ML20137P609

Forwards Response from C Grimes Re TS Issue.W/O Encl

ML20137P609
Person / Time
Issue date:	05/24/1996
From:	Wiens L NRC (Affiliation Not Assigned)
To:	Mark Miller NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
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 9704090282
Download: ML20137P609 (24)

Text

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From: Leonard Wens j N f4-To: ATD1.ATP1.MSM Mark Miller, RII i Date: 5/24/96 7:31am

Subject:

TSISSUE Mark: Attached is the response from Chris Grimes. Chris tends to be a little curt, but he indicates that although not explicitly contained in one statement, the position is covered in existing guidance. I will look into it a little more myself, if for no other reason than to get a little smarter in this area myself. (Like i said, I have been asked similar questions before without having a good answer. Its about time i did I guess) -

Sorry. An old guy like me takes time to leam these new-fangled systems.

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9704090282 970407 PDR FOIA BINDER 96-485 PDR .,

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October 28, 1994 UP 4 i i

l MEMORANDUM T0: Bruce A. Boger, Acting Director I Division of Reactor Projects, RII . l FROM: John A. Zwolinski, Deputy Director (original Signed By)

Division of Reactor Projects I/II, NRR ,

SUBJECT:

REQUEST FOR ASSISTANCE IN ADDRESSING ISSUES REGARDING ST.

LUCIE UNITS 1 AND 2 REFUELING PROCEDURES - TIA 94-023 - 1 (TAC NOS. M90425 AND M90246) l We have completed the review of TIA 94-023 concerning two issues relating to  !

St. Lucie's refueling procedures. The first issue concerns whether it is the j intent of the licensee's technical specifications to have a licensed operator present as an observer during crane operation and fuel movement. The second .

concerns whether the Recommended Move List is part of the refueling procedure,  !

and subject to the license 9's approval and change process.

The review was performed by the Human Factors Assessment Branch. The  ;

evaluation of these issues is attached. ,

1 Docket Nos. 50-335 50-389 '.-

Attachment:

Safety Evaluation Report "..

cc w/ attachment: _

J. Norris &

R. Cooper, RI -

E. Greenman, RII r-W. Beach, RIV o l

Distribution Docket File PDII-2 RF SVarga JZwolir..ki RPrevette DVerrelli, RII Document Name: G:\TIA.JAN Toreceiveacopyofthisdocument,indicateinthgbox: "C" - Copy without/

attachment / enclosure "E" - Copy _wi,th,attachmen,t/Tenclosure "N" - No copy /

OFFICE LAIPOII*2:DRPE l6 PM:PDII*/:DRPE \ / l5 AD/* %/A -(;,0RPE lB UU:UKPt j/ l NAME EDunnington EG JNorrisi 4 \/ MW 4 ni JLwoli ns kA ~

DATE 10/ M /94 10/ M/9A 10/ 2.flI/96 10/ /97 j OFFICIAL RECORD COPY l 1

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.# o g UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION l

,. $ WASHINGTON, D. C. 20066 a

se,e* .

l HUMAN FACTORS SAFETY EVALUATION REGARDING REFUELING PROCEDURES i f.QB ST. LUCIE UNITS 1 AND 2 (TAC Nos. M90425 and M90246) ,

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1.0 INTRODUCTION

1 i The Human Factors Assessment Branch has reviewed the memorandum from Bruce.

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Boger to Gus Lainas, dated September 19, 1994. The memorandum addressed two

specific issues of concern with St. Lucie's refueling procedures. The first issue is whether it is the intent of the licensee's technical specifications l .to have a licensed operator present as an observer during crane operation and

' fuel movement. The second is whether the Recommended Move List is part of the 4

refueling procedure, and subject to the licensee's Technical Specifications .

(TS) requirement for review and approval of changes to procedures.  ;

2.0 EVALUATION First Issue: St. Lucie's Technical Specifications In the September 19 memorandum, Region II requested NRR interpretation of St.

Lucie's Technical Specification 6.2.2.d.

Technical Specification (TS) 6.2.2.d states:

ALL CORE ALTERATIONS shall be observed by a licensed

! operator and supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to

. . Fuel Handling who has no other concurrent

responsibilities during this operation. The'SR0 in

I charge of fuel handling normally supervises from the '

control room and has the flexibility to directly i supervise at either the refueling deck or the spent fuel  ;

pool. I l 10 CFR 50.54(m)(2)(iv) requires that "each licensee shall have present, during alteration of the core of a nuclear power unit (including fuel loading or transfer) a person holding a senior operator (SRO) license or senior operator license limited to fuel handling (LSRO) to directly supervise the activity and, during this time, the licensee shall not assign other duties to this person." The NRC interpretation of "directly supervise the activity" is that the SRO will supervise at the location of the activity of core alteratior.s and fuel movement.

In contrast to this requirement, the licensee's technical specification allows

for observation of core alterations and fuel movement by a Reactor Operator and supervision by an SRO or LSR0 who may supervise from the control room.

This position was confirmed in a letter from Region II to the licensee dated l September 30, 1981.

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, NRR has no regulatory basis for interpreting the licensee's technical specifications regarding their requirement for a licensed operator observer.

2 There is no requirement for such an observer in the regulations. However, i

because the regulations require direct supervision by an SRO or LSR0 who has no other concurrent duties, NRR believes that the licensee should modify their 1 technical specifications to bring them into compliance with the regulations.

NRR recommends that the region along wit!. NRR Projects request that the licensee modify their technical specification accordingly. Standard Technical Specification Section 5.2.2 provides an acceptable example. If the licensee does not choose to amend their technical specifications, we are prepared to Iupport a compliance backfit.

Second Issue: Refueling List as Part of the Procedure The second issue is whether the Recommended Move List is part of the refueling procedure, and subject to the licensee's TS requirement for review and approval of changes to procedures.

In previous cases dealing with this question, the NRC has determined that the ,

fuel movement list is part of the refueling procedure and any changes to the I

movement list must go through the licensee's procedure change process.

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SL.sFP I, ,

y MEMORANDUM T0: Joseph W. Shea, Project Manager Project Directorate I-2 Division of Reactor Projects - I/II.

Office of Nuclear Reactor Regulation

-FROM: Jan A. Norris. Sr. Project Manager Project Directorate II-1 Division of Reactor Projects - I/II '

Office.of Nuclear Reactor Regulation. ,

SUBJECT:

ST. LUCIE UNITS 1 AND 2 - SPENT FUEL P00L' SURVEY.

This memorandum provides the information requested by the February 8,1996, memorandum from John-Stolz regarding a review of.the spent fuel pool practices and current licensing basis.

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Attachment:

St. Lucie SFP Survey g.

Docket Nos. 50-335 and 50-389 '

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St 5FP2 , .

ST. LUCIE SPENT FUEL P0OL SURVEY A .- Spent Fuel Pool (SFP) System Design UNIT l' The system is composed of heat exchanger, filter ion exchanger, pump suction strainer ion exchanger strainer, pumps, piping and valves. The system has only one train of components. The cooling portion of the fuel pool system is a closed loop system consisting of two half capacity

- pumps and one full capacity heat exchanger. For normal refueling discharge conditions, one fuel pool pump and the fuel pool heat '

exchanger are in service. During abnormal refueling conditions, such as full core discharge, two fuel pool pvnps and the heat exchanger are in service. The-system is manually controlled from a local control panel.

High fuel pool temperature high and low fuel pool water level, and a low fuel pool pum) discharge pressure alarms are announciated in the control room. Mateup to the fuel pool comes from the refueling water tank. The heat exchanger is cooled by component cooling water. The system is designed to provide a minimum of 9 feet of water above the top of the fuel during handling and storage operation.

