Plant.Yr=;+, R ~L. ) Arik Date-I / N / ~ NPS Signature b s-Authonzation Date I/O/k H Cancellation Asirersn i (NPS/ANPS) Date / / Reasoir ) i l l 1 .i e 1 e f i l t

7 ~ P g114 of 20 ST. LUCIE UNIT 1 OPERATING PROCEDURE NO.1-0250020, REVISION 35 - M c i BORON CONCENTRATION CONTROL - NORMAL OPERATION M*9; rg.

8.0 INSTRUCTIONS

(continued) 8.4 (continued) IE 3. Enter the number of gallons to be added into the PMW Batch Integrator:4 and set desired flow rate on FRC-2210X (Makeup Water Flow).

4.. Start one Primary Water Pump if not running.

5. Place V2512 in the OPEN position. 6. Place Mode Selector switch in DILUTE and observe flow indication of FRC-2210X. 7. Monitor VCT level to ensure tank does not fill up to high level alarm. For extended dilutions, match makeup flow with charging flow using the PMW makeup flow controller to prevent over-filling the VCT while diverting letdown. 8. Upon completion of dilution, retum V2512 control switch to AUTO or CLOSED position. 9. Retum Mode Selector Switch to AUTO or MANUAL k Ensure that the desired reactivity change occurs. 10. 'b 8.5 MangalMode of Qperation \\

i. Mad Bld y

A$ Determine the desired volume to be added to the VCT and calculate the g proper blend ratio using the most recent chemistry boron samples of the N 1A or 1B BAMT and the RCS. If the chemistry sample for the RCS is not available then use the boronometer reading. ri e i 1 l \\ 1

s /* ...P ge_15 of 20 -l i ST. LUCIE UNIT 1 OPERATING PROCEDURE NO.1-0250020, REVISIO.N 35 BORON CONCENTRATION CONTROL - NORMAL OPERATION 4 -2 i E, s,

8.0 INSTRUCTIONS

(continued) i E ?- i 8.5 (continued) D , : ~S h6 i .NGLTE Bil The following formulas can be used to determine volume and blend ratio. Remember to make note of the current totalizar readings. Volume to be added = desired VCT level % - actual VCT level % X 33.8 gal /%. i 4 i Blend ratio = BAMT Concentration divided by RCS Concentration minus one M-1 } RCS i i I 8 g Ensure Mode Select switch is selected to MANUAL 'I C.@ Place FRC-2210Y and FRC-2210X to manual and close FCV-221 FCV-2210X by taking the controller output to zero. D $ Ensure 1A or 1B primary water pump is running. I E $ Ensure tho' BAM pump recire valves V2510 and V2511 are open. i N 6 @ Start either the 1 A or 1B BAM pump.

-@ Open the Boric Acid Makeup isolation valve FCV-2161.

N. @ Ensure FCV 2210X, Reactor Makeup valve, selector switch is in AUTO. I f-@ Ensure FCV-2210Y, Boric Acid valve, selector is in AUTO. 9 3*@ If blending directly to the VCT, then open V2512, Reactor Makeup I stop valve. V ~ K $ If direct path to the charging pump suction is desired, tilen open valve j V2525, Boron Load Control Valve. A fa*

Pzge 16 of 20 ST. LUCIE UNIT 1 OPERATING PROCEDURE NO.1-0250020, REVISION 35 'y o BORON CONCENTRATION CONTROL - NORMAL OPERATION g i i )

8.0 INSTRUCTIONS

(continued) o %), 8.5 (continued) h \\ %is CAUTION E f To preclude lifting the VCT relief valve while using V2525, do not allow the g combined PMW and boric acid flowrates to exceed the running charging - g pump (s) capacity. i f M Adjust FRC-2210X and FRC-2210Y to the desired flow rates. i g a i j /y m s Monitor VCT level for increase. M The addition of Boiic Acid should be completed before the PMW, such that, the total blend volume remaining allows for at least 30 gallons of primary makeup water alone, to flow through the lines and flush out any remaining boric acid. A M d$ When the desired amount of Boric Acid has been add @ lace the 1 selector switch for FCV-2210Y to CLOSE. d O When the Boric Acid and water flow totalizers show that the proper O amounts have been added to the VCT, then close V2512 or V2525, which ever was used. snp.A,,, 4 Bw/u'n.p aJ O'(S7 Place the runriing BAM pump switch to AUTO, j P. $ Close FCV-2161. I G.8. Close FCV-2210X. R. $ Monitor for any abnormal change in temperature. Check Boronometer for undesirable change in Boron Concentration. lhIW Yh 0 51 0 8,y).3 h ~ 3

ST. LUCIE UNIT 1 i OPERATING PROCEDURE NO. 1-0250020, REVISION 35 d BORON CONCENTRATION CONTROL - NORMAL OPERATION 8.0 INSTRUCT 10M: (continued) 8.5 (continued) 2. Manual Dilution ? fiQIE VCT level equates to 33.8 gallons per percent of scale on LIC-2226, VCT Level. A. Determine the desired volume of water to be added. B. Ensure the Make-up Mode Selector switch is selected to MANUAL - e-i C. h Ensure that FRC-2210X, Make-up Water Row, is in MANUAL and reduce the controller output to zero (0). I D. Ensure that FRC-2210Y, Boric Acid Flow, is in MANUAL d D and reduce the controller output to zero (O). E. Ensure that FCV-2210Y, Boric Acid Valve, selector is in N CLOSE. 3g F. Ensure that either the 1 A or the 1B Primary Make-up Water Pump is running. 4 G. Place FCV-2210X, Reactor Make-up, selector switch in AUTO. i H. H diluting to the VCT, Ihan OPEN V2512, Reactor Make-up Water Stop Viv. I. H diluting directly to the suction of the charging pumps, Iban OPEN V2525, Boron Load Control Valve. i a .___.__________,um - +. -, _...

