ML043430148
| ML043430148 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 11/29/2004 |
| From: | Silberg J Entergy Nuclear Operations, Entergy Nuclear Vermont Yankee, ShawPittman, LLP |
| To: | Atomic Safety and Licensing Board Panel |
| Byrdsong A T | |
| References | |
| 50-271-OLA, ASLBP 04-832-02-OLA, RAS 8939 | |
| Download: ML043430148 (22) | |
Text
'RAS RAS 8939 g November 29, 2004 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION DOCKETED Before the Atomic Safety and Licensing Board USNRC November 29, 2004 (3:38pm) n tOFFICE OF SECRETARY In the Matter of ) RULEMAKINGS AND
) Docket No. 50-271 ADJUDICATIONS STAFF ENTERGY NUCLEAR VERMONT )
YANKEE, LLC and ENTERGY ) ASLBP No. 04-832-02-OLA NUCLEAR OPERATIONS, INC. ) (Operating License Amendment)
(Vermont Yankee Nuclear Power Station) )
)
ENTERGY'S RESPONSE TO NEW ARGUMENTS RAISED IN VERMONT DEPARTMENT OF PUBLIC SERVICE'S REPLY TO ENTERGY'S ANSWER TO THE DEPARTMENT'S REQUEST FOR LEAVE TO FILE A NEW CONTENTION Applicants Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc.
(collectively "Entergy") hereby respond to new arguments raised in the Vermont Department of Public Service's ("DPS") Reply' to Entergy's answer in opposition 2 to DPS's request for admission of a new contention. 3 Although 10 C.F.R. §2.309(h)(3) does not contemplate a response to a reply to an answer to a request to file a new contention, Commission case law permits such a response when the reply introduces new arguments that were not raised in the petitioner's initial filing. ClevelandElectric Illuminating Co. (Perry Nuclear Power Plant, Units 1 &2), LBP-82-89, 16 NRC 1355, 1356-57 (1982). That is precisely the situation here.
Notwithstanding the DPS Reply's attempt to buttress its arguments by the use of unsupported and inappropriate accusations of "distortions," "misrepresentations," "stretch[ing of] facts," "mischaracteriz[ing of] arguments," and "mislead[ing] this Board", DPS Reply at 1, 2, 1 "Vermont Department of Public Service Reply to Answer of Applicant to the Department's Request for Leave to File a New Contention" (November 17, 2004) ("DPS Reply").
2 "Entergy's Answer to Vermont Department Of Public Service Request For Leave To File A New Contention" (November 10, 2004) ("Entergy's November 10 Answer").
3 "Vermont Department of Public Service Request for Leave to File a New Contention" (October 18, 2004)
("DPS Request").
lep 4esC >/-o 0,41 O;
4, 5, etc., the fact is that the DPS Reply sets forth bases and arguments that cannot be found in the proposed contention itself or elsewhere in DPS's initial Request. The new arguments should be rejected since they are raised, without justification, for the first time in DPS' Reply and also because they fail to otherwise present an admissible issue.
I. BACKGROUND Proposed DPS Contention 6 states:
The Application for Amendment, Including All Supplements Thereto, Fails to Comply With 10 CFR 50 Appendix R, Specific Requirements, Paragraph L(2)(b) Because It Does Not Verify The Assumption, Used For Purposes of the Safe Shutdown Capability Analysis (SSCA), that the Reactor Core Isolation Cooling (RCIC)
System Can Be Made Operable In Sufficient Time To Permit The Operator To Perform the Required Actions Before Core Uncovery.
DPS Request at 1. As the text of proposed Contention 6 and the supporting arguments originally offered make clear, the exclusive focus of the proposed contention is that Entergy's application does verify that there is sufficient time for operator actions to prevent core uncovery. The text of the proposed contention itself states that the application fails to comply with Commission regulations "Because It Does Not VeriSj The Assumption" of sufficient time for operator action.
Id. (emphasis added). DPS supports the proposed contention by correctly quoting Entergy's response to the NRC's request for additional information (DPS Ex. 38) for the proposition that Entergy "is in the process of verifying this assumption of' adequate time for operator action and that "[t]his verification" will be completed by December 1, 2004. Id. at 2 (emphasis added).