UNIT 2 The system is composed of heat exchangers, filter, ion exchanger. pump suction strainer. ion exchanger strainer, pumps, piping and valves. The system has only one train of components. The cooling portion of the i fuel pool system is a closed loop system consisting of two half capacity pumps and two full capacity heat exchangers. Full capacity condition

- corresponds to the design condition of a full core placed in the spent

! fuel pool seven days after reactor shutdown, in adtches, the most recent )

of which has been cooling for 90 days. For normal refueling discharge l conditions, one fuel pool pump and the fuel pool heat exchanger are in  ;

service. During abnormal refueling conditions, such as full core discharae. two fuel pool pumps and the heat exchanger are in service.

The system is manually controlled from a local control panel. High fuel pool temperature. high and low fuel pool water level, and a low fuel pool pump discharge pressure alarms are announciated in the control room. Makeup to the fuel pool comes from the refueling water tank. The j

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" heat exchanger is cooled by component cooling water. The system is I designed to )rovide a minimum of 9 feet of water above the top of the fuel during landling and storage operation.

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St. Lucie 1 i a B.

SUMMARY

OF CLB REQUIREMENTS RE: SPENT FUEL POOL OECAY HEAT REMOVAL / REFUELING OFFLOAD PRACTICES 1 1. Technical Specification limits are provided for:

TS 3.9.3: 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> minimum decay time.

TS 3.9.5: Direct communications between the control room and the refueling station during

core alterations.

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, TS 3.9.6: Mardputator crane shall be used to move fuel assemblies and be operable.

TS 3.9.7: Crane travel with heavy loads (> 2000 lbs.) over irradiated fuel is prohibited.

i TS 3.9.11: Minimum water level 23 feet above the top of irradiated fuel in the SFP.

TS 3.9.12: At least one fuel pml ventilation system shall be operable.

TS 3.9.13: Maximum load for the spent fuel cask crane shall not exceed 25 tons.

. l TS 3.9.14: Decay fuel assemblies for 1180 hours0.0137 days <br />0.328 hours <br />0.00195 weeks <br />4.4899e-4 months <br /> (1490 hours0.0172 days <br />0.414 hours <br />0.00246 weeks <br />5.66945e-4 months <br /> for >one-third core) prior to i

.j movement of the spent fuel cask into the fuel cask compartment, u

2. The fuel pool system is designed to provide shielding for irradiated fuel so that personnel dose rates do not exceed 2.5 mrom/hr; maintain pool temperature below 150 'F under offload conditions; maintain purity and clarity of the SFP, refueling cavity, and refueling )

water tank water; and maintain water level 9 feet above the irradiated fuel during transfer operations. )'

3. Design heat load for the normal batch discharge case assumes 18 batches of 80 assemblies l discharge to the SFP in 18 month intervals, followed by a discharge of 80 assemblies after i 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> of decay. With a single pump and heat exchanger in operation, the system can l maintain SFP temperature below 134 'F. Time-to-boil assuming cooling was completely '

lost at '.he maximum temperature is 13.43 hours4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br />. A full capacity pump is available should the first pump fait. [FSAR Section 9.1.3.2] Normal discharge heatload is 16.42 x 108 l Btu /hr. (Staff rerack safety evaluation dated 3/11/88] l 1  !

Does the licensee consider the failure of the fuel pool heat exchanger credible. Having a  !

l single heat exchanger does not provide for single failure (this component is passive)

The FSAR describes the normal batch discharge case as a one-third offload. In March the l licensee plans to offload a full core. Is it and abnormal offload or do they "normally" off I load a full core. If so, change the FSAR. I

$ 4. Design heat load for the abnormal batch discharge case assumes 18 batches of 80 assemblies discharge to the SFP in 18 month intervals, followed by a discharge of 217 assemblies with 169 hour0.00196 days <br />0.0469 hours <br />2.794312e-4 weeks <br />6.43045e-5 months <br /> of decay. With both pumps and a single heat exchanger in operation, the system can maintain SFP temperature below 151 'F. Time-to-boil assuming cooling was completely lost at the maximum temperature is 5.04 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The capability to

- withstand a single failure criteria was not assumed (FSAR Section 9.1.3.21. The heat load for the abnormal case is 33.70 x 108 Btu /hr. (Staff rerack safety evaluation dated 3/11/88)

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The staff also accepted a single failure of the SFPCS pump with a full core in the SFP. The maximum temperature reached 167 *F under these assumptions. [ Staff rarack safety evaluation dated 3/11/88)

The licensee's FSAR does not describe this scenario (full core offload - single failure). The licensee should be questioned whether a SFPCS single failure under full core offload heatload is part of their licensing basis.

5. The storage capacity for Unit 1 SFP is 1706 fuel assemblies (7 2/3 cores).

6. Boron concentration shall be maintained at a minimum of 1720 ppm.

7. The spent fuel poo'is designed to withstand the steady state water temperatures of 217 'F.

8. Makeup sources to the SFP are from: refueling water storage tank, city water storage tank, city water storage tanks via portable fire pump, and the primary water tank. A seismic Category I source is available from the intake cooling water inter-tie (salt) at 150 gpm (FSAR 9.1.3.4.5].

Lining up seismic makeup using the refueling water tank is complicated. The PM should review the procedure to ensure it exists and that the licensee trains on it periodically.

9. No other implicit or explicit prohibitions exist within the CLB against performing a full core offload for any given refueling outage.

Discrepancies:

None. However, see comments in each section above.

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St. Lucie 2

. -l B.

SUMMARY

OF CLB REQUIREMENTS RE: SPENT FUEL POOL DECAY HEAT REMOVAL / REFUELING OFFLOAD PRACTICES- .

1. Technical Specification limits are provided for: )

TS 3.9.3: 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> minimum decay time.

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' TS 3.9.5: Direct communications between the control room and the refueling station during core alterations.

TS 3.9.6: Manipulator crane shall be used to move fuel assemblies and be operable.

TS 3.9.7: Crane travel with heavy loads (>1600 lbs.) over irradiated fuel is prohibited.

- TS 3.9.11

Minimum water level 23 feet above the top of irradiated fuel in the SFP.

TS 3.9.13: Maximum load for the spent fuel cask crane shall not exceed 100 tons.

< TS 3.9.14: Decay fuel assemblies for 1180 hours0.0137 days <br />0.328 hours <br />0.00195 weeks <br />4.4899e-4 months <br /> (1490 hours0.0172 days <br />0.414 hours <br />0.00246 weeks <br />5.66945e-4 months <br /> for >one-third core) prior to movement of the spent fuel cask into the fuel cask compartment.

The PM should ask why there isn't a fuel building ventilation TS similar to Unit 1 TS 3.9.12.  ;

2. The fuel pool system is designed to provide shielding for irradiated fuel so that personnel dose rates do not exceed 2.5 mrem /hr; maintain pool temperature below 150 'F under ,

offload conditions; maintain purity and clarity of the SFP, refue'ing cavity, and refueling water tank water; and maintain water level 9 feet above the irradiated fuel during transfer

• operations, i 3. Design heat load for the normal batch discharge case assumes 11 batches of 80 assemblies i' discharge to the SFP in 18 month intervals, followed by a discharge of 80 assemblies after i

. 130 hours0.0015 days <br />0.0361 hours <br />2.149471e-4 weeks <br />4.9465e-5 months <br /> of decay. With a single pump and heat exchanger in operation, the system can

- maintain SFP temperature below 131 'F with a CCW temperature of 100 'F. Time-to-boll  ;

assuming cooling was completely lost at the maximum temperature is 12.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. A full capacity pump is available should the first pump fall. [FSAR Section 9.1.3.11 Normal ,

discharge heatload is 16.42 x 108 Stu/hr.

