3F0997-14, Application for Amend to License,Consisting of License Amend Request 218,addressing Analysis Rev for Makeup Sys Letdown Line Failure Accident as Discussed in FSAR

Application for Amend to License,Consisting of License Amend Request 218,addressing Analysis Rev for Makeup Sys Letdown Line Failure Accident as Discussed in FSAR

ML20216F841
Person / Time
Site:	Crystal River
Issue date:	09/09/1997
From:	Cowan J FLORIDA POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20216F846	List: 3F0997-14, Application for Amend to License,Consisting of License Amend Request 218,addressing Analysis Rev for Makeup Sys Letdown Line Failure Accident as Discussed in FSAR ML20216F848
References
3F0997-14, 3F997-14, NUDOCS 9709120122
Download: ML20216F841 (18)

Text

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Florida Power COMPOMAiKiN g het los 60 302 September 9,1997 3F0997-14 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

License Amendment itequest 218 Revision of the Makeup System Letdown Line Failure Accident Analysis

Dear Sir:

Florida Fower Corporation (FFC) hereby submits a request for an amendment to its Facility Operating License No. DFR-72 for Crystal River Unit 3 (CR 3). The attached license amendment request addresses an analysis revision for the Makeup System Letdown Line Failure Accident as discussed in the Final Safety Analysis Report (FSAR). It has been determined that the change to the analysis involves an unreviewed safety question and, therefore, requires approval by the Nuclear Regulatory Commission (NRC).

In the original analysis, the event was modeled as being terminated by an automatic isolation of the failed letdown line on low reactor coolant system pressure. The revised analysis has modeled the event as being terminated by manual operator action to isolate the line.

This change was initially performed in April 1996 under the requirements of 10 CFR 50.59,

" Changes, tests and experknents,* using the guidance for safety evaluations at that time, and was determined not to involve an unreviewed safety question. The CR-3 FSAR was revised to incorporate these changes in Revision 23 without requiring prior NRC review and approval.

Upon further review of the change to the Makeup System letdown Line Failure Accident analysis (completed in August 1997), FPC has concluded that this change involves an unreviewed safety question and requires NRC approval. This conclusion is based upon the replacement of the automatie isolation with the manual isolation of the failed line and the increase in radiological dose as discussed herein.

9709120122 970909 lllllll!l kl$.ll,lll-ll PDR ADOCK 05000302 i p PDR CRYSTAL RIVER ENErt0Y COMPLEX

• 15760 W. Power Line Street + Crystal River, Florida 34428 4 708 - (352)7954486 A Florida Progress company 120000 2

U. S. Nuclear Regulatory Commission 3F099714 Page 2 ot'3  ;

l Additionally, the accident analysis provided in FSAR Section 14.2.2.6 is being changed to clarify the bsis of the postulated break consistent with the requirements of Regulatory Guide I 1.70, Table 15-4, item 7 and is presented to demonstrate that the dose consequences from a postulated break in the letdown line outside containment remain below the 10 CFR 100 limits.

A description of the changes along with the 10 CFR 50.92(c) evaluation and conclusion that the proposed changes do not involve a significant hazard is provided in Attachment A. I Marked up changes to the FSAR that are proposed in this submittal are provided in Attachment B. Attachment C provides a listing of the acronyms and abbreviations used in the i submittal.

The manual operator action to isolate the line performs the same function as previously provided in Table 3A, item OA 2, of Technical Specnication Change Request Notice 210.

llowever, its specific use in the Makeup System Letdown 1.ine Failure Accident analysis was not discussed in that submittal.

Florida Power Corporation requests that this amendment be approved by November 10, 1997, to support issuance of Revision 24 of the FSAR in time for CR 3 restart in early December, if you have any questions regarding this submittal or the schedule, please contact Mr. David Kunsemiller, Manager, Nuclear 1.icensing at (352) 563-4566.

Sincerely,

/k 0M John Paul Cowan Vice President Nuclear Production JPC/dah Attachments cc: Regional Administrator. Region 11 NRR Project Manager Senior Resident inspector

' l'IT to NRC leuer dated June 14,1997,

• Technical Sircification Change Request Notice 210" l3f%9710l

U. S. Nuclear Regulatory Commission  ;

3P099714 i Page 3 of 3.  !