The "supporting evidence" for the proposed contention further states that Entergy "has yet to verify how much time it will take to implement those procedures". Id. (emphasis added). There is not a suggestion in the DPS Request that DPS seeks to challenge the results of the verification rather than the fact that verification has not taken place. In its November 10 Answer, Entergy opposed admission of proposed Contention 6 by addressing DPS's claim that verification had not yet taken place.
2
II. NEW ARGUMENTS MADE II THIE DPS REPLY, AND RESPONSES THERETO In contrast with the proposed contention's exclusive focus on the fact that verification had not yet taken place, the DPS Reply attempts to morph the proposed contention into an attack on the results of the verification. DPS's shift is made clear throughout its Reply:
"Applicant equates a commitment to perform verification with actually fulfilling the regulatory requirement." DPS Reply at 3.
"Performing a verification is not the end in itself." Id.
"It is the results of a verification that are controlling . . ." Id.
[I]t is necessary to see if the verification process was properly conducted and to see if it demonstrates compliance with the regulatory standard." Id. at 4.
These are new arguments and newly asserted claims. DPS provides neither justification why these new arguments could not have been set forth in its original Request nor seeks, as required, leave from the Board to advance them. Houston Lighting & Power Co. (Allens Creek Nuclear Generating Station, Unit 1), ALAB-565, 10 NRC 521, 523 n. 11 (1979) (substantive alterations of proposed contentions can only be done with leave of the Board.) The new arguments in DPS's Reply, therefore, should be rejected by the Board as improperly raised.
Furthermore, these new arguments do not give rise to an admissible contention, since they are speculative concerns for which DPS has cited no support. See Duke Energy Corp.
(McGuire Nuclear Station, Units 1 and 2), CLI-02-14, 55 NRC 278, 293 (2002).
In its Reply, DPS challenges Entergy's objection that the original contention is based on information not materially different than information previously available. DPS Reply at 6. .This challenge, too, takes the form of a new argument:
Furthermore, at hearing DPS will contest the method Entergy uses to make it's
[sic] determination. For example, Entergy does not use the 2-sigma value for decay heat uncertainty which it has stated it uses in other analyses. Nor does 3
Entergy use the design basis value of 85 degrees F for initial service water temperature, but rather a less conservative value of 80 degrees F. At hearing DPS will show Entergy has even less time to perform the required actions than claimed.
DPS Reply at 7. None of these assertions can be found in DPS' original filing. That is reason enough to disregard them. But the allegations are also without citation or support 4, further reason to disregard them.
The final reason to disregard these late, unsupported allegations is that they are simply wrong. The current calculation of record for Extended Power Uprate ("EPU") conditions uses both values that DPS says are not used, ie., the design basis value of 850 F for initial service water temperature and the 2 sigma value for decay heat uncertainty 5 . While an earlier non-EPU calculation used the 800 F initial RHR service water temperature and did not include the 2 sigma uncertainty6, that calculation was superceded in 2000 for current licensing basis conditions 7 and in 2003 for EPU application by the calculation cited in footnote 5.
Finally, DPS argues that "[a]n analysis that is found to be out of date and non-conservative, that reduces the time margins available for event mitigation by 50%, is definitely something that must either be remedied or the license amendment application must be denied."
DPS Reply at 8. Here too DPS is incorrect. For this statement, DPS relies on DPS Exhibit 40 (NRC's November 5, 2004 Engineering Inspection Preliminary Results at A-7). DPS Reply at 6.
However, DPS' selective quotation of the Preliminary Results omits several key passages which neutralize whatever support DPS seeks to draw from the document. First, DPS omits the NRC's 4Neither the Reply nor its accompanying Affidavit of William K. Sherman provide even a single reference to support these statements.
5See Attachment I hereto, Vermont Yankee Calculation VYC-2306 (approved August 29, 2003), "Torus Temperature for Appendix R Events at EPU Conditions", pages 15,17.
6 See Attachment 2 hereto, Vermont Yankee Calculation VYC-1 628D (approved November 5, 1998), "Torus Temperature Response to Appendix R and Station Blackout Scenarios". pages 43-44.
