ML20137E529

Notice of Violation from Insp on 961206.Violation Noted: in April 96 Licensee Made Change to Facility Described in Fsar,Which Involved Three USQ W/O Prior Commission Approval

ML20137E529
Person / Time
Site:	Crystal River
Issue date:	03/12/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20137E528	List: IR 05000302/1996012 ML20137E529
References
50-302-96-12, 50-302-96-19, EA-96-365, EA-96-465, EA-96-527, NUDOCS 9703270311
Download: ML20137E529 (10)

Text

.

NOTICE OF VIOLATION I

Florida Power Corporation Docket No. 50-302 Crystal River Nuclear Plant License No. DPR-72 l Unit 3 EA 96-365,96-465, 96-527 During NRC inspections completed on December 6,1996, violations of NRC requirements I were identified. In accordance with the " General Statement of Policy and Procedures for I NRC Enforcement Actions," NUREG-1600, the violations are listed below:

A. 10 CFR 50.59, " Changes, Tests and Experiments," provides, in part, that the licensee may make changes in the facility or procedures as described in the safety analysis report (SAR) without prior Commission approval, unless the proposed change involves  !

a change in the Technical Specifications (TS) or an unreviewed safety question I (USQ). A proposed change shall be deemed to involve a USQ if the probability of  !

occurrence of a malfunction of equipment important to safety previously evaluated in j the SAR may be increased, if a possibility for an accident or malfunction of a different I type than any evaluated previously in the SAR may be created, or if the margin of  !

safety as defined in the basis for any TS is reduced.10 CFR 50.59 further requires j that a written safety evaluation be documented providing the bases for a determination that the changes do not involve a USQ. l The TS bases for TS 3.8.1, AC Sources - Operating, states that the service rating of the emergency diesel generator (EDG) is, in part, 3251 to 3500 kilowatts (KW) on a cumulative 30 minute basis.

The Final Safety Analysis Report (FSAR), Rev.19, dated December 21,1994, Section 8.2.3, Sources of Auxiliary Power, provides the load ratings for both EDGs, including a 2851 - 3000 KW cumulative 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating and a 3251 - 3500 KW l cumulative 30 minute rating. (The maximum load rating shown for any period of time is 3500 KW). It also states that the "A" EDG auto-connected load is within the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating at one minute into the scenario. FSAR, Rev. 20, dated April 1,1994, Table 8-1, Emergency Diesel Generator "A" Auto & Manually Connected Loads, lists the largest auto-connected load as make-up pump 1 A (615.5 KW). This FSAR l information remained current through 1996. '

The FSAR, Rev.10, dated July 1,1988, Table 8-1, Emergency Diesel Generator "A" Auto & Manually Connected Loads, lists the largest auto-connected load as make-up pump 1A (615.5 KW). This FSAR information remained current through 1990. l The FSAR, Rev.10, dated July 1,1988, Section 10.5, Emergency Feedwater (EFW)

System, states that upstream of the turbine-driven emergency EFW pump turbine steam supply line, there are redundant, normally closed direct current (DC) motor operated valves (ASV-5 and ASV-204) which are opened upon actuation from the emergency feedwater initiation and control (EFIC) system. FSAR, Rev. 8, dated l l July 1,1987, Section 7.2.4, Emergency Feedwater Initiation and Control, states that i

the EFIC trip module located in the "A" cabinet actuates the "A" train of EFW (motor-

! driven pump) and the trip module located in the "B" cabinet actuates the "B" train of EFW (turbine-driven pump)(EFP-2). This FSAR information was the first description of the EFIC system, and it remained current through 1992. Section 7.2.4 of the Enclosure 2 9703270311 970312 PDR ADOCK 05000302 G PDR

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

Notice of Violation 2 FSAR, was revised on January 17,1993, Rev.18, as follows

"The trip module located in the "A" cabinet starts the "A" train motor-driven EFW pump and the "B" train turbine-driven EFW pump. The trip module located in the "B" cabinet starts only the "B" train turbine-driven EFW pump. The starting of both EFW pumps on "A" train EFIC actuation is necessary to assure that the turbine-driven pump will be operable in the event of a failure of the ES "B" 250/125V DC system coincident with a loss of offsite power and a [ engineered safeguards) actuation. Under this scenario, EFP-2 will be relied upon to share the emergency feedwater load with the motor driven j emergency feedwater pump in order to decrease the electrical load on diesel generator EDG-3A." This FSAR information remained current through 1996. l