STNTE OF FIDRIDA COUNTY OF CITRUS l t

John l'aul Cowan states that he is the.Vice President Nuclear Production for Morida Power Corp > ration; that he is authorim! on him part of said company to sign and file with the Nuclear Regulatory Commission the infonnation attached hereto; and that til such statements made and matters f.et forth therein are true and correct to the best of his knowledge, infonnation, and belief, e

r i

R$?.bf .)

John Paul Cowan Vice President -

Nuclear Pnxtuction Sworn to and subscribed before me this '7M3 day of) 'e6L ; _

,1997, by John Paul Cowan.

sn hl '

SWtWTc'of Notary Public State ' "' ~~~

LYNNE 8. SMITH W % # CC 614000 DMES. Deesmher it.1980 (Print, type, or stamp Commissioned Name of Notary Public)

Personally Produced Known -OR- Identification

,, 4 .- e e

1 FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUh115ER 50-302/ LICENSE NUh111ER DPR-72 l

l ATTACllh1ENT A l

LICENSE Ah1ENDh1ENT REQUEST 218 AIAKEUP SYSTEh! LETDOWN LINE FAILURE ACCIDENT i

i l

l l

l

L U. S. Nuclear Regul: tory Conunission AttLchment A 3F099714 Page 1 of 14 i A'ITACilMENT A 1,1 CENSE AMENDMENT REQUEST 21N MAKEUP SYSTEM l.ETDOWN 1,INE FAILURE ACCIDENT LICENSEE IX)CUMENT INVOINEDt Crystal River Unit 3 (CR-3) Final Safety Analysis Report (FSAR)

PORTIONS: FSAR Sections 5.4.4.2,14.2.2.6, and 14.3t FSAR Tables 14 24, 14-41, 14-42, and 14-43

SUMMARY

OF CilANGESt Changes to the FSAR are being proposed to reflect the revised analysis for the Makeup System 1.ctdown 1.ine Failure Accident. The revised analysis incorporates a manual operator action in place of the automatic action to iso' ate the failed letdown line. Other changes are being proposed to clarify that this accident is a hypothetical event that is presented only to demonstrate that the dose consequences are below 10 CFR 100 limits.

Prior to Revision 23 of the FSAR, the Makeup System letdown Line Failure Accident analysis assumed that the loss of reactor coolant outside containment would be terminated by -

an automatic Engineered Safeguards Actuation System (ESAS) isolation signal when the low s reactor coolant system (RCS) pressure setpoint was reached. The automatic isolation signal would close the containment isolation valves for the letdown line and terminate the leakage.

During Refueling Outage 10, it was identitled that symptom oriented actions in the Emergency Operating Procedures (EOP) could prevent the automatic isolation of the letdown line assumed in the event-oriented analysis. EOP-3, ' Inadequate Subcooling Margin," directs the oisrator to ensure full high pressure injection (IIPI) system now upon a loss of subcooling margin (LSCM). This step includes manual initiation of IIPI if not already operating, in the case of a certain sired line failure, the operation of IIPI would repressurize the RCS, thereby preventing a decrease to the low RCS pressure ESAS setpoint (the revised analysis shows that the low RCS pressure ESAS setpoint would not be reached even assuming no llPI actuation). Thus, automatic isolation of the letdown line would not occur. In order to resolve this discrepancy, EOP-3 was revised to require earlier manual isolation of the letdown line to resolve this scenario. The analysis for the Makeup System Letdown Line Failure Accident was revised to account for the change in actions and the FSAR was revised to reflect the revised analysis in Revision 23 (Reference 1).

Subsequent to the submittal of Revision 23 of the FSAR, it has been determined by review pursuant to 10 CFR 50.59 that this change involves an Ur reviewed Safety Question (USQ) and requires NRC approval. This conclusion is based upon the replacement of the automatic isolation with the manual isolation of the failed line and the increase in radiological dose caleubted in the revised analysis. While the radiological dose in the revised analysis remains

.very small-(approximately 1% of the 10 CFR 100 limits), the increase was considered

U. S, Nuclear Regul: tory Commission Attachment A 3F099714 Page 2 of 14 significant because the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose at the Exclusion Area Boundary (EAB) tripled from the previous analysis, in addition to the changes provided in Revision 23 of the FSAR, additional changes are being proposed to correct errors, provide added information, and revise the existing FSAR wording to better integrate the revised analysis.