7See Attachment 3 hereto, Vermont Yankee Calculation VYC-2120 (approved June 20, 2000), "Torus Temperature for Appendix R Events with a RHR Service Water Temperature of 850F", pages 13, 16.
4
conclusion that "[t]his finding is of very low safety significance because at current licensed operating conditions, sufficient margin would remain to ensure that the core would not be uncovered during the analyzed event." DPS Exhibit at A-7. The Preliminary Results go on to observe that "the loss of margin may have prevented the operators from initiating the RCIC system in sufficient time to prevent core uncovery under EPU conditions" but significantly qualifies this observation with the introductory phrase, "Had thisfinding not been identified".
Since the finding has been identifled (by the Preliminary Results), the sole predicate for DPS' conclusion disappears. Again, rank speculation that provides no support for the proposed contention. Here to the issue has as a factual matter been resolved. 8 III. CONCLUSION For the reasons stated above and in Entergy's November 10 Answer, DPS's request for the admission of proposed Contention 6 should be denied.
Respectfully submitted, Ja }ilberg Matias F. Travieso-Diaz Douglas J. Rosinski SHAW PITTMAN LLP 2300 N Street, N.W.
Washington, DC 20037-1128 Tel. (202) 663-8063 Counsel for Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc.
Dated: November 29, 2004 8See Attachment 4 hereto, "Apparent Cause Evaluation Report, CR Number CR-VTY-2004-2552" (noting questioning of margin for RCIC initiation during 2004 NRC Engineering Asscssment, page 1, and completion of steps "to regain lost margin in the current timeline," page 3).
5
November 29, 2004 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safetv and Licensing Board In the Matter of )
) Docket No. 50-271 ENTERGY NUCLEAR VERMONT )
YANKEE, LLC and ENTERGY ) ASLBP No. 04-832-02-OLA NUCLEAR OPERATIONS, INC. ) (Operating License Amendment)
(Vermont Yankee Nuclear Power Station) )
)
CERTIFICATE OF SERVICE I hereby certify that copies of "Entergy's Response to New Arguments Raised in Vermont Department of Public Service's Reply to Entergy's Answer to Vermont Department of Public Service Request for Leave to File a New Contention" were served on the persons listed below by deposit in the U.S. Mail, first class, postage prepaid, and where indicated by an asterisk by electronic mail, this 29th day of November, 2004.
- Administrative Judge *Administrative Judge Alex S. Karlin, Chair Lester S. Rubenstein Atomic Safety and Licensing Board Panel Atomic Safety and Licensing Board Panel Mail Stop T-3 F23 Mail Stop T-3 F23 U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Washington, D.C. 20555-0001 ask2Oinrc.gov lesrrr(msn.com
- Administrative Judge Atomic Safety and Licensing Board Dr. Anthony J. Baratta Mail Stop T-3 F23 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Mail Stop T-3 F23 Washington, D.C. 20555-0001 U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 aib(Snrc.gov
- Secretary Office of Commission Appellate Att'n: Rulemakings and Adjudications Staff Adjudication Mail Stop 0-16 CI Mail Stop 0-16 Cl U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Washington, D.C. 20555-0001 secv(anrc.gov, hearinfzdocket(inrc. gov
- Sarah Hofmann *Brooke Poole, Esq.
Special Counsel *Robert Weisman, Esq.
Department of Public Service *Marisa Higgins, Esq.
112 State Street - Drawer 20 Office of the General Counsel Montpelier, VT 05620-2601 Mail Stop 0-15 D21 Sarah.Hofmann(Th)state.vt.us U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 bdpi.nrc.gov, rmnv(Tnrc.gov, rnch5(nanrc.gov
- Anthony Z. Roisman *Jonathan M. Block National Legal Scholars Law Firm 94 Main Street 84 East Thetford Rd. P.O. Box 566 Lyme, NH 03768 Putney, VT 05346-0566 aroisman(avalley.net ionb()sover.net
- Raymond Shadis New England Coalition P.O. Box 98 Shadis Road Edgecomb ME 04556 shadis(eprexar.com Matias F. Travieso-Diaz Document #: 1445547 v.2 This document contains Vermont Yankee proprietary information. This information may not be transmitted in whole or in part, to any other organization without permission of Vermont Yankee.
