ML20058H319
| ML20058H319 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/01/1990 |
| From: | Mroczka E NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20058H326 | List: |
| References | |
| B13627, NUDOCS 9011150172 | |
| Download: ML20058H319 (24) | |
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, General Offices iSeide'n Street. Berlin, Connecticut : .
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HARTFORD. CONNECTICUT 06141 o270 L L U C'N' M [7dZZ *
(203) 665-5000--
November.1,'1990--
Docket No.-50-423-B13627 .
Re: 10CFR50.90 U.S. Nuclear Regulatory. Commission Attention: Document Control Desk Washington,.DC 20555: j .
Gentlemen:
Millstone Nuclear' Power Station,, Unit ~No. 3 Proposed Change to-Technical: Specifications. '
Cycle 4 Reload Submittal Pursuant to 10CFR50.90, Northeast ' Nuclear Energy Company (NNECO).hereby; proposes to amend Operating License NPF-49 ; by1 incorporating-' the t attached proposed changes into the : Tech %al Specifications of: Millstone Unitn No,' 3.
These proposed changes revise numerous? Technical Specifications in support of y refueling nd operation; for Millstone Unit No. 3- with- thel VANTAGE 5. Hybrid !
(SH) improved fuel design.- " -
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The proposed Technical Specification changes; for Millstone Unit No. 3 Cycle 4 primarily result from changes in three areas:- (1) change in; fuel design,= (2)- -
use of improved . analytical methodologies,1 and L(3); associated. fuel-/ core-related changes:
- 1. Chanae in Fuel Desian ,
Millstone Unit No. 3 is currently operating in Cycle 3!wita Westing i house 17 x 17 standard (STD). fueled core. Millstone Unit a. 3 Cycle 4-and subseouent cort loadings will have fuel? assemblies: that incorporate -
the low-pressure drop zircaloy grid and the intermediate; flow mixer (IFM) grid. This upgraded featgrq is known as VANTAGE 5H with.IFM and has been .
submitted as a -
WCA!'-10444-P-A.S2pddendum .The . VANTAGE , - to sthe " VANTAGE SH tReport has- 5 Reference Core Report," '
received : generic i j (1) " VANTAGE 5H Fuel Assembly," -WCAP-10444-P-A, Addendum 2, , April 1988 and '
letter from W. Johnson (Westinghouse): to M. W. Hodges:(NRC),' Supplemental Information for WCAP-10444-P-A, . Addendum 2, ", VANTAGE 5H Fuel Assembly,"
dated July 29, 1988.
-(2) " VANTAGE 5 Fue' t.:sembly Reference' tore Report , -WCAP-10444-P-A, September 1985 9011150172 903it,, .
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U.S. Nuclear Regulatory Commission B13627/Page 2 November 1, 1990 approval.(3) The VANTAGE SH fuel assembly design evolved from the current VANTAGE 5, optimized fuel Assembly (0FA) and Standard Fuel Assem-bly designs. In addition to the above-mentioned VANTAGE SH design features, Millstone Unit No. 3 Cycle 4 reload will also contain several VANTAGE 5 design features and other upgraded fuel design features used in the Cycle 3 core. Millstone Unit No. 3 Tycle 3, Region 5 fuel has already incorporated the VANTAGE 5 integral fuel burnable absorbers (IFBAs), the VANTAGE 5 axial blanket design, the VANTAGE 5 extended burnup design, the VANTAGE 5 reconstitutable top nozzle (RTN) design feature, as well as debris filter bottom nozzles (DFBNs), snag-resistant grids, and standardized fuel pellets.
By s,Per NRC|letter dated December 30, 1983]], Westinghouse requested NRC to review the topical report WCAP-10444, " Westinghouse Reference Core Report, VANTAGE 5 Fuel Assembly." The NRC Staff reviewed WCAP-10AM er.d con-cluded that the Westinghouse topical report was an acceptable reference to support plant-specific application of VANTAGE 5 provided certain conditions were addressed. The conditions are addressed for the Millstone Unit No. 3 application requesting the use of VANTAGE SH fuel assemblies. Attachment 3 provides responses to those conditions.
- 2. hg of Imoroved Analvtical Methodoloaies The existing thermal-hydraulic analysis of the 17 x 17 STD fuel used in the Millstone Unit No. 3 core is based on the standard thermal and hydraulic methods and the W-3 (R-Grid) DNB correlation as described in the Millstone Unit No. 3 Final Safety Analysis Report (FSAR). The DNB analysis of the core containing both 17 x 17 STD and VANTAGE SH fuel assemblies ha correlations (vg genthe modified revisedto incorporate Thermal Designthe WRB-1 (RTDP),
Procedure and WRg DNB and (3) A. C. Thadani (NRC) letter to R. A. Wiesemann (Westinghouse), " Acceptance for Referencing of Topical Report WCAP-10444-P-A, Addendum 2, VANTAGE SH Fuel Assembly," November 1, 1988, and Clarifications on the Safety Evaluation of the Topical Report WCAP-10444-P-A Addendum 2, January 5, 1989.
(4) VANTAGE 5 Fuel Assembly Reference Core Report, WCAP-10444-P-A, September 1985.
(5) New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles With Mixing Vane Grids, WCAP-8762-P-A, July 1984.
l (6) Revised Thermal Design Procedure, WCAP-ll397-P-A, April 1989. i
. a U.S. Nuclear Regulatory Commission:
813627/Page3 November 1, 1990 improved THINC IV Moaeling.I7I . The new DNB correlaiions' take credit for the significant improvement in Lthe'. accuracy of the. critical heat flux .
predictions over previous DNB correlations. The W-3 correlation: and the. ,
]i standard methods continue to~ be' used when conditions. are outside- the limit of the WRB-1 or WRB-2 DNB correlation and of.the RTDP.- As a result-of the new DNB- correlations' ; improved accuracy, confidence 1at .a-' 95/95i level that the- limiting power rod will not experience- DNB is provided- 1 with a limiting DNBR value of-1.17_ versus the existing 1.30; i
- 3. Associated Fuel-/ Core-Related Chanaes
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Numerous. changes -are .being made whichtare closely related sto the change j
in fuel type. Three of these changes represent physical changes ' to ;be i made at the plant. These three changes are:-
- a. Thimble plug deletion--these . devices. will be ' removed' over Lseveraic !
cycles.