Changes are proposed for FSAR Sections 5.4.4.2 and 14.2.2.6.1 to clarify the basis for the Makeup System Letdown Line Failure Accident. As evaluated in the Pipe Rupture Analysis Criteria Outside the Reactor Building, the high energy portion of the letdown line outside containment is not sub. ject to a high energy line break. The Makeup System Letdown Line Failure Accident is a postulated break consistent with the requirements of Regulatory Guide 1,70, Table 15-4. Item 7 and is presented to demonstrate that the dose consequences from a postulated break in the letdown line outside containment remain below the 10 CFR 100 limits.

CilANGE TO Tile FSAR 1pwrlylon of chamm The FSAR changes are provided in Attachment H. Changes to the FSAR addressed in this submittal include the following:

Change _

FSAR Section _

Description Revision 23 changes 14.2.2.6 During Refueling Outage 10, the accident 14.3 analysis for this accident was revised to Tables 14-41, 14-42, correct for differences between the 14-43 accident analysis and the plant EOPs. The FSAR was revised in Revision 23 to add the revised analysis.

~ Correct Table 14 43 Table 14 43 An error made in transferring values from value the analysis to FSAR Table 14-43.

Changes are being made to reflect the correct values.

Correct description of 14.2.2.6 The changes made in Revision 23 of the assumed flow FSAR indicated that the analysis was performed assuming full IIPI llow; however, the analysis assumed full makeup flow, Changes are being made in the FSAR to reflect the correct wording from the revised analysis.

Typos and additional 14.2.2.6, 14.3 Changes are being made to correct other information Tables 14-24, 14-41, typographical errors and to add additional 14-4 2 information from the revised analysis.

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U. S. Nuclear Regulatory Conunission Attachment A 3F099714 Page 1 of 14 Change FSAR Section _

Description Revision to integrate 14.2.2.6 Revision 23 of the FSAR did not revise the original and revised existing information for the original analysis in FSAR Tables 14-41, 14-42, analysis, but added additional information 14-43 from the revised analysis. Changes are being made to integrate the revised analysis to replace the original analysis descriptian and results.

Clarify requirements 5.4.4.2 Changes are being made to update the for dynamic and FSAR with the results of the Pipe Rupture environmental effects report regarding the letdown line outside for this accident containment. Ilased on the Standard Review Plan and Generic letter 8711, this line is not subject to a high energy line break.

Clarily basis for 14.2.2.6.1 The description of the Makeup System Makeup System letdown Line Failure Accident is being lxtdown Line Failure clarified as discussed in FSAR section Accident 5.4.4.2. This accident is presented only to demonstrate that the dose consequences from a postulated break remain below the 10 CFR 100 limits.

Reason For Request This change incorporates the results of the Makeup System letdown Line Failure Accident as revised in Framatome Technologies, incorporated (FTI) Summaries 861257374-00 and 861257374-02, "CR-3 RiiLAPS Ixtdown Line lireak," (References 2 and 20). The revised analysis incorporates a manual operator action in place of the automatic action to isolate the failed letdown line. It has been determined that the change to the analysis involves an unreviewed safety question and, therefore, requires NRC approval.

Secondly, the Makeup System letdown Line Failure Accident, as described in the FSAR, is being clari0ed to note that the letdown line outside containment is not subject to a high energy line break based on criteria defaults in Generic Letter 8711 (Reference 17). The accident analysis provided in FSAR Section 14.2.2.6 is a postulated break consistent with the requirements of Regulatory Guide 1.70, Table 15-4, item 7 and is presented to demonstrate

< that the dose consequences from a postulated break in the letdown line outside contd. ment remain below the 10 CFR 100 limits.

U. S. Nuclear Regulatory Commission Attachment A l 3F099714 Fage 4 of 14 l Justifkation For u.-.a Revision to Accident Amlysis Prior to Revision 23 of the PSAR, the Makeup System Letdown Line Failure Accident analysis assumed that the leakage via the pipe failure would be terminated by an automatic ESAS isolation signal when the low RCS pressure setpoint (1350 psig) was reached. Re.

evaluation of this accident has determined that RCS pressure would not decrease to the setpoint j for the automatic isolation. Therefore, the analysis has been revised to include a manual I operator action to isolate the failed line.

Original Analysis The Makeup System Letdown Line Failure Accident was initially analyzed in HAW 1521.

