ML20024G209
| ML20024G209 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 09/30/1977 |
| From: | Wachter L NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20024G208 | List: |
| References | |
| A00L-770930, AL-770930, NUDOCS 9102080330 | |
| Download: ML20024G209 (13) | |
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UNITED STATES NUCLEAR REGilATORY CotMISSION NORTilERN STATES POWER COMPANY P MONTICELLO NUCLEAR GENERATING PLANT h Docket No. 50- 263 A
REQlfEST FOR AMENDMENT TO }
'. OPERATING LICENSE NO. DPR-22 h
............................. 4 L-(License Amendment Request Dated September 30, 1977) lF l;
Northern States Power Company, a Minnesota corporation, requests authorization for changes to the Operating License and the Technical Specifications as rhown on the actachments labeled Exhibit A, Exhibit B,
] and Exhibit C. Exhibit A describes the proposed amendment to the Exhibit B l: Operating License along with the reason for the amendment.
l describes the proposed changes to the Technical Specifications along with reasons for the change. Exhibit C is a set of Technical Specification pages incorporating the proposed changes.
l This request contains no restricted or other defense information.
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NORTilERN STATES POWER COMPANY l
i By d8MM
/M L J Wachter
\ ice President, Power Production &
System Operation 30th . day o f September , 1977 , before me a On this_
notary public in and for said County, personally appeared L J Wachter, Vice President, Power Production 6 System Operation, and first being duly sworn acknowledged that he is authorized to execute this document in behalf of Northern Staten Power Company, that he knows the contents thereof and tha t to the best of his knowledge, information and belief, the statements made in-it are true and that it is not interposed for delay.
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0 p DENISE E. HALVORSON [
- yy wo ARY PUBUC - M!NNESOTA ;
A I;:(M HENNEPIN COUNTY ::
My Comm;4sen Fxp;res Oct 10.1981 y 9102030330 yyog3f ~ ~ n w :- :^ :-
DR ADOCK 05000263 PDR
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3 EXHIBIT A i
i MONTICELLO NUCLEAR GENERATING PIANI j
g" DOCKET No. 50-263 LICENSE NO. DPR-22 '
A LICENSE AMENDMENT REQUEST i k .
DATED SEPTEMBER 30, 1977 )
2 PROPOSED CHANGES TO PROVISIONAL OPERATING LICENSE DPR-22 ',
Pursuant to 10CFR50.59, the holders of Provisional Operating License DPR-22
- hereby propose the following change to License DPR-22.
PROPOSED CHANGE Delete the following provision from the license which was added by a March 11, 1977 Nuclear Regulatory Commission Order for Modification of License:
"(1) As soon as possible, the licensee shall submit 4 a re-evaluation of ECCS cooling performance calculated in accordance with General Electric i Company's Evaluation Model approved by the }
c NRC staff and corrected for the errors descri-bed herein and any other corrections in the Model of which the licensee is aware at the l
[ time the calculations are performed." j REASON FOR CHANGE
^
The re-evaluation of ECCS cooling perfomance has been submitted pursuant to ttis paragraph, making the referenced paragraph obsolete. The re-evaluation was provided under cover letter dated September 15, 1977 from Mr L 0 Mayer l
_(NSP) to Mr V Stello (USNRC). The re-evaluation report is entitled, " Loss-of-Coolant Accident Analysis Report for Monticello Nuclear Generating Plant; NEDO-24050, September, 1977." 1 l-
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-SAFETY EVALUATION _
The ' safety eval ution for the proposed amendment is embodied in the re-evaluation q report referenced above and the numerous references documented therein. h n H j
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5, EXHIBIT B }i MONTICELLO NUCLEAR GENERATING PLANr DOCKET No. 50-263 LICENSE NO. DPR-22
>1 LICENSE AMEt'DMENT REQUEST y"
DATED SEPTEMBER 30, 1977 PROPOSED CHANGES TO TE.CHNICAL SPECIFICATIONS APPENDIX A 0F PROVISIONAL OPERATING LICENSE DPR-22 Pursuant to 10CFR50.59, the holders of Provisional Operating License DPR-22 hereby propose the following change to the Appendix A Technical Specifications. .f
- 1. LIST OF FIGURES (Pages vii and 1891)
PROPOSED WANGES Replace the words " Maximum Average Linear Heat Generation Rate versus l Average Exposure Monticello 70230 Fuel" with the word " Deleted". Like-wise, all information in Figure 3.11.1-B should be replaced by the words "This Figure has been Deleted".
REASON FOR CHANGES The 70230 fuel type is no longer ured in the Monticello reactor and therefore no replacement figure for the thermal limits for that fuel type is presented.
