3F0298-17, Forwards List of Commitments,Conclusions Presented in 980130 Telcon,Description of long-term CAs Required to Resolve Conflicts within FSAR & Summary of Previously Docketed Correspondence Re Protection Against Effects of LOCA

Forwards List of Commitments,Conclusions Presented in 980130 Telcon,Description of long-term CAs Required to Resolve Conflicts within FSAR & Summary of Previously Docketed Correspondence Re Protection Against Effects of LOCA

ML20202H887
Person / Time
Site:	Crystal River
Issue date:	02/18/1998
From:	Rencheck M FLORIDA POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
3F0298-17, 3F298-17, TAC-M96604, NUDOCS 9802230039
Download: ML20202H887 (12)

Text

- - - - -

e9 CORPORATION Florida Power

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w. om n February 18,1998 3F0298-17 U.S. Nuclear Regulatory Conunission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Protection Against Dynamic Effects of LOCA - Request for Additional Information (TAC No. h196604)

References:

1.

FPC to NRC letter (3F0198 34), dated January 23,1998, " Protection Against Dynamic Effects of LOCA Request for Additional Infermation (TAC No.

h196604)"

2.

FPC to NRC letter (3F0198-31), dated January 28,1998, " Protection Against Dynamic Effecs of LOCA - Request for Additional Information (TAC No i

h196604)"

3.

FPC to NRC letar (3F0298-08), dated February 3,1998, " Licensee Event Report (LER) 50M2/98-001-00: Systems, Structures and Components Were Not Protected From the Dynamic Effects of a Loss of Coolant Accident"

Dear Sir:

BIR On January 30, 1998, Florida Power Corporation (FPC), NRC Region 11, and NRC NRP EliE.

personnel participated in a teleconference to discuss recent discoveries made by FPC during g

investigation of the effects of pipe whip and jet impingement resulting from loss of coolant n==.

accidents (LOCA).

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G.

During the teleconference, FPC was requested to provide further discussion of the identified g,

impacts from possible pipe whip or jet impingement induced failures for any identified Post g~,

Accident Sampling System (PASS) components, Post Accident hionitoring (PAhi) instrumentation, and any other structures, systems, and components (SSCs) described in the improved Technical Specifications (ITS). FPC stated in the teleconference that four specific SSCs included in ITS and one additional PAh! instrument (hot leg level) not included in ITS were affected by the dynamic effects of certain specific LOCA events.

After the teleconference. FPC determined that_ the hot leg level instrument was also required to be operable to meet the requirements of the ITS. 1lowever, the conclusions presented in the teleconference that all affected SSCs required by the ITS were operable remain valid. These conclusions are discussed in Attachment B.

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U.S. Nuclear Regulatory Commission 3F029817 Page 2 Attachment C describes the long term corrective actions required to resolve the identified conflicts within the Final Safety Analysis Report (FSAR) and previous dor.keted correspondence regarding design criteria for protection of SSCs from the dynamic effects of a LOCA, commitments to Regulatory Guide 1.97, Revision 3, and other design criteria.

Attachment C also summarizes the previous docketed correspondence related to this issue to clarify the ;urrent overall corrective actions to be implemented.

]

Attacliment A provides the list of commitments made in this submittal.

If you have any questions regarding this letter, please contact Ms. Sherry Bernhoft, Manager, 1

Nuclear Licensing at (352) 563 4566.

Sincerely, Fodt Mtteit Rta4e(4r5cid-w M.W. Rencheck Director, Nuclear Engineering and Projects MWR/gew Attachments cc:

Regional Administrator, Region 11 Senior Resident inspector NRR Project Manager

U.S. Nuclear Regulatory Commission Attachment A 3F0298-17 Page1of1 ATTACilhlENT A I,1ST OF COMMITalENT_S The following table identifies those actions committed to by Florida Power Corporation in this document. Any other actions discussed in the submittal represent intended or planned actions by Florida Power Corporation. They are described to the NRC for the NRC's information and are not regulatory commitments. Picase notify the Manager Nuclear Licensing of any questions regarding this document or any associated regulatory commitments.