" Crystal River Unit 3 - Cycle 2 Reload Report," which was provided to the NRC in Reference 3. HAW 1521 was reviewed by the NRC in the Safety Evaluation Report (SER) approving Amendment No.19 to the Technical Specifications (Reference 4). The Makeup System Letdown Line Failure Accident analysis was incorporated into the FSAR in 1982 in Revision 0 of the updated FSAR.

This analysis assumed a failure of the letdown line just outside the containment but upstream of the letdown control valves. The RCS would depressurire since flow from the broken line would exceed the makeup capability. Ultimately, the depressurization that resulted from the break would lead to an automatic reactor trip on low RCS pressure.

The loss of reactor coolant would continue until the RCS pressure dropped below the pressure for actuation of the ESAS. At this time, the letdown line isolatiot, valves would .

be automatically isolated. Closure of the isolation valves would stop the release of reactor coolant and fission products to the Auxiliary Hullding, thus terminating the loss of coolant phase of the accident. This sequence of events is provided in Table 1.

The calculation of the response of the RCS to the break in the letdown line and the total

. reactor coolant mass that passes through the break and into the Auxiliary Building was performed using the CRAFr2 computer code. The total mass of reactor coolant calculated to escape through the break and be released to the Auxiliary Hullding was 45,760 pounds (Ibh The dose rates calculated from this release are listed in Table 2.

EOP-3 Actions During Refueling Outage 10, a discrepancy was discovered between the assumed sequence 1

of events in the FSAR analysis and the actual response to a letdown line failure. Ilad an actual letdown _ line failure occurred resulting in a IMCM, EOP-3, " Inadequate Subcooling Margin," would have directed the plant operator to ensure full lip! system flow, This step includes manual initiation of Ilpi if not already operating, in the case of certain sized line failures, the operation of Ilpl would repressurize the RCS, thereby preventing a decrease to the low RCS pressure ESAS setpoint, Thus, automatic isolation of the

U. S. Nuclear Regulatory Conunission Attachment A 3F099714 ' Page 5 of 14 letdown line would not have occurred as assumed, althcugh later steps in !!GP 3 would have directed the operator to isolate the letdown line as a possible leakage pathway, llowever, this action may not have occurred as soon as the automatic isolation assumed in the FSAR accident analysis. The !!OP 3 actions are listed in Table 1. (The revised analysis shows that the low RCS pressure liSAS setpoint would not be reached even assuming no llPl actuation.)

The sequence of steps in I!OP 3 had been implemented based upon the guidance in the Abnormal Transient Operating Guidelines (ATOG) and the llabcock & Wilcox (Il&W)

Owners Group (Il&WOG) limergency Operating Procedures Technical llases Document.

The !! ops are based upon a symptom-oriented response rather than focused on mitigating specific accidents. For a 13CM, the actions in liOP 3 utilize the philosophy that as long as the RCS remains subcooled, adequate core cooling can be assured. The lack of l subcooling margin is assigned top priority requiring treatment shead of other abnormal l heat transfer symptoms, liased on this philosophy, the ATOG describes the llPI system as being initiated early in the event (maintain adequate core cooling) prior to closing the letdown line isolation valve (isolate possible RCS leakage paths).

The philosophy and sequencing of steps in the ATOGs were reviewed and approved in the NRC SI!R provided by Generic letter 83 31 (Reference 5). In the SIIR, the NRC recognized the advantage of the symptorn-oriented approach in that it implicitly covered varlota multiple failure events without the need to specify all failure combinations. A drawback was noted that recovery may be delayed and unnecessary operator actions may be taken. Ilowever, since a balancing of ATOG against alternative approaches showed the selection to have a sound basis, the overall ATOG approach was determined to be acceptable for il&W plants.

The ATOG review by the NRC included the actions taken in response to a IJCM. As reviewed in the S!!R, the third action for this event was to initiate llP' and was followed by the fourth and last mitigating action to isolate potential RCS leaks (including closure of the letdown isolation valve).

The il&WOG submitted a copy of the !!&WOG limergency Operating Procedures Technical 11ases Document (TilD), which consolidated the ATOG technical bases in l addressing the NRC SI!R, to the NRC in Reference 6. The CR-3 plant specific ATOG l was submitted to the NRC in Reference 7 as part of its Procedure Generation Package (PGP) submittal, Subsequently, the NRC issued a SliR on the PGPs in Reference 8. The FPC ATOG/TilD is based on the ll&WOG ATOG/TilD.

