Tech Specs Change Request 30,Amend 2,to Amend DPR-50,App a to Conservatively Account for Maximum Effects of Fuel Rod Bowing Possibly Encountered During Cycle 2 Operation. Certificate of Svc EnclML19260A070 |
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Three Mile Island ![Constellation icon.png](/w/images/b/be/Constellation_icon.png) |
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Issue date: |
04/02/1976 |
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Arnold R METROPOLITAN EDISON CO. |
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ML19260A067 |
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References |
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NUDOCS 7910290694 |
Download: ML19260A070 (10) |
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Category:TECHNICAL SPECIFICATIONS
MONTHYEARML20211M6651999-09-0101 September 1999 Errata Page 4-45,reflecting Proposed Changes Requested in ML20211D1551999-08-20020 August 1999 Proposed Tech Specs Pages,Revising Degraded Voltage Relay as-left Setpoint Tolerances ML20210J1261999-07-29029 July 1999 Proposed Tech Specs Revising ESF Sys Leakage Limits in post-accident Recirculation Surveillance TSs ML20195E6201999-06-0404 June 1999 Proposed Tech Specs,Modifying Conditions Which Allow Reduction in Number of Means for Maintaining Decay Heat Removal Capability During Shutdown Conditions ML20206R1171999-05-13013 May 1999 Proposed Tech Specs Section 3.1.1,incorporating Administrative Updating & Changing Bases Statement ML20205H0781999-04-0101 April 1999 Proposed Tech Specs Adding LCO Action Statements,Making SRs More Consistent with NUREG-1430,correcting Conflicts or Inconsistencies Caused by Earlier TS Revs & Revising SFP Sampling ML20203B0511999-02-0202 February 1999 Proposed Tech Specs Expanding Scope of Systems & Test Requirements for post-accident RB Sump Recirculation ESF Systems & Increasing Max Allowable Leakage of TS 4.5.4 for Applicable Portions of ESF Systems Outside of Containment ML20196H5361998-12-0303 December 1998 Proposed Tech Specs Reflecting Decrease in RCS Flow Resulting from Revised Analysis to Allow Operation of Plant with 20% Average Level of SG Tubes Plugged Per SG ML20196G4861998-12-0303 December 1998 Non-proprietary Proposed Tech Specs,Consenting to Transfer & Authorization for Amergen to Possess,Use & Operate TMI-1 Under Essentially Same Conditions & Authorizations Included in Existing License ML20196F8661998-11-25025 November 1998 Proposed Tech Specs Revised Pages for TS Change 277 Changing Surveillances Specs for OTSG ISI for TMI Cycle 13 RFO Exams Which Would Be Applicable for One Cycle of Operation Only. with Certificate of Svc ML20154Q6271998-10-19019 October 1998 Proposed Tech Specs Adding Operability & SRs for Remote Shutdown Sys Similar to Requirements in NUREG-1430, Std Tech Specs - B&W Plants, Section 3.3.18 ML20154P8661998-10-19019 October 1998 Proposed Tech Specs,Providing Allowable RCS Specific Activity Limit Based on OTSG Insp Results Performed Each Refueling Outage ML20249B2421998-06-11011 June 1998 Proposed Tech Specs Re Alternate High Radiation Area Control ML20217J8201998-03-25025 March 1998 Proposed Tech Specs Page 6-1,reflecting Change in Trade Name of Owners & Operator of TMI-1 & Correcting Typo ML20217E5311998-03-23023 March 1998 Proposed Tech Specs Pages for Section 3.1.2 to Incorporate New Pressure Limits for Reactor Vessel IAW 10CFR50,App G for Period