B13067, Application for Amend to License DPR-65,revising Tech Specs Re Cycle 10 Reload Operation & Entry Into Mode 4 After Feb 1989 Refueling Outage

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Application for Amend to License DPR-65,revising Tech Specs Re Cycle 10 Reload Operation & Entry Into Mode 4 After Feb 1989 Refueling Outage
ML20206J220
Person / Time
Site: Millstone Dominion icon.png
Issue date: 11/15/1988
From: Mroczka E, Romberg W
NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:
NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20206J223 List:
References
B13067, TAC-68360, NUDOCS 8811280101
Download: ML20206J220 (16)
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P.O. BOX 270 H ARTFORD. CONN ECTICUT 06141-0270 k k J [.U. ((',7[',7C (203) 665-5000 November 15, 1988 Docket No. 50-336 B13067 Re: 10CFR50.90 U.S. Nuclear Regulatory Commission Attention: Document Control De', Washington, DC 20555 Gentlemen: Millstone Nuclear Power Station, Unit No. 2 Proposed License Amendment Change Cycle 10 Reload (TAC No. 68360) Pursuant to 10CFR50.90, Northeast Nucle:ar Energy Company (NNECO) hereby proposes to amend its Operating License OpR-65 by incorporating the changes identified in Attachment 1 into the Techt.tcal Specifications of Millstone Unit No. 2. In a letter dated June 3,1988,(1) a schedule was proposed whereby NNEC0 would submit supporting reports and analyses when received from Advanced Nuclear Fuels, Inc. (ANF). This effort would provide the NRC Staff an opportunity to plan, budget, NNEC0 provided, andinschedule support a letter dated for the reload 26,1988,(sgheduled August a preliminary descrip- for early 1989. tion of the supporting re,corts and analyses, and a description of the major identified at that time. By letter dated Technical August Specif30,1988,Igtion changesNNEC0 provided the NRC Staff, as additional i description of the ANF designed fuel, analytical basis, design criteria, drawings, results of the fuel design analysis, and operating experiences with - fuel having similar design characteristics. All other supporting (1) E. J. Mroczka letter to U.S. Nuclear Regulatory Comission, dated June 3, b 1988, "Millstone Nuclear Power Station, Unit No. 2, Cycle 10 Reload g License Amendment Schedule." -o "8 (2) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, dated $b August 26, 1988, "Millstone Nuclear Power Station, Unit No. 2, g Description of Cycle 10 Analysis Package." -o gg (3) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, dated a August 30, 1988, "Millstone Nuclear Power Station, Unit No. 2, fuel Q Design Report for Cycle 10 Reload." de 

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f 1 U.S. Nuclear Regulatory Commission B13067/Page 2 November 15, 1988 analyses (4)(5)(6)(7)(8)(9) needed to justify this license amendment change request have been docketed prior to this license amendment change request to allow the NRC Staff adequate review time to support Millstone Unit No. 2 Cycle 10 reload. INTRODUCTION Cycle 10 at Millstone Unit No. 2 will be the first cycle to use fuel designed and fabricated by ANF. ANF has also provided the plant's safety analysis. The safety analysis and the Technical Speci'ications for the plant will require numerous revisions due to the follot S changes:

1. The cycle length will be increased from a current typical length of about 350 full-power days to approximately 420 full-power days. This is accom-plished, in p.irt, by a change in the fuel management to a low leakage core. This change in the fuel management results in several Technical Specification changes.
2. The plant's safety analysis through Cycle 9 was performed by Westing-house. Beginning with Cycle 10, the analysis will be done by ANF. This results in a change in the methodologies used for the analysis. This (4) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, dated l September 1, 1988, "Millstone Nuclear Power Station, Unit No. 2, large l Breck LOCA/ECCS Analysis."

(5) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, dated September 19, 1988, "Millstone Nuclear Power Station, Unit No. 2, Disposition of Chapter 15 Events." (6) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, dated  ! October 7, 1988, "Millstone Nuclear Power Station, Unit No. 2, Small Break LOCA Analysis." (7) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, dated October 17,1988, "Millstone Nuclear Power Station, Unit No. 2, Cycle 10 Safety Analysis Report." (8) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, dated October 27,1988, "Millstone Nuclear Power Station Unit No. 2, Cycle 10 steam Line Break Analysis." (9) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, dated October 28,1988, "Millstone Nuclear Power Station, Unit No. 2, Cycle 10 Analysis of Chapter 15 Events, Supplement 1." l

f U.S. Nuclear Regulatory Commission B13067/Page 3 Novenber 15, 1988 . different methodology results in several changes to the Tec'nical Speci-fications.