4. Design heat load for the full core discharge case assumes 11 batches of 80 assemblies '

discharge to the SFP in 18 month intervals (the most recent has deceyed 90 days),

followed by a discharge of 217 assemblies with 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> of decay. With both pumps and a single heat exchanger in operation, the system can maintain SFP temperature below 148

'F with a CCW temperature of 100 *F. Time to-boil assuming cooling was completely lost 4

at the maximum temperature is 3.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. The capability to withstand a single failure  :

criteria was not assumed. The heat load for'the abnormal case is 30.3 x 10' Btu /hr [FSAR Section 9.1.3.11. A single failure was analyzed for this heat load case. The maximum pool temperature under full core offload heatload was found to be less than 160 'F [FSAR 9.1.3.3).

5. Piping and components in the SFPCS are Quality Group C, seismic Category 1, designed for a temperature of at least 200 'F. [FSAR Section 9.1.3.2.1, and Table 9.16.]

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6. Normal makeup sources to the SFP are from the refueling water storage tank and the primary water tank. Three million gallons of makeup are also available from the city water storage tanks, condensate storage tank, domineralized water tank, and others. A seismic -

Category I source is also available through hoses from the intake cooling water header.

'lFSAR Section 9.1.3.3.11

7. No implicit or explicit prohibitions exist within the CLB against performing a full core offload for any given refueling outage, d-Discrepancies:

None T

y August 6, 1996 .

'o MEMDRANDUM T0: Jon R. Johnson, Acting Director

. Division of Reactor Projects Region II FROM: Frederick J. Hebdon, Director /s/

Project Directorate II-3 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation i

SUBJECT:

TECHNICAL ASSISTANCE REQUEST (TIA 95-013) IN ADDRESSING ISSUES RELATING TO THE' ADEQUACY OF A 50.59 EVALUATION AT ST. LUCIE UNIT 2 (TAC NO. M93372) ,

4 In a memorandum dated August 28, 1995, NRR assistance was requested in evaluating the acceptability of a 50.59 evaluation supporting isolation of a diesel generator fuel oil transfer system leak at St. Lucie Unit 2. .In addition, several generic questions concerning the relationship between Probabilistic Risk Assessment (PRA) evaluations and 10 CFR 50.59 requirements

were presented for NRR respon te.

f The Probabilistic Safety Assessment Branch, NRR, has completed its review of these issues. A discussion of these issues and NRR's response to your-questions is contained in the attached memorandum dated July 30, 1996. The positions 1 stated in the attachment have been reviewed by the Office of the i General Counsel and they have no legal objection to these positions.

l Docket No.: 50-389

Attachment:

As Stated cc w/ attachment: R. Cooper, RI W. Axelson, RIII

-J. Dyer, RIV $.

Contact:

L. Wiens, NRR\PDII-3 *

! 415-1495 7e

Distribution @

j' Docket File St. Lucie Rdg. SVarga i JRoe JZwolinski JFlack, SPSB AChaffee Klandis, RII DOCUMENT NAME: G:\STLUCIE\TIA13.RES

To receive a copy of this. document, indicate in the box: "C" = Copy without attachment / enclosure "E" = Copy with attachment / enclosure "N" = No copy orrica Pon-3/LA . _ , lo Pou 3/Pa a i IF Poli 3/o 1 10. I maar setevron e twi.n. /e rw. noon T7 f_

oare os/wm os/6 /w" os/ o /* NJ l

OFFICIAL RECORD COPY

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MEMORANDUM TO: Fred: rick Hebd:n, Director l Project Directorate II-3 i

Division of Reactor Projects I/II ,

FROM: Eherd J. Butcher, Chief Probabilistic Safety Assessment Branch Division of Systems Safety and Analysis

SUBJECT:

RESPONSE TO REQUEST FOR ASSISTANCE IN ADDRESSING ISSUES REGARDING ST. LUCIE EMERGENCY DIESEL GENERATOR FUEL OIL

• 1 TRANSFER SYSTEM LEAK ISOLATION AND USING OPERATOR ACTION IN

.! PLACE OF AUTOMATIC ACTION (TIA 95-013)

Plant Name: St. Lucie Unit 2 Utility: Florida Power & Light Co.

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. Docket No.

58-389 i TAC No.: MR3372-l Project Manager: Leonard A. Wiens Review Branch: SPSR Review Status: Complete-The attachemat to this memorandum is our response to TIA 95-013. It contains our responses to the specific questions raised by Region II regarding the 10 CFR 50.59 FPL Safety Evaluation (JPN-PSL-SENS-95-013), and the application of PRA methodology and related issues. If you have any questions regarding our. response to the TIA request or regarding the licensee's PRA assessment which was included in the TIA, please contact John Schiffgens at 415-1074 (E-mail: JOS), or John Flack at 415-1094 (E-mail JHF). In addition, we are in the process of developing a formal position on the use of PRA in the 10 CFR 50.59 process which will be sent to you in a separate memorandum.

Attachment:

As stated DISTRIBUTION Decket File SPSB File LWiens

• SEE PREVIOUS CONCURRENCES.

DOCUMENT NAME: G:\STLUCIEQ.TIA e.ea r . c r o n. n n . .een - me m e.

0FFICE SPS8:DSSA lE SPSB:DSE.I E SPSB:DSSA E SPSB:DSSA E DDSSA lE SRosenberg* JSchiffgens* JFlack* EButcher* GHolahan*

NAME DATE 5/29/96 5/29/96 5/29/96 5/31/96 6/11/96

+ RL \ /

/ l l E OFFICE PEC h HPM lE OGC /v4VI, lE I NAME DMatthews M6er AAD4f_ /

DATE (f/3K/96 1/s t96 /VI / /96 / /96 0FFI L RECORD COPY ATTACHMENT Obhb [f

p f~% \ UNITED STATES

. y N,UCLEAR REEULATCRY COMMISSION WASM?480 TON, D.C. asseMem

,,,,, July 30, 1996 4

MEMORANDUM T0: Frederick Hebdon, Director Project Directorate II-3 Division of Reactor Projects I/II

] FROM: Edward J. Butcher, Chief Probabilistic Safety Assessment Branch Division of Systems Safety and Analysis

SUBJECT:

4 RESPONSE TO REQUEST FOR ASSISTANCE IN ADORESSING ISSUES REGARDING ST. LUCIE EMERGENCY DIESEL GENERATOR FUEL OIL t

TRANSFER SYSTEM LEAK ISOLATION AND USING OPERATOR ACTION IN 1 PLACE OF AUTOMATIC ACTION (TIA 95-013) i Plant Name: St. Lucie Unit 2 l Utility: Florida Power & Light Co.

i Docket No.: 50-389 l TAC No.: M93372

Project Manager

Leonard A. Wiens Review Branch: SPS8

! Review Status: Complete The attachment to this memorandum is our response to TIA 95-013. It

contains our responses to the specific quer' tans raised by Region II regarding the 10 CFR 50.59 FPL Safety Evaluation (JN-PSL-SENS-95-013), and the application of PRA methodology and related issues. If you have any questions regarding our response to the TIA request or regarding the licensee's PRA assessment which was ecluded in the TIA, please contact John Schiffgens at 415-1074 (E-mail

JOS), or John Flack at 415-1094 (E-mail JHF). In addition, we are in the process of developing a formal position on the use of PRA in the 10 CFR 50.59 process which will be sent to you in a separate memorandum.

i

Attachment:

As stated i  ;

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RESPONSES TO SPECIFIC TIA 95-013 OUESTIONS l 1. Is the attached 10 CFR 50.59 FPL Safety Evaluation (JPN-PSL-SENS-95-013) considered acceptable?

