ML20212F267
| ML20212F267 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/31/1997 |
| From: | Cowan J FLORIDA POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20212F271 | List: |
| References | |
| 3F1087-18, NUDOCS 9711040236 | |
| Download: ML20212F267 (19) | |
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Docket 0 2 October 31,1997 3F1097-18 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
License Amendment Request 214 Revision 0
" Request for Decay 1-leat Removal Requirements in Mode 4"
References:
- 1. FPC letter dated June 14,1997 (3F0697-10) " Technical Specification Change Request Notice 210."
- 2. FPC letter dated July 18,1997 (3F0797-10) ' Technical Specification Change Request Notice 213."
- 3. FPC letter dated August 10,1990 (3F0890-05) " Technical Specification Change Request No.174, AdditionalInformation."
- 4. FPC letter dated July 29 1997 (3F0797-21) " Technical Specification l
Change Request No.
- 209, Rev 1.
Post Accident Monitoring Instmmentation."
Dear Sir:
~ Florida Power Corporation (FPC) hereby submits License Amendment Request (LAR) 214, Revision 0 regarding proposed amendments to Operating License No. DPR-72 for Crystal River Unit 3 (CR-3).
The LAR requires operable equipment necessary to mitigate the consequences of Loss-of-Coolant Accidents (LOCA) postulated to occur in Mode 4.
Raskuround CR-3 Improved Technical Specification (ITS) 3.7.5, EFW Systems, presently requires 3l-operable Emergency Feedwater (EFW) trains in Modes 1, 2, and 3, but does not require Emergency _ Feedwater in Mode 4.'
The Bases for ITS 3.5.3, ECCS-Shutdown, state that sufficient time exists for_ manual operator actuation of the required Emergency Core Cooling CRYSTAL RIVER ENERGY COMPT.EX 15760 W. Power Line Street + Crystal River, Florida 34428-6708 - (352)795-6486 A Florida Progress Comparty 97i10505t36971031 h hfhflfflllfg1g-l PDR ADOCK 05000302 a
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tSystem (ECCS) to mitigate the consequences 'of a Design Basis Accident (DBA) in Mode 4.
- ITS13.5.3 requires one train of ECCS operable _ in Mode = 4. but cllows the Low Pressure 1 Injection (LPI) train to be aligned in the Decay lleat' Removal (DIIR) mode of operation, in Reference 1 FPC submitted Technical Specification Change Request Notice (TSCRN) 210--
which documented the results of the small break LOCA analysis to the' NRC. Although this -
knalysis involves accident scenarios occurring in Mode 1, the iesults led FPC to e',41uate -
3 whether decay heat removal requirements specified by the current CR-3 ITS were sufficient' for Mode 4.
Comequently, FPC committed in Reference 1~ to Lcomplete an engineering
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- evaluation of decay heat requirements in Mode 4 prior to restart, but noted that the evaluation would not affect TSCRN 210.
In Reference 2, FPC submitted TSCRN 213 to establish the requirements for a Low Temperature Overpressure Protection (LTOP) System as required by 10 CFR 50.60 to the zNRC. That submittal also referred to the engineering evaluation of decay heat requirements in
' Mode 4 ~and noted that any-resulting revisions to the ITS would preserve the - LTOP requirements proposed by TSCRN 213.
Results of Evaluatjen
-FPC reviewed the present and paFt llCensing basis of CR-3 and determined that LOCAs are postulated to occur in Mode 4 and that operator actions are relied upon for LOCA mitigation.
?
Information available at the t!me CR-3 was converting to the ITS indicated that approximately 30 minutes were available to manually initiate ECCS following a Mode 4 LOCA (see Reference 3). The supporting analysis for Reference 3 did not consider reactor coolant losses due to fiashing and carryout, and heat from metal surfaces of reactor vessel internals.
'Recently, FPC, -in conjunction viith Framatome Technologies, Inc. (FTI), performed a re-evaluation of DilR requirements for Mode 4. Since CR-3 LOCA analyses are performed
-e with initial conditions at full power, the recent evaluation is based on extrapolations of existing analyses which assume a representative time for plant shutdown from full power to Mode 4 and plant configurations unique to Mode 4 at CR-3L Additionally, m' considered the reactor coolant losses due to flashing and carryout, and heat from reactor vessel internals in the current evaluation. This is an important distinction from the previous Mode 4 evaluation since
- these reactor coolant losses decrease the available time for operator actions.