of Applicability Through 17.7 EFPY ML20217G0201997-10-0303 October 1997 Proposed Tech Specs Re Revised Pages 4-80 & 4-81 Previously Submitted ML20211F3781997-09-24024 September 1997 Proposed Tech Specs Revising Steam Line Break Accident Dose Consequence ML20211C3421997-09-19019 September 1997 Proposed Tech Specs Pages 3.8-3.9b to TS Section 3.1.4 Providing More Restrictive Limit of 0.35 Uci/Gram Dose Equivalent I-131 & Clarifying UFSAR Analysis ML20211C2431997-09-19019 September 1997 Proposed Tech Specs Re Decay Heat Removal Sys Leakage ML20210K0011997-08-14014 August 1997 Proposed Tech Specs Revising TMI-1 UFSAR Section 14.1.2.9 Environ Dose Consequences for TMI-1 Steam Line Break Analysis ML20141J4051997-08-12012 August 1997 Proposed Tech Specs Revising Surveillance Specification for Once Through Steam Generator Inservice Insp for TMI-1 Cycle 12 Refueling (12R) Exams Applicable to TMI-1 Cycle 12 Operation ML20198E7941997-07-30030 July 1997 Proposed Tech Specs Incorporating Addl Sys Leakage Limits & Leak Test Requirements for Systems Outside Containment Which Were Not Previously Contained in TS 4.5.4 Nor Considered in TMI-1 UFSAR DBA Analysis Dose Calculations for 2568 Mwt ML20151K2071997-07-25025 July 1997 Revised TS Page 6-19 Replacing Corresponding Page Contained in 970508 Transmittal of TS Change Request 264 ML20141E1491997-05-0808 May 1997 Proposed Tech Specs,Consisting of Change Request 264, Incorporating Addl NRC-approved Analytical Methods Used to Determine TMI-1 Core Operating Limits ML20140E2471997-04-21021 April 1997 Proposed Tech Specs 3.3.1.2,changing Required Borated Water in Each Core Flood Tank to 940 ft,4.5.2.1.b,lowering Surveillance Acceptance Criteria for ECCS HPI Flow to 431 Gpm & 3.3.1.1.f Re Operability of Decay Heat Valves ML20134M1211997-02-0707 February 1997 Proposed Tech Specs,Incorporating Certain Improvements from Revised STS for B&W plants,NUREG-1430 ML20133D2821996-12-24024 December 1996 Proposed Tech Specs 3.15.3 Re Auxiliary & Fuel Handling Bldg Air Treatment Sys ML20132F3301996-12-16016 December 1996 Proposed Tech Specs,Reflecting Change in Legal Name of Operator of Plant from Gpu Corp to Gpu Inc & Reflecting Plant License & TS Registered Trade Name of Gpu Energy ML20135D0991996-12-0303 December 1996 Proposed Tech Specs Incorporating Certain Requirements from Revised B&W Std TS,NUREG-1430 ML20135C5321996-12-0202 December 1996 Proposed Tech Specs Re Relocation of Audit Frequency Requirements ML20128H4121996-10-0303 October 1996 Errata to Proposed Ts,Adding Revised Table of Contents & Making Minor Editorial Corrections ML20117H0451996-08-29029 August 1996 Proposed Tech Specs,Consisting of Change Request 257, Incorporating Certain Improvements from STS for B&W Plants (NUREG-1430) ML20113D1971996-06-28028 June 1996 Proposed Tech Specs,Consisting of Change Request 259, Allowing Implementation of Recently Approved Option B to 10CFR50,App J ML20112C8791996-05-24024 May 1996 Proposed Tech Specs Re Pages for App A.Certificate of Svc Encl ML20101R1571996-04-10010 April 1996 Proposed Tech Specs,Revising Addl Group of Surveillances in Which Justification Has Been Completed ML20100J6931996-02-22022 February 1996 Proposed Tech Specs,Consisting of Change Request 254, Revising Proposed TS Page 4-46 on Paragraph 4.6.2 That Provides Addl Testing Requirements in Case Battery Cell Parameters Not Met ML20095J2311995-12-21021 December 1995 Proposed Tech Specs,Raising