SUMMARY

OF PLANNED CHANGES TO TECHNICAL SPECIFICATIONS l l The changes to the Technical Specifications are summarized below. They are described in more detail in the section entitled DETAILED DESCRIPTION OF CHANGES. l

1. Linear Heat Rate (Soecification 3.2.11 The maximum allowable linear heat rate (LHR) is being reduced from its current value of 15.6 kW/ft to 15.1 kW/ft. However, two of the currently required uncertainty factors included in the calculation of this value will no longer be required for ANF fuel. These two uncertainty factors are the flux peak augmentation factor (an axially varying correction) and the axial fual densification and thermal expansion uncertainty factor (a constant value of 1.01). The net effect of the reduction in the maximum LHR and the removal of the two uncertainty factors is a small increase in the actual allowable LHR.
2. Total Integrated Radial Peaking Factor, FT r (Specification 3.2.3)

This value is being increased from its current value of 1.537 to 1.61. A relateC change is to the allowable power versus F given in Technical Specification Figure 3-2-3b. These changes are ossible because the NRC-approved ANF methodology fcr calculating the departure from nuclear boiling ratio (DNBR) allows substantially increased margin when compared to the Westinghouse DNBR methodoloqy currently being applieu to Millstone Unit No. 2.

3. Total Planar Radial Peaking Fa: tor, Ffy (Specification 3.2.2)

This value is being removed from the plant's Technical Specifications.  ; 

                                                                                            ~

This deletion is possible because ANF's 3-D power distribution method-ology (Advanced Nuclear Fuels Corporation- "Setpoint Methodology for C.E. Reactors: Three Dirensional Axial Power Distribution Generation," ANF-507(P) Addendum 1, dated June 1988), does not require it.

4. Moderator Tamoerature Coefficient Soecification 3.1.1.4)

Both the most positive and most negative moderator tempr.'ture coeffi-cients (MTC) are being changed. The most positive allowaole MTC with power less than or equal to 70% will change from 0.5 x 10 8 AK/K/'F to 0.7 x 10 8 AK/K/'F. The most negative allowable HTC at rated thermal power will change from -2.4 x 10 4 AK/X/'F to -2.8 x 10 8 AK/K/'F. Both

1 . U.5. Nuclear Regulatory Commission B13067/Page 4 November 15, 1988 of these changes are necessitated by the increased cycle length. The most positive MTC for power levels greater than 70% power is not changed for Cycle 10.

5. Shutdown Maroin (Soecification 3.1.1.1)

Tu shutdown trugin required for Modes 1 through 4 is being changed from 2.9% AK/K to 3.6% AK/K. This change is necessitated by the impact of the more negative MTC on the steamline break transient. The increase in the shutdown margin can be accommodated because the change in the fuel management strategy (to a low leakage core) results in higher calculated rod worths and hence larger shutdown margins.

6. Low Steam Generator Trio Setooint (Soecification 2.2.1)

The current setpoint is 500 psia. This is being changed to 680 psia. This change is necessary to assure adequate peutection against the asymmetric power transient resulting from the closure of a single HSIV. DETAILED DESCRIPTION OF CHANGES , This section gives a detailed description of each of the proposed ::hanges and ! the basis for that change. l 1. Index l The unrodded planar radial peaking factor (Fvv) is being removed from the Technical Specifications. Therefore, refehhces to it are deleted on pages II, V, and XI of the index.

2. [Lefinitions The unrodded planar radial peaking facter (F ) is being removed from the Technical Specifications. The" N-- its efinition (Number 1.24) on page 1-5 is being deleted from .ai . w icn.
3. Reactor Trio Setpginti_{Soecifica h e.2.1)

The steam generator low pressure trip setpoint Oable 2.2-1, page 2 4, item 7) is being raised 180 psi from its current valv? of 500 psia to 680 psia. The allowable value is also being raised 180 psi from its current value of 492 psia to 672 psia. This change is necessary to assure adequate protection against the asymmetric secondary side tran-sient resulting from the closure of a single HSIV. Footnote 2 of Table 2.2-1 (see page 2 5) is also being changed to allow the steam generator low pressure trip to be manually bypassed up to 780 psia. This is also an increase of 180 psia from its current value. An additional restriction i; being placed on the bypass to assure that 

        ]i U.S. Nuclear Regulatory Commission B13067/Page 5 November 15, 1988 t
    ,            all the control element a semblies (CEAs) are interted for the trip to be manually byparsed.           This assures that the reactor is in a safe condition whenevar the trip is bypassed.
4. Limitina Safety System Settina (LSSS) Bases There are five changes to the bases. They are:
a. The DNB correlation used by Westinghouse is W-3. ANF uses the XNB ,

correlation. Therefore, the reference on page B 2-1 to the DNB correlation used is being changed. 