ID Number Commitment Due Date c

3F029817-1 FPC will evaluate the conflicting design basis requirements m the FSAR May 29,1998 and previous docketed correspondence in accordance with the CR 3 corrective action program to determine the correct design criteria for protection of safety-related SSCs against the dynamic effects of a LOCA inside containment, including Regulatory Guide 1.97 required instrumentation.

FPC will then initiate additional ei.gineering and/or licensing corrective actions, if required, and initiate revisions to the affected FSAR discussions as necessary.

U.S. Nuclear Regulatory Commission Attachment B 3F0298-17 Page I of 5 NITACilSIENT H IhlPACT OF POSTULATED DYNAhllC EFFECTS OF LOSa OF COOLANT ACCIDENTS ON SAFETY RELATED STRUCTURES, SYSTEAIS, AND COSIPONENTS INSIDE CONTAINhlENT IMut During investigation of the possible dynamic effects of a loss of coolant accident (LOCA) on safety related structures, systems, and components (SSCs) inside containment, Florida Power Corporation (FPC) discovered that certain postulated breah of Reactor Coolant System (RCS) piping would possibly damage or otherwise render nonfunctional several important pieces of equipment as a result o' pipe whip or jet impingement effects. FPC had initially identified that Post Accident Samr.ing System (PASS) components, Post Accident hionitoring (PAhi) instrumentation,

,md other SSCs identified in the improved Technical Specifications (ITS) were potentially affected.

FPC has since evaluated each of these potential targets as documented in a Deficiency Ikport (DR)

I made available to the NRC Resident inspector to ensure sufficient justification for continued operation exists.

J'pt Aceldent Sampline System Components FPC has determined that no PASS components are affated by postulated pipe whip or jet impingement effects as documented in the DR. Ilowever, a normal water sample line from the pressuriier could be affected by jet impingement from a postulated pressurizer spray inne break at one of several locations or a postulated pressurizer surge line break at a single k) cation. This sample line is used as described in current plant procedures to obtain a pressurizer water sample for further processing by the PASS. Ilowever, this specific sample point is not a required PASS parameter to be taonitored after an accident.

Post Aceldent Stonitoring_(PAND Instrumentation Three specific PAhi instruments could be affected by postulated pipe whip or jet impingement effects as documented in the DR.

1. Reactor Vessel Level (RC-164A-TEl)

RC-164A-TEl is a temperature element used to compensate the reference leg for the " A" channel (RC-164A LRI) of the Reactor Vessel Level Indication System (RVLIS). There are two channels of RVLIS (RC-164A-LR1 and RC 16411-LRI) qualified as Type 11, Cater v" 1 instruments according to Regulatory Guide 1.97, Revision 3 requirements. Both <

channels are required to be operable in hiodes 1,2 and 3 as described in ITS 3.3.*

3t Accident hionitoring Instrumentation) and ITS Table 3.3.17-1.

Imoact Eval.nali2D A circumferential terminal end break at the reactor vessel nozzle of the core flood line (west) could result in jet impingement on a conduit (RCR-315) potentially rendering RC-164A TEl

U.S. Nuclear Regulatory Commission Attachment Il 3F0298-17 Page 2 of S nonfunctional. This, in turn, would affect the accuracy of RC-164A LR1 by as much as 10.9%

of full scale (by changing the temperature compensation of the level signal).

RC 164A LR1 is not used to mitigate or limit the consequences of a LOCA. It can be used to trend reactor vessel level, but only without significant RCS flow through the core. The primary indicators used to ascertain adequate core cooling following a LOCA are RCS temperature and RCS pressure.

These two parameters are used to develop degress of subcooling which are used in the Emergency Operating Procedures (EOPs) to trigger specific operator actions. The subcooling margin monitors are not affected by the dynamic effects of a LOCA inside containment. Therefore, RVLIS need not be declared inoperable according to I'lS 3.3.I7,

2. Ilot i et Level (RC-163A-LTI and RC-163A Tliu RC-163A LTl is a level transmitter used for the "A" channel of hot leg level indication (RC-163A LRI). RC-163A-TEl is a temperature element used to compensate the reference leg for the "A" channel hot leg level transmitter (RC 163A-LTl). There are two channels of Nt leg level (RC-163A LR1 and RC 163B LRI) qualified as Type B, Category 1 instrume..'s according to Regulatory Guide 1.97, Revision 3, requirements.