Low Voltage Action Level to 105 Volts DC ML20092K5471995-09-20020 September 1995 Revised Tech Spec Pages 4-31,4-32 & 4-33,incorporating Change in TS Section 4.4.1.5 ML20091L2911995-08-23023 August 1995 Proposed Tech Specs Page 6-11a,incorporating Ref to 10CFR20.1302 ML20087F4921995-08-11011 August 1995 Proposed TS Section 3.2 Re Makeup,Purification & Chemical Addition Sys Requirements ML20085N1601995-06-22022 June 1995 Proposed Tech Specs Revising Replacement Pages in Package & Remove Outdated Pages,In Response to NRC Request for Addl Info ML20084L4351995-06-0101 June 1995 Proposed TS 5.3.1.1,describing Use of Advanced Clad Assemblies ML20084N3501995-06-0101 June 1995 Proposed Tech Specs,Deleting RETS & Relocating TSs Per Guidance in GL 89-01 & NUREG-1430 ML20084B3361995-05-24024 May 1995 Proposed Tech Specs Re Change in Surveillance Test Requirements for source-range Nuclear Instrumentation ML20083N7841995-05-17017 May 1995 Proposed Tech Specs,Consisting of Change Requests 252, Removing Chemical Addition Sys Requirements from TS to COLR ML20079A7811994-12-23023 December 1994 Proposed Tech Specs Page 3-32a ML20069A6781994-05-20020 May 1994 Proposed Tech Specs,Supporting Cycle 10 Control Rod Trip Insertion Time Testing ML20065M6831994-04-19019 April 1994 Proposed Tech Specs,Reflecting Deletion of Audit Program Frequency Requirements ML20065K0451994-04-11011 April 1994 Proposed Tech Specs Reflecting Relocation of Detailed Insp Criteria,Methods & Frequencies of Containment Tendon Surveillance Program to FSAR & Providing Direct Ref to Existing Tendon Surveillance Program ML20073C7731994-03-22022 March 1994 Proposed Tech Specs Re Control Rod Trip Insertion Time Testing 1999-09-01
[Table view] Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML20211M6651999-09-0101 September 1999 Errata Page 4-45,reflecting Proposed Changes Requested in ML20211D1551999-08-20020 August 1999 Proposed Tech Specs Pages,Revising Degraded Voltage Relay as-left Setpoint Tolerances ML20210S7691999-08-12012 August 1999 Rev 10 to 6610-PLN-4200.02, TMI Emergency Dose Calculation Manual ML20210J1261999-07-29029 July 1999 Proposed Tech Specs Revising ESF Sys Leakage Limits in post-accident Recirculation Surveillance TSs ML20195E6201999-06-0404 June 1999 Proposed Tech Specs,Modifying Conditions Which Allow Reduction in Number of Means for Maintaining Decay Heat Removal Capability During Shutdown Conditions ML20206R1171999-05-13013 May 1999 Proposed Tech Specs Section 3.1.1,incorporating Administrative Updating & Changing Bases Statement ML20206R6531999-05-13013 May 1999 Rev 39 to TMI Modified Amended Physical Security Plan ML20205H0781999-04-0101 April 1999 Proposed Tech Specs Adding LCO Action Statements,Making SRs More Consistent with NUREG-1430,correcting Conflicts or Inconsistencies Caused by Earlier TS Revs & Revising SFP Sampling ML20204B5291999-03-12012 March 1999 Rev 9 to 6610-PLN-4200.02, TMI Emergency Dose Calculation Manual (Edcm) ML20203B0511999-02-0202 February 1999 Proposed Tech Specs Expanding Scope of Systems & Test Requirements for post-accident RB Sump Recirculation ESF Systems & Increasing Max Allowable Leakage of TS 4.5.4 for Applicable Portions of ESF Systems Outside of Containment ML20196G4861998-12-0303 December 1998 Non-proprietary Proposed Tech Specs,Consenting to Transfer & Authorization for Amergen to Possess,Use & Operate TMI-1 Under Essentially Same Conditions & Authorizations Included