Based on the recent evaluation, one train of the ECCS:and'one source of feedwater to the i
4 steam generators must be operable to_ mitigate a small break Mode 4 LOCA (less than 0.5 ft').
4 1 Further, the time that would be available for operator actions to preclude core uncovery for a
- small ' break ~ LOCA -is~ in excess <of - 10 minutes.
LAR 214 adds one train of EFW and -
associatal equipment,Lsuch as the EFW Tank, in Mode'4 to the existing requirement for one-p ECCS traini 1 The recent evaluation also determined that the mitigation of a large break LOCA occurring in L
F Mode 4 requires one law Pressure Injection (LPI)' train aligned to the Borated Water Storage
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1 (U. S. Nuclear Regulatory Commission" j 3F109718 :
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' aligne:1 in'the DHR mode of operation. In this plant configuration and:with a large break-
- LOC _A', the recent analysis indicates the time available for the operator to realign the LPI train-
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from DHR2and initiate the ECCS mode of operation does not have sdequate conservatism to
' ensurei the1 core remains covered 'and prevent _ possible pump _ damage.
To minindze the operator burden and improve the operator response -time, the proposed ITS chnges will require that in Mode 4, one LPI train will be dedicated to the PWST.
- Aldiough the licensing basis postulates a Mode 4 LOCA, there are minimal driving forces to
'cause a large break of the reactor primary piping at the lower temperatures and pressures of Mode 4.
Therefore'Jit is reasonable to infer that reactor primary piping degradation would
- occur at a decleased rate, and operators'would have additional time to be alerted and respond
- to such sn impending LOCA. : These operator actions woulu include attempting to identify and -
- isolate the source of the leak and ensuring RCS. inventory is -maintained througi, normal makeup. Further, _CR-3_ ITS 3.4.12 requires _tiie plant ta be placed in Mode 5 if pressure iboundary leaks exist or if identified or unidentified leakage exceeds certain limits. These -
' actions are a part of existing CR-3 procedures. Also, emergency operating procedures require
,nitiation of LPI to maintain the core covered with reactor coolant if a lear did propagate into a LOCA in Mode 4, Consideration of the decreased rate of reauor primary piping degradation in Mode 4 provides additional time for the operator to respond to'a postulated large break of reactor coolant primary piping. -Simulator exercises in wh4ch the operators responded to both small and large break LOCAs confirm the operator capability to implement the abnormal and emergency operating procedures required and maintain the core covered with i
For RCS attached piping (i.e., other than primary piping), the evaluation assumed that the j;
postulated Mode 4 LOCA was the result of a instantaneous pipe rupture. The worst-case pipe would be the decay heat drop line based on size and locatian. The evaluation determined that the time,:vailable for operator action to maintain the core covered with reactor coolant is 7.5 1
minutes. -Simulator exercises confirm the operator capability to implement the required emergency operating procedures within the required time.
o The evaluation of the Mode 4 DHR requirements is based, in part, upon analpes conducted by L
FTI. The FTI analyses are complete, but have not been formally issued pending quality
- assurance verification, which is expected to be completed by the end of November 1997. FPC does not expect the conclusions of this LAR to_be affected by the FTI analyses verification.
LTherefore, FPC is submitting this LAR, at this time, to maximize the time available for NRC review.-_-If any changes are necessary, FPC will promptly notify the NRC.
changes due to other outstanding: license amendment change requests. FPC has determined 1that not outstanding _CR-3. license amendment request is -affected by the changes proposed
-herein. - Spuitically, the proposed ITS changes do not affect.TSCRN 210 since TSCRN 210 p
- addresses Mode 1 LOCAs and = LAR 214 addresses Mode 4 LOCAs. LTOP requirements l proposed by;TSCRN 213 to deactivate HPI under 'certain circumstances are also preserved-isince operator actions proposed by LAR 214 and the available time to complete those actions i
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- Uf S. Noc!sar Rtgulatory Commission 13F1097-18' Page 4; w
- s provide fo'r reactivating HPI, if necessary.; Prior to issuing the proposed license amendment, j FPC will coordinate the actual page changes of the CR-3 ITS with the NRC Project Manager.