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- b. Rod cluster control assembly -(RCCA L parked ; position 1 change--the full-out position of the RCCA' bank)s will- vary tin nthe range- of- 1 222-231 steps withdrawn. T
the boron concentration maintained in the RWST will increase from the range 2300-2600 ppm to 2700-2900 ppm. '
{
One of these changes involves a. change in the plant procedures: i Relaxed axial _ offset control' (RA00)--the - plant willL utilize this -!
method of control which allows a wider axial offset: band. compared to 1 constant axial offset control which is currently. being' used.
The remainder of 'the changes. in. this category are modifications; to? the i safety analysis procedures or inputs:- ;
- a. Increase.in FAH and F' l
- b. Reactivitycoefficien9 changes.
- c. RTDP.
d.
Revised non-RTDP uncertainties. !
- e. Reduced shutdown margin..-
- f. Modified OTAT and OPAT trip setpoints.
A description of the proposed Technical Specification changes related to$ the VANTAGE SH fuel upgrade is provided in Attachment 1. InJaddition, two.addi- _
tional changes to the Millstone. Unit No. 3 Technical- Specifications 1 described l l
1 (7) Improved THINC IV.Hodeling for PWR Design, WCAP-876_2, April -10, 1978. ;
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U.S. Nuclear Regulatory Commission ,
B13627/Page 4 November 1, 1990 in the attachment are not directly related to the Cycle 4.rel'oad.. The firsti set of changes are administrative in naturei Since they do not1 reduce the effectiveness of the existing or proposed Technical Specifications, they dot -
not involve a significant hazaro., consideration. The other set of changes are ' '
due to cable insulation. resistance effects on:thet engineered safety features 3 (ESF) instrumentation. (Table 3.3-4 of Technical . Specifications). .The'- H revised pages of the Technical Specifications are provided in Attachment 2., j The Plant Safety Evaluation (8) (PSE) provides- the safety evaluation for the region by region reload transition from the Millstone Unit-No. 3 Cycle:3 core with 17 x 17 STD fuel assemblies to a core containing;the- VANTAGE SH' upgraded features described above. This safety evaluation . includes the , mechanical,- ;
nuclear, thermal and hydraulic; and accident. evaluations. -This'avaluation'was- ;
accomplished utilizing the methodology described in WCAP-9273-A, " Westinghouse :
Reload Safety Evaluation Methodology." !
As requested by the NRC Project Manager l for Millstone ' bit: No; 3,- two copies l of the PSE are being forwarded directly to him. Using .the RTDPyrequires- a j review of temperature, pressure, power, and flow uncertainties;used in the safety evaluations. For Millstone Unit' No. 3, the uncertainties are calcu -
j q
lated based on the installed plant instrumentation 'or -special' test equip]tand ;l Mills has been prepared to describe the uncertainty evaluation. ;Since.this' report >
contains information proprietary to Westinghouse Electric Corporation,31t-will be submitted under a separate cover.
SAFETY ASSESSMENT
[
The PSE prepared by Westinghouse for Millstone Unit: No. 3 oincludes'. accident' evaluations. Specifically, Westinghouse - reviewed all the : FSAR . Chapter ,15 transients to determine which events need to be reanalyzed for Cycle 4. For each of the potentially impacted non Loss-of-Coolant AccidentJ(non-LOCA)' tran- <
sients, consideration was given to effects :of the VANTAGE 5H" fuel! design ,
features and modified safety analysis assumptions discussed in Section-5.1 of the PSE. As dictated by event-specific sensitivities, a decision was made' for ;
each transient with -regard to the need for= formal analysis, asfopposed? to- ,
simply evaluating the impact of the- subject features Landiassumptions. An additional issue that requires consideration forEthe Millstone Unit 'Not 3 ;
j analyses is the possible difference in relative' behavior 'of- the three loop _
cases as compared to those with four loops' operating. For certain events, -
previous licensing basis analyses ; clearly demonstrate that four-loop results' (8) Plant Safety Evaluation for Millstone. Generating: Station,- , Unit 3, VANTAGE SH Fuel Upgrade, August 1990.
(9) WCAP-12621, " Westinghouse Revised Thermal Design Procedure Instrument Uncertainty Methodology for Millstone 3," August -1990.
U.S. Nuclear Regulatory Commission -
B13627/Page 5 November 1, 1990 a
bound those for three loops operating.. In those cases, an' evaluation of--
three-loop operation; was performed' for the -VANTAGE - 5H transition, even11f -
analysis was determined to 'be~ necessary for! the ' associated four-loop case.
lable 5.1.2-1 of the PSE documents for each of the ' potentially. impactedt non LOCA transients whether an analysis was performed' or - an. evaluation wasi sufficient to assess .the _ impact of the VANTAGEiSH- transition.- In ad11 tion,: a Section 5.2 ofl the PSE describes the :large break and small-break Loss-of- 1 Coolant Accident (LOCA) analysis.- The' boron dilution analysis . is beingt finalized at this time. The Technical Specification changes .related to' this - 1 j
analysis and the results .of the analysis wille be provi.ded-in a future submit- _!
tal. The proposed changes to Technical S)ecifications resultLin'some changes , i in the consequences ' of the design bas's accidents. . It is Edifficulte to- 1 directly compare- the results of the previous analyses to the1results ofs the- j revised analyses, because of changes 1in methodology.: For three y non LOCA i accidents (locked rotor transient for four-loop.'and ,three-loop,' and: inadver-i 1 tent ECCS Operation), the consequencesL have somewhat1 increased; however they3 l are within the applicable acceptance criteria, o Large-break LOCA analyses were performed for a complete = spectrum' of breaks for- '
VANTAGE SH fuel and the limiting case'with standard fuel. Thesecanalyses were-performed for both four-loop-and three-loop; operation _ JJustification for- use J of approved LOCA methodology for the four-loop case _in three-loop operation is given in paragraph 5.2.2 of the:PSE. Peak clad temperature '(PCT);inlallLcases j analyzed remains below the 2200*F acceptance limit with the worst' case;being '
2133'F versus 2132*F in the previous analyses.
1 Small-break LOCA. analyses were performed for four-loop operationo ' nly, as the 4 three-loop ' operation case is bounded bythe four-loop = case.- .PCTJ for Lthe small-break LOCA analysis with- four loops in. operation: was_.1890'F versus 1 1483'F in the previous analysis. The consequences, however, ;are still' calcu-lated to be within the acceptance criteria for'PCTs.
There are two aspects of these changes that have a potential: effect on the radiological consequences of analyzed accidents.