Ily implementing the ATOGs for CR 3, the isolation time for the specific let< lown line failure event may have been extended beyond that assumed in the FSAR analysis. When the difference between the l'SAR analysis and the sequence of actions in liOP 3 were

( recognited, !!OP 3 was revised to rehicate the isolation of the letdown line to an earlier l step. This revision was determined to be prudent in order to minimite any radioactive release from this potential accident even though the letdown line isolation thne would

U. S. Nuclear Regulatory Commission Attachment A 3F099714 Page 6 of 14 have to be significantly extended before the released dose would approach the limits of 10 CFR 100. Additionally, the placement of the manual action to isolate the letdown line corresponds to the NRC position in Reference 9 for a ' simple' operator action.

Revised Analysis Accordingly, the analysis for the hiakeup System Letdown Line Failure Accident was revised to bound the actions specified in EOP 3. This analysis is contained in FTI Summaries 8612573744)0 and 861257374-02 (References 2 and 20). Based on this analysis, the RCS will depressurize following the letdown line failure since flow from the broken line will exceed the makeup capability. Ultimately, the depressurization that results from the break will lead to an automatic trip of the reactor on low RCS pressure.

Upon indication of a LScht due to the break, EOP 3 will be entered. in order to maximize the leak ge calculated in this accident, full makeup flow was assumed. The isolation of the letdown line as directed by E0P-3 was assumed 10 minutes following LSChi. The sequence of events for this analysis are outlined in Table 1.

This analysis bounds manual operator actions that isolate the letdown line within 10 minutes following LSChi as long as llPI has not significantly affected the RCS pressure within that time frame (i.e., the analysis is valid if letdown is isolated prior to llPI actuation, or letdown isolation occurs soon enough after llPI actuation that ilPI has not yet significantly affected RCS pressure). Currently, EOP-3 isolates the letdown line by remote manual operation of the isolation valves prior to a manual llPI hdtiation; however, a proposed revision discussed in Technical Specification Change Request Notice 210 (Reference 10) would manually initiate llPI on a LScht (if not already actuated automatically), followed immediately by manual reactor building isolation and cool'.ng (RillC) actuation (this function automatically isolates the letdown line).

The revised analysis was performed by Fri using RELAPS. The RELAPS computer model is a generic representation of the ll&W designed lowered-loop plants. The use of RELAPS for loss of coolant accidents (LOCA) has been reviewed and accepted by the NRC (Reference 11). The break area corresponds to the double-ended failure of the 2%

inch letdown line just outside the reactor building. The total mass of reactor coolant calculated to escape through the break and be released to the Auxiliary Building is 114,000 paunds mass (Ibm).

The revised analysis updated the off site dose release based upon the latest fuel and fuel cycle designs. After correcting the source terms, the integrated mass release during the event was normalized to the dose per 100,000 lbm of letdown lost to the Auxiliary iluilding. As shown in Table 2, the radiological doses for this event have slightly increased, llowever, the doses continue to be much less than the limits of 10 CFR 100.

l The assumptions used in the analysis are included in FSAR Table 14-41. FSAR Table 14-42 identifies the activity released to the environment and FSAR Table 14-43 contains a i

U. S. Nuclear Regulatory Commission Att:chment A 3RW9714 Page 7 of 14 summary of resultant dues. The dose rates calculated from this release are also listed in Table 2.

Additionally, Revision 23 of the MAR incorrectly reflected the revisev analysis of this accident. FSAR Revision 23 ind cated that the analysis assumed full llPI flow was initiated upon the LSCM followed b;i isolation of the letdown line 10 minutes later. The actual assumption in the revised analysis is that full makeup flow occurs prior to the isolation of the letdown line. The propos:d FSAR revision to section 14.2.2.6 removes the incorrect statement.

CadGcation of BasisJcLAtkilyJiis Section 15 of Regulatory Guide 1.70, Revision 3 (Reference 12) includes breaks in lines connected to the RCS that carry reactor coolant outside containment as a typical initiating

. event which should be considered in Chapter 14 of the FSAR. Table 15 4 of Regulatory Guide 1.70 indicates that this event should be evaluated for dose consequences. A rupture in the letdown line was not included in the original FSAR, but was first evaluated in the Cycle 2 Reload Report (Reference 3) to address that specific requirement in Regulatory Guide 1.70.