SAFETY EVALUATION This change only removes cbsolete caterial from the technical specifications.
- 2. SPECIFICATION 3.11.A and BASIS (Pages 189B,189C and 189F)
PROPOSED CHANGES On Page 189B, add the sentence "When core flow is less than 707. of rated flow, the APLHGR shall not exceed 907, of the limiting value shown in Figures 3.11.1." as shown in Exhibit C. Move overflow naterial to the next page, 189C.
On Page 199F 3 er.11ce Reference 6 with the following refe::ence, " 'Re' ision of Low Core P t.u 3ffects on LOCA Analysis for Operating BWR's,' R L G:idley'(GE) to D G Eisenhui (USERC), September 28, 1977."
REASON FOR CHANGE The recent ECCS re-analysis shows that the most limiting break size is a break smaller than the maximum design brenk area. The low flow ECCS limitations have likewise been re-analyzed and new ifmits have been established. The new limits are bounding for the postulated loss of coolant accident over the entire break spectrum. !
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SAFETY EVALUATION i The document referenced abcve as proposed Basis Reference 6 is the safety y evaluation for this change. MonticelloisidentifiedasPlantBinthatreport.) l
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- 3. FIGURES 3.11.1-A. C and D and BASIS (Pages 189F,189H,189J and 189K) l
.P_ROPOSED CBAUG_E_S
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Replace existing Figures 3.11.1-A, C and D of the Techni:a1 Specifications, fg j Pages 189H, 189J and 189K with the corre:ponding proposed figures in-3 , ,
cluded in Exhibit C Replace Refcrence 4 on Page 189F with the reference,
" '1,oss-of-Coolant Accident Analysis . Report for Monticello Nuclear Generating ' ; ;
Plant,' NEDO-24050, September,1977, L 0 Mayer (NSP) to V Stello (USNRC) I September 15, 1977". J J.
REASON FOR CHANGE I The ECCS themal limitations for the fuel types in service at Monticello i have b,en re-evaluated using a newly approved model. The thermal limits j are being adjusted and documented accordingly.
l.
SAFETY EVALUATION The document reference above as a proposed Basis Reference 4 is the i safety evaluation for this change.
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EXIIIBIT C !N 1, lY LICENSE AMENDMENT REQUEST kM, DATED SEPTEMBER 30, 1977
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'1his exhibit consists of the following pages revised to incorporate all of the proposed Technical Specification changes:
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189B 1890
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s2 LIST OF FIGURES Page No.
Figure No.
2.1-1 Deleted 2.3.1 APRM Flow Referenced Scram and Rod Block Trip Settings 12 2.3.2 Relationship Between Peak IIeat Flux and Power for Peaking Factor of 3.08
'46 4.1.1 'M' Factor - Graphical Aid in the Selection of an Adequate Interval Between Tests 74 4.2.1 System tinavailability Sodium Pentaborate Solution Volume - Concentration Requirements 92 3.4.1 Sodium Pentaborate Solution Tempetature Requirements 93 3.4.2 3.6.1 Change in Charpy V Transition Temperature versus Neutron Exposure 122 3.6.2 Minimum Temperature versus Pressure for Pressure Tests 122A 3.6.3 Minimum Temperature versus Pressute for Mechanical lleatup or Cooldown Following Huclear Shutdown 1225 3.6.4 Minimum Temperature versus Pressure for Core Operation 122C 4.6.1 Deleted 4.6.2 Chloride Stress Corrosion Test Results e 500*F 123 4.8.1 Off-gss Storage Tank Gross Activity Limits 176A 3.11.1-A Maximum Average Linear lleat Generation Rate versus Planar Average Exposure Monticello 8D219 Fuel 18911 3.11.1-B Deleted 1 189I vii REV .;
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3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.11 REACTOR FUEL ASSEMBLIES 4.11 REACTOR 'TEL ASSEMBLIES
- Applicability Applicability The Limiting Conditions for Operation The Surveillance Requirements apply to associated with the fuel rods apply to tha parameters which monitor the fuel those parameters which monitor the fuel rod operating conditions.
rod operating conditions.
Objective Objective The objective of the Limiting Co'nditions The objective of the Surveillance Requirements for Operation is to assure the perfor- is to specify the type and frequency of surveil-mance of the fuel rods. lance to be applied to the fuel rods.
Specifications Specifications A. Average Planar Linear Heat Genera- A. Average Planar Linear Heat Genera-tion Rate (APLHGR) tion Rate (APLHGR)
During power operation, the APLHGR The APLHCR for each type of fuel as a for each type of fuel as a function function of average planar exposure shall of average planar exposure shall not be determined daily during reactor operation exceed the limiting value shown in at >25% rated thermal power.