                 .) . The 95/95 DNBR limit has a different value for the ANF methodology                      ;

than for the Westinghouse methodology. The Westinghouse PNB corre- , lation has a 1.30 DNBR value at the 95/95 limit. The corresponding 95/95 value using the ANr methodology is 1.17. Therefore, the value of 1.30 given for the acceptance criteria on pages B 2-1, 3, 5, 6, and 8 is being changed to 1.17.

c. The steam generator operating pressure of 815 psia given on page B 2-5 is being deleted since it is no longer valid or neces-  ;

sary. A qualitative statement without a specific number is consid-eied acceptable.

d. The ANF analysis of the thermal margin / low pressure trip assumes a 72 psi allowance which consists of a 22 e i spressure measurement ,
error and a 50 psi time delay allowance. The Westinghouse analysis i as;umed a 67 psi allowance with a 22 psi measurement error and a 45 '

psi time delay allowance. The changes from the Westinghouse values to the ANF values are made on pace B 2-7.  ; i

e. The steam generator low pressure trip setpoin+ on page B 2-5 is being changed from the curreat value of 500 psia to the proposed r

, value of 680 psia. l

5. Shutdown Marain (Soecification 3/4.1.1.11  ;

1 The shutdown margin assumed for Modes 1 through 4 in tlie Cycle 10 analy-  ! sis is 3.60% AK/K. The corresponding Cycle 9 value is 2.90%. Therefore, ( 4 elue is changed in three locations on page 3/41-1. '

6. w ig_r_hw t  : .1 c'eg_C2 efficient (Soecification 3.1.1.41  ;
                  -                         n r flodes 1 and 2 in the Cycle 10 analysis is being                I c,',,        .

C :le 9 value. T!.c new most positive MTC value when the  ! p s. v or equal to 70% is 0.7 x 10 4 AK/K/'F. This corri- F 3 , 9 value of 0.5 x .10 4 AK/K/*F. The new most negative t ve%e at - <wer is -2.8 x 10 5 AK/K/'F. The corresponding Cycle 9 [ i

t h i e U.S. Nuclear Regulatory Commission B13067/Page 6 Novetber 15, 1988 value was -2.4 x 10 4 AK/K/'F. These two changes are made to page 3/4 1-5. e 7. Total Planar Radial Peaking Factor, Ffy (Specification 3/4.2.2) The ANF analysis methodology does not require this parameter to be monitored. Therefore, this entire section (pages 3/4 25 through 3/4 2 8) is being deleted frcm the Technical Specifications. The axial shape index (ASI) tents currently in this section are, however, also required for monitoring of the LHR. The LHR tent that is assumed in the ANF analysis (current figure 3.2-2a) will therefore be moved into the LHR specification. Note that several of these deleted pages are planned for rgion currentlyin our proposed Cycle 9 coastdown Technical Specification change under NRC Staff review.

8. Linear 4 eat Rate (Soecification 3/4.2.11

! The maximum LHR is being changed from 15.6 kW/f*, in the current Technical Specification to 15.1 kW/ft. This new value is that assumed by ANF in their LOCA and setpoint analyses. Also, the end of cycle coastdown restrictions are no longer required. The LHR value is currently given as a constant value on Figure 3.2-1 on page 3/4 2-3. This figure is being deleted and the constant value of 15.1 kW/ft is instead written into the text on pages 3/4 2 1 and 3/4 2 2. The definition of the LHR as the sum of the fuel, clad and moderator heat rates, which is currently on Fig-

ure 3.2-1, is also written into the text. Inclusion of the limit in the

! text and the deletion of the figure is being done for simplification and ! does not affect the interpretation of the Technical Specification. - The Technical Specifications that apply when the LHR is being monitored using the excore detectors are being changed. The current Technical SpecificationreferstotheASItents(currpntFigures3.2-2aand3.2-2b ! on pages 3/4 2-7 and 3/4 2-7a) i' the F specification for the LHR limits. There is only one tent applicable N Cyclo 10. This applicable I tent is the same as that currently given as Figure 3.2-2a. This figure is now being ircluded, as Figure 3.2-{, in the LHR Technical Specifica- ! tion. This is necessary because the F yy specification is being deleted, i I (10) E. J. Nroczka letter to U.S. Nuclear Regulatory Commission, dated September 20, 1983, "Millstone Nuclear Power Station, Unit No. 2, l Proposed Changes to Technical Specifications."