During the teleconference on January 30, 1998, FPC mistakenly presented this instrument as not being required to satisfy ITS 3.3.17. This mistake was due to the wording in ITS Bases 3.3.17, which does not clearly specify hot leg level as one of the instmments required to satisfy -

ITS 3.3.17, item 4 (Reactor Coolant Inventory). This is a shortcoming of the Crystal River Unit 3 (CR-3) ITS llases, and will be corrected so that each required instrument will be trecified by FPC instrument number within the ITS Bases section for ITS Table 3.3.17-1.

After further research of the licensing docket for PAM instrumentation, FPC determined that instrumentation required to satisfy ITS 3.3.17, item 4 (Reactor Coolant Inventory) consists of both reactor vessel level (RC-164A LR1 and RC-164B LRI) and hot leg level (RC-163A-LR1 and RC 16311 LRI). This is consistent with the statement in the ITS Bases that "the instrument channels required to be OPERAllLE by this LCO are those parameters identified during the CR 3 specific implementation of Regulatory Guide 1.97 as Type A variables and non Type A, Category I variables "

Upon discovery of this mistake, FPC initiated action to correct the discrepancy in the ITS Bases in accordance with the CR-3 corrective action program. As part of this review, FPC has determined tiet both reactor vessel level and hot leg level instrumentation are included in the applicable,rveillance procedures required by ITS Surveillance Requirement (SR) 3.3.17.2, and bo.. channels of each of these instruments are operable. Therefore, both channels of reactor vessel level (RC-164A LR1 and RC-164B-LRI) and both channels of hot leg level (RC-163A-LR1 and RC-163B-LRI) are required to be operable in Modes 1,2, and 3 as described in ITS 3.3.17 and ITS Table 3.3.171.

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U.S. Nuclear Regulatory Commission 1 3F029817 Page 3 of 5 Imnact E' valuation The same circumferential terminal end break at the reactor vessel nozzle of the core Dood line (west) discussed in item 1 above could result in jer

'ningement on the instrument tubing to RC-163A LTl potentially rendering the level trans.

nonfunctional. The same break could also result in jet impingement on a conduit (RCR 3L. to RC 163A TEl potentially rendering the temperature element nonfunctional. This, in turn, would impact the accuracy of RC-163A.

LR1 by as much as 10.9% of full scale (by changing the temperature compensation of the level signal).

RC-163A LR1 is not used to mitigate or limit the consequences of a LOCA. The primary indicators used to ascertain adequate core cooling following a LOCA are RCS temperature and RCS pr.

ire. These two parameters are used to develop degrees of subcooling which are used in the EOps to trigger specific operator actions. The subcooling margin monitors are not affected by the dynamic effects of a LOCA inside containment. Therefore, RC-163A LRI need not be declared inoperable according to ITS 3.3.17.

3. Pressurizer 1 evel (RC-!-LT3)

RC 1 LT3 is a level transmitter used for the "B" channel of pressurizer level indication (RC LR3). There are two channels of pressurizer level (RC-1 LRI and RC I LR3) qualified as Type D, Category 1 instruments according to Reguhtory Guide 1.97, Revision 3, requirements. Both of these channels are required to be operable in Modes 1, 2, end 3 as described in ITS 3.3.17 and ITS Table 3.3.17-1.

In addition, a third non safety related pressurizer level instrument is available.

Imnact Evaluation ITS 3.4.8 (Pressurizer) requires the use of pressurizer level indication to determine pressurizer operability during Modes I, 2, and 3.

ITS SR 3,4.8.1 requires verification of pressurizer water level every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Since the pressurizer level instrumentation is only used for determining pressurizer operability during normal plant operations for meeting the requirements of ITS 3.4.8, consequential failures of pressurizer level instruments during a LOCA are not of concern in meeting the requirements of this specific ITS.