1. Contrary to the above, in April 1996, the licensee made a change to the facility as described in the FSAR, which involved three USQs, without prior Commission approval. Specifically, the modification, installed by Modification l Approval Record (MAR) 96-04-12-01 changed the EFW initiation logic to allow the motor-driven EFW pump to provide all EFW during certain analyzed accidents which increased the calculated post-accident motor-driven EFW pump load from about 616 KW to about 666 KW. As a result, the "A" EDG accident loads were in excess of the limits specified in FSAR Section 8.2.3, TS 3.8.1 Basis (3500 KW limit), TS surveillance requirement (SR) 3.8.1.11 Basis (3100 KW one-minute load), and TS SR 3.8.1.8 Basis (616 KW largest single post-accident load that could be rejected). This change reduced the margin of safety as defined in the FSAR and three TS Bases, resulting in three USQs. ,

The 10 CFR 50.59 safety evaluation for this modification was inadequate in i that it did not address electrical loading effects on the "A" EDG and did not recognize the USQs. (01012) I

2. Contrary to the above, in April 1996, the licensee made a change to a procedure as described in the FSAR, which involved three USQs, without prior Commission approval. Specifically, Emergency Operating Procedure EOP-13, EOP Rules, was changed by Rev. 2 to require operators to take manual  ;

control of the motor-driven EFW pump to increase EFW flow under certain conditions, resulting in an increase in EFW pump load from about 666 KW to about 713 KW. As a result, the "A" EDG accident loads were in excess of the limits specified in the FSAR and TS 3.8.1 Basis (3500 KW limit), TS SR 3.8.1.11 Basis (3100 KW one-minute load), and TS SR 3.8.1.8 Basis (616 KW largest single post-accident load that could be rejected). This change reduced the margin of safety as defined in the FSAR and three TS Bases, resulting in three USQs. The 10 CFR 50.59 safety evaluation for this procedure change was inadequate in that it did not address electrical loading effects on the "A" EDG and did not recognize the USQs. (01022)

3. Contrary to the above, in June 1990, the licensee made a change to a procedure as described in the FSAR, which involved a USQ, without prior Commission approval. Specifically, Operating Procedure OP-402, Makeup and Purification System, was changed by Rev. 64 to allow operators to select, for Engineered Safeguards, the swing "B" makeup pump to either EDG. This resulted in an increase in the largest single post-accident load on the "A" EDG, from 616 KW ("A" makeup pump) to 691 KW ("B" makeup pump). The

l Notice of Violation 3 l

)

691 KW exceeded the largest single post-accident load, that could be rejected by the "A" EDG, specified in the FSAR (616 KW) and in TS SR 3.8.1.2.2 i (515 KW). This change in the largest single post-accident load required a TS change which was not made, and therefore resulted in a USQ. The 10 CFR 50.59 safety evaluation for this procedure change was inadequate in that it did not address electrical loading effects on the "A" EDG and did not recognize the USQ. (01032)

4. Contrary to the above, in May 1987 and in March 1992, the licensee made changes to the facility as described in the FSAR, which involved a USQ, without prior Commission approval. Specifically, modifications

)

TMAR 87-10-09-01 and MAR 87-10-09-01A changed the EFW system '

electrical power supply for the turbine-driven EFW pump alternate steam l admission valve, ASV-204, from "B" train to "A" train DC power and changed the automatic opening of ASV-204 from a "B" train to an "A" train EFW l initiation signal. The change introduced a USO in that, in certain accident scenarios, a failure of the "B" battery would cause the turbine-driven EFW pump to go to runout with its flow control valves failed fully open which would increase the probability of fai!ure of the turbine-driven EFW pump. If the event were also concurrent with a loss of offsite power, the "B" EDG would not operate (due to failure of the "B" battery). Also, the licensee's design basis relied on the turbine-driven EFW pump to share the EFW flow requirements with the motor-driven EFW pump in order to maintain the "A" EDG within its loading limits. The plant operated in various modes from 1987 through April 1996 with this design. The 10 CFR 50.59 safety evaluations for the TMAR and MAR were inadequate in that they did not address hydraulics, potential net positive suction head (NPSH) problems, a resulting potential increase in the probability of a malfunction of the turbine-driven EFW pump, or consequential effects on the "A" EDG; and did not recognize the USQ. (01042)

5. Contrary to the above, in May 1996, the licensee made changes to the facility, which involved a USQ, without prior Commission approval. Specifically, the 10 CFR 50.59 safety evaluation for MAR 96-04-12-01 (installed in May 1996) was inadequate in that the safety evaluation did not identify that removal of the automatic open signal from valve ASV-204 increased the probability of occurrence of a malfunction of equipment important to safety and therefore was a USQ. Removal of the automatic open signal from valve ASV-204 disabled one of the two automatic steam supplies to EFP-2, which reduced the reliability and increased the probability of a failure of EFP-2. (01052)