i .- I U. A. Nuclear Regulatory Commission B13067/Pagu 7 November 15, 1988 3 The specific changes to the Technical Specifications for.LHR monitoring . using the excore detectors are as follows:

a. Item a of Technical Specification 3.2.1 was required because, for certain F T values, the maximum allowable power was less than 100%

of the rand thermal power. This is no longer true with the pro-poted Technical Specifications. Therefore, Item a can be deleted.  ; I

b. Item b of Technical Specification 3.2.1 is combined with the intro-  :

ductory sentence. References to 100% of the allowable power limit i are deleted as discussed in item 1 above. The two separate condi- i tions following the "either" are separated as Items a and b. < References to Technical Soecification 3.2.2 are changed to reference  ! Figure 3.2-2 as discussed above.  ;

c. Reference to Technical Specification 3.t-2 in Technical Specifica-  !

tion 4.2.1.2.b (page 3/4 2-2) is being che med to refer to Fig-ure 3.2-2 as discussed above.  : i

c. Item c of Technical Specification 4.2.1.2 (page 3/' 2-2) is being deleted. It requires verification every 31 days of operation within the limits of Figure 3.2-2. However, this is monitored continuously -

by the power ratio recorder per Technical Spec fication 3.2.1. i Therefore, this requirement is redundant to a more restrictive , i requirement and can be deleted. . These changes to the excore monitoring of the LHR result in a substantial

simplification of the section. The Technicalc.pecifications for opera- '

tion using the excore detectors will now be similar in format to the incore Technical Specification. I There are also two penalty factors (see Technical Specification 4.2.1.3 on page 3/4 2-2) included in the calculation of the LHR using the incore detector monitorino system that are not applicable to the ANF fuel assemblies. These are (1) Item 4.2.1.3.b.1, the flux peaking augmenta-tion factor as shown in Figure 4.2-1 and (2) Item 4.2.1.3 b.4, the LHR uncertainty factor of 1.01 due to axial fuel densification and thermal expansion. The justification for the removal of dese factors is as follows:

a. The flux peaking augmentation factors are a result of tne fuel pellets interacting 'sith the cladding prior to the time of maximum
fuel densification. This can cause the pellets to "hang up," with gaps potentially opening up between the individu11 oellets. Local l power peaking iacreases can result in the area around the pellet 1

gap. The fuel design report shows that there can be no interaction between the pellet and the clad prior to the time of maximum pellet densification in the ANF fuel. Therefore, this penalty is not l applicable to the ANF fuel. 1 

             - . - - _ ~ - . . - _ . . _ . _ . . . _ . _ , . . _ _ _ . , _

    --- ---     )

U.S. Nuclear Regulatory Commission B13067/Page 8 November 15, 1988 

                              .           The LHR uncertainty factor due to axial densification and thermal expansion is included in the engineering uncertainty factor of 1.03 (see Item 4.2.1.3.b.3 on page 3/4 2-2) in the ANF methodology as described in the fuel design report. Therefore, it does not need to bo included again for the ANF fuel.

. The removal of these :enalties only applies to the ANF fuel. They should still be applied to tie Westinghouse and Combustion Engineering fuel wPen used in the core. Therefore, a footnote is applied to these two factors (Technical Specifications 4.2.1.3.b.1 and 4.2.1.3.b.4 on page 3/4 2-2) 1 specifying that they only apply to the non ANF (Batches "A" through "L") J fuel. This footnote is also added to Figure 4.2-1 (page 3/4 2-4) which j specifies the values of the peak augmentation factors. Note that several of there affected pages are planned for revision in our proposed Cycle 9 coastdown License Amendment change currently under NRC Staff review. ]

9. Total Integrated Radial Peaking Factor--Ff (Specification 3.2.3)

T The Cycle 10 analysos assume 1.61 for F . This compares to the Cycle 9 value of 1.537. Therefore, the value 5f 1.537, which occurs twice on i page 3/4 2-9, is changed to 1.61. T There exceeds is athe figure full which sower gives maximum thevalue. maximum Thisallowable is currently thermal power Fig;re 3.2-3bwhen F[ i which is in Sect < on 3.2.2. Since Section 3.2.2 is being deleted, the current figure is being replaced by a new Figure 3.2-3. This figure will

be tdded as a new page, 3/4 2 9a. The values in this new figure are l found in the Analysis of Chapter 15 Events (Supplement 1). The reference to the figure in Section 3.2.3.a of page 3/4 2-9 is being changed to
refer to this new figure, i
10. Azimuthal Power ' lilt--Tq (Specification 3.2.4) and the j The references current Technicalto the total planar Specification Section radial 3.2.2peaking factor are deleted f (Fh) thisrom l

section since they are being deleted from the Technical Specifications. l This change affects page 3/4 2-10. In addition, therp are two references to the total integrated radial . peaking factor (F ) on this page. The current Technical Specifications ! have them capitalized, implying that they are a defined term. In fact,

only F These two terms are different. In order ko,removf not F; ispotential any a defined tern.

confusion between them, references to the l 

                           "Total Integrated Radial Peaking Factor" are no longer all capitalized.