4 A;TSCRN;209 (Ref. 4). adds, in.part, instrumentation providing indication of reactor coolant Subcooling to ITS 3.3.17, PAM Instrumentation. -Mitigation of Mode 4 LOCAs would include
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- indication of reactor coolant subcooling. Implementation of LAR 214 is dependent upon the' 7
use of the subcooling instrumentation and the NRC's prior approval of TSCRN 209.
j Submittal Format The LAR consists of several attachments that should be considered by the NRC in' support of the tiroposed LAR. A summary of these attachments follows:
attachment A - List of Commiingn11
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Tie attachment provides the list of commitments made in this submittal.
Attachment B - License Amendment Reauest 214. Revision 0
--The attachment;provides the description, justification, and evaluations of the proposed
- li;cnse amendment. FPC has also included in the attachment its:
'e
- 10 CFR 50.92(c) evaluation and conclusion that the proposed ITS changes
'and operator actions do not involve a significant hazard, and 10 CFR 50.22(c)(9) evaluation and conclusion that the proposed ITS' changes and operator actions meet the eligibility criteria for categorical exclusion and <10 not require an environmental assessment or environmenial impact statement.
- Attachment C - Redline / Strikeout ITS & Bases Paggg Ptoposed changes to the ITS and Bases pages are provided-in this attachment. To assist the NRC in reviewing the L'AR, the deletions are shown in strikeout font, and new or-revised 'information is shown by shaded fent.
Attachment D - Revision BrtITS & Bases Paaes The attachment provides a copy of the ITS pages with revision bars indicating the proposed changes.'
TAttachment E = Ooerator Actions _.
LThe attachment identifies the operator actions necessary to implement the proposed ITS.
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FPC requests specific NRC review of the operator actions presented in this' attachment.as
'an integral part of the license amendment review.
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- 3F1097 tPage 5 Attachmerit_E-l Safety Assessment 4
The attachment documents FPC's. approach to addressing the Mode 4 LOCAsc The assessment includes background information, a description of analyses demonstrating that sufficient time is available'for operators to mitigate Mode 4 LOCAs, and a discussion of.
simulator exerciser demonstrating that the operating procedures at CR-3 are adequate for mitigation.
FPC respectfully requests that the NRC promptly consider this proposed license amendment arut provide its approval as soon~as possible. To allow as much time as possible for NRC review, FPC will forego the usual 30-day implementation period and will complete its written implementation plan tr parallel with NRC review. Pending NRC approval of LAR 214, FPC
.will administratively implement LAR 214 prior to entry 'into Mode 4.
Administrative -
limplementation is conservative in that the requirements of LAR 214 equal or exceed the current Technical Specifications requirements by increasing the equipment required to be operable in Mode 4, thereby, reducing operator burden and decreasing the time to initiate LPl.
To facilitate prompt NRC review and approval of this submittal, FPC suggests a meeting be
' held about the second week'of November 1997 in the NRC's Rockville, MD offices. During this meeting, FPC anticipates presenting the evaluation methodologies and proposed operator actions to the NRC.
If you have any questions concerning this proposed license amendment, please contact Mr. David Kunsemiller, Manager, Nuclear Licensing at (352) 563-4566.
Sincerely, j$hM John Paul Cowan Vice President Nuclear Production
-JPC/ mal cc:
Regional Administrator, Region 11 Senior Resident inspector NRR Project Manager JAttachments:
A. List of Commitments B. License Amendment Request 214 C. Redline / Strikeout ITS & Bases Pages D. Revision Bar ITS & Bases Pages E. Operator Actions F. Safety Assessment
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U. S. Nucicar Regulatory Conunission 3F109/-18 Page 6 STATE OF FLORIDA COUNTY OF CITRUS John Paul Cowan states th:.t he is the Vice President, Nuclear Pnxtuction for Florida Power Corporat!on; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the infonnation attached hereto, 2nd that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information, and belief,
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ggf&: _ _ J John Paul Cowan, Vice President Nuclear Production Swcrn f.o and subscribed before me this 3/3/ day of (Je/cher" 1997,by John L - Cowan.
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Signature of Notary Public State of Florida n
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(Print, type, or stamp Commissioned Name of Notary Public)
Personally Produced Known
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FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 i
ATTACllMENT A i
LIST OF COMMITMENTS i
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A*ITACHMENT A LIST OF COMMITMENTS 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. Please notify the Manager, Nuclear Licensing of any questions regarding this document or any associated regulatory commitments.
ID Nmber Comrnitment Due Date 3F1097181 FIC does not expect the conclusions of Prior to issuance of the this IAR to be affected by the FTI license amendment analyses verification. If any changes are resulting from LAR 214.
i necessary, FPC will promptly notify the
- NRC, 3F109718-2 FIC will coordinate the actual page Prior to issuing the changes of the CR-3 Technical proposed license Specifications with the NRC Project amendment.