Figures 3.11.1. When core flow is less than 90% of rated core flow, the APLHGR shall not exceed 95% of the limiting value shown in Figures 3.11.1.
When core flow is less than 707. of rated core flow, the AFLHGR shall not exceed 907. of the limiting value shown in Figures 3.11.1. If any time during operation it is determined thau the ,
timit for APIlIGR is being exceeded, action shall be initiated within 15 189B "
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3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENfS minutes to restore operation to within the prescribed limits. Surveillance and corres-ponding action shall continue until reactor .
. operation is within the prescribed limits.
I If the APUIGR is not returned to within the >
prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown l condition'within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
B. Linear Heat Generation Rate (UlGR) B. Linear Heat Generation Rate (UlGR)
During power operation, the UlGR as a function The UIGR as a function of core height of core height shall not exceed the limiting shall be checked daily during reactor value shown in Figure 3.11.2. If at any time operation at 2 257. of rated thermal during operation it is determined that the power.
limiting value for UIGR is being exceeded, action shall be initiated within 15.minutca to restore operation to within t.he prescribed limits. Surveillance and corresponding action shall continue until reactor operation is with-in the prescribed limits. If the UIGR is not is returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutuown cor.dition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
189C 3.11/4.11 REV E q
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Bases 3.11 (continued)
C. Minimum Critical Power Ratio (MCPR)
The ECCS evaluati0n presented in Reference 4 assumed the steady state MCPM prior to the .
postulated loss-of-coolant accident to be 1.18 for all fuel types. In addirion, ~9- CCCS analysis presented in Reference 6 assumed an initial MCPR of 1.24 for reduceJ ' w rca-ditions. The Operating MCPR Limit of 1.38 for 8x8 fuel and 1.29 for 7x7 fuel 2 decenained from the analysis of transients discussed in Bases Sections 2.1 and 2.3. By r s %t w.ng an operating MCPR above these limits, the Safety Limit of 1.06 (T.S.2.1.A) applica de tc all fuel typer* is maintained in the event of the most limiting abnormal operational transient.
For operation with less than rated core flow the Operating MCPR Limit is adji. sted bf multiplying the above limit by Kr. Reference 5 discusses how the transient ar.alysis done at rated conditions encompasses the reduced flow situation when the proper Kf factor l 1s applied.
Those abnormal operational transients, analyzed in FSAR Section 14.5, which rerult in an automatic reactor scram are not considered a violation of the LCO. Exceeding MCPR limits in such cases need not be reported.
References
- 1. " Fuel Densification Effects in General Electric Boiling Water Reactor Fuel," Svplements 6, 7, and 8, NEDM-10735, August, 1973.
- 2. Supplement I to Yechnical Report on Densification of General Electric Reactor Fuels, December h 14, 1974 (USAEC Regulatory Staff)
- 3. Communication: VAMoore to IS Mitchell, " Modified CE Model for Fuel Densification,"
Docket 50-321, March 27, 1974.
- 4. " Loss-of-coolant Accident Analysis Report for the Monticello Nuclear Generating Plant,"
NEDO-24050, September 1977, L 0 Mayer (NSP) to V Stello (USNRC), Septembar 15, 1977.
- 5. " General Electric BWR Generic Reload Application for 8 x 8 Fuel," NEDO-20360, Revision 1, November 1974.
- 6. " Revision of Low Core Flow Effects on LOCA Analysis for Operating BWR's," R L Gridley (GE) to D G Eisenhut (USNRC), September 23, 1977.
3.I1 BASES 189F
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i 50~2LT ! [ ( NRC DISTRIBUTION Pon PART 50 DOCKET MATERIAL '""""" TO* FROM: ,; V Stello oAra or cocVMENT horthern States Power Company Minneapolis, Mn 9,39,77 j L 0 Mayer DAT,,,C,,v,,10-3-77
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'casen Prion ENCLOSURS 3 6 /4 Dea 6 i j notarized 9-30-77. . . . . . trans th . following: Licnese # DPR-22 Amend:
f proposed change to tech specs concerning incorporation of ECCS re- k analysis....................................... i. p a 12p 3p . , PLANTMont NAME:icello
. 10- -77 ehr S AFETY kd [Ndd FOR ACTION /INFORMATION ! BRANCH CHIEF: (7) JO46//S A INTERNAL OISTRIBUTION .AEG FM T atN FDR ICE (2)
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