The first is a chan in the calculated amount of fuel that ex)eriences; cladding -failure as gea- result of a locked rotor accident during. t1ree-loop: ,
0)eration. The value for Cycle 3 was 1 -percent, whereas -for- the proposed i c1ange, the value is estimated to be 3.3 percent.' The current design; basis accident assumed 4.4 percent cladding failure. Hence, the~ proposed. change is- <
bounded by the current calculation and there are no increased' consequences' .
The second aspr.:t is a change in the radial. peaking. factor. For.theLVANTAGE :
SH fuel _upgradt., the new peaking factor of_ l.7.is assumed-in the PSE. 'This is- (
different from the current - peaking factor of?l.65. The' only. accidentT for which a peaking factor is used to calculate radiological consequences is; the fuel-handling accident. Hence, this accident was reanalyzed. ;
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U.S. Nuclear Regulatory Commission B13627/Page 6 November 1, 1990 The TACTIll computer code was used for this evaluation. Due to the differ-ences in the codes used by -Northeast Utilities and Stone and Webster,- the fuel-handling accident was also evaluated using the peaking factor of 1.65 in-order to determine the effect of'the change, The following assumptions were used in this calculation:
- 1. Activity released is one assembly plus ' 50 rods of . another - (314 . total rods).
- 2. All of the -gap activity in the damaged rods is relea' sed and consists fof 10 percent noble gases (30 percent Kr 85) and 10 percent' iodines.
- 3. Iodine. species: 75 percent elemental, 25 percenttorganic.
3 Pool decontamination factor: 100,
- 5. Exhaust is via the ventilation vent on the Turbine Building.
- 6. EAB X/Q: 4.3(-4)sec/m3,
- 7. Minimum time after shutdown for transfer of fuel is 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.
- 8. Fuel building filter efficiencies:- 95 percent.
A fuel-handling accident in containment is -not . analyzed -because:it would be less severe than the fuel accident:in the fuel building due to:the isolation capability of the containment. Ta'ule 1 shows: the results of the doses due to a fuel-handling accident with the old peaking factor l of 1.65 and the new peaking factor of 1.7. The increase in the dose is only 3 percent. The' doses-are well within the Standard Review Plan acceptance criteria of 75 REM to the thyroid and 6 REM to the whole body. The new doses are alsoLat,or belows the doses that the NRC reviewed from the previous Stone and Webster calculation included in the existing FSAR.
In summary, the proposed changes will: somewhat' increase; the consequences of some of the design basis transients.- However, in all . cases, the ; changes result in a calculation of acceptable consequences.. Therefore,' even' though there is-an increase in consequences, there;is still>no impact on the protec-tive boundaries.
In a letter dated February -26, 1990, ;NNECO 'submittedL a~' proposed amendment request that would allow NNECO--to operate Hillstone-Unit No. 3'with a maximum-containment - pressure of 14.0 psia- during Modes.1 ' through' :4. ..The. safety analysis for the proposed change was performed ati maximum allowable contain-ment pressure of :14.2 psia.- LAs a part of the< VANTAGE:5H fuel upgrade- for Millstone Unit No. 3, Cycle 4, .an evaluation was performed for the effects of increasingthe containment pressure to 14.2 psia. - The . evaluation concluded that the . increase in the containment pressure will have no . adverse impact on the mass and energy release data reported in.the-Millstone Unit No 13 FSAR.
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1-U.S. Nuclear Regulatory Commission B13627/Page 7 November 1, 1990 SIGNIFICANT HAZARDS CONSIDERATION '
-l NNECO has reviewed the proposed changes in accordance with"10CFR50.'92(c) and-- 1 has concluded that the changes do not involve a significant hazards'considera- 'l tion. The basis. for -this conclusion is that the three criteria! off 10CFR50.92(c) are not' compromised. The ' proposed: changes do 'not involve a significant hazards consideration because the changes would not: ,
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- 1. Involve a significant increase in _the probabilityLor_ consequences off an- 1 accident previously analyzed. .;
. . - q To determine any potential impact, the proposed changes can; be grouped. i into three general categories. These are:-
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- a. Changes to Technical Specifications (Cycle l4- reload) due- toi change:
in fuel design'and use of improved analyticalf methodologies ~. ;1
- b. Changes that are not related to the Cycle 4 reload and are adminis-trative in nature such as redrawing :the existing figures in -Techni-_ l cal Specifications and renumbering of1 pages' from 6 the - existing ,!
Technical Specifications to allow removal of' " intentionally blank" .
pages from the _ Technical Specifications. '
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- c. Changes due to cable insulation resistance: effects on ESF w *trumen-tation.
Each of these groups of changes is discussed in more det' ail below.
)
- a. As discussed above, Westinghouse reviewed. all FSARL Chapter 15 l accidents and transients to - determine which eventso needed to be- -
reanalyzed for Cycle 4,- assuming a mixed core or a core containing:
only VANTAGE SH fuel. The-PSE documents;for each of the potentially ;
impacted transients or accidents whether-an- analysis was performed a or an evaluation was sufficient to assess the impact of the VANTAGE 5H transition. The. evaluation of the impact -of the proposed changes. j!
is discussed in Section:5.9.3 of the PSE.. The evaluation addressed 1
a full core of VANTAGE SH as well as , transition cores consisting'of ,
VANTAGE 5H and standard fuel. Four-loop ' and three-loopl operation' !
also was addressed. No increase-in the probability of-occurrence:of any accident was- identified, but ' extensive reanalysis, as- described -
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in the PSE was required-to demonstrate; compliance with the proposed 4 changes to the Technical Specifications.. These reanalyses. applied.
methods which have been. previously found acceptable by'the NRC._ The-results, which includes ' transition core effects, show' changes in _
consequences of accidents previously analyzed. lHowever, .as stated !
in SAFETY ASSESSMENT above, the increase in consequences are not significant and the results- are clearly within pertinent acceptance- s criteria, q
U.S. Nuclear Regulatory Commission B13627/Page 8 November 1, 1990 The increase in RWST boron concentration has been _ evaluated' with: I respect - to its effect on components in .a postaccidentienvironment.-
The corrosion of aluminum, zinc, and stainless" steel components are' '
affected- by ' the - pH = of their environment;- The volume and , NaOH concentration of the Chemical ~ Addition: Tank have been modified to' neutralize the effect of the higher boric acid concentration coming:
from the RWST.; 1There is thus no effect .on' the- performances of components containing these materials. '
b.- The renumbering of pagesu or ~ redrawing of- existing ifigures t for clarity purposes do ~ not reduce the effectiveness - of the Technical-Specifications-. Also, these changes do not affect the existing: or ?
proposed limiting conditions. for operation or surveillance' require-ments. Therefore,- there is no impact on. the desigt basis accidents.-
- c. The _ design basis accident which credits . safety injection - (SI)> 'on -
pressurizer pressure low and steam line pressure. low:is a steam line break. Since the s'afety analysis SI'setpoint:for a steam line break -
is not changed, there is.no increase:in-the consequences. -.Since~the' !
engineered safety features system will_ function ^ as before, there is 1 no increase in the probability- or' consequences: of - ai steam line
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break.