Manager.
3F1097-18 3 FPC will administratively implement Pending NRC approval of LAR 214.
LAR 214 and prior to entry into Mode 4.
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FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 ATTACIIMENT B LICENSE AMENDMENT REQUEST 214 REVISION 0 l
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A'ITACIIS!ENT 11 1,1 CENSE AMENDMENT REQUEST 214, REVISION 0 This attachtnent provides:
a description of the changes to the Technical Specifications and Bases, e
the reason for the changes, e
the justification for the changes, e
a detennination of no significant hazards, and e
an environmental impact evaluation.
e Itedline/ Strikeout pages and Revision Dar pages of the Technical Specifications and Bases are included in Attachments C and D.
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U.S. Nuclear Regulatory Commission Attactunent B 3F109718 Page 2 PART I - EFW REQUIREMENTS FOR MODE 4 OVERVIEW:
These are the changes resulting from the engineering evaluation of the requirements for decay heat removal capabilities in Mode 4.
LICENSEE DOCUMENTS INVOLVED: Technical Specincations & Bases PORTIONS:
B 3.3.11 - Haws for EFIC Instrumentallnu Applicability - Manual EFW actuation in Mode 4 3.3.12 - EFIC Manual Initiation Applicability - Manual EFW actuation in Mode 4 Condition E - Relocate manual FFV ectuation Condition F - New condition in he EFW actuation The Bases of this specincation ha & wised accordingly.
113.3.14 - Bases for EFIC-EFW Vector Va!ve Logic Action A.1 - EFW System Reference 3.3.17 - PAM instrumentation Applicability - Mode 4 applicability Condl tion E - Mode 5 shutdown The llases of this specification are also revised accordingly.
3.4.5 - RCS Loor+ Mode 4 LCO - Available RCS loops Action A - Inoperable DilR loop Action B - Inoperable RCS loops The Bases of this specification are also revised accordingly.
3.5.3 - ECCS-Shutdown -
SR 3.5,3.1 - LPl/D11R operability The Bases of this specification are also revised accordingly.
3.7.4 - TBVs Applicability - Mode 4 applicability Condition B - Inoperable Turbine Bypass Valves (TBV).
The Bases of this specincation are also revised accordingly.
U.S. Nuclear Regulatory Commission Attachment B 3F109718 Page 3 3.7.5 - EFW System Title - Title change The Bases are also revised accordingly.
3.7.6 - Emercency FeedwatrLTank Applicability Mode 4 applicability Condition B Inoperable Emergency Feedwater Tank The Bases of this specincation are also revised accordingly.
3.7.18 - EFW System - Shutdown New specincation and bases to address the requirements for EFW in Mode 4.
SUMMARY
OF CIIANGES The Technical Specifications and associated Bases were revised to require in Mode 4 one operable EFW train and associated equipment, including the EFW Tank, manual EFIC actuation instrumentation for EFW, PAM instrumentation, and the TBVs. Additionally, the Technical Speci0 cations and associated Bases are revised to require in Mode 4 an LFI train, dedicated to the BWST, and to reflect that the available loops for decay heat removal do not include this dedicated LFI train. Editorial changes were made to clarify the description of Mode 4 accidents requiring ECCS injection, and to revise the title of LCO 3.7.5.
TECilNICAL SPECIFICATION 3.3.11 - EFIC Instrumentation 3.3.12 - EFIC ManualInitiation 3.3.14 - EFIC-EFW Vector Valve Logic l
3.3.17 - FAM instrumentation
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3.4.5 RCS loops-Mode 4 l
3.5.3 ECCS Shutdown 3.7.4 TBVs 3.7.5 EFW System Emergency Feedwater Tank l
3.7.6 3.7.18 - EFW System - Shutdown Ikscription of Reauest Bases B 3.3.11, Applicability - Manual EFW actuation in Mode 4 The Bases are revised to refer to Specification 3.3.12 for manual actuation features of EFIC and to clarify that the automatic actuation features of EFIC instrumentation are not required operable in Mode 4.
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U.S. Nuclear Regulatory Commission Attachment B 3F109718 Page 4 TS 3.3.12, Applicability - hiamial EFW actuation in hiode 4 TS 3.3.12, Condition E - Relocate manual EFW actuation TS 3.3.12, Condition F - New condition for manual EFW actuation The Applicability is revised to require manual EFW actuation in hiode 4.