~ ~~ 4 ST. LUCIE UNIT 1 OPERATING PROCEDURE NO. 1-0250020, REVISION 35 BORON CONCENTRATION CONTROL - NORMAL OveMATION

8.0 INSTRUCTIONS

(continued) 8.5 (continued) e 2. (continued) i CAUTION To preclude lifting the VCT relief valve while using V2525, do NOT allow the PMW flowrate to exceed the running charging pump flow rate. J. Adjust FRC-2210X to the desired flowrate. N q K. 11 necessary to maintain the desired VCT level, Ihan h divert the letdown flow to the WMS by placing V2500, VCT Divert Valve, in the WMS position. 3 l L Whan the desired VCT level is reached, lhen: @I 1. Return V2500, VCT Divert Valve, to the AUTO position. M l 2. Ensure that V2500 indicates CLOSED. Q M. When the desired amount of PMW has been added, Iban n place the FCV-2210X selector switch in the CLOSE position. N. CLOSE V2512 or V2525, whichever was used. O. Ensure that FRC-2210X is in MANUAL and reduce the controller output to zero (0). P. Monitor for unexpected results: 1. Abnormal change in the RCS temperature. 2. Undesired change in the RCS boron concentration by boronmeter indication. n i 4

u. ,a aa.... -.... ,.-.x. ..n.--. 5 i ST. LUCIE UNIT 1 i OPERATING PROCEDURE NO. 1-0250020, REVISION 35 BORON CONCENTRATION CONTROL - NORMAL OPERAT10N

8.0 INSTRUCTIONS

(continued) 8.5 (continued) 3. Manual Boration } HQIlii I VCT level equates to 33.8 gallons per percent of scale on LIC-2226, VCT Level. A. Determine the desired volume of boric acid to be added. N M B. Ensure the Make-up Mode Selector switch is selected to ) 3 h. MANUAL. ( I C. Ensure that FRC-2210X, Make-up Water Flow, is in } g MANUAL and reduce the controller output to zero (0). ON D. Ensure that FRC-2210Y, Boric Acid Flow, is in MANUAL g and reduce the controller output to zero (0). i E. Ensure that FCV-2210Y, Boric Acid Valve, selector is in. CLOSE. Q F. Ensure that either the 1 A or the 1B Primary Make-up { N Water Pump is running. l E Whlie it is acceptable to use either BAMT for RCS boration, it is preferable to operate the BAM Pump for the BAMT NOT designated as ' Tech Spec'. G. START either the 1 A or the 1B BAM Pump. H. Place FCV-2210Y, Boric Acid Valve, selector switch in AUTO. i 1. OPEN FCV-2161, Boric Acid Make-up Isolation. J. H borating directly to the VCT, Ihan OPEN V2512, Reactor Make-up Water Stop Viv. + P S i 4 ~ e -

1 ST. LUCIE UNIT 1 1 OPERATING PROCEDURE NO. 1-0250020, REVISION 35 ) BORON CONCENTRATION CONTROL - NORMAL OPERATION 8.0 INSTRUCT 10NS: (continued) 8.5 (continuedh 3. (continued) K. H borating directly to the auction of the charging pumps, Iban OPEN V2525, Boron Load Control Valve. ? L Adjust FRC-2210Y to the desired flowrate. M M. H necessary to maintain the desired VCT level, Ben ~g divert the letdown flow to the WMS by placing V2500, VCT Divert Valve, in the WMS position. I N. Eben the desired VCT level is reached, Iban: 1. Retum V2500, VCT Divert Valve, to the AUTO D position. b 2. Ensure that V2500 indicates CLOSED. 4 lQ O. Whan the desired amount of boric acid has been added, rs Den place the FCV-2210Y selector switch in the CLOSE i position. i j P. CLOSE FCV-2161, Boric Acid Make-up Isolation. I CAUTION To preclude lifting the VCT relief valve while using V2525, do NOT allow the PMW flowrate to exceed the running charging pump flow rate. Q. STOP the running BAM pump and place the selector switch in the AUTO position. 4 l F 8 w

ST. LUCIE UNIT 1 OPERATING PROCEDURE NO. 1-0250020, REVISION 35 BORON CONCENTRATION CONTROL - NORMAL OPERATION

8.0 INSTRUCTIONS

(continued) 8.5 (continued) 4 3. (continued) \\ R. 11 flushing the CVCS piping following boration is desired, Iban: 1. . Place FRC-2210X, Make-up Water Flow, controller q. In AUTO. h CAUTION g To preclude lifting the VCT relief valve while using V2525, do NOT allow the PMW flowrate to exceed the running charging pump flow rate. 2. Adjust FRC-2210X to the desired flowrate to flush the lines with a total of at least 30 gallons of PMW. D 3. When the desired amount of PMW has been k added,Ihan place the FCV-2210X selector switch in the CLOSE position. 4. Place FRC-2210X in MANUAL and reduce the i \\ controller output to zero '(0). S. CLOSE V2512 or V2525, whichever was used. i T. Ensure that FRC-2210Y, Boric Acid Flow, is in MANUAL and reduce the controller output to zero (0). U. Monitor for unexpected results: 1. Abnormal change in the RCS temperature. 2. Undesired change in the RCS boron concentration by boronmeter indication. 1 4 1 6 s mec ,,w --w.-

~ OL5-PR/P.SL-1_. Revision 67 December,1995 l Page 1 of 101 PSL FLORIDA POWER & LIGHT COMPANY NUCLEAR ENERGY DEPARTMENT L ST. LUCIE PLANT Pnoctount PnobuCTION PREPARATION. REVISION.' REVIEW / APPROVAL OF PROCEDURES

1.0 APPROVAL

Reviewed by Facility Review Group 1/3019J,,E.,, Approved by J.H. Barrow (for) Plant General Manager 2/319J,5,_ Revision 67 Reviewed by FRG 12/8 19,25, Approved by J. Scarola Plant General Manager 12/819Et

2.0 PURPOSE

2.1 This procedure provides administrative guidance for the preparation, review, approval and revision of all plant procedures and letters of instruction, for use at the St. Lucie Plant. 2.2 This procedure defines the instructions that shall be used by St. Lucie Plant personnel to assure conformance with NRC Regulatory Guides 1.33 and 1.68, NUREG-0737 and the Site Quality Manual (SOM 2.1 'and 5.0). S_,,, OPS DATE DOCT PROCEDURE DOCN 01-6-1 SYS COMP COMPLETED ITM 67