No. The attached 10 CFR 50.59 FPL Safety Evaluation is not considered acceptable. ,

The 50.59 evaluation prepared by FPL for St. Lucie is not acceptable because it involves an unreviewed safety question. An unreviewed safety question exists because the proposed change introduces a new procedure and associated malfunction of a different type (operator error) and involves an increased probability of the malfunction of equipment important to safety (mechanical valve failure to open). Specifically, the 2B EDG fuel oil isolation valve, which is a manual valve, is

, normally in a LOCKED OPEN position and requires no change-of-state for EDG operation. The proposed change involves operating with this valve in the closed position and opening it manually as needed. With the valve in the closed position, two new failure modes exist for the fuel oil supply system: failure of the operator to open the fuel oil manual isolation valve, and mechanical failure of the valve to open. One failure mode results in a malfunction of a different type, introducing operator error where no operator action was required before. The other increases the probability of malfunction of the valve, since the

- probability of failure to open is greater than zero, where it was zero ,

before. Both increase the probability of malfunction of the 2BEDG. l In the evaluation, JPN-PSL-SENS-95-013, Rev. O, page 8, FPL improperly answered the question "Does the proposed activity increase the 4

probability of occurrence of a malfunction of equipment important to ,

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' safety previously evaluated in the SAR7" by stating that "the l compensatory actions assure the reliability of the EDG fuel oil supply."

In general, the introduction of compensatory measures suggests that l there is an unreviewed safety question for which compensation is needed, hence, a 50.90 submittal should be prepared by the licensee and evaluated by the staff to determine whether the compensation is 1 adequate. Frequently, however, licensees refer to risk reducing features that are an integral part of the change as compensatory measures. For example, introducing operator instructions for a newly i instituted manual operation should not be considered a compensatory action nor should new administrative controls intended to assure sufficient time to perform the action. Although NRC Inspection Manual, "Part 9900: 10 CFR Guidance," provides some limited guidance on compensatory actions, the staff is in the process of better defining

-what constitutes appropriate use of compensatory measures in 10 CFR 50.59 safety evaluations. In this case, the change consists of the licensee introducing a procedure and operating restrictions to replace an automatic " supply on demand" condition. The " compensatory" actions

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were intended to make the procedure as reliable as the original

'l configuration, however, they did not.

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j The licensee also stated that "the failure of the EDG fuel oil isolation valve is possible" and provided.a quantitative assessment. Its i 10 CFR 50.59 evaluation quantified the Gange in frequency of the loss i of the 8 side electrical bus in ccr. junction with a LOOP initiating J event. The result obtained was a 6% increase in the frequency per year

. of the loss of the 283 4.16ky bus in conjunction with a LOOP. The .

. report does not provide sufficient' detail on model and assumptions to evaluate the licensee's analysis. It should be noted that for an appropriate analysis-(i.e., one intended to demonstrate compliance with the provisions of 10 CFR 50.59), the change should have been assessed in

' terms of the probability of malfunction of equipment g.t the probability of occurrence or the consequences of an accident. In this case, the licensee should have explicitly evaluated the probability of malfunction

, of the B EDG.

2. From a PRA perspective, is it possible to completely mitigate a risk, l once introduced?

I

. Yes. Not only can an introduced risk be mitigated, i.e., reduced, it

can have a positive safety impact, i.e., the risk can be made lower than j it was originally. It is often a matter of economic balance; how much will it cost to reduce the risk. It is frequently possible to put in place equipment, change equipment configurations, and/or change procedures so as to effectively and satisfactorily mitigate risk (e.g.,

to mitigate the increase in risk associated with an increase in the probability of equipment malfunction or accident initiation) in a cost

< effective manner when the introduced risk is fully understood. This is significant with regard to 50.90 submittals. 10 CFR 50.59 evaluations I

are concerned with identifying unreviewed safety questions, i.e., with deciding whether a proposed change (a) may increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated, (b) may create a

, possibility for an accident or malfunction of a different type than any evaluated previously, or (c) reduces a margin of safety as defined in the basis for any technical specification. That is, 10 CFR 50.59 .11

concerned with whether there may be a decr2ase in safety not with how

[ "larae" it may be.

i 3. Is the licensee's position (that the risk of operator failure / error can F

be mitigated, probabilistically, through procedures and training) valid?

Yes. Operator failure probabilities can be reduced through the use of

improved procedures, proper training, increased knowledge, etc.

However, it should be noted that although the probability of human error can be reduced or mitigated through procedures and training, it cannot be reduced to zero.-

4 It should also be noted that for a given task, hardware is usually more reliable than human action, and the uncertainty associated with quantifying human action is usually greater than the uncertainty

associated with hardware reliability. For example, if an automatic I-1

. l

hardware action is substituted with a human action, the point estimates i of the failure probabilities for the human performance are generally

, greater than for the hardware performance and there is usually more uncertainty associated with the human error probability than there is i- with the hardware failure probability.

Do probabilistic estimations of operator error rates presuppose the existence of procedures and training and if so, can one then take credit for them in a deterministic mitigation of risk?

, In human reliability analysis (HRA), performance shaping factors (PSFs)

modify human error probabilities by accounting for the impact of various

! factors on operator actions. PSFs include procedures and training among other factors- such as stress, environmental conditions, etc. However, i

analysts will sometimes use " screening" values for human error rates. -

i These screening values are usually bounding " guesses" and may not

include performance shaping factor aspects.

Whether or not credit can be taken for the existence of procedures and training in a " deterministic mitigation of risk," is outside the purview of PRA. If by " deterministic mitigation of risk" is meant " evaluation of the mitigation of risk using techniques other than probabilistic,"

l one should be able to take credit for procedures and training in assigning an " effectiveness measure" to operator actions. The difficulty would be in devising a " measure" and applying it systematically in a deterministic framework.

4. Can 10 CFR 50.59 requirements (that the probability of failure of components important to safety not be increased if no unreviewed safety question is deemed to exist) be satisfied if new failure mechanisms are added to a previously reviewed system?

A proposed change, test, or experiment (CTE) can not be made under the provisions of 50.59 if it involves an unreviewed safety question. The stated change, resulting in the introduction of a new failure mechanism (e.g., replacing a manual valve with an MOV), would involve an unreviewed safety question, because it may result in a malfunction (of equipment important to safety) of a different type than any evaluated previously in the safety analysis report. In addition, a change which introduces a new failure mechanism, may increase the probability of malfunction of equipment (e.g., a train or system) important to safety previously evaluated, and thereby also constitute an unreviewed safety question.

5. -PRA insights are beginning to provide a more structured evaluation process for proposed changes to facilities and, as a result, are showing that changes (in a 10 CFR 50.59 context) present finite, although small, increases in the probabilities of failures. Is there a threshold value

iA -

I of increased probability (representing " negligible" or " insignificant" increases) below which 10 CFR 50.59 criteria (for demonstrat' ng that j unreviewed_ safety questions do not exist) are satisfied?

I No. According to the rule, the proposed CTE must not, nor have a l

. credible potential to,' result in a finite increase in the probability of failure in order for it to be implemented under the provisions of

• 1 10 CFR 50.5g.