On the basis of this review, NNECO concludes that there is' no:significant increase in the probability or consequences of an accident. previously analyzed. '
- 2. Create the possibility of- a new or different kind of accidentt from any previously analyzed.
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- a. Evaluations have been performed on the fuel assemblies, RCCAs, 'and other sifsty-related equipment tot confirm that their function and reliabilit/ are not - negatively impacted due to the? upgrade to VANTAGE SH and other changes described;in AttachmentL1. The conclu-sion was made that there was no negative impact; on safety l as a- ,
result of the proposed changes. :Noinew accident: scenarios, failure mechanisms, or limiting single failures areLintroduced as a' result of-the fuel transition. The ~ presence of : VANTAGE -5H fuel assemblies..
in the core or the ~ revised . analytical assumption Lhave no adverse y effect and do not challenge the performance of any Lother ' safety-- ;
related system. Therefore,, NNEC0 concludes' that the proposed j
changes do not create any new 'or Ldifferenti kindoof accidente from those previously analyzed.
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- b. Since these changes do not affect plant t operation, the potential for:
an unanalyzed ' accident is not created; No new failure imodes are introduced.
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- c. The proposed changes have not modified the plant response and no new failure modes are introduced. Therefore, the potential for an unanalyzed accident is not created.
- 3. Involve a significant reduction in a margin of safety.
- a. The upg.ade to VANTF~ 5H fuel and the other changes discussed in Attachr ent I werr valuated against the applicable acceptance criteria. The cri ia are discussed in Section 5.9.2 of the PSE and ar e summarized velow.
l (1) j~uel-related criteria:
(a) DNBR greater than safety analysis limit (b) PCT less than 2200'F for LOCA (c) Fuel centerline temperature less than 4900*F (B0L), 4800*F (E0L)
(d) Average fuel pellet enthalpy less than 200 cal /gm for rod ejection (e) Fuel melting limited to 10 percent for rod ejection (f) Remainder of 10CFR50.46 criteria (clad oxidation, hydrogen generation, coolable geometry, long-term cooling)
(2) RCS oressure boundarv-related criteria:
(a) Pressure less than 110 percent for Condition It and III events (b) Pressure less than 120 percent for Condition I\ events (3) Containment oressure:
(a) Pressure less than design pressure A significant increase in DNB margin is provided by the IFM grids which are a feature of VANTAGE 5H. Additional DNB margin is created by the use of the WRB-2 correlation and the use of the revised thermal design procedure. This additional DNB margin can then be used to increase the FAH limit as well as remove thimble tube plugging devices. The non-LOCA analyses confirm that the DNB design basis is met for Standard 17 x 17 and VANTAGE 5H fuel.
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4 November 1, 1990 The non-LOCA analyses ccafirmed that the acceptance criteria for-rod ejection were met. The results also- showed that peak RCS pressure remains below 110 percent of design forLall events- . l Margin to PCT is obtained primarily through the use of> improved LOCA!
evaluation models (BASH for 1arge-break LOCA analysis andiNOTRUMP for small-break LOCA analysis). - Also, the- zircaloy. grids and IFM grids used in VANTAGE SH fuel provide additional PCT margin. -This i margin can then be used to implemer' the changes _ described ~ in Section 5.0.1 of the PSE. The .LOCA' analysis 1 considered four- and' three-loop- operation ~ as well as transition; core effects.- The q results for Standard fuel provided- the most -limiting PCT.: . The LOCA'- 3 analysis demonstrates that the PCT acceptance criteria: contained .in '
10CFR50.46 of 2200*F - continues ' to be met as well as --the _ criteria -
related to clad oxidation and maximum hydrogen generation.
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LOCA-related analyses demonstrate that - thet margin 7 of . safetyLwith~
respect to blowdown reactor vessel and loop forcesf is- preserved, .
thus satisfying the 10CFR50.46 criteria that the core remain- amena- 3 ble to cooling .after a - LOCA. Long-term cooling and post-LOCAL j subcriticality concerns are- satisfied by increasing ~ the. RWST' boron' _-
concentration. This in turn affects the concern'with. boron-precipi-- I tation. The analysis shows that hot-leg'switchover must bel accom-plished 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> after accident initiation,' which.is acceptable.
Mass and energy release from LOCA and steam line1break' are. not '
adversely affected by any of. the changes. discussed i_n Secticn 5.3.of 1 the PSE. Therefore, the input to the containment and subcompartment; analysis continues to be valid. j In summary, performance ofianalyses; and evaluationsL for'tbe' ' upgrade to VANTAGE SH and associated changes.have confirmed that tae operat-ing envelope defined by the technical specifications continues to' be bounded by the revised analytical: basis, which in no case exceeds- 1 the acceptance limits. Therefore, the margincof safety provided by the analyses in L accordance with these acceptance - li aits' is not:
reduced.
b.- Since the proposed changes do not affectL the_ consequences of any accident previously analyzed, there is no reduction in the margin.of safety. -
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- c. The proposed changes do not. increase the' consequences of anLaccident-previously analyzed, and_'the protective boundaries are.not ' affected by these: proposed changes. Therefore, there is no reduction:in the . 1 margin of safety.
1 Moreover, the Commission: has' provided guidance coricerning the ' application Dof the standards in 10CFR50.92 by providing ~ certain examples '(March 6, 1986,' !
51FR7751) of amendments that are considered not ' likelyc to- involve 'a
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l U.S. Nuclear Regulatory Commission -l B13627/Page 11 i November 1, 1990 ,i l
l l significant hazards consideration. . Example'-(iv) provides<Ithat a:significant- '
l hazards consideration-finding is unlikely for: j
! A change which either may result -in"some increase to the probabil -l ity or consequences of a previously analyzed accident'or may reduce 1 in some way a safety margin, but where..the results of _the change 1 l are clearly within -all acceptable criteria with> respect . to thel -
M system or compnent specified in the ' StandardL Reviews Plani . for: l I example, a change resulting from a small refinement of a previously j used calculation-model or design method.