VY CALCULATION TITLE PAGE VYC-2306 0 N/A NIA VY Calculation Number Revision Number Vendor Calculation Number Revision Number
?,I.- TOnrSi Te.mnterature.for Annendix R Events. nt PM Cotinditinns QA Status: 0 SC D NNS El OQA Operating Cycle Number* N/A
- The Opcrating Cycle Number should only be entered here if the results of the calculation only apply during a specific operating cycle otherwise enter 'N/A".
Calculation Supports A Design Change/Specification? El Yes 0D No N/A a _J( VYDCMIMMtMISpec No.
Implementation Required? O Yes ,I'No Calculation Done as a Study Only? E] Yes 0 No Safety Evaluation Number: N/A Superseded Calculation Number, Title and Revision: N/A For Revisions: List CCNs, IlIs, or SAs incorporated/supcrseded by this revision: N/A Computer Code(s): GOTHIC 7T0M2 Are there open items in this calculation/revision? B3 Yes [] No Review and Approval: (Print and Sign Name)
Prrnarer Illiane Schor ) K k4 Date, 8/zq/a 2c m3 I
Preparer Alan Robertshaw (Sections 2.2. 3.2. 4.2 & 5.2) aL& , QA44 Date: . A( ai-st DOOM lnterdiscipline Reviewer(s): _ 1 A Date:
1.-lnAn-,nt V2v- cl Alnn DnhPrtthnw - r-n~t-i 4-. i ,"
UdPnn1P,,,nr RPviPw#,.mM(r t1_Mn,-_Stqhnr
...... .I---- - _.Vqexvion.m 3.-7-I42~&r7S2'h (~/~ IfTiie+ r)ftt- 9 1-2-C? 1-2-cyn"Z,
__1 Approved: *Jc s G rle elCy Dlate-4.v. ,
z, , ,
o Accepted (only for AP 0017 calculations performed by vendors) 0 N/A Date:_
Final Tumover to DCC (Section 2):
- 1) All open items. if any, have been closed.
- 2) Implementation Confirmation (Section 2.3.4) Total No. Pages in Package (including all attachments)
D Calculation accurately reflects existing plant configuration, 121 (confirmation method indicated below)
El Walkdown El As-Built input review El Discussion with (Print Name)
OR El NIA, calculation does not reflect existing plant configuration
- 3) Resolution ofdocuments identified in the Design Output Documents Section of VYAPF0017.07 has been initiated as required (Section 2.3.6, 2.3.7)
/ I Printed Name Signature Date Page 1 of 58 Pages*
- For calculations performed using AP0017 this is the number of pages in the body of the calculation.
For vendor calculations, this is the number of p3ges of AP0017 forms added.
(Title page, review forms, data sheets, 50.59, etc.)
VYAPF 0017.01 AP 0017 Rev. 8 Page I of I LPC #2
Torus Temperature for Appendix R Events, at EPU Conditions.
Calculation Number: VYC-2306 Revision Number. 0 CCN Number: N/A Page 15 of 58 Table 1 :Vessel and Core Initial Conditions Parameter Nominal Analysis Basis Comments Value Value Power 1912 MWth 1950 MWth 120% power + 2%
calorimetric uncertainty Decay Heat ANS 5.1 ANS Reference 3 indicated that (Decay Heat from Reference 5) 5.1+2sigrna the 2 sigma uncertainty is needed to compensate for neglecting other actinide than the ones in the ANS 5.1 standard. appropriate to be used Vessel Pressure 1020 psia 1045 psia Reference 19 Appendix R analysis for EPU which determrines the time to vessel uncovery uses a vessel pressure of 1035.5 psia (see References 18) as an upper bound of the nominal pressure.
The use of J045 psia for the vessel pressure for the torus temperature Appendix R analysis is judged to be conservative.