5. The response to a related TIA from Region II, transmitted via letter i from you to Edward Greenman dated June 23,.1993, stated in part that "latt has no particular objection to the use of PRA in 10 CFR-50.59 evaluations but recomannds that it play a supportive role in conjunction i with other inputs, such as engineering judgement and operating experience." In the given case at St. Luc'e, when PRA insights provide j

i information counter to (as opposed to supportive to) the 10 CFR 50.59

- conclusions, is it appropriate to accept deterministic conclusions over  ;

, the PRA-indicated 1,ncrease in probabilities of failure? 1 In general, when there are differences in conclusions based on

$ deterministic considerations compared to those based on probabilistic considerations, the solution is Ag1 to simply accept one over the other but to determine the reason for the differences. Fundamentally, such l analyses, if done properly, should complement each other, the latter being an extension of the former. In this regard, it should be kept in  :

e mind that engineering judgement about components and systems is .

incorporated in PRA models, as is operating experience and associated

, data. The inputs to different assessments need to be consistent if the

outputs or tesults are to be consistent. Frequently, differences can be  !

i reconciled by identifying and evaluating assumptions incorporated in the j assessments.

i 1

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g y ...,,% j UNITED STATE 3

" NUCLEAR REGULATORY COMMISSION ,

hwAsniwarow. o.c. somewaei '

. E' I

%,*****/ 1 July 12,1996  :

t l

.: )  !

Stewart D. Ebneter, Regional Administrator, Region II ,

iEMORANDUH FOR:

William T. Russell, Director FROM: Office of Nuclear Reactor Regulatio .

SUBJECT:

INSTITUT!Ni N+1 AT ST. LUCIE f.

ThisreferstoyourmebiorandumdatedJune 28, 1996, in which you requested my concurrence in returning resident inspector staffing at St. 4Lucie j

ito N+1.

I have reviewed your request and agree that based on plant 4

performance, the exemption t6 N+1 staffing at St. Lucie that was approved i l

my memorandum dated October'2,1995, should be rescinded. l cc: J. Taylor, EDO T. Martin, RI H. Miller, RIII L. Callan, RIV V. McCree, EDO F. Hebdon, NRR l l

, )

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O FACSIMILE TRANSMITTAL

^

U. S. NUCLEAR REGULATORY COMMISSION REGION II ATLANTA, GEORGIA ENFORCEMENT AND INVESTIGATION COORDINATION STAFF To: CEM ail C EP2 p.As

SUBJECT:

1 E L bl b b (301+415-3431) C WF (301+415-2260)

C OTHER DATE:n i 5 m

^

T, U C.\ ( $ W i

i NO. OF PAGES !b + TRANSMITTAL SHEET FROM: BRUNO URYC, DIRECTOR, EICS OFFICE: (404) 331-5505-

. FAX: (404) 331-0426 g INTERNET: BXU@NRC. GOV N 3

SENT BY: TIAE:

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9 p 9 1 I

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• i STAR 951048 i

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. 07-15-1995 10:35AM St 8.ucte Residrnt Office 1 407 461 4622 P.03

'.' - ST.LUCIE ACTION REPORi mTAm y , c,gm IDENTIFICATION SECTION }- 4 , g,3 PersonW frAledne 3 ot M S [ Q b- - mrrW.Q.C R MEW eGsm 9k .5-

[ Description l AF 1/2 ooIoi LS~A "IbarA %E 9 A A % 3 une /~2.

E-oci1tr Ter A Lomumoc 9M h- VrEreMh p ht w co.ibo er m. Tw othe.r TMr h * -- c.'"'*"'"' o Fc.V crt-t A / t E

&c r-cA crr- 1 A /iR Tas, ss A chwsneiJMus- inSMuelNoSed M M hmes Tu Tut Ir As.T@,s. TW N .buh __

imlen I'aub nnd ot. Twr Vetvr I Mtry Nef Tric .s cry Ttte- TY.g E NBtuw ok T%tst AA5 To 7tess Tr+e- 51tect Tme Test'. opermorwortaround C Yes No O De, lec Correspondence 4, Audk Report, k h l Dnueno e, personnel obsensnan, one.)

4 i

4ot ASME XI

._ L _, 0 m

s. me cause of candloon a knom. _

4. Rosemmendenen a coneet and deperiment veeponsit>ie. _ _ 7 b~M - O $ / MW )

/ . , -

mn ; Heads:,;,w//M MM Dets Y / 5 ,'/foo you roovire soorovalto dose? $Yes O No

--f REVTEW/ APPROVAL ,j

1. Assigned Depenmort. W/ K p- On/'M Grw 5 66. udy e' - ,

ANii 2 isi 8 operandive eeeemeata auir o: JPN O oes O

2. Reviews STAO

3. NP 700 0 HPES O tHe O T**'*=' 8****** * "*a" O

4. Evolunden due try1./.[_f.2d.M to: I .1 inteleis:

E 5. Conecthre measures completed try _ 1 / . Extended to: 1 l_ inftels: j s wi7so 4 s )

2 1 2

6. Isinomamodebokf?C Yes [No G [ ] [ [ [ -

commengg. Os vch t V YCl? sa Pc n c e 70 ,

<~,onitte r 7Mtf k4'Ac ric C. fusrwFr NM l ADwJrD TM4r ryi3_ we s 2bd . -

l

~

Date _ 9 / 6116 Doyou require approvalto cose? e8 L_lNo i

1. N 0.e., AP 0010721,'NRC RegJired Non Routne Nodnestons & Repor: - O vee O no AP 0005782. ' Plant Guide to Reportin0 ErMronmental Non-Corrp0ances and SignBcant Events' )

seomein, Repo+.e see ue, .everm g

2. Ehent Type 14 -

senature omne

3. Securtly overt C Yes O No O POP

)

(Rev.1 Nee)

(Gh1648.WPG) i I

E 07-15-1996 10:36AM St 1.ucie Residrnt Office 1 407 461 4622 P.04

1. # I *--- O'*

• D sTMeR*5IOW$

oats laeviewwTw.melAssiniones l ooer

% coCN i

l A. ?-_ _ ~ ,lz Z of a non confom*g hem is h queston and operespils req *ed gyg

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I M essessnosneeded Ereon O COMP rm jgg g mised g. ggnsm is que.nonepa (as sound coneson).

Technical O G.so.80lesu. O H. ASME)0 issue n =v me* peveuou)w cheew ane.,to mes.

Rossentan tendred. Oste ] / - Mode condWonalreiseas to use asis Oyes O No (supportng domsneniston require @

ohpossoned my vermed by eonusemenerkimpowHowtoPmegnesponsej nesponsasePerson In#aao FC - / - 9 $~- 2.<S Wns (A/&wica Ao 5coconnrw Z.ro / ~ 6 0/ Beed 0 tl.'i A m T/zh e- Lu ous_e Cw e r Pmur m f+rre urn) //svr Tun amco Le L.) . w Gr Ema n powe n L &c Alen. Asn < i .+.1 ie-

1. W ouisse vehisse used consessed

! Requted Dess conoceve Ansons Aeolensdio Dais (PCM Wonunter)

SNo REF

1. PCL '%nO4 _ - _,,J J Q Q F99 *%*5=b(de \t l_t Ib*5 L . J. J.- O C -.

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~~f Cf.OSKOVTSECTION  ; .

1. Air conome seson. ,r,i Inm. peore below; tt /[b8/b Y/,MA ignedo ,anmem l

z inne.n. os. ore r u =n.or,.n.e M Je UL ,diik G ,o,e.,e,.no

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15 IO N st s.ucie Restd:nt Offico i 407 461 4622 P.05

' QI 5 PR/PSL 1 Revision 62 l i .

May,1995 l Page 83 of 96 l

P!GURE 1 1 PROCEDURE CHANGE / REVIEW REQUEST (Sheet 1 of 2)

PROCEDURE TkLE Svr vn lln orea $n b SSW2h ___

_Presem Rev. No. # ?