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This example appears applicable to the portion of the proposed license > amend- ,
ment request involving a change in the- fuel ' design: and . improved canalytical a methodology. As discussed above, for non-LOCA accidents, the consequences y have somewhat increased. However, they nare within the: applicable acceptance . 0
, criteria, ~ Similarly, - for LOCA accidents, . the ' proposed. changes: resultiint an i increase in the consequences (PCT);' however in: all cases, they; remain within-the acceptance criteria for. PCT. For. a fuel-handling accident, the. increase- j in the- consequences (Table .1) is not significant. Therefore, the proposed changes do not involve a significant hazards consideration. For other Techni-cal Specification changes', there- is no increase fin . the' consequences ofs an accident previously analyzed. Therefore, based on .the .above, NNECO: concludes 1 that the proposed Technical Specification changes .do not involve a significant- I hazards consideration.
i ENVIRONMENTAL CONSIDERATION NNEC0 has . reviewed accidents analyzed in. thet Millstone' Unit No. 3.FSAR with' l respect to - the radiological source term and radiological; consecuences; in :
association with the transition to VANTAGE 5H fuel'.- Included in tiis ' review is consideration of extended fuel burnup- of 45,000 MWD /MTUtforEthe? batch
! average discharge. This extended burnup has aipeaktfuel rod averageLburnup1of j
< 60,000 MWD /MTV. The existing Millstone- Unit No. '3 Technical Specification - l permits use of reload fuel with aL maximum 1 nominal en'richment; of, 5.'0 weight - ,
percent U-235. This allows a burnup level to 50,000 MWD /MTUL (60,000 MWD /MTU 1 peak rod burnup). The safety consideration.= associated with reactor operation' with higher enrichment and extended irradiation has been evaluated by:the, NRC- 1
- and the ;
safety (;gt)affThe concluded proposedthat suchtochanges change . VANTAGEwould 5H "not ' adversely fuel will' : affectthe operate within: plant ;
j existing Technical Specification requirements; therefore, the existing source- j terms are applicable. For those events in Chapter 15 whichihad to be reevalu-ated or reanalyzed as a- result of the proposed changes, no . increase;in"the- j, amount of fuel failures was calculated. Since there was no; increase in' failed- 1 fuel and- the fission product source terms'~are; unaffected; there is ..no l
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- (10) D. H. Jaffe letter to E. J. Mroczka, Millstone '3--Issuance of Amendment ~,
dated July 28, 1989. '
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B13627/Page12 1 November 1, 1990 j radiological consequences resulting from .the proposed -tanges. l As demon-strated .above, the proposed amendment request does not' involve: a significant ,
hazards consideration. The pro)osed amendment request also does not result in '
4 a significant increase . in inc ividual- or -cumulative occupational ~ radiation - i exposure.- This is supported by the NRC as documented in a public notice,.
" Extended Burnup Fuel Use in- Commercial LWRs; Environmental: AssessmentJ and. ,
finding of No Significant Impact," dated February 23, - 1988, : and 6 supplemented by the NRC in a public notice for Millstone Unit No.- 3 ~(54FR27082). ~ There-- -1, fore, NNECO concludes that there are no significant radiological or nonradio- ,
logical environmental impacts associated.with the proposed amendment request.
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The Millstone Unit No. 3 Nuclear Review Board has' reviewed ;and-approved the
-i attached proposed revisions and has concurred:with the above. determinations, j 1
In a letter dated - January 20, 1988, the NRC requested l that. for ' future i Millstone Unit No. 3 reload core cycles, NNEC0 provide an expected anticipated- !
transients without scram (ATWS) . moderator temperature coefficient; (MTC) at 1 equilibrium Xenon conditions, pending the Staff's evaluation' of a' forthcoming y Westinghouse Owners Group (WOG) response to- a Staff request for information'on 1 ATWS MTC dated June 12, 1987. On March 1, 1989,:the WOG submitted to the NRC: !
a report, WCAP-11993, " Joint Westinghouse Owners, Group / Westinghouse P*ogram: i Assessment of Compliance with ATWS. Rule-Basis for Westinghouse PWRs," (nonpro- ;
prietary). This report presents results of ;a L. joint WOGaand LWestinghause !
effort to quantify - the frequency of ~ core damage resulting: from ATWS for :
Westinghouse pressurized water reactors .(PWRs) and demonstrates compliance. '
with the ATWS Rule as specified in SECY-83-293 for JWestinghouserPWRs. 1 Although the WOG -submittal demonstrates compliance withithe ATWS Rule, a
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plant-specific response to Millstone Unit No. '3 is provided below.
For the Cycle 4 core, full-power BOL equilibrium.Xe MTC is -7.2. pcm/*F' which is more negative than the -5.5 pcm/*F assumed in Westinghouse Let- i ter NS-EPR-83-2833 (E. P. Rahe, Westinghouse,c to S. J. Chilk, NRC, October 3, 1983).
The proposed amendment request-needs to be approved to support Cycle.4 opera-tion and prior to entry into Mode 4 'after the February 1991 refueling ~ outage.
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NNEC0 requests that these proposed changes be approved ,and; effective by :
March 6, 1991. This would allow specific applicable . Technical Specifications : l to be in place prior to any affected Mode change.; 1 0
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L U.S. Nuclear Regulatory Commission l i B13627/Page 13-November 1, 1990 In accordance with- 10CFR50.91(b),- we are providing the -State of Connecticut -
with a copy of this. proposed amendment. q Very truly yours,- -
NORTHEASTNUCLEARENERGYl COMPANY' FOR: E. J. Mroczka q Senior Vice President- j BY: c ,
- C. F Sears j
- Vice President a
cc: T. T. Martin,-Region I Administrator l
D. H. Jaffe, NRC Project Manager, Millstone Unit- No.:3 W. J. Raymond, Senior Resident Inspector,. Millstone Unit Nos.1, 2, and 3 l Mr. Kevin McCarthy, Director Radiation Control Unit a Department of Environmental Protection :
Hartford, CT 06116-1 STATE OF CONNECTICUT)
. ) ss. Berlin COUNTY OF HARTFORD -)
Then personally ' appeared before me, C. F. Sears, who. being duly sworn, did 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 Licenseetherein,- and that the i
statements contained'in said information are true- and correct to theobest of -
-his knowledge and belief. :
% UAnm; ~
Notar9 Public i
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4 U.S. Nuclear Regulatory Comission B13627/ Table 1/Page l' November 1,.1990
- Table 1' EAB Dose in REM Due to-a Millstone Unit No. 3- l
- Fuel-Handling Accident i 1.65 Peakina Factor 1.7 Deakina Factor-- '
.l Thyroid 7.38(+0) 7.59(+0) .