The methods and general criteria used to postulate and protect pipe rupture effects cutside the reactor building at CR 3 are described in the report " Pipe Rupture Analysis Criteria Outside the Reactor llullding Crystal River Unit 3." This report was submitted to the NRC in Reference 13 (March 31,1989) and Revision I was submitted in Reference 14 (December 18, 1989). NRC reviewed this report and found it acceptable as documented in Reference 15 (April 11,1990).

The NRC approved pipe rupture report concluded that the high energy portion of the letdown line outside containment is not subject to a high energy !!ne break. The NRC Standard Review Plan Section 3.6.2 (Reference 16) allows the establishment of a "No Break Zone" if certain criuria are met. Alsa, per Generic letter 8711. " Relaxation in Arbitrary Intermediate Pipe Rupture Requirements," (Reference 17) ebitrary intermediate piping breaks need not be postulated if certain criteria are met.

Revision 1 of the Pipe Rupture Analysis established the "No Break Zone" (NBZ) for the letdown line piping between the wntainment isolation valves for this specific containment penetration. The NBZ was based upon the methods specified in the Standard Review Plan.

Additionally, the report determined that the stress in the piping from the con ainment to the manual isolation valves downstream of the block orifice and letdown controi valves is low enough that an arbitrary intermediate break need not be considered in this section of the letdown line. Therefore, a break in the high energy portion of the letdown line outside containment is not considered a credible event.

The NRC acceptance of Revision 1 of the Pipe Rupture Analysis in Reference 15 stated, "We have reviewed the revised sections of the submitted report and have determined that the revised sections incorporate the statements, conditions, and criteria required by our letters of

U. S. Nuclear Regulatory Commission Attachment A 3F0997-14 Page 8 of 14 September 28 and November 9,1989, and are acceptable." Based on the requircraents of the Standard Review Plan and Generic letter 87-11, designing for the dynamic or environmental effects of a high energy line break in the letdown line outside containment is not required.

p Therefore, the Makeup System Letdown Line Failure Accident is prer ated caly to

% demonstrate that the dose consequences from a postulated break in the letdot. . line outside containment remain below the 10 CFR 100 limits. Relative to dose consequences, the hypothetical break in the letdown line bounds other postulated breaks 'm lines connected to the RCS that carry reactor coolant outside containment. The proposed changes to FSAR Sections 5.4,4.2 and 14.2.2.6.1 are provided to clarify the requirements for this analysis.

LER 97-015-00

'. Licensee /Ivent Report (LER) 97 015-00 (Reference 19) reported that the inboard containment isolation valves for the letdown line were not designed to close against the maximum differential pressure anticipated for ht e valves to perform their safety function. Corrective actions associated with this LER will ensure that the inboard and outboard isolation v.ilves for the letdown line containment penetration fully meets their design requirements for containment isolation.

q NO SIGNIFICANT IfAZARDS CONSIDERATION:

An evaluation of the proposed license amendment has been performed in accordsnee with 10 CFR 50.91(a)(1) regarding significant hazards considerations, using the standards in 10 CFR 50.92(c).

Criterion 1 Does not involve a significant increase in the probability or consequences of an accident previously evaluuted.

I This change involv > a p sion to the analysis for the Makeup System Letdown Line Failure Accident. YN & asG analysis assesses the resultant change in consequences of 21s event based on the actions specified in EOP-3 to manually isolate the letdown line fciture. No changes have beca made to any precursors to this event. Therefore, the probability of an accident previously evaluated has not been increased.

This change has resulted in an increase in the calculated doses due to the greater iclease of reactor coolant prior to rannination of the leak. Although the doses have increased, they remain significantly les" than the limits of 10 CFR 100. These doses also remain lower than the resultant doses ior the design basis LOCA.

The revised analysis evaluates the consequence:. of this accident based on the replacement of the automatic isolation of the letdown line with a manual operator action to isolate the letdown line. This action was added to EOP-3 when it was identified that the manual

initiation of the IIPI system directed by the EOP would interfere with the automatic

U. S. Nuclear Regulatory Commission Attachment A 3F0997-14 Page 9 of 14 isolation signal assumed to terminate this event. Manual initiation of the IIPI system for a LSCM cvent is consistent with the symptomatic philosophy of the E0Ps. This philosophy is utilized in order to manage a wide range of event / leaks (Tat wouU be indicated by a LSCM. Early initiation of the llPI system is intended to ensee ndequate core cooling as the primary concern during a LSCM event.