PROCEDURE NUMBER ~ /- Oo Id(2.f A /R62 Q 1.1 Response to INPO. QA. or 31.5 PC/M of STAR

1. .61M2n Q 1.6 ProceduralIr piri.wat /R62 f or Recuest NRC Requirements

' 01.2 Cycle specifir:/ documented Setpoint Revision. or Vendor Reymon g 1.7 Tech. /R62 Manua i

Q 1.3 Tech Spec. Rev4.lconsing Requirement

, g 1.8 Technically incontet /R62

' (Ust Tech. Spec. Amendment No. below)

/R62 a 1.9 Other

' 01 A Affects Plant Operability / Availability e if ro talled e nation: N!vt l loclude Mys, and source (t.e., PC/M e. STpfr e),e D$r ASW

< I 1lfoaU dn$ be D/'tCsMou .s ra e /~*Sescs n su e ve d6,,'e e l .

fQg l l

Attach a copy of the affected pages of the present revision. Changes should be legible i 2

RED Ift directly on the affected page. If extensive additions are focuired use addPJonal  ;

' sheets as necessary; clearfy indloats the proper placement on the appropriate page of the l revision. Highlighters. conwetion fluid or any other obilteration meterial should not be used. _I a f

a. le this change to a un:t specific procedure? fYes _No cvwiis picc+Are 2

X.,,,yes ,,No

b. 't yes, has a PCR for the opposite unt been sutmtted?

If no, explain

' c. Does this PCR Incorporate a T.C.7 Des _No T.C. s (if applicable) Mf .2 N l

' d. Does this PCR teference any enemical? _Yes M l If yes, Chemical Control Supv. Review l '

3. Periodic Review (Check if Mahle)

Check below only if performhg a review where no changes are necessary.

O Review perforrned, no changes (FRG not required).

. 4. Reauested by:

J,,, <e.,, Phone J an .

~(print name):

Date: ~T/A/ I~

IAtft %4e'h j (Signature):

/d 7-Subcommitteed by: U@d- _ -

Date- 9/ .

Date: / / _,

Dept. Head:

)

Prionty:(determined by Dept. Head) 01 02 03 04 Required by /J (if app 8 cable)

I cATE DOCIf DOCN__

sYS I

ITM a

1 i i' a i - i is =i . , ,

l

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

Page 32 ~ 236 .-

a ST.Lb . UNIT 1 e

• ADMINISTRATIVE PROCEDURE NO. 1-0010125A, REVISION 39 SURVEILLANCE DATA SHEETS J DATA SHEET #8A VALVE CYCLE YEST - NOf6 CHECK vat.VES (Page 5 of 9)

[j Note: S in required stroke time column indcates ESFAS Valve.

SaSefact-

.. e, -- ,,e,,,e ve,s.

Eserdeo Emeer Mestaura Enter esa ftestored aselem Vetus Actuut Reghod Acted Required I.V. Reedred Test -

Vahm No. Desedpass (kelais) Tkne Tims OC PostAuri Peelson (tnants) After Teel Masted Resnaes rSE-274* $@ kto A H, Analyrer Cleeed B

^

FSE-27-s* Sample kee 9 H, Aneiper Cassed B 2 O NfA NfA I3C-27-10* Semple iseen 8 H Analyter C30 sed B Y . rse.27 ti- se been A H. Analpa 2 O t9A I NTA j 2 C FC u

a rcvas ie- ser se,e,as uneincasem 2 c rc w a O b,) mot c rc

= reve>ir- sirsanvastawi.e na. 2 ca eed a q j E 2. s r -

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07-15-1996 10:30AM St Lucie Residant Office  ; 407 461 4622 P.08 4 -

' 015 PR/PSL 1 Revision 62 May,1995 Page 83 of 96 FIGURE 1

PROCEDURE CHANGE / REVIEW REQUEST _

I (Sheet 1 of 2)

PROCEDURE TkLE JurMtlNAf9 AYA J PP Ptosent Rev. No. #f 3 PROCEDURE NUMBER '2_ - 9 # / S r 2 f A g.1.5 PCM or STAR /R62

1. Res E U. O 1.1 Response to INPO, CA. or /R62 NRC Requirements C 1.6 Proceduralimprovement for Recuest O 1.2 Cycle Spect'ic/ documented Setpoint Revision. O 1.7 Tech. Manual Req or Vendor Reveen /R62 O 1.3 Tech Spec. Rev/Ucensing Requirement ,

l (Ust Tecn. Spec, Amenoment No. becw) 01.s Technica:ly inconect /R62 Q 1.9 Other /R62 O 1.4 Affects Plant Operability /AvaQM!ity Include depnption and yource (i.e PCM #. SJAR #), if her give deta: led exp ShuId nef kr a te rre dition e (M) DSor 'fo l

1 s ur ve r /fa ,e zil 37v l~9CofL% '

i E

' Attach a copy of the affected pages of the present revision. Changes should be legible in RED ink directly on the affected page, if extensive additions sre required use additional sheets as necessary; clearty indicate the proper placement on the appropriate page of the old rey.sion. Highlighters, correct!cn fluid or any other obilteration material shou!d not be used.

i

&[Yes __,No comm0n procedure

2. a. is this change to a unft specific procedure?

b. If yes, has a FCR for the opposite unit been submitted? _fYes .,,,,No If no, explain __

l _No T.C. e (if applicable) 1- f l~-/ 7 3

c. Does this PCR incorporate a T.C.? 5Yes _

d. Does this PCR reference any enemical? .,,_Yes If yes, Chemical Contel Supv. ReWew

3. PeriodkReview (Checkif applicaDie)

Check below only if performrig a review where no changes are necessary.

O Review performed. no changes (FRG not required).

~

4 Recuested by:

J dd [

(pnnt name): dr1 k4 Pem Phone (S8gnature): MM N /M Dato: 7. / /,, J Y Subcommitteed by: "Wt'h Date: 9._/ E/D Dept. Head: Date: / /__

Priert!y: (deterruned by Dept. Head) 01 02 03 04 Required by / / _ (if app!! cable) s_.,, OPS DATE f oocT M

l sYs CCidP tN

9 9

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STAR 951063 1

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g, gwn y.g e (cg. . ;i S. Senser eveN ' Yes No ("] POP , ;i , ,

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- + .6

i - 15-1996 ist41AM St Lucle Reeident OffIc2

• 407 461 4622 P.12

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inter 0ffloe C+,ir; -Cm j

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f October 5,1995 if pese:

I To: Paul Pulford .

Ops Test *g I pe= J les oseenment:

- .h3:rg

(..jp M

seet: STAR 951063 Response ,

i I have reviewed tlw test and saveillance procedwas permining to the inser fp i

penpa and valves r k both Units 1 and 2. I fond no fath . . . .. . .,

I TheplusdoesroiminelypuformPMspriorto schedialedWsavoillaises.M 2 ..

l

' as a matter of convenienses since a saveDianos re is required ibliowing the PMM.9y . . .

' scheduling the PM work to be perhnned prior to'this'acheduled survelusdos'this iiue i

swwillanoes performed is reduosd. His is desirable because this Iknits the number P. ' .of demands placed on the pumps, the amount of operecer manpower needed to support' ,,, i l

and reduces the unavailabilty for pumps which are taken out of servios to perkn,aSthe'.

l -

surwillance. Attached is a listing of the PMs performed prior to scheduled survaillana=a . . . . . . ,

l ' l Dess PMs are in two maior categories; pumps and ihns, and valves and AFW Terry Turbine.

l 1 have myiewed these PMs with the applicable SCE and Predictive Mainamanana Engineers l

and have conectively coene to the hilowing two conclusions. ...