  't e 1

U.S. Nuclear Regulatory Cormission B13067/Page 9 November 15, 1988

11. Reactor Protective Instrumentation (Soecificatior 3.3.1) 2 There is a footnote on Table 3.3-1 (see Footnote B on page 3/4 3-4) which  !

allows the steam generator low-pressure trip to be bypassed under certain  ! conditions. This corresponds to the footnote used in the reactor trip i setpoirt specification (see Item 3 above)._ This footno'e is also changed as di. cussed in Item 3. The reasons for this change is a same as given in Item 3.  !

12. Incore Detectors (Soecification 3/4.3.3.2)  :
Th re are three references to the unrodded planar radial peaking factor .

(F v) in this section. These references are on pages 3/4 3-30 and 3/4  ; 3- 1. Since F T is being removed from the specifications, references to  ; it are inapprokiste and are being deleted. . I

13. Soecial Test Exceptions--Group H31aht and Insertion Limits (Soecifica- -

tion 3/4.10.2) l , There are two changes being made to this Technical Specification. Both changes are on page 3/4 10 2. These changes are discus.ed below,

a. Test excepti s are currently allowed from Technical Specifica- .

tion 3.2.2 (F y). However, this Technical Specification is being ' removed. The efore, the exception is no longer reqtired and is being deleted. Test exceptions are also currently allowed from Technical Specifica-b. tion 3.1.3.2. However, this Technical Specification no longer . exists. The Technical Specification dealt with the part length  ! control rods. Both the part length rods and the Technical Specifi-  ! cation were removed for Cycle 2. Therefore, references to it are  ; being deleted. l 1 1

14. Baset for Reactivity Control Syste.ns (Soecification 3/4.1) i i

The shutdown margin requirement for Modes 1 through 4 is being changed to 3.6% AK/K from its current value. This is consistent with the changes J described above. This change is made once on both oages B 3/41-1 and . i 8 3/4 1-2. Page B 3/41-A also rLfers to radial peaking factors. Since one of these  ! two factors is being deleted, the word ' factor" is no langer pluralized. i j 15. Bases for Power Distribution Limits (Specification 3/4.2) 

!                      The bases for the LHR is being changed in two ways.                                                                                                                                                                              Both changes are to    I l                       page B 3/4 2-1.

t, pr

l U.S. Nuclear Regulatt y Commission  : B13007/Page 10 November 15, 1988

a. When the LHR is monitored by the excore detectors, the allowable  !

radial power distribution is currently given by the total planar radial peaking factor of Technical Specification 3.2.2. This is being changed to the Total Integrated Radial Peakirg Factor of Specification 3.2.3. >

b. The end of cycle coastdown restrictions are bein3 removed. There-fore, the basis for this restriction is also being deleted.

The bases for F T p T and T on pages B 3/4 -1 and B 3/4 2.2 are being i changed to reNy, e r ail r%ferences to F yanj Technit:a1 Specifica-tion 3/4.2.2, where the F inits are given F is now used to monitor both the LHR (in place o}YF ) and the departu[e from nucleate boiling l (DNB) limits. The two sepa e functions of F r are therefore now con- ^ i nected with the word "and." Note that several of these deleted pages are planned for revision in our ' proposed Cycle 9 coastdown License Amendment change curtontly under NRC Staff review.  ;

16. Bases for Reactor Coolant System (Srecification 3/4.4.1) f The reference te the DNB limit of 1.30 on page B 3/4 41 is being changed I to 1.17. ,

SAFETY ASSESSMENT Chapt transients to 

 !                                ANF determine             has reviewed                                           all theneed which events      Standard to be Review reanalyzed                  Planfor (SRP)

Cyc le 10gg This evalua-l - tion has shown that all the nonradiological events currently in Chapter 14 of r ti.e FSAR require reanalysis. These events were reanalyzed by ANF. The  ! results were then compared to the current analysis contained in Chapter 14 of l 

,                                  the FSAR. The results of this comparison were then used to determine if there                                                                                                                                 :

was an increass in consequences of design basis accidents.  ; i l These proposed changes to the Technical Specification limits result in some [ changes in the sansequenct of the design basis accidents. It is difficult to directly compare the Cycle 9 analysis results (performed by Westinghouse) to  ! the Cycle 10 results (performed by ANF) because methodology differeaces between the two ver. dors can result in calculated consequence differences not I attributable to the Technical Specification change. Nonetheless, a reasonable comparison can be made, m - (11) E. J. Mroczka letter to U.S. Nuclear Reguiatory Commission, dated j September 19, 1988, "t.111 stone !bclear Power Station, Unit No. 2, Disposition of Chapter 15 Events."