Vessel Level 162 inches 172 inches Reference 19 Analysis value conservatively accounts for 3 inches increase above normal (uncertainty and operational fluctuations) and 7 inches for dimensional uncertainties. The reviewer comments (Reference 19, page65) identified that the normal level could be as high as 165 inches. Hence the analysis could be performed at a vessel level of 175 inches. An additional 3 inches in level lead to an increase in the suppression pool temperature of 0.2 OF which was judged to be insignificant.
MSIV closure time 3-5 seconds 3 seconds The Tech Spec minimum value closure retains more energy in the vessel Core Flow Rate 48.0E6 lb/hr 51 .36e6 lb/hr Includes ICF of 7%. VY operates with only 5% 1CF, however the analysis would allow a 7% increase in core flow. ICF conditions lead to
. higher stored energy.
Initial Feedwater 7.8786e6 lb/hr 8.076e6 lb/hr Feedwater flow consistent Reference 20 (TE 2003-020)
Flowrate with power level, matches Steady State steam flow rate.
Initial feedwater 393.5-393.6 0F 393.9 OF Consistent with 20% See discussion in TE 2003-020.
temperature power uprate +2%(
calorimetric uncertainty (Reference 20)
Torus Temperature for Appendix R Events, at EPU Conditions.
Calculation Number: VYC-2306 Revision Number. 0 CCN Number: NIA Page 17 of 58 Table 2 - ECCS Initial Conditions and Parameters Parameter Nominal Analysis Basis Comments Value Core Spray Flow Curve of flow Same as The core spray flow rate The Core Spray System will be vs. vessel-torus nominal. used in the Appendix R used for level control only after the AP. analysis will be used. feedwater is depleted.
The flow rate is determined function of During the alternate shutdown the vessel-torus AP. cooling a flow of 3500 gpm at 100 (consistent with the psid will be used, provided that LOCA analysis) (OPL4 adequate NPSH is available to
-Reference 21) support this flow (open item).
RHR Flow (t>600 7000 gpm 7000 gpm Nominal RHR flow used The Appendix R analysis is seconds) for the Appendix R performed assuming nominal application. conditions.
Assumption unchanged Based on the results of the for EPU analysis, the available RHR pump can be split between vessel and torus cooling.
RIIR Hx Tube N/A 5% Allowable plugging Plugging margin RHRSW Flow 2950-3140 2950 gpm Assumption unchanged The Appendix R analysis is gpm (per OP- for EPU performed assuming nominal 2124 - Rev 49 conditions.
Reference 10 RHRSW Inlet 32-85 OF 85.1 OF Refcrence 13 Maximum allowable Service Temperature Water Temperature iAY ORIGINAL: PAGE Rev. l:PAGE I of I of
- i PAGES PAGES Rev. 2
- PAGE I of PAGES Rev.3: PAGE I of PAGES
- I QA RECORD? NO. 13.C16.036; 13.C09.001 X YES NO Safety Class/P.O. NO. (if applicable) SC I; 00241.00.0090.00.0000 YANKEE ATOMIC ELECTRICCOMPANY CALCULATION/ANALYSIS FOR TITLE Torus Temperature Response to Appendix R and Station Blackout Scenarios PLANT Vermont Yankee CYCLE N/A CALCULATIONNUMBER VYC-1628D Supplement to VYC-1628 Rev. 0 KEYWORDS: GOTHIC60a, Wetwell, Torus, SBO (Station Blackout), LPCI (Low Pressure Coolant Injection), CS (Core Spray), CSCS (Core Standby Cooling System), ECCS (Emergency Core Cooling System)
NPSH (Net Positive Suction Head)
COMPUTER CODES: GOTHIC60a, EXCEL 97 EQUIP/TAGNOs.: P-1-IA, B, C, D; P-46-IA, B SYSTEMS: RHR, CS, Primary Containment
REFERENCES:
See page 45 FORM WE-103-1
- For individual reviewer scope, see pgs. 51 &-55. Revision 5 CDF - VYC- 628D App. R & SDO Scenitaios November 5,1998
- fyrm-em
Torus Temperature Response to Appendix R and VYC-1628D Rev. 0 Station Blackout Scenarios Page 43 4.0 Conclusions The two most limiting, previously analyzed App. R scenarios were examined. For the most limiting App. R scenario, a variety of input sensitivities were performed. The most limiting, previously analyzed SBO scenario was examined. Peak torus temperatures for these cases were determined. A summary of the results is provided in Table 6. Run 01 through Run 04 analyze the same scenario but with modified input assumptions. Run 05 and Run 06 present additional scenarios, an App. R scenario and a station blackout scenario, respectively.