A circumferential pressurizer surge line break could result in jet impingement on the instrument tubing to RC-1 LT3 potentially rendering the level transmitter nonfunctional. The failure of the measured variable reference leg will result in pressurizer level indication failing to the low, off-scale position.

110 wever, this break would also result in voiding of the pressurizer.

Therefore, monitoring of pressurizer level is not required during this specific event.

In addition, RC-1-LR3 is not used to mitigate or limit the consequences of a LOCA. The EOPs do not require the use of pressurizer level indication to perform any action necessary for mitigation of a LOCA. Therefore, RC-1-LR3 need not be declared inoperable according to ITS 3.3.17.

U.S. Nuclear Regulatory Conunission Attachment B 3F029817 Page 4 of 5 Dihtr improved Technical Specification Structures. Systems, and Components A total of five specific SSCs required to be operable by ITS could be effected by postulated pipe whip or jet impingement as described in the DR, including the three PAM instruments discussed previously, FPC evaluated each SSC using the specific ITS bases and the guidance provided in Generic Letter 9118 to determine if adequate justification for continued operation exists with these ITS SSCs potentially impacted by the dynamic effects of a LOCA inside containment. Generic Letter 91 18 Section 6.3 (Technical Guidance), describes acceptable methodologies to be used for treatment of consequential failures in operability determinations.

An important element in determining operability !s the ability _of the SSC to perform its specified safety functions during-normal operation, as well as t nder accident conditions.

Generic Letter 91 18, Section 6.3.2 (Consequential Failures and Operability Determinations), states in part. "Where consequential failures would cause a loss of function needed for limiting or mitigating the effects of the event, the af fected SSC is inoperable...." This evaluation is document &d in the DR.

As discussed in the previous section, none of the potentially affected PAM instruments are relied upon to limit or mitigate the consequences of the specific event which could potentially render them nonfunctional (specific LOCA scenarios). In addition, as further discussed below, the other two SSCs required to be operable by ITS also are not reliea upon to limit or mitigate the consequences of the specific LOCA scenarios which could render them nonfunctional. These other two SSCs are RCV Il (Power Operated Relief Valve (PORV) bk)ck valve), and RC 14A DPT2 and RC-14A-DPT3 (two of three RCS How signals for the "A" loop of the Reactor Protection System (RPS)).

1. PORV likick Valve (RCV-II)

RCV-11 is used to isolate the PORV should the PORY become inoperable during normal operation, as required by ITS 3.4.10 (Pressurizer PORV). The PORV is normally closed with RCV 1I normally open, himaet Evaluation A break in the pressurizer spray line could result in a pipe whip or jet impingement affecting the junction box that supplies power to RCV-11, potentially rendering RCV 11 nonfunctional.

RCV-1i fails as is on loss of power, and therefore would not impact operation of the PORV, llowever, this break would also result in reduction of RCS pressure. Therefore, operability of the PORY is not required during this specific event.

The 10RV is not required to mitigate or limit the consequences of a LOCA. In addition, the PORY is not credited for any other design ba:.is accidents. CR-3 accident analyses take credit for the RCS safety valves instead of the PORV for limiting maximum RCS pressures. Even though PORY use is not specifically credited in LOCA analyses, the EOPs direct the use of the PORV as a method to reduce RCS pressure whenever pressurizer spray is not available.

RCV-li is also not required to mitigate or limit the consequences of t LOCA. EOPs direct the use of RCV-1i for managing the effects of the small break LOCA which would be created if the PORV inadvertently stuck open after manual or automatic actuation due to high RCS pressure, llowever, the failure of RCV-11 to close is acceptable as the resulting small break

y U.S. Nuclear Regulatory Commission Attachment B 3F029817 Page 5 of 5 i

i LOCA is' bounded by the specific LOCA analyses which have been performed for the facility.

Therefore, RCW11 need not be declared inoperable according to ITS 3.4.10.