Ql.5-PR/Psu.1 -) Revision 67 December,1995 Page 41 of 101

5.0 INSTRUCTIONS

(continued) 5.12 (continued) 2. Controlled vendor technical manuals may be utilized as references to safety or non-safety related NPWOs to provide technical guidance.(e.g., DWGs, specifications, torque values, dimensional information, nitage/ current values, etc.) to supplement an invoked plant approved procedure / guideline or the work scope / instructions without prior FRG Review / Plant General Manager approval. In this case, the vendor's step-by-stop maintenance instructions are not being used. 1 3. Changes to technical manuals received from the vendor or changes initiated by FPL shall be forwarded to PEG /JB for review and approval. 4. N'ew tschnical manuals received from vendors shall be nu controlled in accordance with O! 6-PR/PSL-1. The maintenance and preventive maintenance requirements specified in 5. technical manuals shall be considered when writing maintenance - procedures. Vendor recommendations for preventive maintenance i activities or frequencies contained in these Vendor Tech. Man'uals may be deviated from, provided a technical review is performed by the respective maintenance engineering group. 6. Distribution of revisions to vendor technical manuals shall be main by the information Services Supervisor or designee. 5.13 Adherence to Procedures: 4 1. A strict adherence to procedural requirements - Verbatim Compliance - is j the policy expected and required of all St. Lucie Plant personnel. 2. A procedure shall be performed in a step by step manner, with each step being completed prior to the performance of the next step, unless exceptions allowed by the procedure or as specified by this procedure. A. Procedures and Instructions of an ddministrative nature (Quality instructions, ADMs, etc.) shall not be violated, but step by step Implementation is not required. By nature, these types of procedures and instructions often do not lead themselves to sequential implementation. B. Procedures and instructions that are of a technical nature shall followed sequentially except as specifically allowed by approved plant procedures. i

__...m.. _.__ QI 5-PR/PSLs1_ Revision 67 December,1995 Page 42 of 101

5.0 INSTRUCTIONS

(continued) 4 5.13 (continued) ~

2. '(continued)

] B. (continued) 1. Required sign-offs and data entries shall be made as each step is 2 ' performed. 2. If a procedure step cannot be completed as written, or if in the l Judgement of the individual performing a procedure, completion of a specific stop(s) could result in an unsafe condition (e.g., personnel injury, damage to equipment, conditions outside the limits of the procedure etc.), conduct of the procedure shall be stopped, the system / components placed in a safe condition and the Nuclear Plant Supervisor shall be notified. 3. Deviation from Procedure Valve Checklists may be made guvided the deviation is noted in ink on the applicable valve alignment and is approved (initialed and dated) by the Nuclear t Plant Supervisor. 3. Personnel shall not give directions, guidance, recommendation,' or i clarifications which conflict with approved procedures. 4. Adherence to procedures shall be accomplished by use of one of the 3 following methods: A. Method 1 - Prtx=riure Present Durino Performance of Activity: The types of procedures that shal! be present and referred to directly are: 1. Those procedures developed for extensive or complex jobs where rdiance on memory cannot be trusted. 2. Tasks which are infrequently performed. 3. Tasks which must be performed in a specified sequence and/or { which verification is documented by initial or signature. ) i i .\\

4 . OLS-PR/PSL-1 l Revision 67 { L December,1995 Page 43 of 101 - 1

5.0 INSTRUCTIONS

(continued) 5.13 (continued) i B. Method 2 - W.norization: Method by which the procedural steps for the required actions are committed to memory. This method does not permit any deviation from the Procedural Adherence Policy. Procedures for which actions should be committed to memory are 1. immediate Actions in Emergency Operating Procedures and Off Normal Operating Procedures. 2. Procedures for which actions may be committed to memory are routine procedural actions that are frequently repeated and may not require the procedure to be present during performance of the activity. However, copies of procedures shall be available to the user at his/her work location for reference during performance of the tank, if necessary. 5. Procedural adherence may be accomplished by use of a Temporary Change, if necessary. 6. When used in a procedure the word "shall" is used to denote a requirement, the word "should" to denote a recommendation and the word "may" to denote permission, neither a requirement nor a recommendation. 7. Independent Verification: A. Independent Verification has been defined in ADM-17.06, " Independent Verification." Definitions of Independent Verification should not be added to procedures as they may conflict with the guidance outlined in ADM-17.06. l

_Page.1 oL174_.. FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT - ADMINISTRATIVE PROCEDURE NO. 0010120 REVISION 79 3 a 1 f. g

1.0 TITLE

2 CONDUCT OF OPERATIONS u ~ 2.0 REVIEW AND APPROVAL: i Reviewed by Plant Nuclear Safety Committee 1/17 19 I Approved by J. H. Barrow (for) Plant General Manager 1/22 1975 i Revision 79 Reviewed by Facility Review Group 12/21 19,25_ Approved by J. Scarola Plant General Manager 12/21 19,25_

3.0 SCOPE

3.1 Purpose

[ This procedure defines the responsibilities and conduct of the Operations Department during the performance and documentation of all departmental activities. This procedure provides instruction to ensure *. hat plant operations are conducted in an effective, consistent, professional and businesslike manner as per the operating license, plant procedures and applicable regulatory requirements. This procedure applies to all persons in the Operations Department. It identifies operational requirements and management policies necessary to ensure the daily conduct of plant operations is consistent with good operational and engineering practices. 9 S__, OPS DATE DOCT PROCEDURE DOCN 0010120 SYS COMP COMPLETED ITM 79 j l

Page A1 of.174 l ST. LUCIE PLANT ADMINISTRATIVE PROCEDURE NO. 0010120. REVISION 79 CONDUCT OF OPERATIONS APPENDIX D l CREW RELIEF / SHIFT TURNOVER (Page 5 of 5) ) i 1. (continued) D. Instruction for an Interim or Short Term Relief / Shift Tumover. 1. If a specific watchstander requires a short term relief for a period of less ' than 2 hours, then the following instructions provide the minimum requirements for shift relief: a. General watchstation status. l b. Off-normal conditions. I c. Tests in progress. 2. The applicable unit ANPS shall be notified immediately after the shift tumover has been completed. 3. If an individual is expected to be absent for period of greater than 2 hours, then an Individual Relief / Split-Shift Tumover shall be performed. i l i e s I