Condition E is revised to relocate manual EFW actuation to new Condition F.
The Bases are revised accordingly.
Il 3.3.14, Action A.1 EFW System Reference The reference to the LCO for EFW System is revised to reflect the revised title for LCO 3.7.5 and to correct the LCO number.
TS 3.3.17, Applicability - Afode 4 Applicability TS 3.3.17 Condition E - Alode 5 Shutdown The Applicability is revised to require PAh! Instrumentation in hiode 4.
Condition E is revised to required shuidown to hiode 5. The Bases are revised accordingly.
3.4.5, LCO - Ava'lable RCS k> ops 3.4.5, Action A - Inoperable DilR loop 3.4.5, Action H - Inoperable RCS loops The LCO and Action statements are revised to reflect that only one DilR loop would be available in hiode 4 for removal of core decay heat, sin'e the remaining DilR loop would be in standby LPl. dedicated to the llorated Water Storage Tank.
The Bases are revised accordingly SR 3.5.3.1, LPI/DilR operability Bases H 3.5.3, Accidents requiring ECCS injection The note associated with the Surveillance Requirement is removed to require a standby LPl.
dedicated to the Borated Water Storage Tank. The Bases are revised accordingly and to clarify the description of hiode 4 accidents requiring ECCS injection.
TS 3.7.4, Applicability - Mode 4 Applicability TS 3.7.4, Condition H - Inoperable THVs The Applicability is expanded to require operable TBVs in hiode 4. Condition B is revised to require plant shutdown to blode 5 within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> in the event that the required action and completion time for inoperable TBVs could not be satisfied.
The Bases are revised accordingly.
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U.S. Nuclear Regul: tory Commission Attachment B 3F1097-18 Page 5 TS 3.7.5, Title - EFW System The title of LCO 3.7.5 is revised to "EFW System Operating" to distinguish it from the new proposed LCO 3.7.18. "EFW System Shutdown." The Bases are revised to reflect the new
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title and reference LCO 3.7.18.
TS 3.7.6, Applicability - Mode 4 Applicability TS 3.7.6, Condition H - Inoperable Emergency Feedwater Tank The Applicability is revised to require an operable Emergency Feedwater Tank in Mode 4.
Condition B is revised to require plant shutdown to Mode 5 within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> in the event that the required action and completion time for inoperable Emergency Feedwater Tank could not be satisfied. The Bases are revised accordingly.
TS 3.7.18, EFW System - Shutdown A new specincation and bases are issued to address the requiremenia for EFW in Mode 4.
This specification requires immediate action to restore an inoperable motor driven EFW train to operable status or plant shutdown to Mode 5 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if restoration can not be reasonably accomplished.
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Reason for Rtquest improved Technical Specification (ITS) 3.7.5, EFW Systems, requires operable Emergency Feedwater trains in Modes 1,2, and 3, but does not require Emergency Feedwater in Mode 4.
FPC provided the NRC with Technical Specification Change Request Notice (TSCRN) 210 which documented the results of the small break LOCA analysis. Although this analysis involves accident scenarios occurring in Mode 1, the results led FPC to evaluate whether DilR requirements specified by the current CR-3 ITS were sufficient for Mode 4.
Consequently, FPC committed in TSCRN 210 to complete an evaluation of DilR requirements in Mode 4 i
prior to restart, but noted that the evaluation would not affect TSCRN 210.
FPC also provided the NRC with.TSCRN 213 to establish the requirements for a Low Temperature Overpressure Protection (LTOP) System as required by 10CFR50,60. That submittal also referred to the evaluation of DilR requirements in Mode 4 and noted that any resulting revisions to the ITS would preserve the LTOP requirements proposed by TSCRN 213.
The evaluation has determined that the mitigation of a small break LOCA occurring in Mode 4 requires one train of Emergency Cooling (ECCS) and one source of feedwater to the steam generators. The current CR-3 ITS for Mode 4 require one train of ECCS, but do not require EFW, The proposed Technical Specification changes would require one train of EFW and j
associated equipment, such as the EFIC Manual Initiation, in Mode 4.
The Bases of ITS 3.3.11 are also revised to clarify that the operability of the manual Emergency Feedwater
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l U.S. Nuclear Regulatory Commission Attachment B 3F1097-18 Page 6 Initiation and Control (EFIC) actuation features in Mode 4 are addressed by revised ITS 3.3.12.
