Text

Proposed Tech Specs for Cycle 11 Operation,Including Changes to Referencing WRB-1 DNB Correlation,Accumulator Spec & Peaking Factor & Deletion of Third Line Segment of K(Z) Curve
	ML20140A731
Person / Time
Site:	Prairie Island
Issue date:	01/13/1986
From:	; NORTHERN STATES POWER CO.
To:	;
Shared Package
ML20140A716	List: ML20140A711; ML20140A718; ML20140A731; ML20140A747; ML20140A751; ML20140A775; ML20140A781;
References
	NUDOCS 8601230359
	Download: ML20140A731 (48)
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Exhibit A Prairie Island Nuclear Generating Plant License Amendment Request Dated January 13, 1986 Evaluation of Proposed Changes to the Technical Specifications Appendix A of Operating License DPR-42 and DPR-60 Pursuant to 10 CFR Part 50, Sections 50.59 and 50.90 the holders of Operating Licenses DPR-42 and D?R-63 hereby propose the following change to Appendix A, Technical Specifications:

1. Reference WRB-1 DNB Correlation Proposed Change Add reference to the URB-1 DNB correlation for Westinghouse fuel and identify that the W-3 DNB correlation will be used for Exxon fuel.

Pages affected: TS.2.1-1, 2, 3 TS.2.3-6, TS.3.10-8, TS.3.10-13, Figure TS.2.1-1 (The bases on existing page TS.2.1-3 will be moved to proposed page TS.2.1-2. Therefore, there is no page TS.2.1-3 in Exhibit C.)

Reason For Change In our April 19, 1985 submittal revising Topical Report NSPNAD-8102 Rev 3, we requested approval of methodology changes which would allow the WRB-1 correlation to be used for Westinghouse fuel. The proposed changes are consistent with the methodology changes.

The reference to Westinghouse fuel does not include the Westinghouse Standard fuel (LOPAR, Low Parasitic), it only refers to the Optimized Fuel Assemblies (OFA). At this time, we have no plans to reuse old Westinghouse standard fuel. Prior to using these assemblies, we would complete the necessary analyses. If the analyses support a 50.59 type evaluation, we would use the old Westinghouse standard fuel and update the Technical Specification Bases in the next Technical Specification change to be submitted. If the review cannot support a 50.59 evaluation, an amendment would be proposed prior to use of this fuel.

Significant Hazards Evaluation The proposed changes are related to methodology changes currently under review by the Staff. While these changes may change the consequences of a previously-analyzed accident or may change a safety margin, the results are clearly within all acceptable criteria.

For these reasons, operation of the Prairie Island Nuclear Generating 1

8601230359 860113 2 g,t PDR ADOCK 050 P

P4 Plant in accordance with this proposed change will not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

2. Temperature Coefficient Changes Proposed Change Delete reference to the " moderator temperature coefficient" in Specification 3.1.F.1.

In place of the existing restriction on isothermal temperature coefficient, require the isothermal temperature coefficient to be below 5 pcm/ F when below 70% power and negative above 70% power.

Change the associated bases.

Replace the existing action statement, specification 3.1.F.3, with the Standard Technical Specification action statements, specification 3.1.1.3. Except the requirement to submit a special report in 10 days has been changed to allow 30 days to submit the report.

Pages affected: TS.3.1-17 and 18 Reason For Change The existin5 specifications refer to moderator temperature coefficient for one fuel type and isothermal temperature coefficient for another.

This change will simplify the specifications by only referring to one temperature coefficient for both fuel types; the isothermal temperature coefficient.

Fuel economics are significantly affected by the isothermal temperature coefficient limitations. In order to operate the plant more economically a larger (more positive) temperature coefficient is specified.

The existing action statement for the temperature coefficients is unclear and is being replaced with the applicable Standard Technical Specification action statement.

Thirty days to submit the special report is requested for two reasons.

10CFR50.73 allows 30 days for all Reportable Event reporting.

Secondly, the isothermal temperature coefficient is measured at the beginning of the cycle during physics testing. This is a very busy time for the engineers that would write this report.

l A-2

N, 4 Significant Hazards Evaluation Exhibit H contains the Safety Evaluation for the Isothermal Temperature Coefficient ' changes.

These temperature coefficient changes have also been found to be conservative for Unit 2 Cycle 10 (current cycle in operation).

The addition of the Standard Technical S . cifications action statement constitutes an additional limitation not presently in the

. specifications.

While the temperature coefficient revision may change the consequences of a previously-analyzed accident or may change a safety margin, the results are clearly within all acceptable criteria.

For these reasons, operation of the Prairie Island Nuclear Generating .

Plant in accordance with this proposed change will not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

3. Accumulator Specification Change Proposed Change Change the volume requirement for the accumulators from "between 1250 and 1282.9" to "1270 +20" in specification 3.3.A.l.b.(2).

Pages Affected: TS.3.3-1 Reason For Change This change is being made in an effort to make the accumulator volume setpoint easier for operators to use without significantly changing accumulator volumes. The recent LOCA analyses have used a nominal water volume of 1266.5 cubic feet (Exhibit E, Table 14.6 2).

The 20 cubic foot plus or minus tolerance will provide more allowance for drift of the level instruments as well as more operational flexibility.