Prior to the addition of the EOP step to manually isolate tiu letdown line, the EOP directed actions towards locating and isolating the source of the leak resulting in the LSCM. Ilowever, due to the potential signincance of the letdown line failure which can result in RCS leakage outside the reactor building, the manual action was added early in EOP-3 to isolate the letdown line. This action is proactive in ensuring early isolation of the potential leakage path and is consistent with the concept of a " simple" operator action (Reference 9).

Crediting a manual operator action instead of the automatic isolation introduces the possibility of a malfunction of a different type (i.e., operator error). The revised analysis assumes that operator action to isolate the letdown line occurs 10 minutes following a LSCM. Although the probability of operator error during this action may be greater than the probability of the failure of the automatic function, the consequences of this error would be small. Several iridications would be available to the operator to identify the continued loss of coolant through this line. As discussed above, the radiological dose calculated by this event remains a small fraction of the limits of 10 CFR 100. Therefore, adequate time would exist for the identification of sn operator error and correction of this mor before any significant increase in the consequences of this event would occur.

Alitionally, the probability for operator error in this event is considered to be small due o the extensive training plant operators receive regarding the EOPs and the simple nature of the action. Validation of the required actions in the EOPs, including isolation of the letdown line, is performed on the plant simulator to ensure the validity of the EOPs as well as to ensure that these actions can be performed as required.

The clarification added to FSAR Section 5.4.4.2 and 14.2.2.6.1 reDects the previously approved evaluation for pipe rupture criteria outside the reactor building for CR-3. A break in the high energy portion of the letdown line outside containment is not considered a credible event. This accident is presented only to demonstrate that the dose consequences from a postulated break in the letdown line outside containment remain below the 10 CFR 100 limits.

Based on the above, this change does not involve a significant increase in the c.onsequences of an accident previously evaluated.

i U. S. Nuclear Regulatory Commission Attachment A 3F099714 Page 10 of 14 Criterion 2 Does not create the possibility of a new or diferent kint qf accident pom any accident previously evaluated.

This change does not involve any modilication to the plant nor a change in the operation of the plant prior to the postulated failure of the letdown line. This change only evaluates the radiological dose consequences of the actions taken following the line failure. The addition of the action to manually isolate the letdown line for a LSCM event is consistent with the need to isolate potential RCS leakage paths and replaces the automatic isolation that was previously assumed to occur. Therefore, this change does not create the possibility of a new or different kind of accident.

Criterion 3 Does not involve a significant reduction in the margin ofsafety.

This change does not result in a reduction to the margin of safety as defined in the Ilases for any Technical Specifications. As discussed above, the radiological doses for the revised analysis have increased but remain a small fraction of the 10 CFR 100 limits.

ENVIRONMENTAL IMPACT EVALUATION:

10 CFR 51.22(c)(9) provides criteria for and identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not (1) involve a significant hazards consideration, (2) result in a significant change in the types or significant increase in the amounts of any efiluents that may be released off site, or (3) result in a significant increase in individual or cumulative occupational radiation exposure. FPC has reviewed this license amendment and believes it meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(c), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the proposed license amendment.

==

Conclusions:==

1. The proposed license amendment does not involve a significant hazard as described previously in the evaluation.

2. As discussed in the No Significant llazards Consideration, tnis change does not result in a significant change or significant increase in the radiological doses for the Makeup System Letdown Line Failure Accident. The proposed license amendment does not result in a significant change in the types or a significant increase in the amounts of any effluents that may be released off site and does not involve irreversible environmental consequences beyond those already associated with the Final Environmental Statement.

U S. Nuclear Regulatory Commission A".achment A 3F0997_ Page 11 of 14

- 3. The proposed license . amendment does- not result in a significant increase to the individual or cumulative occupational radiation exposure. This change does not affect the occupational radiation exposure but revises the radiological doses analyzed for an

- . accident as described in the FSAR safety analyses.