The recent evaluation also determined that the mitigation of a large break LOCA occurring in Mode 4 requires one Low Pressure Injection (LPI) train aligned to the Borated Water Storage Tank (BWST). Ilowever, the current ITS require one LPI train to be cperable, bat allows that train to be aligned in the DilR mode of operation. In this plai.t connguration and with a large break LOCA, the recent analysis indicates that the time available for the operator to realign the LPI train from DIIR and initiate the ECCS mode of operation to prevent core temperature excursions and possible pump damage for certain line breaks does not have adequate conservatism. To minimize the operator burden and improve the operator response time, the proposed ITS changes will require one LPI train in Mode 4, dedicated to the BWST. The remaining DilR train would be available for DilR mode of operation.
Additionally, the description of accidents stated in the Bases of CR-3 ITS 3.5.3 is revised by replacing ' Design Basis Accident (DBA), LOCAs and Design Basis Events (DBE)' with
' accidents requiring ECCS injection.' The Basis to Technical Specification 3.5.2, ECCS-Operating, describes the function of ECCS is to protect the reactor core after three different accidents: LOCA, strtn generator tube rupture, and steam line break. Such accidents have been analyzed for CR-3 with initial conditions in Mode 1. The postulated occurrence of such accidents in Mode 4 is not the basis of original ECCS design.
litalmalion of Regttest FPC reviewed the present and past licensing basis of CR-3. This review included docketed correspondence, Technical Specifications and related license amendments, and engineering documents. This review of the CR-3 licensing basis concluded that:
(1) LOCAs are postulated to occur in Mode 4, (2) Certain automatic Engineered Safeguards Actuation System actuations are not available in Mode 4, and operator actions are relied upon for LOCA mitigation, and (3) No safety analyses are performed with initial conditions in Mode 4.
ITS 3.5.3, ECCS-Shutdown, currently states that sufficient time exists for manual operator actuation of the required ECCS to mitigate the consequences of a DBA in Mode 4. ITS 3.5.3 requires one train of ECCS operable in Mode 4, but allows the LPI train to be aligned in the DilR mode of operation. Further, ITS 3.7.5, EFW Systems, req" ires operable Emergency Feedwater trains in Modes 1,2, and 3, but does not require Emergency Feedwater in Mode 4.
U.S. Nuclear Regulatory Commission Attachment il 3F1097-18 Page 7 The current CR 3 ITS are based on NUREG 1430, B&W Standard Technical Specification.
Information available at the time CR-3 was converting to the ITS (see FPC letter dated August 10,1990, 3F0890-05) indicated that approximately 30 minutes was available to manually initiate ECCS following a Made 4 LOCA. The analysis for supporting Reference 3 did not consider RCS losses due to Hashing and carryout, and heat from metal surfaces of reactor vessel internals.
Recently, FPC, in conjunction with Framatome Technologies Inc. (FTI), performed an evaluation of DilR requirements for Mode 4. Since CR-3 LOCA analyses are performed with initial conditions at full power, the Mode 4 evaluation is based on extrapolations of existing Mode 1 analyses, which assume decay heat levels and plant configurations unique to Mode 4 at C R 1 Additionally, FPC considered the reactor coolant losses due to Hashing and carryout, and heat from reactor vessel internals in the current evaluation. This is an important distinction from the previous Mode 4 evaluation since these reactor coolant losses decrease the available time for operator actions.
The initial operating conditions and plant connguration in Mode 4 were used in the current evaluation. The CR-3 ITS denne Mode 4 as having an average reactor coolant system (RCS) temperature between 200*F and 280*F and the core subcritical. Normal makeup and letdown are used to control pressuriter level. Core decay heat was calculated assuming that Mode 4 is achieved in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following a reactor trip from 102 percent of full power of 2568 Mwt.
This point in time was selected because it is the shortest amount of time the ITS specify for placing the plant in Mode 4 from full power to comply with an action statement.
For the purposes of analysis, manual actuation of the required ECCS functions was assumed to mitigate the LOCA occurring in Mode 4.
Certain automatic engineered safeguard features would be disabled in Mode 4, and reactor building pressure actuation is conservatively assumed not to occur. Also, the core Dood tanks would be isolated prior to entering Mode 4.