Page 42 of 174 - ST. LUCIE PLANT ADMINISTRATIVE PROCEDURE NO. 0010120. REVISION 79 CONDUCT OF OPERATIONS APPENDIX E NOTIFICATION OF OPERATIONS SUPERVISOR /FPL MANAGEMENT (Page 1 of 3) 1. The Nuclear Plant Supervisor is. responsible for notifying higher station authonties and appropriate station personnel. Advance notification should be made when possible. The following situations require prompt, vert >al notifications: Notify the Operations Supervisor for the following situations: ~ A. Any event that would cause entry into an Emergency Operating Procedure (EOP). B. Any event requiring phone call notification to the NRC. C. Any event that will generate an LER. D. Inadvertent radioactive liquid or gaseous release. E. Major equipment failure or malfunctions. F. Unexplained or unplanned reactivity changes. G. Forced power reduction. H. Major personnel injury or radiation overexposure. l. Any LCO that would require unit shutdown within the next 24 hours. J. Any operational event that generates an in House Event (IHE) Report AND causes heightened awareness to FPL sources offsite. K. Any release that is or is potentially, damaging to the environment. L Load restrictions or inability to meet load dispatcher requiroments. This includes, but is NOT limited to the following: 1. A planned power escalation is unexpectedly halted for any reason and can not be resumed within one hour. 2. If at a power level less than 100 percent, any unexpected condition that would prevent a future power escalation and can not be resolved within two hours. 3. If at a power level less than 100 percent and the plant is unable to support an unexpected request for more power from the load dispatcher.

P ge 48 of 174 ST. LUCIE PLANT ADMINISTRATIVE PROCEDURE NO. 0010120. REVISION 79 CONDUCT OF OPERATIONS APPENDIX F LOG KEEPING (Page 2 of 9) 2. Chronoloaical Loos: i A. Lreg books and/or computerized logs shall be maintained at the RCO, NO/SNPO, NTO/NPO and ANPO normal stations. Entries are to be in concise and complete enough to reconstruct the events of the shift. Particular ~ attention should be made to the entries pertaining to any abnormal condition that occurs. Times for each entry shall be as near correct as possible using military time. The entnes are to be made in chronological order, 1. Evolutions, manipulations and operations that are performed, observed and monitored by operators NOT actively assuming the responsibilities of 4 a particular watch station shall be recorood in the applicable watch station chronological log and initialed by that operator. The operator should notify the responsible watchstander of the log entry. i 2. When it is necessary to insert additional information after the fact, Then the entry shall be recorded with the actual time of occurrence, the words ' Late Entry in parenthesis, and the information to be logged. Example: 1234 Started the 1 A EDG for survelliance run 0827 (Late Entry) Filled the 1 A2 SIT with the 1B HPSI Pump in accordance with OP 1-0410021 1345 Secured the 1 A EDG. Surveillance run SAT. 3. When it is necessary to correct information recorded in error, then " o entry shall be recorded with the actual time of occurrence, the words j " Corrected Entry' in parenthesis, and the information to be logged. 5 Example: 1234 Started the 1B EDG for surveillance run 1345 Secured the 1 A EDG. Surveillance run SAT. 1234 (Corrected Entry) Started the 1 A EDG for surveillance run 4 4. Entries in the RCO log should include, but are NOT to be limited to, the following: a. Conditions at the beginning of each watch. i i l

Page 47 of 174 ST. LUCIE PLANT ADMINISTRATIVE PROCEDURE NO. 0010120, REVISION 79 CONDUCT OF OPERATIONS i APPENDIX F i LOG KEEPING (Page 3 of 9) 2. Chronolooical Loos: (continued) i A. (continued) 4. (continued) ~ b. Significant changes in plant conditions.- Examples: 1. Mode changes. 2. Load changes. 3. Reactivity changes. 4. Startups and Shutdown. 5. Time of Reactor criticality. c. Any new condition that would limit unit generation. t Examples: 1. Condenser back pressure at administrative limits. 2. Chemistry parameters limiting operation. d. Special tests, including periodic and surveillance testing, for major equipment. Examples: 1. Start and stop times for periodic or surveillance tests and outcome (SAT or UNSAT), for major equipment. i s i 2. Post maintenance testing and outcome, for major equipment.

e. ' Control problems associated with major equipment or systems.

Examples: 1. Changes in plant work arounds. b 1 1 1

P ge 70 of 174 ST. LUCIE PLANT ADMINISTRATIVE PROCEDURE NO. 0010120, REVISION 79 ) CONDUCT OF OPERADONS APPENDIX M PROCEDURAL COMPUANCE AND IMPLEMENTATION (Page 1 of 6) 1. Controlled procedures are available in both Control Rooms and shall be implemented and complied with in accordance with the instructions provided in 015-PR/PSL-1, " Preparation, Revision, Review / Approval of Procedures." 2. In the event of an emergency where procedural guidance does NOT exist or in ~ which a specific emergency is NOT addressed by an approved procedure, then Operations personnel shall take action to protect the health and safety of the public, minimize personnel injury, and damage to the facility. 4 3. Numerous tasks performed by the operators are repetitive and routine in nature. These tasks come under the guidance of the memorization method of adherence to procedures in accordance with 015-PR/PSL-1, " Preparation, Revision, Review / Approval of Procedures," and may be performed from memory. These tasks, which are listed in the following sections, are considered to be skill of the trade for qualified operators. Each listed task shall have one or more of the below justification reasons: i 4 (A) Task is routine and not complex - satisfactory completion assured by routine training and observation. (B) Task is routine and has a low level of complexity - satisfactory completion assured by completion of verification checklist and independent verification. (C) Posted instructions in place as reference. 3 (D) Satisfactory completion assured by multiple levels of review and/or feedback from system. A. General Control Tasks 1. Racking IN and OUT of 6.g KV,4.16 KV, and 480V breakers. (B) 2. Tuming ON and OFF 480V MCC breakers. (D) 3. Writing clearances and NPWOs. (A,D) 4. Changing chart paper. (A,D) 5. Placing controllers in MANUAL or AUTO. (A,D) l