2. RCS Flow (RC-14A DPT2 and RC-14A DPT3)

RC 14A DIT2 and RC 14A-DPT3 provide RCS Dow signals for the "A" loop of the RPS, channels 2 and 3, respectively. All four chaniels of the nuclear overpower RCS How and measured axial power imbalance RPS instrumentation for ca:h RCS loop (RC-14A DPT1, RC-14A DIT2, RC-14A DPT3 and RC-14A DPT4 for RCS loop "A") are required to be operable in Modes I and 2 as described in ITS 3.3.1 (RPS Instrumentation) and ITS Table 3.3.1 1.

ITS 3.4.1 (RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits) requires the use of RCS How indication to determine if the minimum RCS flow rates assumed in the departure from nucleate boiling ratio (DNBR) analyses are maintained during Mode 1. ITS SR 3.4.1.3 requires verifying RCS total Dow rate every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Since the RCS flow instrumentation is used for determining acceptable flow rates only during normal plant operations for meeting the requirements of ITS 3.4.1, consequential failures of RCS How instruments during a LOCA are not of concern for meeting the requirements of this specific ITS.

Impact EvaluatinD i

A circumferential pressurizer surge line break could result in jet impingement on the instrument tubing to RC 14A DPT2 and RC-14A DFT3 potentially rendering the instrumentation nonfunctional.

The nuclear overpower RCS flow and measured axial power imbalance RPS trip function is not required to mitigate or limit the consequences of a small break LOCA. Instead, the low reactor pressure RPS trip backed by a diverse high reactor building pressute RPS trip is credited in the small break LOCA analyses. Therefore, RC-14A DPT2 and RC-14A DPT3 need not oe declared inoperable according to ITS 3.3.1.

Conclusions As described above, the effects of pipe whip and jet impingement frem postulated LOCAs on PAM instrumentation and other SSCs identined in the ITS have been evaluated, and justification for continued operation exists to support the safe operation of the facility as described in the FPC approved DR.

As previously committed to in letter dated January 23,1998 FPC will submit a license amendment request providing justification for use of Generic Letter 87-11 and NUREG-CR-2913 as the licensing bases for CR-3 and/or propose additional plant modifications that will be completed prior to restart from Refueling Outage 11 in a separate submittal.

This separate submittal will specifically address the affected PAM instrumentation and other SSCs identified in the ITS which could possibly be rendered nonfunctional as the result of pipe whip or jet impingement from postulated RCS breaks.

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U.S. Nuclear Regulatory Commission Attachment C 3F029817 Page 1 of 4 l

ATTACllMENIC CONFLICTS IIETWEEN DESIGN CRITERIA FOR PROTECTION OF SYSTESIS, STRUCTURES AND COAIPONENTS FROh! TIIE DYNAhllC EFFECTS OF A LOCA INSIDE ( JNTAINSIENT AND REGULATORY GUIDE 1.97 LICENSING HASIS REQUIRE 51ENTS hin During investigation of the possible dynamic effects of a loss of coo' ant accident (LOCA) on safety related structures, systems, and components (SSCs) inside containment, Florida Power Corporation (FPC) discovered that conflicts exist within the Final Safety Analysis Report (FSAR) and previous docketed correspondence. These connicts include FSAR descriptions that imply that failure of one train of a safety-related system induced by either pipe whip or jet impingement from a LOCA is acceptable, provided that the redundant train is unaffected by the same event. in addition, these FSAR descriptions also conflict with the current FPC commitments for compliance with Regulatory Guide 1.97, Revision 3, for Post Accident Monitoring (PAM) instrumentation, which do not allow failures of qualified instrumentation caused by the anticipated environmental conditions aner a LOCA event.