6. Contrary to the sbove, in 1994, the FSAR was revised in Rev. 21, dated December 1,1994, Section 4.3.10.1, Boron Dilution, to add information on the boron precipitation methods following a loss of coolant accident ((LOCA), and the 10 CFR 50.59 evaluation was inadequate in demonstrating that a USQ did not exist. Specifically, after identifying deficiencies in the active methods (decay heat drop line and the pressurizer auxiliary spray line) used for boron precipitation control, the FSAR and Design Basis Documents were inappropriately changed to specify flow through gaps in the reactor vessel intemals (a passive method) as the first and preferred method. This departed

i i

Notice of Violation 4 l 4

)

from the original licensing basis of the plant. Also, flow through reactor vessel intemal gaps had been identified as acceptable to the NRC (Letter dated  !

March 9,1993) only as a backup method and not as the primary method.

(01062) l These violations represent a Severity Level 11 problem (Supplement 1).

B. 10 CFR 50, Appendix B, Criterion lil, Design Control, requires in part, that measures be established to assure that applicable regulatory requirements and the design basis, l as defined in 10 CFR 50.2, Definitions, and as specified in the license application, are I correctly translated into specifications, procedures, and instructions. In addition, 10 CFR 50, Appendix B, Criterion Ill, requires that design control measures provide for verifying or checking the adequacy of design, by individuals other than those who performed the original design. It also requires that design control measures shall be applied to items such as the following: reactor physics, stress, thermal, hydraulic, and accident analyses.

The licensee's Quality Program commitments, as described in Table 1-3 of the FSAR, states that in all cases, the design verification shall be completed prior to relying on the component, system, or structure to perform its safety-related function.

Contrary to the above, measures were not established to assure that applicable regulatory requirements and the design basis were correctly translated into  ;

specifications, procedures and instructions in the following examples:

1. The design basis information from calculation E91-0026, approved by the licensee's engineering group on May 11,1989, was not adequately translated into design documents, in that, the FSAR, Enhanced Design Basis Document, and the TS Bases were not updated to state that the turbine-driven EFW pump (EFP-2) was assumed to be running when the motor-driven EFW pump (EFP-1) tripped automatically at 500 psig reactor coolant system pressure.

2. Design basis information was not correctly translated into the design input requirements for MAR 96-04-12-01, "ASV-204 EFIC Auto Open Removal," in that the previous credit being taken for EFP-2 operating after EFP-1 automatically tripped when RCS pressure decreased to 500 psig during a LOCA concurrent with a loss of offsite power (LOOP) and failure of the train B vital battery was not recognized in the preparation of the MAR. As a result, MAR 96-04-12-01, which was installed in May 1996, removed the train "A" EFW initiation and control (EFIC) automatic open signal from valve ASV-204, one of the two steam supplies to EFP-2, which would have prevented EFP-2 from automatically starting during certain accident scenarios. The design basis was not met from May 1996 through September 1996. In an event, there would have been no EFW for the period of time between the 500 psig actuation signal and when RCS pressure is reduced below the low pressure injection pump shutoff head (approximately 185 psig), when EFW is no longer required for residual heat removal.

I Notice of Violation 5

3. On December 6,1994, the design basis was not correctly translated into procedures in that, Calculation M94-0056, performed to generate Procedure OP-103B, Curve 15, Nuclear Closed Cycle Cooling System (SW)

Heat Exchanger Fouling versus Ultimate Heat Sink (UHS) Temperature, did not correctly model the heat input to the SW Heat Exchangers from the Reactor Building Fan Coolers. As a result, Curve 15 allowed a larger number of SW Heat Exchanger tubes to be blocked, which could have resulted in the SW Heat Exchanger outlet temperature exceeding the 110 F limit during '

accident conditions and the system not being capable of removing design basis heat from safety-related equipment.

4. Regulatory requirements were not translated into procedures and the licensee failed to provide measures to verify the adequacy of design by an individual other than those who performed the original design. Specifically, Engineering Procedure NEP-210, Modification Approval Records, Rev.15, dated 4 January 16,1996, was inadequate in that it allowed unverified calculations to l be relied upon to support modification installation and retum to service. As a ,

result, REA 96-047, EDG Loading Case Study, was not verified and was used  !

to support modification MAR 96-04-12-01 approval in April 1996 which contributed to the introduction of three USQs related to EDG loading.