in Existing License ML20196H5361998-12-0303 December 1998 Proposed Tech Specs Reflecting Decrease in RCS Flow Resulting from Revised Analysis to Allow Operation of Plant with 20% Average Level of SG Tubes Plugged Per SG ML20196F8661998-11-25025 November 1998 Proposed Tech Specs Revised Pages for TS Change 277 Changing Surveillances Specs for OTSG ISI for TMI Cycle 13 RFO Exams Which Would Be Applicable for One Cycle of Operation Only. with Certificate of Svc ML20154P8661998-10-19019 October 1998 Proposed Tech Specs,Providing Allowable RCS Specific Activity Limit Based on OTSG Insp Results Performed Each Refueling Outage ML20154Q6271998-10-19019 October 1998 Proposed Tech Specs Adding Operability & SRs for Remote Shutdown Sys Similar to Requirements in NUREG-1430, Std Tech Specs - B&W Plants, Section 3.3.18 ML20154D5491998-10-0101 October 1998 Cancellation Notification of Temporary Change Notice 1-98-0066 to Procedure 6610-PLN-4200.02 ML20206C0911998-09-0101 September 1998 Rev 17 to Odcm ML20249B2421998-06-11011 June 1998 Proposed Tech Specs Re Alternate High Radiation Area Control ML20216E9751998-04-13013 April 1998 Emergency Dose Assessment Users Manual, for Insertion Into Rev 7 of Edcm ML20216E9491998-04-0909 April 1998 Rev 7,Temporary Change Notice 1-98-003 to 6610-PLN-4200.02, Edcm, Changing Pages 2 & 57 & Adding New Emergency Dose Assessment Users Manual ML20217J8201998-03-25025 March 1998 Proposed Tech Specs Page 6-1,reflecting Change in Trade Name of Owners & Operator of TMI-1 & Correcting Typo ML20217E5311998-03-23023 March 1998 Proposed Tech Specs Pages for Section 3.1.2 to Incorporate New Pressure Limits for Reactor Vessel IAW 10CFR50,App G for Period of Applicability Through 17.7 EFPY ML20202B2061998-01-30030 January 1998 Rev 7,Temporary Change Notice 1-98-0013 to 6610-PLN-4200.02, Edcm ML20198T4721997-12-31031 December 1997 TMI-1 Cycle 12 Startup Rept ML20217G0201997-10-0303 October 1997 Proposed Tech Specs Re Revised Pages 4-80 & 4-81 Previously Submitted ML20211F3781997-09-24024 September 1997 Proposed Tech Specs Revising Steam Line Break Accident Dose Consequence ML20211C2431997-09-19019 September 1997 Proposed Tech Specs Re Decay Heat Removal Sys Leakage ML20211C3421997-09-19019 September 1997 Proposed Tech Specs Pages 3.8-3.9b to TS Section 3.1.4 Providing More Restrictive Limit of 0.35 Uci/Gram Dose Equivalent I-131 & Clarifying UFSAR Analysis ML20210K0011997-08-14014 August 1997 Proposed Tech Specs Revising TMI-1 UFSAR Section 14.1.2.9 Environ Dose Consequences for TMI-1 Steam Line Break Analysis ML20141J4051997-08-12012 August 1997 Proposed Tech Specs Revising Surveillance Specification for Once Through Steam Generator Inservice Insp for TMI-1 Cycle 12 Refueling (12R) Exams Applicable to TMI-1 Cycle 12 Operation ML20198E7941997-07-30030 July 1997 Proposed Tech Specs Incorporating Addl Sys Leakage Limits & Leak Test Requirements for Systems Outside Containment Which Were Not Previously Contained in TS 4.5.4 Nor Considered in TMI-1 UFSAR DBA Analysis Dose Calculations for 2568 Mwt ML20151K2071997-07-25025 July 1997 Revised TS Page 6-19 Replacing Corresponding Page Contained in 970508 Transmittal of TS Change Request 264 ML20217M7251997-06-22022 June 1997 Rev 16 to Procedure 6610-PLN-4200.01, Odcm ML20141E1491997-05-0808 May 1997 Proposed Tech Specs,Consisting of Change Request 264, Incorporating Addl NRC-approved Analytical Methods Used to Determine TMI-1 Core