Description of Design Hasis and Licensine Hnsis Gnflicts Portions of the FSAR describing protection against the efTects of pipe whip and jet impingement imply that the initiating LOCA event can cause consequential damage to safety related equipment as long as the damage is limited to only one train of redundant safety related equipment. Specifically, FSAR Section 4.2.6.4 states the following:

"The efTects of pipe rupture inside the containment have been considered. Generally, physical separation is employed to assure that a single incident will not damage both portions of redundant safety related piping or equipment" llowever, this statement s in conflict with other established requirements that the dynamic effects i

of a LOCA will not resun in the failure of safety related equipment to perform their specified safety functions. In accordance with the principal design criteria contained in FSAR Section 1.4, these specified safety functions must be assured given any postulated single failures in addition to any consequential failures induced by the initiating event. Specifically, the CR-3 principal design criteria specified in FSAR Section 1.4.40 states " protection for engineered safeguards shall be provided against dynamic effects and missiles that might result from plant equipment failures," and FSAR Section 1.4.42 states " engineered safety features shall be designed so that the capability of each component and system to perform its required function is not impaired by the effects of a loss of coolant accident (LOCA)." These design criteria cannot be met if the initiating event (LOCA) causes consequential damage to an SSC which prevents a single train of a safety related system from performing its intended safety function.

FPC provided commitnents regarding installation and qualification of PAM instrumentation meeting the design basis requi ements of Regulatory Guide 1.97, Revision 3, in letters to the NRC dated August 21,

U.S. Nuclear Regulatory Conunission Attachment C 3F029817 Page 2 of 4 1984, Noveniner 15,' 198, arxl March 27,1986. Where determined by FPC to be technically justified, exceptions to tu requir ments of Regulatory Guide 1.97, Revision 3, were requested. The extent of compliance arxl ain :.emptions approved by the NRC were described in the SER issued by the NRC by letter dated Jure 16, 1987.

The following section of Regulatory Guide 1.97 (page 1.97 2) provides the basis for when an instmment is required to function:

"It is essential that the required instrumentation be capable of surviving the accident environment in which it is kicatal for the length of time its function is required. It could therefore either be designated to withstimd die accident environment or be protected by a kical protection environment."

There were no exemptions requested by FIC or approved by the NRC in their SER for Regulatory Guide '.97, Revision 3, to this requirement.

Upon approval of the design basis for the FAM instnnnentation in the SER, the FSAR was not revised to reflect the impact of die liccasing commitments to maintain the committed design features required by Regulatory Guide 1.97, Revision 3. Specifically, discussions concerning protection of safety-related SSCs from the dynamic effects of a LOCA inside containment in the FSAR were not reviewn! and revised as necessary to specifically note that PAM instrumentation requires this protection.

Physical protection of PAM instrumentation from the postulated dynamic effects from any other design basis accident was also not addressed in a revision to the FSAR, As a result of these identified conflicts, FPC will evaluate the connicting design basis requirements in the FSAR and previous docketed correspondence in accordance with the CR 3 corrective action program to determine the correct design criteria for protection of safety related SSCs against the dynamic effects of a LOCA inside containment, including Regulatory Guide 1.97 required instrumentation. FPC will then initiate additional engineering and/or licensing corrective actions, if required, and initiate revisions to the affected FSAR discussions as necessary.

Eunmlary of Previous and Current Com!pitments By letter dated December 22, 1997, the NRC requested that FPC clarify whether the dynamic effects of a pressurizer surge line LOCA on systems and components other than the SW System had been evaluated.

FI C responded to this additional request by letters dated January 23, 1998, and January 28,1998, in these responses FPC stated that a review of postulated LOCAs in the reactor building had been performed which fully evaluated postulated breaks in the high energy portions of the Core Flood (Cb System lines, Decay lleat (Dil) System drop line, Makeup and Purification (MU) System letdown line, Auxiliary Pressurizer Spray (APS) line, and the Pressurizer Spray line.

From this review, FPC identified SSCs that were potential targets of pipe whip or jet impk ement t

from a LOCA for these affected lines, and has since dispositioned each of these targets to ensure adequate justification for continued operation exists. Attachment 11 of this letter addresses specific SSCs from this review involving Post Accident Sampling System components, Post Accident Monitoring instrumentation, and other SSCs identified in the ITS.

U.S. Nuclear Regulatory Commission Attachment C 3F029817 Page 3 of 4 The disposition of t'he potential targets of a LOCA event inside the reactor building as described above has been documented in a Deficiency Peport (DR) made available to the NRC Resident inspector, in addition, the nonconformances identified in the DR were reported to the NRC in LER 50-302/98-00100 (Reference 3 to this letter).