REFERENCES t

1. Fit to NRC letter, " Final Safety- Analysis Report Revision 23,* dated November 18,1996 [3F1196-01]

2. Framatome Technologies Incorporated Summary 86-1257374-00, "CR-3 RELAPS i Ixtdown Line Break," June 28,1996

3. FPC to NRC letter, " Environmental Impact Appraisal and Balance of Plant Review-for Cycle 2 Reload and Power Level Upgrade," dated February 28, 1979

, . [3F0279-10]

~ 4. NRC to FPC letter, " Amendment No.19," dated July 3,1979 [3N0779-01]

5. Generic Letter 83 31, " Safety Evaluation of ' Abnormal Transient Operating Guidelines'," dated September 19,1983

6. B&W Owners Group to NRC letter, "B&W Owners Group Emergency Operating '

Procedures Technical Bases Document, BW Document No. 74-1152414-00," dated September 11,1985

7. FPC to NRC letter, "NUREG-0737, item I.C.1, Guidance for the Evaluation and Development of Procedures for Transients and Accidents," dated March 25, 1983 13F0383-31]

8. NRC to FPC letter, " Safety Evaluation for the Crystal River Unit 3 Procedures Generation Package (TAC No. 44292)," dated April 6,1990 [3N0490-05]

9. NRC to FPC letter, Long-term modifications regarding ECCS Small Dreak Analysis problem, dated September 26,1978 [3N0978-071

10. FPC to NRC letter, " Technical Specification Change Request Notice 210," dated June 14,1997 [3F0697-10]

11, NRC to Framatome Technologies Incorporated, " Acceptance for Referencing of Topical Report BAW-10192-P, 'BWNT Loss of Coolant Accident Evaluation Malel for Once-Through Steam Generator Plants' (TAC No. M89400)," date February 18, 1997

12. Regulatory Guide 1.70, " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, LWR Edition," Ra'sion 3, November 1978 w -

-U. S. Nuclear Regulatory Commission AttT.hment A-3F0997-14 Page 12 of 14

13. - FIC to NRC letter, "lligh Energy Line Break Outside Reactor Building Criteria,"

dated March 31,1989 [3F0389-19] -

-14. FPC to NRC letter, "lligh Energy Line Break (IIELB), Revision of Design Criteria.

and Schedule Update," December 18,' 1989 [3F1289 ll]

15. NRC to FPC letter, Crystal River Unit 3 - liigh Energy Line Break (IIELB) Criteria for Analysis of Piping Outside Containment (TAC No. 69533) " dated April 11, 1990 [3NN90-10]

16. NUREG-0800, " Standard Revi:w Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition," dated July 1981 :

17. Generic Letter 87-11, " Relaxation in Arbitrary Intermediate Pipe Rupture Requirements," dated June 19,1987

18. Safety Evaluation of the Crystal River Unit 3, Docket No. 50-302, dated July 5,1974

19. lFPC to NRC letter, " Licensee Event Report (LER) 97-015-00," dated July 11, 1997

[3F0797-24]

20. Framatome Technologies Incorporated Summary 86-1257374-02, "CR-3 RELAPS Letdown Line Break," September 4,1997 Y

i i

U. S. Nuclear Regulatory Commission Attachment A 3F0997-14 Page 13 of 14 Table 1 Sequence of Events for Makeup System Letdown Line Failure Accident Sequence PREVIOUS CURRENT appox, time ANALYSIS EOP Sequence ANALYSIS (seconds) FSAR Revision 22 (Pre-revision) FSAR Revision 23 t=0 Letdown line failare - Letdown line failure - Letdown line failure -

event initiation event initiation event initiation t = 261 Reactor trip on low Reactor trip on low pressure pressure t = 330 Reactor trip on low pressure t = 570 RCS saturates - RCS saturates -

LSCM LSCM t = 745 ESAS signal - reactor building isolation on low RCS pressure t = 752 Letdown isolation valve fully closed -

event terminated t = l l70 Operator action to close letdown valve (10 minutes f ' lowing LSCM) - event terminated symptom- Operator action to based close letdown valve to isolate possible RCS leakage source - event terminated

U. S. Nuclear Regulatory Commission Attachment A 3F099714 Page 14 of 14 4 Table 2 Revision of Radiological Dose Rates Associated with the Makeup System Letdown Line Failure Accident Total Integrated Filtration Thyroid (REM) Whole Body (REM)

Doses Emclency (10 CFR 100 Limit = 300) (10 CFR 100 Limit = 25)

FSAR Rev 22 FSAR Rev 23 FSAR Rev 22 FSAR Ret 23 2-hour Dose at the 90 % 0.115 0.304 0.066 0.096 Exclusion Area -

Boundary 0% 1.15 3.04 0 067 0.099 90 % 0.0101 0.027 0.0058 0.008 30-day Dose at the Low Population Zone 0% 0.101 0.27 0.0059 0.009

_ _ _ _ _ _ - - _ _ _ _ . _ - _ _ - _ .