Table 6 App. R and SBO Torus Temperature Case Results Summary ID Title Peak Torus Temperature RunOl App. R Case 3a, Fully Conservative 187.50 F @ 20,140 sec RunO2 App. R Case 3a, Run 01 + Reduced Decay Heat Uncertainty 186.20 F @ 20,080 scc RunO3 App. R Case 3a, Run 02 + Reduced RHR and RHRSW Flow Uncertainty 184.0 0 F @ 19,840 sec RunO4 App. R Case 3a, Run 03 + Reduced SW Temperature Uncertainty 181.9 0 F @ 19,680 sec:
RunO5 App. R Case 4, Fully Conservative 178.5 0 F @ 6,574 sec r RunO6 SBO Scenario 3, Fully Conservative 176.80 F @ 24,000 sec For Appendix R scenario Case 4 (Run 05) and the station blackout scenario (Run 06), the results show peak torus temperatures under 182.1 0F using the same input parameters as the LBLOCA and SBLOCA torus temperature analyses. For Appendix R scenario Case 3a, the peak torus temperature results are shown lo be under 182.1 0F using a revised set of input parameters (Run 04). As these input parameters are appropriate for use in Appendix R safe shutdown analyses, Run 04 is proposed as the revised basis for this Appendix R scenario. Therefore, it is concluded that the App. R and SBO events are bounded by the large break LOCA peak torus temperature results.
The limiting Appendix R scenario uses several input assumptions different than used in Ref. 5:
- the decay heat values used in this analysis are cycle independent as opposed to those used in Ref. 5 that were cycle dependent; this is a more conservative assumption than wvas used Ref. 5;
- the decay heat input for Run 04 does not include the 2a decay heat uncertainty. The use of this decay heat assumption is justified in Section 3.1.2 and is a less conservative assumption than was used in Ref. 5; CDF .VYC-162SD App. R& 530 Scenarios November 3, 1998
VYC- I 628D Rev. 0 Torus Temperaturc Response to Appendix R 2nd Page 44 Station Blackout Scenarios percentile the assumed river water temperature in this calculation is 80°F that reprcscnts a > g9 th is a less value for this environmental condition. The use of this value isjustified in Section 3.1.4 and for river conservative assumption than was used in Ref. 5. Note that plant design and operating limit water temperature remains 85 F. The input assumption used here does not change this limit.
0 basis For the last two items, the basis for the acceptability of using inputs that deviate from design accident input parameters is the statement in 10 CFR 50 Appendix R para. III.L.6 that states:
to meet "Shutdown systems installed to ensure postfire shutdown capability need not be designed
... other design basis accident criteria..."
and 3.1.4. The The values used for these two input parameters are specifically justified in Sections 3.1.2 specific input resulting analysis is still conservative even while it minimizes the conservatism on these parameters.
NovcmEkr 4, 1998 CDF- VYC 162D A,-p. R A SDOSc4nar.1s This document contains Vermont Yankee proprietary information. This information may not be transmitted in whole 1 I or in part, to any other organization without permission of Vermont Yankee.
VY CALCULATION TITLE PAGE 2120 0 N/A N/A VY Calculation Number Revision Number Vendor Calculation Number Revision Number
Title:
Torus Temperature for Appendix R Events with a RHR Service WVatcr Temperature of 85 V' I QA Status: [RI SC [ NNS ] OQA Operating Cycle Number Cycle Independent Calculation Supports A Design Change/Specification? EYes [1K No VYDCIMMTMSpec No.