4 U.S. Heclear Regulatory Commission + B1306)/Page 11 November 15, 1988 ,

1. The methodology to calculate the DNBR is different between the two vendors. Westinghouse uses a deterministic methodology, which assumes that all the uncertainties are biased in the worst ;rection. ANF also uses a deterministic methodology, but supplements it with a statistical combination of uncertainties when the limits are not met using the deterministic calculation. It is concluded that the need to rely on statistical methodologies to show acceptable results is indicative of a potential increase in consequences.  ;
2. The rod ejection accident show.i an increase ir, consequences due to the  :

increase in rod worth related to the increase in shutdown margin, a9d also due to the increase in the allowable radial peaking factor, F' r These two changes result in the following increases in consequences:

a. The deposited enthalpy for the Cycle 10 analysis is calculated to be 240.6 cal /gm. The corresponding Cycle 9 value is 174.7 cal /gm. 1he  :

Cycle 10 consequences remain below the acceptance criteria of L 280 cal /gm.  ;

b. The Cycle 10 analysis shows that up to 11.5% of the core wil1 exceed the DNBR limit and is assumed to experience fuel failures. The previous radiological consequences analysis assumes 10% fuel fail-ure. This increase from 10% to 11.5% in fuel failures assumed for  !

radiological consequences will increase the off-site doses by approximately 15%. The radiological consequences, even with this i ,! increase in calculated off-site doses, are still well within the acceptance criteria. l

3. The consequences of the rod withdrawal from subcritical accidents are i increased for the same reason that the rod ejection accident consequences i are increased. The change in the peak fuel temperature is estimated to i be in excess of 200'F. The consequences, however, are still calculated to be within the acceptance criteria for fuel centerline melt.

< i

4. The large break LOCA analysis results show an increase in the peak l i cladding temperature of 21'F. However, the proposed reduction in the LHR j

. limit would be expected to result in a decrease in the peak cladding - temperature. The calculated increase therefore must be due to method- ! ology differences. As such, we consider this as a change in the evalua- i tion model as defined in 10CFR50, Appendix K, item II.l.b. This proposed Technical Specification change will increase the consequences of some of the design basis transients as discussed. However, in all cases, the . changes result in a calculation of acceptable consequences. Therefore, even j though there is an increase in consequences, there is still no impact ca the protective boundaries. 4 4 i 

                      .-m--   - _ c. ,_ _ . . . _ . . . . _ _ _                   , , _ _ . , _ . . . _ _ .     . , ,     _ _ . _ _ , . ,

U.S. Nuclear Regulatory Commission B13067/Page 12 , November 15, 1988. SIGNIFICANT HAZARDS CONSIDERATION NNECO has reviewed the proposed amenduent in accordance with 10CFR50.92 and  ! has concluded that it does not involve a significant hazards consideration. 

                                                                                                                                                                           ~

Based upon the information developed as part of the SAFETY ASSESSMENT, and summarized above, NNECO has concluded that these changes would not. 1, Involve a significant increase in the probability or consequences of an  : accident previously analyzed. ' As discussed above, ANF reviewed all SRP Chapter 15 accidents and tran-sients to determine which events need to be reanalyzed for Cycle 10, l assuming a mixed core or a core containing only ANF fuel. As a result, ' ANF reanalyzed all of the nonradiological events currently in Chapter 14 of the Millstone Unit No. 2 FSAR. On the basis of this review, NNECO concludes that there is no significant increase in the probability or consequences of any of these events. I With respect to calculated con:equences, ANF spe:ifically reanalyzed the impact of the events on relevant key parameters associated with the plant response to the event (i.e., assessments of consequences were not restricted to dose assessments). The parameters analyzed all relate to the boundary performance during the accident. The analysis showed that there are some instances in which consequences do increase, at least in terms of the limiting value of the relevant plant parameter, in all . cases, however, the values of the relevant parameters remain below I applicable acceptance criteria and there are no impacts on the protective ' boundaries. NNECO therefore concludes that the propused amendment does  ; not involve a significant increase in consequences of any event previ-outly cnalyzed. Specifically, both l arge- and small-break LOCA safety analyses were . performed to support the proposed amendment. The results of the large-break ECCS analysis indicated that the limiting break size is the 0.6 DECLG breek. The peak cladding temperature (PCT) for this limiting case was calculated to he 2163*F. The PCT for a 12'F reduction in primary coolant temperature was calculated to be 2176'F. These PCT k values remain within the ECCS acceptance criteria of 10CFR50, Appendix K. The small-break LOCA analysis indicated the limiting break size, with . symmetric steam generator tube plugging, is the 1.9% break. The PCT for this case was calculated to be 1811'F with a maximum local cladding oxidation of 4.17%. The results for 3ymmetric steam generator tube , plugging at the limiting break size are simila. to the results for i symmetric tube ) lugging, with the PCT being slightly higher for the j symmetric tube plugging case. Again, these values remain well within the l Appendix K ECCS acceptance criteria, l i