Pcge.71 of 174_ d ST. LUCIE PLANT ADMINISTRATIVE PROCEDURE NO. 0010120, REVISION 79 CONDUCT OF OPERATIONS APPENDIX M PROCEDURAL COMPLIANCE AND IMPLEMENTATION (Page 2 of 6) 1. (continued) B. Reactor Control Operator 1. Divert Letdown to Control VCT level. (A,D) ~ 2. Check Sheet 1 of AP 10010125. (A,B,D) 3. Refueling Operations - movement of machine, etc. (A.D) 4. Adjusting Main Generator loading, including Megavars and Megawatts (manipulation of C9: controls). (D) 5. Swapping Auxiliary and Start-up Transformers. (D) 6. Adjusting CEA position (eg. ASI control). (D) 7. Manipulation of control valve. .DVs, FCVs) to control Heatup and Cooldown rates. (D) 8. Pumping down Reactor Drain Tank. (A,D) 9. Placing CST on recire. (A,D) C. Senior Nuclear Plant Operator 1. Generic Rounds Sheets. (A) 2. Swapping HUTS. (A,D) 3. Blowing down BAMT ievel transmitters. (C) 4. Operator Readings and AP 0010125 checks. (A,B,D) j 5. Recirculating of HUTS, WMTs, and AWSTs. (A,D) 6. Backwashing ICW/CCW strainers. (C)

( QI 5-PR/PSL-1 r. .e4-December,1995 Page 92 of 101 RGURE 4 TEMPORARY CHANGE REQUEST (Page 1 of 3) l A ' Reference information: (Originator to cornpiete) St. Lucie Unit # C8*=od TC # d O l'/ Procedure

Title:

toan ott or opt eA r, o s.s i Procedure Number: Ar oo/o/2o Rev. 74

_

Reason for change: ,aunoro( ~ 4 s4 um as-o,m re sc s ~ros 5 TC D PetcEOL6 7t. 0 -9L - 0I( l l Originator: d 7A c ua/C Phone: 709/ Date: / / J 9/ 96 B Proceoural Controts: (Originator to complete) J Yes No i O E is the intent of the procedure altered? (Tech. Spec. 6.8.3.A) If yes, a TC is NOT f applicable. A PCR is required. O E is this Temporary Change for a one-time use? If yes, this TC can be executed g.ng Jirrlg only. If no, this TC may be used up to 90 days, and the originator of the TC shall i submrt a procedure change musst incorporating this TC at the same time the TC is approved. Department Head or Designee k / / 2 9 / 'l 0 l 4-O E' is this T.C. for a Q.l.? If yes, the Quality Manager or designee at d the Dept. Head or designee who is jurisdictionally responsible for the O.l. shall sign. Quality Manager or Designee / / j Department Head or Designee / / C-Temocrarv chance contents: (originator to complete) i Does this Change: Yes No O E incorporate complex or extensive enanges? If Yes, Subcommittee required. Subcommittee Irvtials O 5 Modify instrument setpoints? O 9 Delete an independent verification? O E Alter a QC holdpoint? 0 2 Modify a procedural step which alters a regulatory requirement as identified in the procedure? O S After the first execution of a procedure? (Preop, LO!) O 3 Addition of any chemicals? NOTE If any of the above enteria are marked yes. prior FRG revow is required. m ) 1

QI 5-PR/PSL-1 Revision 67 - December,1995 Page 93 of 101 RGURE 4 TEMPORARY CHANGE REQUEST (Page 2 of 3) TC # R M j D-10 CPR 50.88 Sersomng Yes No l 1. Does en ehenge represent a change to the lectly as described in the SAR7 2. Does eis change represent a change to procedures as desenbod in me SAR7 3. Is me enange -d wnh a test or expenment not descdbed in 1hs SAR1 [ 4. Could the change anect nucsear salsty a a way not pronously evar.m:ed in 7 l m um / E. Does ine change recure a enange to the Techncel SpeerAcanons? 1 .N.G.T.E. t if the answer to M the above 10 CFR 50.59 screening questions are no, (Questions 1 5), then a safety evaluation is not requirec. AMS Date f /Of[g l STA revow (signature) ' E Does sus enange (NPS to comsfetI) Yes No 1. Compromes the separaton of redundant trans of esapment? 2. Potentady isolate pressure rehels?

/

/ 3. Defeat automanc agnais? 4. Delost rnschancel or elecincel interlocks? t/ 5. Aber the completon of an evoluton oue to an operator work arouna. V if yee to No. 5. aumonzaton from the Plant General Manager of Sete Vice Preeoent shall be obtened Date / / Yes No O I Pnor PRG rowe. remared? !!Q.IE If any of the above criteri3 are rnarked yes, discuss possible alternatives with the e onginator. A // NPS Signature

////Tl Dato

// 2 'f / IS F Fno Revow: Plant General Manager Approval Date I TNs change snaN be rowowed (if pner FRG towow is not FRG Numoer recured) by the Faciidy Rewow Group and approved oy the Plant General Manager untmn 14 days of the aumonaaton case. (Tech. Spec. 6.8.3.C) REJECTED by FRG/ Plant General Manager _ Date I_ I Reason. Resum 1o Ongmator it is tne responetely of the ongenator of Ine regocted temporary change to cancel the change m the apprognate Control Room, destroy all Said copees and helt al suoseguent evolueens uomg tnas temocrary change 7 -~.m

~. QI 5-PR/PSL-1_ Revision 67 December,1995 Page 94 of 101 FIGURE 4 1 l TEMPORARY CHANGE REOJgE (Page 3 of 3) I G Tcs0-%-oN .Anorovat: (This change shall have prior approval by a NPS and one member of the plant j management staff.) (Tech. Spec. 6.8.3.8) Plant Management Staff Si G PatT[/ m Date / /2 9 / 96 u NPS Signature Authonzation Date / /E f'/9 6 H Cancellation Authonzation (NPS/ANPS) Date / / Reason: 1 ( 1 i i