5. As of June 5,1996, design basis information was not correctly translated into TS Surveillance Procedures (SP) SP-324, Containment inspection, SP-341, Monthly Containment Isolation Valve Operability Check, and SP-346,  ;

Containment Penetrations Weekly Check During Refueling Operations.

Engineering Procedure NEP-210, Modification Approval Records was 1 inadequate in that it did not provide sufficient guidance to incorporate containment isolation valve surveillance requirements in the review of modifications and calculations. In 1988,1990, and 1996, modifications were installed that would have required revisions to SP-324,346, and 341, to include certain valves / blind flanges. In 1991, a reanalysis of two containment penetrations resulted in reclassification of the penetrations such that a revision I to SP-341, to include valves / blind flanges in the procedure was necessary. In 1996, a review of surveillance compliance to TSs regarding containment integrity was conducted. This review failed to consider the surveillance requirements for containment penetrations in the context of maintenance l conditions in SP-346. As a result of the modifications, reanalysis and review of i surveillance compliance, SP-341 did not include 18 valves / blind flanges in the monthly performance check; SP-324 did not include nine valves / blind flanges in tile mode 4 to mode 5 surveillance requirement; and SP-346 did not include 55 valves / blind flanges in the surveillance requirement. (02013)

This is a Severity Level til violation (Supplement 1).

Notice of Violation 6 C. 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires that measures be established to assure that conditions adverse to quality, such as nonconformances, are promptly identified and corrected, in the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and the corrective action taken to preclude repetition.

Contrary to the above, the licensee failed to correct conditions adverse to quality and failed to take measures to assure that corrective actions were taken to preclude repetition of significant conditions adverse to quality as follows:

1. Precursor Card 96-2750, dated May 31,1996, and Problem Report 96-0210, dated July 3,1996, identified that changes made to the facility in April 1996 introduced EDG loads that were in excess of the TS limits, a significant condition adverse to quality; however, adequate corrective actions were not implemented. The adverse conditions were not corrected as of October 11, 1996. As a result, the plant operated for several months with USQs related to EDG loading.

2. Problem Report 94-0218, dated June 24,1994, described a problem where engineers failed to address EDG loading effects of several modifications in the 10 CFR 50.59 evaluations; however, the licensee failed to take adequate corrective actions for this significant condition adverse to quality. As a result, in April 1996, the 10 CFR 50.59 evaluation for modification MAR 96-04-12-01 did not address EDG loading effects and MAR 96-14-12-01 was inappropriately installed and placed in operation with USQs.

3. On October 12,1994, the licensee identified that penetrations were not being tested in accordance with TS 3.6.3.3, as reported in Licensee Event Report 94-007; however, the corrective actions taken for LER 94-007, dated November 10,1994, were not adequate to prevent recurrence, resulting in numerous additional valves / blind flanges that were omitted from the surveillance procedures being identified in 1996. (03013)

This is a Severity Level til violation (Supplement 1).

Pursuant to the provisions of 10 CFR 2.201, Florida Power Corporation (Licensee) 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 Re.gionM Administrator, Region 11, and a copy to the NRC Resident inspector at the Crystal River facility, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a " Reply to Notice of Violation" and should include for each violation: (1) the reason for the violation, or, if 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 previously docketed correspondence, if the correspondence adequately addresses the required response. If an adequate reply is not received within the time specified in this Notice, an order or Demand for Information may be issued as to why the licence should not be modified, suspended, or revoked, or why such other action as may be proper should not

Notice of Violation 7 I

be taken. Where good cause is shown, consideration will be given to extending the response I time. l J

l Under the authority of Section 182 of the Action,42 U.S.C. 2232, this response shall be l submitted under oath or affirmation. V 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. If personal privacy or proprietary information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that should be protected

, and a redacted copy of your response that deletes such information. If you request l withholding of such material, you must specifically identify the portions of your response that

~

4 you seek to have withheld and provide in detail the bases for your claim of withholding (e.g.,

explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.790(b) to support a request for withholding confidential commercial or financial information). If safeguards information is

, necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.

Dated at Atlanta, Georgia j this 12th day of March 1997 l I

...os3...

.  ! . i

. i i

i

FPC , 6 l .

$ 1: intribution w/encls: .

i lUBLIC 4

JCallan, 200 ,

Thompsdn DEDR l Jordan,'bEDO , ,

Tracy, OEDO

Reyes R ,  !

- Chandler,Il 0GC l i Goldber'g, .0GC 1 FZimmermanJ NRR Julian, SECY i i Keeling,CA l* nforcement Coordinators

• RI, RIII, RIV

' JLieberman, OE

' CE:EA File (BSummers) (2 letterhead)

Nsatorius, OE  !