Whole Body 4 '. 94 (- 1 ) .5. 09 ('- 1 ) ]
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b Docket'No.~50-423- l B13627 H 1
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1 11 Attachment l' l
q Millstone Nuclear Power Station,; Unit-No. 3 ; 'l Description of Proposed Technical Specification Changes' >
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! U.S. Nuclear Regulatory Commission - I B13627/ Attachment 1/Page 1- R November 1,!1990 i
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Millstone Nuclear Power Station, Unit No. 3 .
Description of Procosed Technical Specification Chanaes. Cycle 4 Millstone Unit No. 3 plans to refuel 'and operateLwith upgraded Westinghouse '
l fuel features (VANTAGE 5 -Hybrid (5H])L and increased. peaking factors andLthe capability of operating with up_ to all of- the : assemblies" thimble- tube' plugs being removed. ,;
3 The transition to and al(l ) VANTAGE y SHLeore . evaluations / analyses in -the plant-Safety Evaluation Report were performed at';a reactor ' thermal' full poweri 7 level of ' 3411 MWt (100 percent _ rated power) forf four-1oop operation, and.
2560 MWt (75 percent rated' power) for three-loop . operation.. -The: following. i conservative assumptions' were made in the; safety: evaluations / analyses:
10 percent uniform steam generator tube plugging with' reactor coolant system (RCS) thermal design flow rates 'of. 378,400 gpm' and 294,900 gpm for: four .and- t three-loop operation, respectively, core bypass flow' increases from 6 to .;
8.6 percent- to account for complete thimble plug. removal' and/the use of sthe i Intermediate Flow Mixers (IFMs),- grids, the 'use of Relaxed Axial ' Offset Control'(RAOC) power distribution for both three-'and'four-loop operation, and +
increases in F and F For four-loop ' operation,ca vessel / core < inlet coolant temperature. 00 557.0h. and a vessel coolant average temperature of 587.l*F were used. For. three-loop operation, a vessel / core. inlet coolant- temperature of 550.2*F and a vessel coolant average temperature of 579.6*F were used. l Beginning with Cycle 4, theN future fycles of operations for.-Millstone .Ungt i
No. 3 will use increased F A and F peak The ' full . power - F l peaking factor design limit Nill _inchease ing factors.from the current -value o- of 1.5 1.70. The maximum four-loop F peaking- factor-limit will increase- from the '
current value of 2.32 to 2.60. gThe maximum three-loop Fn peaking factor limitL ,
will increase from the current value of 2.25 to 3.0 atifhe maximum 75 percent' J rated power. The K(Z) envelopes' will be modified.; These: changes will permit - ,
more flexibility. in developing fuel management schemes (i .e.',< longer l fuel: ;
cycles, improvement of fuel economy and neutron- utilization,~ vessel fluence-4 reduction).
l- The proposed Technical Specification changes have been prepared to support the-Cycle 4 reload. The proposed changes are summarized below:: j i
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f' (1). Plant Safety Evaluation for Millstone Unit 3, _ VANTAGE- 5H Fuel Upgrade, !
August 1990.
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U.S. Nuclear Regulatory Commission- l B13627/ Attachment 1/Page 2 -
November 1, 1990 .[
A. Technical Specification Chanaes Due to the Cycle'4 Reload;
- 1. Definitions--Allowed Power Level .;
ND Definitions'l.43 and 1.44 have been added for. APL - an'd AP!. which .t were included as part of the RAOC specifications. ,
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- 2. Ficures 2.1-1 and 2.1-2. Reactor Core Safety Limits
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The reactor core safety ~ 1imits figures, have' been revised totreflect the Departure: from Nucleate _ Boiling (DNB) correlation used; forrthe standard and VANTAGE SH fuel. ,
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- 3. Table 2.2-1. Reactor Trio System ' Instrumentation Trio J Setooints. .
Functional Units 6 (Overtemoerature AT) and 7 (Overoower AT)
The values of Total Allowance (TA), Z, and sensor 3 error -(S)L for-overtemperature AT (OTAT) and overpower AT (0 PAT) .have been revised due to the new core safety limits'and instrumt.ntation uncertainties.n s In addition, a separate set of- values (TA, 2, S); for: Channels 1 and 1
i 2 (Veritrak) and Channels 3 and 4 (Rosemount) Mare' provided. Fur-:
h ther, the following constants included in. Notes _l, 2, 3, and 4. are revised. -
- a. For OTAT [
c (1) Note 1 Current Revised .
l K1 1.08 ^1.20 K2 0.01313/F' O.02456/F* y Taul 12'sec. 8 sec: !
Tau 4 33 sec. , 20 sec' _ l K3 0.00066/ psi 0.001311/ psi 't f(AI) Penalty . ,
Break Points -30%,.+10%- -26%,.+3% i Negative Slope 3.6%/%AI 3.55%/%AI-Positive Slope 2.0%/%AI' 1.98%/%AT.
o (2). Note 2 Allowable- Value N 2;1%AT '2.7%AT :!
N-1 3.6%AT 2.7%AT L ]
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U.S. Nuclear Regulatory Commission :
B13627/ Attachment 1/Page 3- 'l November 1, 1990 b.. For OPAT (1) Note 3 ;
K6 0.00129/*F 0.00190/'F 1
(2) Note 4 a
Allowable Value .2.8%AT. 2.7%ATv i An asterisk (*) footnote 'on page 2-5 has been revised and move (to'. I page 2-6 as it applies to Functional - Unit 12, Reactor 1 Coolant - 1 Flow-Low. The minimum measured' RCS . flow has been' provided -~~rather- l than - the loop design flow due torth.e use of ~the - revised thermal-design procedure (RTDP). The : calculational method-~ utilized to meet i 1
the DNB design basis is the RTDP-i
- 4. Bases Sections 2.1.1. Reactor Core (Pace B2-1) -and -2.2.1 (Paa ' ;
es B2-5. 82-7)-
q The existing thermal-hydraul.ic analysis-of the standard fuel used in j the . Millstone Unit.No. 3 -core is based on; the standard thermal. and '
hydraulic methods and the W-3 DNB correlation as' described in the' -
Millstone Unit No. 3 FSAR. The DNB analysis l of"the core, containing: l both the standard and VANTAGE SH fuel assemblies .has been modified
. to incorporate the WRB 1 and WRB-2 DNB correlations. .Thei proposed changes to Bases Sections 2.1.1 and 2.1.2 reflect that.