Additionally, IIPI injection would be operable, but may be deactivated under certain conditions in Mode 4 consistent with the requirements of the proposed Technical Specincation Change Request Notice 213 dated July 18, 1997 (3F0797-10) associated with Low Temperature Over-pressure Protection System (LTOPS).
The evaluation also assumes that during a Mode 4 LOCA, offsite power would be available and no si.igle failure uuld occur. A single failure in Mode 4 is not required to be assumed.n accordance with the current Bases of ITS 3.5.3. A loss of power is also not assumed since CR 3 would not be generating electricity in Mode 4 and there would be no reactor and turbine trips resulting from the postulated Mode 4 LOCA to cause grid disturbance.
Attachment F, Safety Assessment, presents the results of the evaluation. A summary of the conclusions is as follows:
For a small break LOCA in Mode 4. LAR 214 would ensure that the manual actuation of IIPI and EFW is avaibble from the Control Room, and can be completed within 10 minutes to maintain the core covered with reactor coolant.
U.S. Nuclear Regul: tory Commission Attachment B 3F109718 Page 8 For a large break LOCA in hiode 4, LAR 214 would ensure that manual actuation of LPI is available from the Control Room, and can be completed within 7.5 minutes to maintain the core covered with reactor coolant.
Although the licensing basis postulates a hiode 4 LOCA, there are minimal driving forces to cause a large break LOCA of the reactor primary piping at the lower temperatures and pressures of hiode 4.
Therefore, it is reasonable to infer that a decreased rate of reactor primary piping degradation would occur, and operators would have additional time to be alerted and respond to such an impending LOCA. This position is supported by the Bases of Specification 3.4.12, RCS Operational LEAKAGE, which states:
... operational LEAKAGE [other than primary to secondary LEAKAGE] is related to the safety analyses for a LOCA in that the amount of leakage can affect the probability of such an event."
The operator actions would include attempting to identify and isolate the source of the leak and ensuring RCS inventory is maintained through normal makeup. Further, CR-3 ITS 3.4.12 requires the plant to be placed in blode 5 if pressure boundary leaks exist or if identified or unidentified leakage exceeds certain limits. These actions are addressed in existing CR-3 procedures. Emergency operating procedures require ir.itiation of LPI to maintain the core covered with reactor coolant if such a leak did propagate into a LOCA in hiode 4.
Consideration of decreased rate of reactor primary piping degradation provides additional time for the operator to respond to e postulated large break of reactor coolart primary piping in hinde 4.
Although no single failure is assumed for the hiode 4 LOCA, two channels of Post Accident Monitoring (PAhi) instrumentation are required operable in Mode 4 by the revised ITS. FPC elected to apply the current requirements for 2 channel operability in Mode 4 for simplicity.
PAM instrumentation required by the operators during mitigation of Mode 4 LOCAs include indication of reactor coolant subcooling.
TSCRN 209 adds, in part, instrumentation providing indication of reactor coolant subcooling to ifs 3.3.17, PAM instrumentation, implementation of this LAR is dependent upon the use of the subcooling instrumentation and the NRC's prior approval of TSCRN 209.
The motor d.iven EFW pump would still be considered operable in Mode 4 when manually secured and locked out with its associated hand switch.
Securing the pump in Mode 4 prevents hiadvertent EFW actuation which could ove:rfill the steam generator and result in an uncontrolled cooldown. Previous simulator exercises have demonstrated the ability of the operators to reactivate the EFW pump, if necessary.
Additionally, the description of accidents stated in the Bases of CR-3 ITS 3.5.3 is revised by replacing ' Design Basis Accident (DBA), LOCAs and Design Basis Events (DBE)' with
' accidents requiring ECCS injection.' The Basis to Technical Specification 3.5.2, ECCS-Operating, describes the function of ECCS is to protect the reactor core after three different accidents: LOCA, steam generator tube rupture, and steam line break. Such accidents have
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U.S. Nuclear Regul: tory Commission Att:chment B 3F1097-18 Page 9 been analyzed for CR-3 with initial conditions in hiode 1. The postulated occurrence of such accidents in hiode 4 is not the basis of original ECCS design.
FPC has evaluated the changes by LAR 214 relative to the currently proposed changes in i
TSCRNs 210 and 213. TSCRN 210 addresses accident conditions in hiode 1 and, therefore, is not impacted by these proposed changes involving hiode 4 accidents requiring ECCS injection.