Por the first category, the PMs are fbr pump oil change out, coupling lubrication and ihn j

lubrication. %smmeconsidend tohaveaminor,ifany,impactontimi f mofdu i j

pumps and fans during the surveillances. Does PMs me performed less Aequently thll l

quarterly surveillances with no indication that the perfbrmance or lack of the activities i

influenced any surveillance. Also, some of the components me run during normal plant operations, and problems with Charging pump accumulator pressure and other panps l fans are detected by Operations during normal observation of the components. Herefore this l

first category of PMs should continue to be perfonned prior to sobeduled surwulances.

l i

De PMs perfbrnwd on FCV471 A. FCV471B, and the AFW Tr'9.Turbins, and ,

Governor valves may have an impact on the performance of these P %e tests perenned on als Unit 2 MFIVs how no ef5 set of the safety related fianction of the "- MFIV surwillanos, nis estegory of PMs with the exception of the MFIVs should be raahlad l 6em just prior to the perfbrmance of the surveillances associated with theos components. l l 1

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o ST. LUCIE UNIT 2 .

ADMINISTRATIVE PROCEDURE NO. 2-0010125A, REVISION 43 -

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SURVEILLANCE DATA SHEETS y DATA SHEET #8B ' h VALVE CYCE TEST - NON@ECK VALVES

{Page 11 of 13) E Note: S in reqthed stroke time Column indicates ESFAS valve.

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, ORA :404-331-6471 1 ** Transmit Conf. Report **

Jul 15 '62 12:08 ORA - - > 8-301-415-3431 tb. 0003 Mode tG%L Time 4'30' Pages 14 Page(s)

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'JNITED STATES y j g /,,

NUCLEAR REGULATORY COMMISSION - ,

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96 JL 10 M1 :46

, July 2,1996 7/

7 MEMORANDUM TO:- Jon R. Johnson, Acting Director / I^) ' '

Division of Reactor Projects, RI'I FROM: Frederick J. Hebdon, Director Project Directorate II-3 Division of Reactor Projects I/II, NRR

SUBJECT:

TECHNICAL ASSISTANCE REQUEST (TIA 96-007) REGULATORY ACCEPTABILITY OF LUBRICATING VALVES PRIOR TO SURVEILLANCE TESTING (TAC NOS. M95274 AND M95275)

In a memorandum dated April 12, 1996, as a result of valve stroke timing practices at the St. Lucie Plants, you requested NRR assistance in evaluating

, the acceptability of lubricating valves prior to the performance of stroke time testing. You also asked NRR to resolve a question as to whether the purpose of the stroke time testing was to demonstrate current and past operability of a valve, current and future operability of a valve, or both.

The Mechanical Engineering Branch (EMEB), NRR, has completed its review of these issues. A discussion of these issues and NRR's response to your j questions is contained in the attached memorandum dated June 24, 1996. I l

, Docket Nos.: 50-335 and 50-389

Attachment:

As Stated )

cc w/ attachment: R. Cooper, RI W. Axelson, RIII J. Dyer, RIV

)

Contact:

L. Wiens, NRR\PDII-3 415-1495 4

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C L% RAD mi~-

/ ,/ 7,

1 .

l June 24, 1996

, MDt0RANDLM TO: Frederick J. Hebdon, Director  ;

Project Directorate II-3 Division of Reactor Projects I/II FROM: Richard H. Wessaan, Chief i Mechanical Engineering Branch Division of Engineering I

SUBJECT:

TECHNICAL ASSIS'ANCE T REQUEST (TIA 96-007)

! REGULATORY ACCEPTABILITY OF PRELUBRICATING VALVES (TAC Nos. M95274/M95275) 3 In a memorandum dated April 12, 1996 Ellis W. Marschoff, Director, Division

of Reactor Projects, Region II, discussed the determination by Region II inspectors that the licensee of the St. Lucia nuclear power plant had lubricated a containment spray flow control valve prior to performing stroke

time testing under Section XI of the ASME Boiler & Pressure Vessel Code. The i Region II inspectors considered this pre-lubrication to result in a nonrepresentative test of valve capabilities.

i

! Region 11 requested the Office of Nuclear Reactor Regulation (NRR) staff to respond to specific questions on the acceptability of the licensee's actions l in pre-lubricating valves prior to testing. Attached is our response to those questions.

2 CONTACT: T. Scarbrough, DE/EMEB

415-2794 Docket Nos.: 50-335 50-389 Attach:nent: As stated

! cc w/ attachment: J. T. Wiggins A. F. Gibson G. E. Grant T. P. Gwynn Distribution:  :

1 Central Files EMEB RF/CHRON LWiens RCroteau

Valve List DOCUMENT NAME

G:\SCARBROU\RHWLURE and PRECOND or. i.e.,y vien .et.cn ne/. net ure n = m. ec te r.c.iv. . e.,r .e eni e.co.ne, insie.t. in en. m c<ew /. .et.en ne/.nei 0FFICE EME9fDC 6 EMEB:DE N E ,

} NAME TScadrY9 RWess W DATE $ /8/96 lW / 96 0FFICIAL llECORD COPY

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ATTACHMENT

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1 REGULATORY ACCEPTABILITY OF PRELU8RICATING VALVES

PRIOR T0 SURVEILLANCE TESTING (TIA 96-007)

Technical Assistance Ra st ,

i In a memorandum dated April 12, 1996, Ellis W. Marschoff, Director, Division l of Reactor Projects,-Region II, discussed the determination by Region 11- l inspectors that the licensee of the St. Lucie nuclear power plant had 1

lubricated a containment spray flow control valve prior to performing stroke-  ;

i time testi under Section XI of the ASME Boiler and Pressure Vessel (B&PV) I j Code. The ion !! inspectors considered this pre-lubrication to result in a nonrepresentative test of valve capabilities. Therefore, Region II requested a response to the following questions:

i 1. Is the practice of lub -icating a valve prior to stroke-time testing I l

acceptable under the regulations?

! 2. Is the purpose of stroke-time testing under ASME Section XI to

demonstrate the current and past operability of a valve, the current and 4 future operability of a valve, or both7

]

l Evaluation

The NRC regulations in 10 frR 50.55a require that nuclear power plant licensees provide valves and pumps within the scope of Section XI of the

! ASME B&PV Code with access to enable the perfomance of inservice testing of those valves and pumps for assessing operational readiness as set forth in Section XI of the ASME B&PV Code. Criterion XI, " Test control," of Appendix 8 l

to 10 CFR 50 requires that testing be performed under suitable environmental conditions. The current Inservice Testing (IST) Programs at St. Lucie Units 1 l

l and 2 are based on the requirements of Section XI of the ASME B&PV Code, 1986

Edition, with approved relief to certain requirements. Article IWV-1000 of i

ASME B&PV Code (1986 Edition),Section XI, states that it'provides the rules

- and requirements for inservice testing to assess operational readiness of

, certain Class 1, 2, and 3 valves in nuclear power plants, which are required to perform a specific function in shutting down a reactor to the cold shutdown

condition, in mitigating the consequences of an accident, or in providing

. overpressure protection.

l Subarticle IW-3417 of the 1986 ASME B&PV Code states that, if a valve fails

to exhibit the required change of valve stem or disk position or exceeds its specified limiting value of full-stroke time by this testing, the licensee shall initiate corrective action immediately with the valve declared i Generic Letter inoperative if the condition is not corrected in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(GL) 89-04, " Guidance on Developing Acceptable Inservice Testing Programs " in Position 8 indicates that, rather than delaying 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the licensee should make a decision on operability when the data is recognized as being within the required action range. GL 91-18, 'Infomation to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming ATTACMENT

J Conditions and on Operability," provides similar guidance on the timeliness of operability decisions based on test resMts. IWV-3417 also requires that the

test frequency be increased if a significantly longer stroke time is observed since the last test. Finally, IWV-3417 requires that any abnormality or i erratic action be reported. The St. Lucie IST Program Plan identifies no

! differences in interpretation of the NRC regulations or, ASME Code when stating

. that the inservice testing in the plan is to be performed specifically to

, verify the operational readiness of pumps and valves which have a specific i function in mitigating the consequences of an accident or in bringing the

reactor to a safe shutdown.