- 5. Section 3.1.1.1. Boration Control' 3 i
This section has been split into two; one-for Modes 1~ and 2 and the
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other ' for . Modes 3 and - 4. The shutdown' margin considered ;in the
-revised -safety analysis was : reduced.. from- 1.6 percent' AK/K1 to-1.3 percent AK/K. Therefore, the; shutdown margin included 1in _Sec- '
tion 3.1.1.1 has been revised. accordingly. Further, Surveil -
1ance 4.1.1.1.1.e .has been' deleted as it appliesf to - Mode 3, 'and -
4 Surveillance 4.1.1.1.2-.has been. revised to . reflect- the : deletion of
. Section 4.1.1.1.1.e. ShutdownL. margin.'in Modes'3'"and-'4 is Jnot 1 changed lhere, but will be changed in a future submittal. !
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U.S. Nuclear Regulatory Commission B13627/ Attachment 1/Page-4 November 1, 1990
- 6. Section 3.1.2.5. Bonted Water Sources--Shutdown The minimum boron: concentration in the refueling water-storage tank;- [
(RWST) is increased froii; ?300 ppm t _2700 ppm. This is-to provide' q adequate shutdown margin.
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- 7. Section 3.1.2.6. Borated Water Sources'--00eratina- 1 A range -in .the - RWST boron concentration has been changed from-
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2300 2600 ppm to 2700 2900 ppm. This is to provide : adequate shut-: j.
down margin. The shutdown! margin included in Action 'a has beenc ;
revised to 1.3 percent 4K/K to match with Section 3.l'.1.1, Boration Control. Shutdown margin in. Modes 3 and 4":will -be changed in1 a future submittal.
- 8. Section 3.1.3.4. Rod Dron Time I The rod drop time has been increased from12.'2 seconds to 2.7Lseconds--
due to the fuel design changes. _ The! effect of-:this incre'ase' on' ;
safety analysis has been considered, i 1
- 9. Section 3.2.1.1. Axial Flux Difference--Four Looos- Oheratina Section 3.2.1.2. Axial Flux Difference--Three Loops Ooeratina-Currently, Millstone Unit No. 3_ uses the constant axial- offset 1 control (CAOC) procedure for the control of axial ' power ' distribution '
(see Specifications 3.2.1.1 and _3.2.1.2). Beginningi withL Cycle 4i Millstone Unit No. 3 will have an. option of, the : relaxed axial offset' control -(RAOC) or base -load. This procedure 1 defines the ' allowed operating space of axial. flux difference (AFD)~ versus; thermal ~ power.
The limits are selected by considering a range- of' axial -xenon-distributions which could occur. as a ' result of large" variations of:
AFD. Also, a direct F surveillance will'now be.used. The' proposed Technical Specificatio8 sections for four'and three 1 cops operating-are consistent with the Westinghouse Standard Technical Specifica-tions(HSTS).
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B13627/ Attachment 1/Page 5-November 1, 1990 l
- 10. Section 3.2.2.1, Heat Flux Hot Channel Factor-F (Z)--Four Loops-n Operatina 1 Section 3.2.2.2, Heat Flux Hot Channel Factor-F (Z)--Three' Loops i 0
Operatina i i As stated above,. Millstone- Unit No. 3 will ' be using1the .RA0C. - The : .;'
proposed Technical Specification surveillance sections. for;four-: and three-loop operation are . based on ' the WSTS. Fn . measurement for three loop operation is now based onra safetyJanal} sis full power of:
75 percent. Previously, all analyses were done at 75 ' percent'except' -l LOCA and a portion of the locked' reactor coolant pump rotor accidents which used 65 percent.
- 11. Section 3.2.3.1 RCS_E,',0w Rate and Nuclear Enthalov Rise Hot Channel Factor--Four loops 00eratina
! Section 3.2.3.2. RCS Flow Rate and Nuclear Enthalov Rise Hot Channel Factor--Four looos'0oeratina-The RCS flow now included in Technical Specification: Sec- 1 tions 3.2.3.1 and 3.2.3.2 is the : minimum . measured flow duetto the-use of RTDP. In addition,'the conservative-value of.the uncertainty; l L for RCS flow for four-loop operation is 2.4 percent flow:versus the previous ~ 1.8 percent and - for three-loop ?operationtisi 2.8 flow.
versus 2.0- percent flow. This .is reflected in:the proposed:Techni-cal Specification changes. ,
- 12. Table 3.2-1. DNB ' Parameters 1
The DNBforparameters pressure) -(indicated four- and three-loop = op RCS T'erStion' has been . modified.!
V to the revised instrument uncertainties. ]
- 13. Talie 3.3-2. Reactor Trio System Instrumentation'Resoonse Times- 4 1
The response time.for.the1 pressurizer water . level-high reactor trip .
has been specified. ' This has 'been credited in. the : revised safety j
analysis' for the Rod Cluster Control Assembly -(RCCA) bank withdrawal at power transient. ;i j
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f U.S. Nuclear Regulatory Commission.
813627/ Attachment 1/Page-6 November 1, 1990 , 1
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l l 14. Section 3.4.1.2. Reactor toolant System--Hot Standby 1
Three reactor coolant loops are now required to be in' operation L in i Mode 3. This is due to the "evised RCCA bank = withdrawal from. a ;
subcritical or. low power conditioii accident. .This -is reflected in i
the proposed Technical Specification change;. i
- 15. Section 3.4.1.6. Reactor Coolant System--Isolated Loon Start-Un' l The boron concentration of the isolated lo' op is ' increased from .s 2300 ppm to 2600 ppm. -Shutdown margin in Modes'5 and 6 are. 'not changed here but will be changed.in a future submittal.