U.S. Nuclear Regulatory Commission B13067/Page 13 November 15, 1988 The steamline break analysis involved a calculation of the expecttd asymmetric thermal hydraulic and neutronic core characteristics resulting from this accident. Specifically, fuel response was evaluated against fuel failure criteria for four scenarios, assuming both availability of off-site power and loss of off-site pwer. The hot zero power (HZP) scenario with loss of off-site power was determined to be the most limiting in this analysis from an MONBR standpoint. In ne scenario evaluated, however, does fuel failure occur *.s a result of penetration of the HDNBR safety limit. An HZP sc.nario with off-site power available was determined to be the most limiting in the analysis from the stand-point of centnrline melt. However, again this case does not represent a significant increase in consequences as there remains margin to the fuel centerline melt limit (maximum LHGR). Finally, a non-LOCA transient event analysis was also performed in support of Millstone Unit No. 2 operation with ANF reload fuel, and a disposition of events for Cycle 10 provided in 'he SAFETY 6 ASSESSMENT. All anticipated operational securrences were shown to result in na signif-icant increase in either DNB, fuel centerline melt, deposited enthalpy, or radiological consequences. The postulated accidents were also shown to meet all appropriate acceptance criteria. With respect to the fuel centerline melt, deposited enthalpy and radiological consequences, the rod ejection and rod withdrawal from subcritical accidents show increases in consequences due to the increase in rod worth related to the tecrease in shutdown margin and also due to tne increase in the allowable radial peaking factor. However, as discussed in the summary of the SAFETY ASSESSMENT above, these increases are not significant in that they Jo not challenge the acceptance criteria for deposited enthalpy, fuel centerline melt, or off-site doses due to fuel failure. With respect to DNB, the increases are not significant in that they do not violate the 95/95 acceptance criteria. In total, therefore, NNECO concludes that the proposed change does not involve a significant increase in consequences of any accident previously analyzed. With respect to probability of an accident previously analyzed, there is no change in the probability of occurrence of any design basis accident. Further, there are no changes or failure modes associated with the proposed amendment that will increase the probability of an accident to the point where it should be considered within the design basis. There-fore, ir this respect, no significant hazard consideration is involved.

2. Create the possibility of a new or different kind of accident from any previously ai,alyzed.

As a result of the proposed use of ANF fuel and the proposed amendment, there will be no change to plant response. The plant will respond for all events in a manner similar to that previously analy2ed. As discussed in the SAFETY ASSESSMENT and above, the only changes identified in the reanalysis of Chapter 14 events relate to the impact of certain tran-

U.S. Nuclear Regulatory Commission B13067/Page 14 November 15, 1988 sients on parameters related to boundary perft. mance. There are no changes to the basic trends the transients follow. Thus, there are no failurr modes asseciated with the proposed change that could represent a new unanalyzed accident. In addition, there is no change in the probability of occurrence of any design basis accident. There are also no changes or new failure modes , associated with the changes that will increase the probability of an accident or transient to the point where it should be considered to be within the design basis. Therefore, NNECO concludes that the proposed

changes do not create any new or differer.t kind of accident from those previously analyzed.
3. Invo*ve a significant reduction in any margin of safety.

4 As discussed above, the accident reanalysis performed to support the proposed change included all nonradioiogical events currently in Chapter 14 of the Vill , tone Unit No. 2 FSAR. The analysis specifically

focused on the impNt of these accidents and transients on key parameters

! related to protective boundary performance. On the basis of this review, NNECO is able to conclude that the proposed amendment does not involve a i significant reduction in any margin of safety.