Calculation Supports An Independent Analysis? ] Yes 2 No VYC-1628D Rev. 0 WRFO-0058Trask 12 Reference Calculation Done as a Study Only? Yes [C] No [El Safety Evaluation Number: NIA Superseded Calculation Number, Title and Revision: N/A Review and Approval: (Print and Sign N G-k4-oo 40 Preparer: Philip J Guimnond Date: 6-ls -a&
Interdiscipline Reviewer(s): Dan Yasi 9 v9 Carl Fag < Date: d/ dbI 6/6L/v-Independent Reviewers(s): Farrokh Seifaee l I Date: '4 11f /0 0 Approved: M. W. Gmyrek . Date:
Open Items Associated with Calculation? E Yes [mK] No Closed 1Total No. Pages in Package (including attachments) 2O7 Installation Verification
[I Calculation accurately reflects plant as-built configuration, OR
[] N/A, calculation does not affect plant configuration
/ l Printed Name Signature Date Page 1 of 7-CX Pages (body of calculation)
I z es166-16?
VYAPF 0017.01 (Sample)
AP 0017 Rev. 6 Page I of I
VY CALCULATION SHEET O Calculation Number: VYC-2120 N/A Revision Number:
Page /3 of 0
_ CCN Number:
Table 1 :Vessel and Core Initial Conditions Parametcr Nominal Analysis Basis Comments Value Value Power 1593 MV~th 1625 MWth 100% power+ 2%
calorimetric uncertainty Decay Heat ANS S.1 ANS 5.1 Based on ANS 5.1 + 2a benchmarking with was also used to ANS 5.1 + 2a Appendix R demonstrate that the was also used assumptions used uncertainty would not by other utilities result in temperature (Pilgrim and exceeding the acceptable Cooper Nuclear value Pover Station)
Vessel Pressure 1020 psia 1045 psia Reference 14 Appendix R analysis which determines the time to vessel uncovery uses a vessel pressure of 1035.5 psia (see References 4 and
- 13) as an upper bound of the nominal pressure. The use of 1045 psia for the vessel pressure for the torus temperature Appendix R analysis is judged to be conservative.
Analysis value represents a 25 psi increase above nominal to account for operational fluctuations and uncertaintiy.
Vessel Level 162 inches 172 inches Reference 14 Analysis value conservatively accounts for 3 inches increase above nominal (uncertainty and operational fluctuations) and 7 inches for dimensional uncertainties and thermal expansion.
More details can be found in Reference 14, Page 65.The reviewer comments (Reference 14) identified that the normal level could be as high as
_ __ 165 inches. Hence the 0 VYAPF 0017.03 (Sample)
AP 0017 Rev. 6 Page 1 of I
VY CALCULATION SHEET Calculation Number: VYC-2120 Revision Number: 0 CCN Number. N/A Page /J of Table 2 - ECCS Initial Conditions and Parameters Parameter Nominal Analysis Basis Comments Value Core Spray Flow Curve of flow Sare as The core spray The Core Spray System vs. vessel-torus nominal. flow rate used in was used for level control AP. the Appendix R only after the feedwater is torus temperature depleted.
(Reference 2)analysis was During the alternate used. The flow rate shutdown cooling a flow is determined as n of 3500 gpm at 100 psid function of the was used. Adequate vessel-torus AP. NPSH is available to (consislent with the support this flow.
LOCA analysis)
(References 5 and
___ ___ 13)._ _
RHR Flow (t>600 7000 gprn 7000 gpm Nominal RHR flow The Appendix R analysis seconds) used for the is performed assuming Appendix R nominal conditions.
application.
RHR Hx Tube N/A 5% Allowable Plugging plugging margin RHRHX Flow 29503140 2950 gpm_ The Appendix R analysis gpm (per OP- is performed assuming 2124 -Rev 49 norninal conditions.
Reference 10 RHRSW Inlet 32-85 °F 85.1 *F Reference 17 Maximum allowable Temperature Service Water Temperature + O.10F.
The additional 0. I'F provides a conservative margin for RHRSW pump heat addition and has negligible impact on peak torus temperature results.
VYAPF 0017.03 (Sample)
AP 0017 Rev. 6 Page I of I APPARENT CAUSE EVALUATION REPORT CR Number: CR-VTY-2004-2552 Assigned Department: Operations Investigator: John Twarog Ext: (802) 258-5467 Date Report Completed:
E The CR was caused by or identified an equipment/component failure.