- 16. Section 3.5.1. Accumulators ~ b The boron concentration in -the accumulators- has- been changed from 2200 to 2600 ppm to 2600; to 2900 ppm. This is consistent with the- '
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revised safety analyses.
l 17. Section 3.5-2. Emeraency Core Coolina Systems A 10 percent reduction in safety! injection (SI)- flow.- h'as been -!
incorporated in the safety _ analyses as well as a 10 gpm SI pump'. flow :
imbalance. This is reflected - in surveillance 4.5.2.f.2 and 7 4.5.2.h.2, respectively. 1 1
- 18. Section 3.5.4. Refuelina-Water Storaae Tank The boron concentration in the RWST is' increased from - 2300 to 2600 pm to 2700 to 2900 ppm. The effect of this. increase on. safety analyses-has been considered, t
.i L 19. $1ction 3.6.2.3m Sorav Additive SystetD 1 i
The proposed change will ; increase the range of acceptab1e-~ sodium ,
hydroxide concentrations in. the chemical addition tank '(CAT) to. 3.4 '
to 4.1 percent from the previous range of 2.4~ to 3.l? percent. . T.he s volume (level) in the CAT is reduced to a range of - 17,760; to ' a 18,760 gallons from the previous range of 18,000 to-19,000; gallons.- i
- 20. Section 3.9.~1.1. Refuelino Operations--Boron Concentration i The proposed change will . increase the boron concentration - in the = I filled portion of the RCS and' refueling; canal during Mode' 6' to 1 correspond with_the new minimum RWST boron concentration.
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.l U.S. Nuclear Regulatory Commission l B13627/ Attachment 1/Page 7 J November 1, 1990~ i 1
- 21. Bases Sections 3/4.1.1.1'and 3/4.1.1.2. Shutdown Marain' .t Bases Section 3/4.1.2. Boration Systems Bases Section 3/4.1.3. Movable Control' Assemblies
- Bases Section 3/4.2. Power Distribution Limits f Bases- Section 3/4.2.1. Axial Flux Difference -
SAses Sections 3/4.2.2 and 3/4.2.3. Heat: Flux Hot Channel Factor and .
ES Flow Rate and Nuclear Enthalov Rise Hot Channel Factor- ,
.i Bases Section 3/4.2.5. DNB Parameters )
Bases Section 3/4.4.1. Reactor Coolant looos and' Coolant Circulation 4
The above bases sections are being revised to reflect.the proposed Technical Specification changes.
- 22. Section 6.9.1.6. Core Operatina-Limits Reoorts Sections 6.9.1.6.a.2 and 6.9.1.6.a.3 have' been revised to correct--
the references for ' specification numbers._ Section 6.9.l.6.a.4; .
6.9.1.6.a.5, and 6.9.1.6.a.6 have been revised - to make them consis- 1 tent with the revised Sections 3/4.2.1.1,:3/4.2.1.2L 3/4.2.2.1, and 3.4.2.2'2.
Section 6.9.1.6.b has been revised to. reflect the- methods used l in '
the plant safety analyses for- the- upgraded VANTAGE SH- fuel. It is l noted that -these methods have been reviewed and approved; previously by the NRC. However,. these methdds will be used for Millstone. Unit -
No. 3 for the first time beginning with the Cycle' 4 reload analyses. "
i B. Other Technical Specification'Chanaes-
- 1. Section l'. Definitions. 1-26. Radioloaical Effluent Monitorina: and .
Offfite Dose Calculation' Manual- (REM 00CM) !
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Reference to Specification 6.161has been changed to 6.13. This is to correct a typographical' error, j l
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u B13627/ Attachment 1/Page 8
. November 1, 1990 l i Fiaures '
- 2. i i
L The following figures are redrawn for clarity. There are no techni-:
. l' l cal changes to these figures.
Fiaure Numbers '
3.4-1 l 3.4-2 3.4-3 j 3.4-4a 3.4-4b .
- 3. Renumberina of Technical = Soecification Paaes j 1
Table ' 2.2-1, pages 2-5 through 2-11,. were' retyped andLrenumbered' to: !
accommodate the proposed changes described above (see Change A.3).
Pages 3/4 2-1 through 3/4 2-24 were ;also retyped to. accommodate the-proposed changes to Section: 3. 2.1.1, : 3._2.1. 2, ; 3. 2,2.1, 3.2.2.2; '
- 3. 2. 3.1, 3. 2. 3. 2, and Table 3.2-1.. Iniaddition, sthese ~pagesfwere- ,
renumbered to deletet the blank 'pages from Technical, Specifications j which were left by Amendment No. 50. Bases Section.pages-BL3/4'2 ~
through B 3/4' 2-6 were retyped and renumbered to ' accommodate? the "
proposed changes described above (see Change A.21)... As, a. result"of the above, appropriate pages of the Technical 'Specifi. cation ~Index -
were revised. ;
C. Technical Soecification Chanaes Due to Cable Insulation Resistance l Effects on ESF Instrumentation l
- 1. Table ~3.3-4. Enaineered Safety Features Actuation System Instrumen--
tation Trio Setooints -i .
'i The proposed changes revise the total allowance (TA), ..Z,' sensor- W error (S), and allowable value in Table 3.3-4.for pressurizer pres- ,
l sure. . low and steam line pressure ; low. In addition, i a= separate i
pressurizer pressure low trip value for Channels 1 and 2 (Veritrak)-
and Channels.3 and 4 (Rosemount).are provided. tThe~ proposed changes are accomplished by adjusting instrument settings and do not'irvolve any physic'al addition, deletion, or; modification ~ to plant. compo- .i nents. The revised pressurizer pressure low and steam line' pressure '
low trip setpoints and other associated values 1 such as total allow- ;
ance, etc., are calculated by the' methodology described in WCAP- 1 10991, '" Westinghouse Setpoint Methodology for Protection . System, I
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U.S. Nuclear Regulatory Commission B13527/ Attachment 1/Page 9 November 1, 1990 .
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Millstone _ Unit No. 3,": which =is :also the method used in calculating q other setpoints in Table _3.314 of Technical Specifications.'. The .;
proposed changes are based'on"recent data which indicates.there is an additional error due to. decreased cable insulation' resistance at -
high temperature. _ The Jnew setpoints betterJ represent --the error 1
L associated with the pressure measurement;- the changes willt. maintain- 1 L the assumed. performance of. engineered ? safety features _ actuation-- >l
! systems. - In addition, an attempt-to regain some lost margin on the: 1 pressurizer pressure, safety' injection (SI) signal w'as'made by taking . j credit - to a lower safety-.; analysis Rsetpoint and L by' splitting the-j specification into a separate Eset of values for the different-
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1 channels of instrumentation, Veritrak and- Rosemount. 'The low steam-line . pressure SI setpoint used in the Lsafety ' analysis . is not affected by the inclusion of instrument cable insulation resistance jl error. '!
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