Pcge 60 of-174- ^ ST. LUCIE PLANT ADMINISTRATIVE PROCEDURE NO. 0010120, REVISION 79 CONDUCT OF OPERATIONS APPENDIX B SHIFT OPERATIONS POLICIES (Page 5 of 8) 4. (continued) 'A. (continued) 4. P - Prove Prove to yourself that the actions that were just performed produced, a. the desired results, b. Observe and verify the following: 1. The task was perfomled correctly. <*e 'qu 2. The actual response was the expected response. 3. The component / system is in the proper configuration to support h} ,6 the intended operation. / ' / 49 4. The proper component was operated. Y 5. Mius Medpulgt .Only licensed operatoD]annit4% late the controls that riirectly A. affect the rea ower level of a reaMfor training purposes. A t manipulate controls only under direct visual supervisiqrtof a ~ consed operator. N 6. Unit Reliability A. The NPS/ANPS should make every effort to prevent putting the plant in a situation where a single failure would jeopardize plant safety or availability. Systems listed under AP 0010142, " Unit Reliability - Manipulation of Sensitive Systems" warrant particular attention. Maintenance or testing should not be allowed on an in-service train or channel with the opposite train out-of service or another channel in Trip, except for Tech. Spec. required surveillances or to prevent a plant shutdown.

g c, cg - Ql'l APPENDIX B SHIFT OPERATIONS POLICIES l l 5. ReactMty Manipulations A. Reactivity manipulations in the course of normal plant operations is defined as the insertion of positive and negative reactivity due to manipulation of the 1 following: i l 1. CEA insertion and withdrawal.

2. Addition of water and/or boric acid to the VCT or Charging Pumps'

[ suction. I 3. Turbine / Generator load changes. 4. Placing a purification lon Exchanger in service, (any time V2520, "lon Exchanger Bypass Valve," position is changed from bypassing the ion

VNOT.se.(o) iv uunung TIDW unwud. DO b" *ehangar(s)),

m / I B. All reactivity manipulation the course of normal plant operations, both positive and negative ve prior approval from the SRO fulfilling the role of the Control Room' and function, except as provided for in step 5.D. i i j C. When reactivity manipulations are being performed, both positive and ~ negative, the SRO fulfilling the role of the Control Room Command function shall directly supervise the manipulation and additionally assume the role of a j reactivity manager, except as provided for in step 5.D. D. In the event of off-normal and emergency conditions, Reactor Control Operators are authorized to perform reactivity manipulations without the presence of and approval of an SRO, ifin his/herjudgement immediate intervention is required to protect the health and safety of the public and/or challenging of plant safety functions. The SRO fulfilling the role of the Control Room Command function shall be notified of the manipulation as soon as possible. E. Crew Relief / Shift Tumover shall NOT take place for Reactor Control, Operators or the Assistant Nuclear Plant Supervisor while reactivity - manipulations are in progress. i 4 1(

TC 0 'l G ~ Ol'I ~ ~ APPENDIX B SHIFT OPERATIONS POLIClfS 5. (Continued) F. Reactivity manipulations shall be performed only by those individuals possessing an active license applicable to the unit on which the manipulation is being performed. The only exceptions are' persons reactivating a license or + in a bonafide training role in pursuit of obtaining a license; they may perform reactivity manipulations under direct visual supervision of a licensed operator with an active license. T 4

~ - Page-40 of;174 ST. LUCIE PLANT ADMINISTRATIVE PROCEDURE NO. 0010120, REVISION 79 CONDUCT OF OPERATIONS. APPENDIX D CREW RELIEF / SHIFT TURNOVER (Page 4 of 5) r j l 1. (continued) C. Instrudions for an Individual Relief / Split-Shift Tumover 1. If a specific watchstation shift is being split by two individual watchstanders, then the following instructions provide the minimum 1 requirements for shift relief: " die off going watchstander shall review applicable plant log sheets to f a. determine the existence of any off-normal condition or trends. g J 1 + / , 9 L' b. The off-going watchstander shall complete the applicable Tumover I C Check Sheet (Data Sheet 1) for their watchstation. fd The off-going watchstander shall verbally transmit and explain the l~A information as recorded on their applicable Tumover Check Sheet c. ] g s' (Data Sheet 1) to the on-coming watchstander. S g f. The on-coming watchstander shall review the following and g k acknowledge that review by initialing Check Sheet 1 of 4, c AP 1(2)-0010125, " Schedule of Periodic Test, Checks, and [p Calibrations." l 1. Applicable Watchstation Chronological Log. 1 2. Applicable Watchstation Operater Log Readings. ) 3. Night Order Book. 4. NPWO, ANPS, and NWE shall review equipment out-of-service log. a f. The applicable unit ANPS shall be notified immediately after the shift J tumover has been completed. ,l e 1 i

i ._ (- D - c g - OjLl . ~~ / y 1 APPENDIX D l CREW RELIEF / SHIFT TURNOVER j 1. C. 1. d. On-coming and off-going control room watchstanders shall conduct a i face-to-face complete walkdown of the RTGBs and control panels. t The on-coming watchstander shall make a chronological log entry e. indicating he/she has assumed the responsibilities of the watchstation. it J 4 i e

~- Page 41 of 174m - ST. LUCIE PLANT ADMINISTRATIVE PROCEDURE NO. 0010120, REVISION 79 CONDUCT OF OPERABONS APPENDIX D CREW REllEF/ SHIFT TURNOVER (Page 5 of 5) 1. (continued) D. Instruction for an interim or Short Term Relief / Shift Tumover. 1. If a specific watchstander requires a short term relief for a period of less ' than 2 hours, then the following instructions provide the minimum requirements for shift relief: gb L A a. General watchstation status. ,.4 b. Off-normal conditions. 4 ~ 4 . [' c. Tests in progress. ?/ h, 2. The applicable unit ANPS shall be notified immediately after the shift tumover has been completed. p( b ,p 3. If an individual is expected to be absent for period of greater than 2 hours,

.y then an IndNidual Relief / Split-Shift Tumover shall be performed.

u 4 i J s l i 1

I ^ ~ - - +* = APPENDIX D ' CREW RELIEF /SHIFTTURNOVER 1. D. 1.

d. Control room watchstanders with the responsibility of the Operator at the j

Controls or the Control Room Command function shall conduct a face-to-face complete walkdown of the RTGBs and control panels with the individual assuming their responsibility. 4 4 j 4 i 1

i Undated ner Amendment'134' dated 3/15/95 DPR-67 Page 1 FLORIDA POWER &' LIGHT COMPANY PE i DOCKET NO, 50-335 ST LUCIE PLANT UNIT NO. 1 FACILITY OPERATING LICENSE l i L% pen"JCDON, 1 '1. The Nuclear Regulatory Commission (the Commission) _ ' having found that: ) A. The application for license filed by Florida Power & Light company (the licensee) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions's rules and regulations set forth in 10 CFR Chapter 1 and all j required notifications to other apencies or bodies have been duly made; j B. Construction of the St. Lucia

Plant, Unit No.