§3 The CR involved one or more HU inappropriate acts.
Problem Statement:
Existing Appendix R alternate shutdown timelines may be too long to support EPU conditions.
Report Narrative:
As a result of the 2004 NRC Engineering Assessment, an inspector questioned how ENVY controls, reviews and documents operator actions and timeline credited in the licensee design basis. The question arose during the discussion of the Appendix WRsafe shutdown timelines.
Background
An Appendix 1R"fire, that requires a plant shutdown from outside the Control Room, requires certain actions to be performed within a certain time. One of these actions is to initiate injection with RCIC prior to uncovering the core.
At current license power limit (CLPL) of 1593 Mth, the calculated time for uncovering the core is 25.34 minutes and injection with the RCIC system must commence within this time (Ref. VYC -1917). The current revision of the Safe Shutdown Capability Analysis (SSCA) references a time to Initiate RCIC injection of -15 minutes. This is based on a series of walk-downs performed in 1996 which document a time of 13.33 minutes (Ref. Memo, W. M. Pittman to L E.
Doane, Timelines - Operator Actions in OP 3126 Rev. 14.). An informational walkdown performed in 2001 resulted in a RCIC initiation time of 18.82 minutes. In 2003, new electrical safety requirements were added (coveralls and face shield) that added an additional 2 minutes to the timeline. This resulted in a total RCIC initiation time of 20.82 minutes which is still less than 25.34 minutes required in VYC-1 917. At our present power level and using the present electrical safety standard, there is approximately -4.5 minutes of margin in the RCIC initiation timeline.
The company is currently evaluating additional safety requirements, i.e. flash suits, and if the use of the flash suits becomes a requirement, then an additional 2 to 4 minutes of time will be required to compensate for dress out and the increased difficulty of working in a flash suit. This would have the potential to invalidate the current timeline.-
A walkdown, utilizing OP-3126 'Shutdown Using Alternate Methods', was performed with an NRC Inspector, the Assistant Operator Manager and a Control Room Supervisor to confirm compliance at the CLPL Although the walkdown wasn't a complete walkdown of the OP3126 procedure, the NRC Inspector concurred with Operations the required action can be was completed within the required time of 25.34 minutes.
It was noted that at EPU conditions there may not be enough margin in the timeline to complete the required actions.
The vessel boil off time under 20% EPU conditions is analyzed to be 21.33 minutes. (Ref. VYC-2270) This is -4 minutes shorter than the timeline for the CLPL and results in a margin of 0.51 minutes using the 2001 timeline with the current safety requirements. This may not be enough margin to support an alternate shutdown scenario at 20% EPU conditions and if the flash suits are required, the margin would be negative.
- .I-
Extent of Condition An Extent of Condition Evaluation was performed to identify other areas where timeliness of Operator's Action is required. This evaluation Is attached as Appendix A to this Condition Report.
Common Cause Analysis:
Not required to be performed since Condition Report is not an Adverse Trend.
RELATED OPERATING EXPERIENCE:
Previous Related Conditions (ENVY):
A search of the PCRS database was performed and the following ENVY Condition Reports were identified:
CR-VTY-1996-0192 "Communications Systems not adequate to ensure timely OP3126 completion"
- CR-VTY-1 995-0533 "Insufficient time to initiate Altemate Shutdown System prior to core being uncovered" Both of these Condition Report identified problems with the timeline associated with an Appendix WR" fire where shutdown is required to be performed outside of the Control Room. CR-VTY-1 995-0533 identified the current 25 minute time requirement.
Related Industry (OE):
A fleet wide conference call was held on August 26, 2004 to discuss the issue of processes other sites utilize to validate timelines associated with required Operator Actions. Additional benchmarking was performed with utilities outside of the Entergy fleet. None of the other sites contacted had a formal program to ensure changes in licensing basis operators actions are identified and evaluated.
Corrective Actions:
Immediate/lnterim Actions Completed Item # Action taken AC-1 Safety Standard Variances (SSV) No. 2004-42 and 2004-43 (attached) have been completed to regain lost margin in the current timeline due to donning PPE.