Operating Limits ML20140E2471997-04-21021 April 1997 Proposed Tech Specs 3.3.1.2,changing Required Borated Water in Each Core Flood Tank to 940 ft,4.5.2.1.b,lowering Surveillance Acceptance Criteria for ECCS HPI Flow to 431 Gpm & 3.3.1.1.f Re Operability of Decay Heat Valves ML20134M1211997-02-0707 February 1997 Proposed Tech Specs,Incorporating Certain Improvements from Revised STS for B&W plants,NUREG-1430 ML20133D2821996-12-24024 December 1996 Proposed Tech Specs 3.15.3 Re Auxiliary & Fuel Handling Bldg Air Treatment Sys ML20132F3301996-12-16016 December 1996 Proposed Tech Specs,Reflecting Change in Legal Name of Operator of Plant from Gpu Corp to Gpu Inc & Reflecting Plant License & TS Registered Trade Name of Gpu Energy ML20135D0991996-12-0303 December 1996 Proposed Tech Specs Incorporating Certain Requirements from Revised B&W Std TS,NUREG-1430 ML20135C5321996-12-0202 December 1996 Proposed Tech Specs Re Relocation of Audit Frequency Requirements ML20128H4121996-10-0303 October 1996 Errata to Proposed Ts,Adding Revised Table of Contents & Making Minor Editorial Corrections ML20117H0451996-08-29029 August 1996 Proposed Tech Specs,Consisting of Change Request 257, Incorporating Certain Improvements from STS for B&W Plants (NUREG-1430) ML20113D1971996-06-28028 June 1996 Proposed Tech Specs,Consisting of Change Request 259, Allowing Implementation of Recently Approved Option B to 10CFR50,App J ML20112C8791996-05-24024 May 1996 Proposed Tech Specs Re Pages for App A.Certificate of Svc Encl ML20138B4641996-05-0606 May 1996 Rev 14 to Procedure 6610-PLN-4200.01, Odcm ML20101R1571996-04-10010 April 1996 Proposed Tech Specs,Revising Addl Group of Surveillances in Which Justification Has Been Completed ML20100J6931996-02-22022 February 1996 Proposed Tech Specs,Consisting of Change Request 254, Revising Proposed TS Page 4-46 on Paragraph 4.6.2 That Provides Addl Testing Requirements in Case Battery Cell Parameters Not Met ML20096F0521995-12-31031 December 1995 TMI-1 Cycle 11,Startup Rept ML20095J2311995-12-21021 December 1995 Proposed Tech Specs,Raising Low Voltage Action Level to 105 Volts DC ML20092M1941995-09-21021 September 1995 TMI-1 Pump & Valve IST Program 1999-09-01
[Table view] |
Text
1
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METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT CC:GANY -
AND PENUSYLVANIA ELECTRIC COMPANY TEREE MILE ISLAND NUCLEAR STATION UIIIT 1 '* LS..+}&
8 .-
Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Reauest No. 30 Amendment No. 2 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island nuclear Station Unit 1. As a part of this request, proposed replacement pages for Appendix A are also ir . 'ided.
METROPOLITAN EDISON COMPANY By ) (
Vice Predhent-Generhion Sworn and subscribed to me this day of , 1976.
...,s.. o.. a ubis tuv.19. H79 Notary Putfic 1479 231 391029 M
l .,
Metropolitan Edison Co. (Met-Ed)
Three Mile Island nuclear Station Unit 1 (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 Technical Specification Cnange Reauest No. 30 Amendment No. 2 The licensee requests that the att' ached changed pages (2-2, figure 2.1-2, and figure 2.3-2) replace page 2-2 figure 2.1-2 and figure 2.3-2 of Change Request No. 30 A=endment #1.
Reason for' Amendment No. 2 This supplement to our previous request is necessary to conservatively account for the maximum effects of fuel rod boving which may be encountered during cycle 2 operation.