1 (facility) has been substantially completed in conformity i with Construction Permit No. CPPR-74 and the application,. i as amended, the provisions of the Act and the rules and regulations of the commission; 1 I C. The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; i i D. There is reasonable assurance: (i) that the activities authorized by this operating license can be conducted e without endangering the health and safety of the public, i and (ii) that such activities will be conducted in i compliance with the rules and regulations of the { Commission; E. The licensee is technically and financially qualified to engage in the activities authorized by this operating license in accordance with the rules and regulations of the Commission; t F. The licensee has satisfied the applicaole provisions of 10 CFR Part'140, " Financial Protection Requirements and Indemnity Agreements," of the Commission's regulations; I G. The issuance of this operating license will not be l inimical to the common defense and security or to the [ health-and safety of the public; I a .2. .. ~. -

_ _ _ _ - ~ _ g Undated ner Amendment 134 dated 3/15/95 DPR-67 'Page 3 ) l (3)~ Pursuant to the'Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use at any time byproduct, source and special nuclear material ~as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as. fission detectors in amounts as required; 't (4) Pursuant to the Act, and 10 CFR Parts 30, 40, and b 70, to receive, possess and use in amounts as required any byproduct source or special nuclear i material without restriction to chemical or i physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, to l

possess, but not separate, such byproduct and special nuclear materials as may be produced by the f

operation of the facility. C. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Sections 30.34 of Part 30, Section-40.41 of Part 40, Section 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified'or incorporated below; 1 (1) Marinum Power Laval The licensee is authorized to operate the facility { at steady state reactor core power levels not in excess of 2700 megawatts (thermal), provided that the construction

items, preoperational

tests, startup

tests, and other items identified in to this license have been completed as specified in Enclosure 1.

Enclosure 1 is an integral part of, and is hereby incorporated in this license. (2) Tarhnical snacifications I The Technical Specifications contained in Appendices ' A and B, as revised through Amendment No. 134 are hereby incorporated in the license. The licensee shall operate the facility in accordance t with the Technical Specifications. 5 [

f s. ~ l DEFINITIONS RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2700 MWt. REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIE shall be the time interval from T when the monitored parameter exceeds its trip setpoint at the channel sensor until electrical power is interrupted to the CEA drive mechanism. REPORTA8LE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50. SHIELD BUILDING INTEGRITY { 1.28 SHIELD BUILDING INTEGRITY shall exist when: Each door is closed except when the access opening is being used i a. for normal transit entry and exit; b. The shield building ventilation system is in compliance with 4 Specification 3.6.6.1, and 2 c. The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERA 8LE. l SHUTDOWN MARGIN 4 1.29 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is suberitical or would be subcritical from its present condition 4 assuming all full-length control element assemblies (shutdown and regulating) are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn. 1 SITE B0UNDARY s site Boundary means that line beyoh which the land or-1.30 property is not owned, leased, or otherwise controlled by the licensee. SOURCE CHECK 1.31 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor _is exposed to a radioactive source. ST. LUCIE - UNIT 1 1-6 Amendment No.27.32.59.85 g,125

DEFINITIONS STAGGERED TEST BASIS 1.32 A STAGGERED TEST BASIS shall consist of: a. A test schedule. for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, and b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval. THERMAL POWER I.33 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. UNIDENTIFIED LEAKAGE I.34 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE. UNRESTRICTED AREA 1.35 Unrestricted area means an area, access to which is neither limited nor controlled by the licensee. UNRODDED INTEGRATED RADIAL PEAKING FACTOR - Fr 1.36 The UNRODDED INTEGRATED RADIAL PEAKING FACTOR is.the ratto of the peak pin power to the average pin power in an unrodded core, exclud n tilt. l l m m-l ST. LUCIE - UNIT 1 1-7 Amendment No. 50,50, 100,125 i s

- - = r el w. j. .._..__--.~.:_.- %)9 v~ -s SY"E S UNITED STATES ( 5 NUCLEAR REGULATORY COMMISSION wasm. crow,o.c. sones \\....e August 28, 1980 NOTE T0: R. Tedesco T. Novak G.~ Lainas 4 1 agree with E. Jordan's memo in that further debate on this issue is probably not warranted Please ensure that your staff at this time. is aware of this interpretation and that this 3 will be the NRC position on this matter at this time. Darrell G. Eisenhut Enclosure E. Jordan [ ec: J. Scinto i

l l

1 l

* ~ ~ ~
* * " " ~ ~

T0*d' 684STEEPOPI c1 N W 0T P66T-4T-0T

'N . t '.._. - g pesi i

/g [Sqj, umTED sTATet NUCLEAR RBOULATORY COMMISSION y

y, a g gyj j wAewmeton,o s.neses %(* GM]. AUG 2 21!80 55 INS #0100 j MEMORANDUM FOR: E. J. Brunner, Chief, RO&NSB, RI R. C. Lewis, Acting Chief, RO&N58, RII R. F. Heishman, Chief,'RO&NSB, RIII G. L. Madsen, Chief, RO&NSB, RIV J. L. Crews, Chief, RO&NSB, RV FROM: E. L. Jordan, Assistant Director for Technical Programs Division of Reactor Operations Inspection, IE j SUSJECT: DISCUSSION OF " LICENSED POWER LEVEL" (AITS F14580H2) 2 l Dating back at least to 1974, there have been many lengthy " discussions" i regarding the exact meaning of " full, steady-state licensed power level" (and similarly worded power limits). We do not believe the real safety benefits d that might be carived from an NRC-wide agreement would be worth the further expenditure of manpower in meetings, etc. that would be required to achieve a consensus. We de reali:e that some comon uniform basis for enforcing maximum Itcensed l power is needed by I&E inspectors. Therefore, until and unless an NRC-wide position is put forward and agreed upon (and as stated, !&E coes not propose to initiate proceedings to that and), I&E will use the following guidance, q The average power level over any eignt hour shift should not exceed the " full l steacy-state licensed power level" (and similarly worded terms). The exact signs hour parieds cefined as " shifts" are up to the plant, but should not be i varied froc cay t: day (tha easiast definition is a narcal snift mannec by a particular " crew"). It is permissible to briefly exceed the " full, stesey-state licensed power level" by as muen as 3 for as long as 15 minutes. In no i case should 105 power be exceseec, out lesser power " excursions" for longer-periods should be allowed, with the above as guidance (i.e.,1% excess for 30