At a result of this review, FPC is conunitting to submit a " cense amendment request providing justification for accepting the current design and/or propose additional plant modifications that will be completed prior to restart from Refueling Outage 11 for the affected PAM instrumentation and j

other SSCs identified in the ITS which could possibly be rendered nonfunctional as the result of pipe whip or jet impingement from postulated RCS breaks as described in Attachment B to this i

letter.

Conclusim15 As described above, the issue of determining and implementing acceptable design criteria for considering the dynamic effects of a LOCA on safety related SSCs inside containment is still being evaluated by FPC in accordance with the CR 3 corrective action program.

Previous commitments made by FPC, and the current commitments addressed in this letter, are sufficient to fully address the immediate corrective actions required to ensure adequate justification for continued operation exists while the long-term licensing basis and design basis corrective actions are evaluated and implemented. These long-term corrective actSns will prevent recurrence of the identified nonconformances, including providing assurance that modifications to the plant can be safely and correctly made according to applicable regulatory requirements and Colninilments.

Previous and current commitments made by FPC to address this issue, and their current status, include the following:

ID NU5 tiler AND DATE CO5thttTAIENT STATUS 3 F0697-13-1 FPC will provide additional information to the NRC Complete, see June 5,1997 regarding the findings for jet impingement effects letter 3F0997-01 between the surge line and the SW piping.

3F0997-01-1 FPC will revise FSAR Section 4.2.6.6, "LOCA Surerseded by September 29,1997 Restraints," to state that longitudinal ruptures were 300198-34-1 and eliminated by following the guidelines of Generic Letter 3F0198-34-2 87-11.

3F0997-01-2 FPC will revise FSAR 14.2.2.5.11 " Reactor Building Superseded by September 29,1997 Subcompartments Pressure Response," to state that the 3F0198 34-1 fluid jet created by a circumferential rupture is limited to a 20" included angle rather than 30" as currently described in the FSAR.

3F1197 29-1 FPC will revise CP-213 to provide additional guidance Complete November 7.1997 regarding the use of new design methodologies and the potential impacts to the established licensing basis.

U.S. Nuclear Regulatory Commission Attachment C 3F029817 Page 4 of 4 ID NUMHER '

' ND DATE COMMITMENT STATUS 3F1197 29-2 FPC will complete an analysis of the pressurizer surge Complete, see November 7,199'/

line jet impingement effects using the original 30" cone letter 3F1297 05 angle and resubmit conclusions to the NRC.

3F1297-05-1 FPC will evaluate several approaches to resolving the May 29,1998 December 13,1997 nonconformance identified between postulated pressu-izer surge line jet impingement effects on SW System piping. The approaches under consideration include hardware modifications, such as a jet impingement shield, re-routing of SW System piping, or installation of additional isolations valves; or a license amendment request to change the CR 3 licensing basis.

FPC will advise the NRC of the planned approach for resolution of this issue.

3F0198-34-1 FPC will submit a license amendment request providing May 29,1998 January 23.1998 justification for use of Generic Letter 87-11 and NUREG-CR 2913 as the licensing bases for CR 3 and/or propose additional plant modifications that will t

be completed prior to restart from Refuel i1 in a separcte submittal.

3F0198-34 2 FPC will, as part of the Large Bore Piping Program, Program will take January 23,1998 evaluate the pipe stresses on reactor coolant system 4-6 years (2-3 piping to ensure the application of GL 8711 for the fuel cycles) to elimination of intermediate pipe breaks remains implement.

acceptable.

3F0298-17 1 FPC will evaluate the conflicting design basis May 29,1998 February 18,1998 requirements in the FSAR and previous docketed correspondence in accordance with the CR 3 corrective action program to determine the correct design criteria for protection of safety-related SSCs against the dynamic effects of a LOCA inside containment, including Regulatory Guide 1.97 required instrumentation.

FPC will then initiate additional engineering and/or licensing corrective actions if required, and initiate revisions to the affected FSAR discussions as necessary.

I