Safety Analysis Justifying Amendment The method of analysis and the effect of fuel rod bow are discussed in the attached B&W Fuel Rod Bov Evaluation. This evaluation has conservatively
- accounted for the effects of the maximum predicted fuel rod bow during Cycle 2 operation. Further, this change does not constitute an unreviewed safety question in that incorporation of rod bow penalties into the TMI-1 Technical Specifications imposes additional, more conservative limits than those required by the original TMI-1 FSAR analyses, and the ECCS analyses requirements of 10 CFR 50.46.
The attached changes to technical specifications incorporate fuel rod bow penalties and, when combined with those requested in our request of February 11, 1976, provide safe operating limits for continued operation of TMI-1 at rated power of 2535 MWt during Cycle 2.
1479 232
a conservative margin to DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant syste= pressure has been considered in determining the core protection safety li=its. The difference in these two pressures is noninally 45 psi; however, only a 30 psi drop vas assumed in reducing the pressure trip set pointe to correspond to the elevated location where the pressure is actually measured.
The curve presented in Figure 2.1-1 represents the conditions at which a mini =u= DNBR of 1.3 is predicted for the maximu= possible ghernal power (112 percent) when the reactor coolant flow is 139.8 x 10+ lbs/h, which is less than the actual flow rate for four operating reactor coolant pumps. This curve is based on the following nuclear power peaking factors (2) with poten-tial fuel densification and fuel rod bowing effects; N N N F = 2.6T; F = 1 78; F = 1 50 q M z The 15 axial peaking factor associated with the cosine flux shape provides a lesser =argin to a DNBR of 13 than the 17 axial peaking tactor associated with a lover core flux distribution. For this reason the cosine flux shape
_- and the associated.F_ =-140 is more -limiting _ and thus the more conservative assu=ption.
. ~ -
The 150 cosine axlar flux shape in conjunction with FAH = 178' define the reference design peaking condition in the core for operation at the maxi =um overpower. Once the reference peaking condition and the associated ther=al-hydraulic situation has been established for the hot channel, then all other combinations of axial flux shapes and their accompanying radials must result in a condition which vill not violate the previously established design criteria on DNBR. The flux shapes examined include a vide range of positive and negative offset for steady state and transient conditions.
These design li=it power peaking factors are the most restrictive calculated at full power for the range frc= all control rods fully withdrawn to maxi =u=
allovable control rod insertion, and form the core DUER design basis.
The curves of Figure 2.1-2 are based on the more restrictive of two ther=al li=its a .d include the effects of potential fuel densification and fuel rod boving;
- a. The 13 DNER limit produced by a nuclear power peaking factor of FN = 2.67 of the combination of the radial peak, axial peak, and position of the axial peak that yields no less than a 13 DUER.
- b. The co=bination of raciel and axial peak that prevents central fuel telting at the hot spot. The limit is 19.6 kW/ft.
Fover peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalance produced by the power peaking.
The specified flow rates for curves 1, 2, and 3 of Figure 2.1-2 correspond to the expected =inimum flow rates with four pumps, three purps, and one pump in each loop, respectively.
1479 233 2-2
~ .
Thermal Power Level, f, UNACCEPTABLE OPERATION .-120 (112) (+35.112)
(-33,I12) *
- 110 OI ACCEPTA8LE 4 PUHP
- 800 OPERATION Kw/ft Limit . Kw/f t Limi t
(-56,90) *- 90 (86.7) (+58.90)
(.33,86.7) (+35,86.7) h __ 80 ACCEPTABLE 3 & 4 PUMP OPERATION
._ 70 ,
( .56.64.7) (+58,64.7
(-33,59.1) __ 60 (59,;) (+35.59.I) -' '-
, . . . .* -: ;-. c -' ' ," T ~. :-!
ACCEPTA8LE _.
- ~~ ~
- 50 2,3 3 4 PUNP - - -
(-46.47) OPERATION ( +47,48)
.- 40
.- 30
.. 20
.- 10 l l I I I t
-60 -40 -20 0 20 40 60 Reactor Power lobalance, 5 i.
CURVE REACTOR COOLANT FLOW (Ib/hr) .