Hnutes,1/3 for one hour, etc., should be allowed).

There are no limits on the numser of times these " excursions" may occur, or the time interval that must separate such " excursions," except note that the above requirement I regarding the eight hour average power will prevent abuse of this allowance. l l CONTACT: H. W. Woods, IE 49-28180 0 0 5 O Y D lfG G l@ ~

CO'd T 101 c o o 2~ AUS zg ng i 4 The above is considered to be within the licensing basis and, therefore, acceptable to us, and it is also fair to the utilities and their ratepayers. J . J rean, Assistant Director for Te ical Programs Division of Reactor Operations Inspection Office of Inspection and Enforcement cc: R. C. DeYoung, IE S. J. Bryan, IE f.Eisenhut,NRR i D. Ross, NRR 4 O e 4 I s j

=k.-

[

g y n. q. I-15.2.4 CHIMICAL AND'VOL1HE CONTROL SYSTl!M MALFUNCTION - BORON DILUTION EVENT i ~ 15.2.4.1 Identification of causes 4 j The chemical and volume control system (CVCS) described in Section 9.3.4 regulates both the chemistry and the quantity of coolant in the reactor i coolant system. Changing the boron conce'ntration in the reactor coolant system is a part of normal plant operation, compensating for long-ters reactivity effects, such as fuel burnup, xenon buildup and decay, and plant l startup and cooldown. For refueling operations, borated water is supplied from the refueling water tank, which assures adequate shutdown margin. An i inadvertent boron dilution in any operational mode adds positive reactivity, l produces power and possibly temperature increases, and, in Modes 1 and 2 I (startup and power operations) can cause an approach to both the DN5R and CIM i limits. 4 Boron dilution is conducted under strict administrative procedures which i specify permissible limits on the rate and magnitude of any required change in 1 horon concentration. Boron concentration in the reactor coolant system can be decreased either by controlled addition of unborated makeup water with a j corresponding removal of reactor coolant (feed and bleed) or by using the deborating ion exchanger. The deborating. ion exchanger is normally used for i j boron removal when the boron concentration is low ((ppa) and the feed-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 instrument provides a continuous measure of boron concentration and high-low boron concentration alaras. p( During normal operation, concentrated boric acid solution is mixed with. desineralised makeup water to the concentration required for proper plant operation and is automatically introduced into the volume control tank in response to a low water level signal from the volume control. To effect ron j dilution, the makeup controller mode selector switch must be set to "Di te" f l, and the desineralized water batch quantity selector set to the desired Q quantity. When the specific amount has been injected, the desineraliser water j,, control valve is shut automatically. Dilution' of the reactor coolcat can be terminated by isolation of the makeup water system, by stopping either the askaup water pumps or the charging pumps, i cr by closing the charging isolation valves. A charging pump must be running in addition to a makeup water pump for boron dilution to take place. The CVG is equipped with the following indications and alarm functions, which will inform the reactor operator when a change in boron concentration in the reactor coolant systen may be occurring: 0, I a) Boronometer high and low alaras and concentration indication b) Volume c trol tank level indication and high and low alarus Q. 15.2.4-1 kr ~~ [ I

I i w. i c) Makeup flow indication and alaras d) Volt ' * ' control tank isolation. - l Changes in boron concentration while the reactor is on automatic control at full power are compensated for by repositioning the CEA's. Ibwever, to assist j the reactor operator in asintaining an adequate shutdown margin, CEA insertion below a position that would provide a minians of one percent shutdown margin-(assuming one stuck CEA) is accompanied by control room alarms. Because of the procedures involved and the numerous alarus and indications available to. ~ j the operator, the probability of a sustained or erroneous dilution is very low. 15.2.4.2 Analysis of Effects and Q>nsequences 15.2.4.2.1 Method of Analysis The time required to achieve criticality from a suberitical condition due to j boron dilution is based on the initial and critical boron concentrations, the j boron reactivity worth, and the rate of dilution. Reactivity increase ratM due to boron dilution are based on the boron werth and the dilution rate. ,a i Cases have been analyz'ed f or all six operational modes, i.e., power operation, startup, hot standby, hot shutdown, cold shutdown, and refueling.* In each case, it is assumed that the boron dilution results from pumping unborated desineralized water into the reactor coolant system at the maximum possible rate of 132 spa (3 x 44 gpa per charging pump) and that the boron concentrations are uniform at all times. 4 The boron dilution rate is calculated by CESEC f or all cases except dihitton. during refueling. CESEC described in Section 15.1.4-1 divides the reactor coolant system into 15 control volumes with the continuity equation beif,';.3 satisfied by all nodes. - The charging rate of non-borated water and the bosva: g ,? content of the system are inputs to CESEC. The maximum d,ilution rate 'h (10.5 ppa / minute) occurs at the initiation of the transient. For dilution during refueling the reactor coolant system is assumed to be one control ..e volume with the boron concentration calculated by: the time race of chansa.o,L boron equals, flow in times the boron concentration 1minus flow out times boron concentration. The uniformity of the boron concentration can be assured for the different 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 systes pumps. Because the baron is chec cally dissolved in the reactor coolant, it will not precipitat e. The only possible means of j obtaining a nonuniform solution is by the addition of domineralized water via the charging pumps. Ibwever, 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 Baference 42. 15.2.4-2 i .}}

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