Specifically, as discussed above, the proposed changes do involve some
nonsignificant changes in consequences. This is reflected in the change j in some parametcrs relative to Technical Specification bases. However, i these changes are largely due to (a) methodo.ogy differences between the accident analyses performed by the previous vendor, Westinghouse, and that of the new vendor, ANF; (b) changes to or deletian of an LCO or
LSSS, resulting in e corresponding change in Technical Specification
bases; and (c) minor clarificctions. However, in all cases where there are increases in the limiting value of a parameter, the value remains below the applicable safety limit and thereforo does not affect the ability of the boundary to perform its function, i The changes in boundary parameters, discussed above, may be summarized as p follows

o The limiting large break LOCA resulted in a PCT of 2163'F. The PCT for a 12'F reduction in primary coolant temperature result-ed in a PCT of 2176'F. These values do not involve a signif-4 icant reduction in margin for any acceptance criterion.

o The limiting sma'.1 break LOCA resulted in a PCT of 1811'F, with symmetric steam generator tube plugging. The PCT with asymmet-ric plugging was slightly lower. These values do not involve a i reduction in any acceptance limit.

l

a 

,                      U.S. Nuclear Regulatory Commission B13067/Page 15 Ncvember 15, 1988 l                                  0     Fuel response was evaluated for the steam line break events.

In no scenario evaluated was fuel failure calculateo to occur as a result of exceeding the MONBR acceptance limit or the maximum LHGR, o The rod ejection and rod withdrawal from suberitical accidents result in increases in relevant parameters due to the proposed increase in rod worth and the increase in the allowable radial peaking factor. However, as discussed above, there is no reduction in targin of safety because the parameters do not i exceed existing acceptance criteria for deposited enthalpy, fuel centerline melt, or for off-site doses due to fuel fail-ure. 

.                                 o     Accidents with DNB criteria may show consequence increases due J

to the need to rely on statistical :nethodologies. However, there is no reduction in the margin of safety because the 95/95 acceptance criteria is met in all cases.

in summary, the proposed amendment does not involve any significant
;                            reduction in a margin of safety and, therefore, does not involve a significant hazard consideration, i                       The Commission has also provided guidance concerning the application of the i                        standards in 10CFR50.92 by providing examples (51 F.R. 7751, March 6,1986).

! The proposed changes described herein, in some respects resemble example (vi). This example involves a change which either may result in some increase to the probability or consequences of a previously analyzed accident or may reduce in - some way a safety margin, but where the results of the change are c1carly within all acceptable criteria with respect to the system or component as specified in the Standard Review Plan (e.g., a change resulting from the application of a small refinement to a previuusly used calculational model or ! design method). As with this example, NNECO's proposed i ondment should be determined to involve no significant hazards consideratio. 4 

                       'he Millstone Unit No. 2 liuclear Review Board has reviewed and approved the i                       attached proposed revisions and has concurred with the above determination!..

The proposed changes need to be approved to support Cycle 10 operation and prior to cntry into Mode 4 after the February 1989 refueling outage. NNECO requests that those proposed changes be .pproved and effective by March 3, 1988. This would allow specific applicable Technical Specifications to be in j place prior to any affected mode change. In accordance with 10CFR50.91(b), we are providing the State of Connecticut with a copy of this proposed amendment. i 

     - _ _ - . -___ _              ,m

U.S. Nuclear Regulatory Conmission . B13067/Page 16 November 15, 1988 

                     -Pu;*suant to the requirements of 10CFR170.1P(c), enclosed with this amendment request is the application fee of $150.                                                ;

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY E J~ m a<tka ^ E. J. Mroczka 4 Senior Vice President [/ By: W. D. Romberg-Vice President Attachment cc: Kevin McCarthy , Director, Radiation Control Unit  ; Department of Ent'ronmu.tal Protection , Hartford, LT 0611o  !

W. T. Russell, Region I Administrator ,

4 D. H. Jaffe, hRC Project Manager, Millstone Unit No. 2 -

W. J. Raymond, Seniac Resident Inspector, Millstone Unit Nos. 1, 2, a.1d 3 STATE OF CONNECTICUT ss, Berlin COUNTY OF HARTFORD l t

Then personally appeared before re W. D. Romberg, who being duly sworn, did i state that he is Vice President of Northeast Nuclear Energy Company, a

Licensee herein, that he is authorized to execute and file the foregoing information in' the name and on behalf of the Licensee herein, and that the ,

statements contained in said information are true and cor ec to the best of his knowledge and belief. C ko,2b Y Pedi i l M ry P lic ~ j w - - a Lviss M:rch 31, 1939 [ l l 

   . o Docket No. 50 *)1 213067 Attachment Hillstone Nuclear Power Station, Unit No. 2 Proposed Changes to Technical Specifications Cycle 10 Reload (TAC No. 68360)

November 1988 r - - p,}}

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