6 I 139.8 x 10 6
2 104.5 x 10 3 65.8 x 10 UMIT l, CYCLE 2 CORE PROTECTION SAFETY LIMITS F i gu re 2.1 -2
Power Level, 7 UN ACCEPTABLE -- 120 OPERATION
(-20,108) .- 110 (108) (+21,108)
..100 +
g l +
N' e i
'/
ACCEPTABLE 90 iPp
// 4 PUNP .-
+N OPERATION l (80,7)
(-46,80) 80 l .. 70 ACCEPTABLE .
. . . . . _. 2.& 4-P. UMP ~
~~
I
'~
-OPERATION,_,_.-- _
E ,0
]~~ '
h- )
0'
(-46,52.7) -- 50 l l
.. 40 I I ACCEPTABLE
-- 30 2,3 & 4 PUMP ( +4 8',2 5. I )-
(-4s,2s.1) OPERATION 20 g 7 O N to
' E Y -- 10 +
11 11 11 l ~
-i
~
i i i i i t
i i
-50 -40 -30 -20 -10 0 10 20 30 40 50
Power imbalance, 7, 1479 235 UNIT I, CYCLE 2 PROTECTION SYSTEM HAXIHUM ALLOWABLE SET PolNTS Figure 2.3-2
" -~
F ; "D PDN EVA!J!ATION
'. ~
An evahation of the effects of fuel rod bowing onTMI-l cycle 2 operation h'as been performed. It is concluded from the results of this evaluation that modifications to the proposed Cycic 2 Technical Specification limits are required to maintain acceptabic fuci design limits with the increase in rcwer peaking associated with fuel rod bow. This report described the analyscs used to evaluate the effects of
, fuel rod bowing and the resulting modifications to the Cycle 2. Technical Specifica-tions.
IDCAL p0WER PEAKING ANALYSIS FOR FUEL PDD TN EFFECTS The tecnhiques used to analy e the local. power peaking effects of fuel rod bowing are similar to those currently used for fuel'densification analyses. Briefly .
these techniques employed the two-dim,nsional transport theory code Dar (discrete
! ordinates technique) to detenaine the clianges in local power in the center fuel rod as a function of the bowing (in both magnitude and amathal orientation) of the
[ surrcunding fuel rods. Analysis was perfomed to establish neutron group structure, quadrature order, mesh line spacing for rectangular rod representation, boundary i
i conditions and buffer zone characteristics surrounding the discrete rod array, and
! the validity of superposition techniques for subsequent Mante Carlo analysis. Results frca these studies support the use of 6 neutron grcups and S4 quadrature for the DUT calculations. The 5x5 fuel red array used 15 mesh lines per fuel cell and was surrcunded by a buffer region of homogeneous fuel. The cnalyses were perfomed for a range of peak magnitudes of deflection based on the predictions in table 1, i
! and for a=uthal orientations of the deflection in 45 increments.
- This model was used to detenaine the % power change per rod as a function of bow rngnitude and angle for each type of rod location in the 5x5 array. This data fonned the input to a B5W Monte Carl'o code. .
In the Monte Carlo analysis, a normal distribution was assumed for the amount of bow and a unifona distribution for the angle of bow. For cach axial intenal L
j47g 2}6
(distance between spacer' 9-), the Monte Carlo code comp-e- th r. total spike on
', ' the central rod 100,000 times to develop a good statistical sample. This data is used to deter =ine power spikes at each axial interval in a minner similar to densification power spike analysis. The power spikes presented in table 1 were calculated such that less than one rod in the core will exceed the given power spike at a 95% confidence level.
EFFECT OF ROD BOW ON DNBR Analysis was performed with the COBRA III-C code to determine the effect of a fuel rod bowing into the hot channel and reducing the flow area of that channel.
The results demonstrate that rod bow of the magnitude predicted is adequately co=pensated for by the flow area reduction factor. Rod bow away from the hot channel was also analyzed. In this analysis, the effect s of a power spike was added'
' to the hot rod in the area of the minimum DNBR. This analysis also demonstrated that the current TliI-1, Cycle 2 DNBR results conservatively accc"nt for the effects of fuel rod bowing.