! ONelson, OE i ayden, OPA -

l aputo, OI '

i i oss, AE00 H8e11.OIG i Evans, RIl l

. ryc, RII - ,

ohnson, RII .

i 1 audon,'RII '

. Landis,RII(IFSActionRequiredh asto, RIT i

i agahayan,'NRR  :

. ebdon,.NRR .

chin, RI! -

l rowley, RII ,

, ellen,'RII i  :

I lark, RII

! rojanowski,RI! .

i RC Resident Inspector .

i N)h.S.

t Nuclear Regulatory Commission 6 45 N. Tallahassee Road C'rystal Riv'er FL 34428 .

*in oreviam concu,rrence asse asun to ptjatle hvieterut annut vpq I j OFFICE RIItDRP Rff

DRE RfItRC IlfftElct RII da L 06.

1M1 Y! At s' "

l j MAhr *JJOHN$0W *JJA W W 5 *CEVH$ BUAVC KTEGERMAN l  ! Daft 03 / / 97 03/ /97 03/ /97 03 / / 97 01 /E / 97

) I nan =s va. un vna u.v vs. on vue on [e) en V

4

' 0FFICIAL RfCORD COPY DCCCHfWT liAME: h\1960PfW.ENF\96465tR.DIR\DisCAET.flN i

-. - - - .- . = . ..-_~ - . - - . - -. . __.

1 1

l FPC 5 Distribution w/encls:

PUBLIC I LJCallan, EDO HThompson, DEDR EJordan, DEDO ,

GTracy, OEDO

( w h * '"' ] I. #W 1Reyes, Rii LChandler, OGC JGoldberg, OGC RZimmerman, NRR l EJulian, SECY I BKeeling, CA Enforcement Coordinators RI, Rlli, RIV JLieberman, OE OE:EA File (BSummers) (2 letterhead) l MSatorius, OE DNelson, OE EHayden, OPA GCaputo, Ol Dross, AEOD HBell, OlG CEvans, Ril l BUryc, Rh JJohnson, Ril JJaudon, Ril l KLandis, Ril (IFS Action Required) l CCasto, Ril LRagahavan, NRR l FHebdon, NRR RSchin, Ril BCrowley, Ril LMellen, Rll KClark, Ril RTrojanowski, Ril NRC Resident inspector U.S. Nuclear Regulatory Commission 6745 N. Tallahassee Road Crystal River, FL 34428

• See previous concurrence page _

SEND TO PUBLIC DOCUMENT ROOM 7 [Y65) w l OFFICE Ril DRP Ri!. ORS Ril.RC RhNIC[ Ril:DRA l lNITIALS \

NAME *JJOHNSON *JJAUDON *CEVANS fu , JLIEBERMAN DATE 03/ / 97 03/ 19 7 03I 197 03 / /97 03 1 I 97 copy, vrs un vrn un vre un vrm M vrn un l OFFICIAL RECORD COPY DOCUMENT NAME: H\1960 PEN ENF\96465CR.DIROISCRETflN V

FPC 5  !

l Distribution w/encis:  ;

PUBLIC l LJCallan, EDO j HThompson, DEDR EJordan, DEDO

! GTracy, OEDO l

LReyes, Ril f LChandler, OGC JGoldberg, OGC RZimmerman, NRR EJulian, SECY BKeeling, CA Enforcement Coordinators l

Rl, Rlll, RIV JLieberman, OE OE:EA File (BSummers) (2 letterhead)

MSatorius, OE

! DNelson, OE EHayden, OPA GCaputo, Ol Dross, AEOD HBell, OlG CEvans, Ril '

BUryc, Ril l JJohnson, Ril JJaudon, Rli KLandis, Ril (IFS Action Required)

CCasto, Ril LRagahaven, NRR FHebdon, NRR  !

RSchin, Rll l BCrowley, Rll LMellen, Rll KClark, Ril RTrojanowski, Ril NRC Resident inspector U.S. Nuclear Regulatory Commission 6745 N. Tallahassee Road Crystal River, FL 34428 m SEND TO PUBUC DOCUMENT ROOM 7 L YES /

l OFFICE Ril:DRP f Ril DRSp Rit.RC Ril ElCS Ril DRA j INITIALS N

-NAME JJOH N JJAUDON CEVAN BURYC JUE8ERMAN DATE 03 /Q / 97 03I 19 7 01'd/97 03 / / 97 03/ I97 i enny, MM M /Cd vrm Ch vrm wo vrm un OFFICIAL RECOR@ DOCUMCl@ FAME: H\1960PEftliNF\96465CR.DIR\DISCRET.RV3 l

l l

Y