. More recent ASME codes and standards have repeated and amplified the importance of evaluating the operability of valves during inservice testing.

For example, Subsection ISTC, " Inservice Testing of Valves in Light-Water j Reactor Power Plants," of the ASME Operation and Maintenance (0Mc) Code states that it establishes requirements for inservice testing to assess the e operational readiness of certain valves and pumps used in nuclear power

! plants. Subsection ISTC 4.2.9 requires that the valve be immediately declared j inoperable if the valve exceeds the limiting values of full stroke time.

Subsection ISTC 4.2.4 also requires that any abnomality or erratic action be '

j recorded and that an evaluation be made regarding the need for corrective j act1on.

s

! The NRC regulations, and ASME codes and standards, clearly indicate that the i' purpose of the inservice tesi.ing programs is to " assess" the operational readiness of the valves and pumps. Article IWA-9000, " Glossary," of ASME B&PV i Code (1986 Edition),Section XI, defines " assess" as determining "by i evaluation of data compared with previously obtained data such as operating data or design specifications.' Nore generally, Webster's II New Riverside

'. University Dictionary defines " assess" as "to appraise or evaluate." If

! maintenance is perfomed prior to inservice testing that ensures the l capability of a valve or pump to operate properly, the licensee's IST protiram j would be unable to evaluate the operational readiness of the component. "his i is reinforced by the requirement in the ASME Code that, if the stroke-time l limits'are exceeded, the condition be corrected or the valve be considered

inoperable. The St. Lucie IST Program Plan intent "to verify the operational readiness" is more specific regarding the purpose of the. testing to determine

the capability of the valves to perform their safety function.

The ASME Code recognizes that routine preventive maintenance will be performed l by licensees. In some instances, this maintenance may occur shortly before a i scheduled test required by a licensee's IST program. The effect of this

! maintenance on the validity of the test to assess operational readiness should be evaluated. In Section 3.5, ' Testing in the As-Found Condition," of

}_ NUREG-1482 (April 1995), ' Guidelines for Inservice Testing at Nuclear Power i Plants," the staff stated that the Code does not specifically require testing to be performed for components in the as-found condition except for safety and

relief valves, but does not define as-found even in the context of safety and relief valves. In NURES 1AA?, the staff noted its belief that most inservice

~

testing is performed in a manner' that ger.erally represents the condition of a

standby component if it were actuated in the event of an accident (i.e., no

pre-conditioning prior to actuation).

i 2 i

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• In NRC Infomation Notice 96-24 April 25 i Case Circuit Breakers Sofore Surv(eillance, Testing," the staff stated s

practice of preconditioning molded-case circuit breakers (for example, by lubricating of the periodic pivottast.

points and manually cycling the breaker) defeats the purpose The staff stated that such preconditioning does not confirm continued operability between tests nor does it provide information on

! the condition of the circuit breaker for trending purposes. The applicable l licensee planned to revise its procedures before the next surveillance test to correct this situation.

j i In ASME Code Case OMi-1, " Alternative Rules for Preservice and Inservice

Testing of Certain Electric Motor Operated Valve Assemblies in LWR Power i Plants (OM - Code - 1995 Edition; Subsection ISTC)," the ASME provides an alternative to the stroke-time testing requirements of the OM Code to assess

' the operational readiness of actor-operated valves (MOVs). The code case uses the same language as the NRC regulations and ASME Code in stating that inservice testing is intended to assess the operational readiness of valves.

i In implementing the code case, the licensee is required to determine the capability of the MOV during inservice testing. The code case requires MOVs to be on cycledintervals.

periodic at least every refueling cycle with diagnostic testing conducted-4 The code case allows grouping of MOVs with the information obtained from individual MOV tests applied to other MOVs in the group.

In Section 3.3, the code case specifically states that maintenance i

4 activities, such as stem lubrication, shall not be conducted if they might invalidate the as-found condition for inservice testing. The performance of maintenance prior to testing would defeat the ability to determine any

degradation in the operation of the tested MOV and to apply the test results to other MOVs within the group. This code case is being endorsed (with certain limitations unrelated to preconditioning) for voluntary use by licensees in a forthcoming generic letter.

In summary, the performance of maintenance en a component to ensure its proper j operation prior to conducting a test negates the validity of the test in assessing the operational readiness of the component. If the maintenance had

, not been safety pertcced, the component may not have been capable of performing its function.

i Clearly, the conduct of maintenance prevents the licensee i

from upon. assessing if the component would perform as design, should it be called j Further, important information on trending of operating parameters for evaluating degradation would not be available.

t

! EMEB Resoonse In response to the specific questions from Region II:

! 1.

- The performance of maintenance that ensures the capability of a valve to satisfy the stroke-time test requirements of the ASME Code provides a falso indication of the operational readiness of the valve. Therefore,

a licensee activity to lubricate a valve prior to stroke-time testing for the principal purpose of satisfying the test criteria at that specific time would not be considered to be within the intent of the NRC regulations under 10 CFR 50.55a or Appendix 8 to 10 CFR 50. It is recognized that routine preventive maintenance, such as valve I

3 I

- _ - - - - .-w-_ -

4 j lubrication, might coincide occasionally with IST program testing. In those cases, the effect of such maintenance needs to be evaluated to I ensure that the ability to assess operational readiness of the valves and to trend degradation in the valve performance are not. adversely

, affected.

2. The NRC regulations, and ASME codes and standards,; require licensees to

estabitsh ST programs to assess the operational readiness of certain

valves and pumps. If a valve falls its stroke-time test, the licensee j is required to declare the valve inoperable. Therefore, the stroke-time test is intended to demonstrate current operability. The licensee evaluates past operability since the previous stroke-time test based in part on the most current test results. The ASME Code prescribes

comparison of stroke-time test data to previous test data so that i

! licensees may obtain an indication that the valve should remain operable until the next test. It is recognized that the stroke-time test is limited in its effectiveness and, as a result, the ASME developed an alternative IST approach for MOVs in ASME Code Case OMN-1.

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From: James D. Dockery (JDD)j01 To: BXU id()fyd, 4/7

- Date Monday, . October 30, 1995 9: 44 pm

' Subjects- PII-95-A-OO26 Bruno,- ' Subject allegation pertains to elleger Gery DNIPPS. found by DOL'WSH Div, to have been discriminated aga!rct by F:&L. Et.

Lucie NP"menegement. " Chill.ng Ef f ect' leite- ar* :v+ to CP&L

' ~10/6/o5 .sith-30 cro res;r-se re .drec. It : my .nde-ctending t~n . y r.0 "oc s e : .; :n ars v. w r ; the -irst tc see . :. c e n s c o e:n reu t : .'s u c h l e t ter t. . I'coticica::e *:rmal interview of AHfor5, ASAD.:but in o-der to (possibly) :.reclude in terviewing hi,m twice. I arr very.2nterested in reviewing the licensee's rendition of'evente. -It would be't. ore efficient tc cfford PHIPAS

- the oppo-tunit, to try'to'refu'te the'IIcensee~s' assertions when !

- interview him.the first ti e.

Irgo. I woulc appreciate if ycu/ycur c4 dice cc.Id Iet 'ne P r;ow wher the licensee -esocnte te Mr..Ebneter*s *etter is 'eteived' .

or, if the-11;:ensee requet,ts an' entensico tc the 50 day -ecponse recuiremert (.J f t*ey're allowed to do thet7). Approct. ate your

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