J '!
TECHNICAL SPECIFICATION CIIANGES The imbalance limits for the reactor protection system are determined to preclude violation of the central fuel melt criteria (DNBR is less limiting in TMI-1, Cycle 2. To deter =ine these li=its, a correlation is developed which includes the following items:
A.
Power distribution effects (all in combination)
- 1. Fuel depletion
- 2. Power transients.
- 3. Full length control rod position (normal and abnormal)
- 4. Axial power shaping rod position (normal and abnormal)
- 5. Fuel dencification spikes B. Uncertainties (all in combination) }k7h 237
- 1. Hot channel faccors
- 2. Nuclear uncertainty
- 3. Spacer grid effect
- 4. Instrumentation (Observabil[tv, cicetronics and calibration).
. Including the power spike impact of rod bow in t he analysis results in the necessity to reduce the RPS inhalance limits (technical specifications 2.1 and
- 2. 3. )
The normal operations limits (technical specirication- 3.5.2) are determined to preclude violation of the ECCS linear heat rate criteria (See lapical Report BAW-10078 Rev.1, " Operational Parameters for B&W Rodded Plants", Proprietary, and BAW-10079, non proprietary), shutdown margin criteria, and ejected rod worth criteria. Evaluation of the impact of the rod bow power spikes on these limits indicates that the rod bow penalty (1.6%) is offset by the conservatism of the current limits. A densified average linear heat rate of 5.80 kw/f t was assumed in the analysis and a slightly more conservative power spike *han required *
(1.022 instead of 1.018 for instance, at the 30" elevation). The actual limiting average linear heat rate is 5.73 kw/ft at a power level of 2535 Mat. These conserv.tisms provide an excess margin slightly greater than the required 1.6%.
f 1479 238
, TABLE 1 FUEL ROD E0W POWER SPIKE PREDICTED FUEL ROD BOW
'AT OPERATI:G CONDITIONS Mid Span Elevation EOC 1 EOC 2 EOC 3 Top 1 5.55 6.45 7.53 2 6.66 '9.16 12.33 .
3 7.71 11.26 15.55 '
4 8.81 13.76 19.21 5 9.46 e 15.00 20.54 - ~
~6 9.23 14.01 19.'86 Bottem 7
,)
9.26 13.66 18.45 - 4
. POWER SPIRE RESULTING FROM PREDICTED FCEL ROD EOT ~
Ed Sp PERCENT POWER SPIRI Elevation EOC 1 EOC 2 EOC 3 Top 1 0.6 0.7 0.8
- 2 0.7 1.0 1.3 3 0.8 1.2 . 1.6 .
4~ 0.9 1.5 2.05 5 1.0 1.6 2.15 I 6 1.0 1.5 2.1 Bottes 7 1.0 1.5 1.7 :(
9 1479 239
.h' u h p>
f a. . . w -J__ h a
U'IITED STATES OF AMERICA ,
k g .jg g y N' m NUCLEAR REGULATCRY COMMISSION M y IN THE MATTER OF 4
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DCCKET NO. 50-289 LICE:ISE NO. DPR-50 FMCPOLITAN EDISON CCIGANY ggp; Qg {
This is to certify that a copy of Technical Specification Change Request No. 30, Amendment No. 2 to Appendix A of the Operating License for Three Mile Island Nuclear Staticn Unit 1, dated April 2,1976 and filed with the U. S. Nuclear Regulatory Cor:sission on April 2,1976, has this 2nd day of April been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania and Dauphin County, Pennsylvania by deposit in the United States mail, addressed as fo11cvs:
Mr. Weldon B. Arehart, Vice-Chairman Mr. Harry B. Reese, Jr.
Scard of Supervisors of Board of County Conmissioners Londenderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Court House Middletown, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 k m 0POLITAN EDISCN CC?GANY By Vice President-Generation 3SSI see 1479 240