The purpose of TSCRN 213 is to ensure that inadvertent actuation of IIPI in certain low temperature conditions does not overpressurize the reactor vessel. FPC intends to implement the de activation of IIPI when in LTOPS conditions by de-powering the four IIPI injection valves using the power selector switches on the Control Board. By applying this LTOPS protection, a minimum of two operator actions would be required to initiate IIPl. These actions are to actuate llPI and then restore power to the llPI valves. These actions are performed within the Control Room.
Simulator exercises of the operator actions have confirmed that the operator response time to activate llPI is sufficient to maintain the core covered with reactor coolant in the unlikely event of a small break LOCA in hiod: 4 while IIPI is deactivated in accordance with LTOPS requirements.
NO SIGNIFICANT IIAZARDS EVALUATION:
FPC has evaluated the provisions in 10 CFR 50.92(c) regarding the proposed operator actions and ITS changes of LAR 214 and TSCRN 209, associated with the subcooling instrumentation, and concludes that no significant hazard is involved. In support of these conclusions, the following evaluation is provided:
- 1. The proposed ITS changes and operator actions involving mitigation of postulated hiode 4 LOCAs will not result in a significant increase in the probability of an accident previously evaluated. The initiators of any accident previously evaluated are not affected by the proposed ITS changes and operator actions involving mitigation of hiode 4 LOCAs.
Consequently, there is no significant impact on any previously evaluated accident probabilities.
The proposed ITS changes and operator actions involving mitigation of blode 4 LOCAs do not result in a significant increase in the consequences of any accidents previously evaluated.
The proposed ITS changes, modifications and operator actions will not adversely affect the integrated ability of any system to perform its intended safety functions. Therefore, the combined ability of these components, systems and actions to mitigate the consequences of a hiode 4 LOCA will continue to be maintained, in fact, the collective impact of these ITS changes and operator actions improve the capability of CR-3 to mitigate biode 4 LOCAs by requiring additional equipment operable in hiode 4, by reducing operator burden, and by decreasing the time to initiate LPl. The proposed ITS changes are either consistent with or exceed the original licensing and design basis for
U.S. Nuclear Regulatory Commission Attachment B 3F109718 Page 10 CR 3. In addition, the ITS changes and operator actions do not affect the onsite or offsite doses which remain well below 10 CFR Part 100 limits.
- 2. The proposed ITS changes and operator actions do not create the possibility of a new or different kind of accident from any accident previously evaluated. Since, the ITS changes and operator actions do not involve a different initiator for any accident previously evaluated, they also do not create any n:w kind of accident.
Mitigation of Mode 4 LOCAs, utilizing manual actions, is already part of the CR 3 licensing basis, hianual operator actions necessary for the mitigation of hlode 4 LOCAs are currently addressed or are being addressed in CR 3 procedures.
- 3. The proposed ITS changes and eperator actions do not involve a significant reduction in the margin of safety for mitigation of Mode 4 LOCAs. In fact, the collective impact of the ITS changes and operator actions represent a improvement in the overall margin of safety to a degree that exceeds the original plant design and licensing bases for mitigation of Mode 4 LOCAs by requiring additional equipment operable in Mode 4, by reducing operator burden, and by decreasing the ".ime to initiate LPl. The proposed ITS changes are either consistent with or exceed the orig nal licensing and design basis for CR-3.
ENVIRONMENTAL IMPACT EVALUATION:
10 CFR 51.22(c)(9) provides criteria for and identification of licensing and regulatory actions eligible fm categorical exclusion from performing an environmental assessment. A proposed amendment to an operating license for a facility requiret no environmental assessment if operation of the facility in accordance with the proposed amendment would not: (1) involve a signincant hazard; (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released off site; or (3) result in a significant increase ir, indisidual or cumulative occupational radiation exposure. As set forth below, the proposed operator actions and ITS changes proposed by LAR 214 and TSCRN 209, associated with the subcooling instrumentation, satisfies the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c),
- 1. The proposed ITS changes and operator actions do not involve a significant hazard as described previously in the No Significant llazards evaluation.
- 2. The proposed ITS changes and operator actions do not result in a significant change in the types, or a significant increase in the amounts, of any effluents that may be released
- offsite,
- 3. The proposed ITS changes a -1 operator actions do not result in a significant increase in individual or cumulative occupational radiation exposure. As previously described in the No Significant llazards evaluation, these changes provide a net improvement in the ability to mitigate the consequences of a Mode 4 LOCA and, therefore, do not significantly affect the onsite or offsite doses which remain a small fraction of 10 CFR 100 limits.
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