Significant Hazards Evaluttion Westinghouse uses the nominal accumulator level as an input for their LOCA analyses. The nominal value proposed is 3,5 cubic feet (approximately a 0.3% ) larger than the nominal level. The maximum water volume will increase with this change from 1282.9 cubic feet to 1290. The associated decrease in gas pressure is not large enough to A-3

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h 1 affect the accumulator performance. These minor changes proposed in Exhibit B and discussed above will have no significant effect of the safety of the plant.

While this change may result in some change in the consequences of a previously-analyzed accident, the results would be clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan.

For these reasons, operation of the Prairie Island Nuclear Generating Plant in accordance with this proposed change will not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

4. Peaking Factor Changes Proposed Change Change the peaking factors limits in Section 3.10 as shown below:

Old Value New Value FQ 2.32 2.30 FdH 1.55 1.60 FdH equation 1+0.2(1-P) 1+0.3(1-P)

Note: The old FQ for Westinghouse fuel was 2.21.

Revise the peaking factor equations on page TS.2.1-2.

Delete definitions of BU(E q) and E j in the specification and references to them in the bases.

Increase the required high neutron flux trip setpoint reduction from 1% to 3.33% for each percent that measured FdH exceeds the limit.

Revise the Bases of Section 3.10 as necessary for the peaking factor changes. The bases have also been edited to remove some outdated information, e.g. deletion of the definition of the term F (z),

q daletion of an description of the FdH uncertainties and deletion of a discussion of rod bow.

Delete Figure TS.3.10-7, renumber the next sequentially numbered curve and delete references to the curve.

Pages affected: TS -x, TS.2.1- 2, TS.3.10- 1, 2, 9,10, 11, 13, 17, and Figures TS.3.10-7 and 8 A-4 u

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Reason For-Channe The proposed peaking factors and equations will provide the plant with more-operating flexibility than previously existed. Both FQ and FdH are being changed in order to obtain a more optimum relationship between the two limits. The FdH equation change will provide a less restrictive limit at reduced powers.

The normalized exposure dependent function, BU(Ej) was used for Exxon Nuclear fuel. Since this value of this function is one for all burnup -

values up to 55 GWD/MTU, it can be deleted.

The increase in the setpoint reduction for.FdH corrects a error in our specifications. If the FdH limit were exceeded by it, for example, ,

power would need to be Jacreased by 3.334, not the 14 currently in our '

specification in order to comply with the limit. .Similarly, the high neutron flux trip setpoint should be reduced 3.33% for each 14 that the FdH limit is exceeded.

The bases are being revised in accordance with these changes and also being simplified by removing outdated and inaccurate information. On page TS.3.10-10, the reference to " experimental error" has been ,

changed to " measurement error"; a more applicable term. The rod bow bases are bei.ig deleted since the rod bow effects are explicitly calculated in our analyses.

Significant Hazards Evaluation Supporting these changes are Exhibits E and H.

The transient analyses have been done using an FQ of 2.32 and FdH of 1.60 (See Exhibit H, Table 4.2). The change in the FdH equation was '

also evaluated (Exhibit H, Section 3.6) and found to conform to acceptance criteria. These evaluations have been performed using-NSP methods which include methodology changes submitted for NRC review on April 19, 1985. The effect of new upper internals was evaluated and found to have only ~a minor effect on the results shown in Exhibit H and will meet all acceptance criteria. These new peaking factor limits were also evaluated for Unit 2 Cycle 10 (currently operating Unit 2 cycle) and found to be acceptable.

A revised large break LOCA analysis using the new peaking factors has -

been performed for Westinghouse fuel including assessments of penalties associated with Upper Plenum Injection and with transition core hydraulic mismatch applicable to a core with Westinghouse and Exxon fuel types. This analysis is contained in Exhibit E. This analysis assumed an FQ of 2.30, FdH of 1.60, the'new upper internals

and the removal of thimble plugs. This analysis will bound the' first ' ,

i cycle of Westinghouse fuel operation for both units, up to a maximum ' ,

peak pin exposure of 22,000 MWD /MTU. This analysis will only'be i needed for one cycle operation. Approval of the new UPI (Upper Plenum ;

Injection) Model is expected for subsequent cycles (refer to our ~

letter dated 3/11/85 titled: Upper Plenum Injection IDCA IL. del A _ _ _ . _ _ _ _ _ . _ . . . . _ . _ . _ _ _ _ . _ . . _ _ _ _ _ _ . . . _ . . _ _ , . _ , _- _ .._ .. _., ,..__.2,_.c,,

Development).

The fuel parameters used as input for the LOCA analysis were generated using the Revised Thermal. PAD Model. Due to the use of the Revised Thermal PAD Model, Westinghouse has evaluated the effect of burnup on peak cladding temperatures (PCT) predicted for the LOCA through the maximum burnup level of cycle 11 using the currently approved LOCA models (1981 EM) as required Reference 18 of Exhibit E. At a peak pin burnup of 22,000 MWD /MTU (maximum burnup for either Unit during cycle 11), the burnup evaluation predicted a PCT of 1934*F compared with a PCT of 2098'F for the beginning-of-life (maximum densification) case, demonstrating that the time of maximum densification remains limiting in terms of peak clad temperature.

A revised large break LOCA analysis has been performed for the worst case break (CD - 0.4) for Exxon fuel. This analysis (submitted November 4, 1985) assumed an 79 of 2.32, FdH of l'.60, thimble plug removal and the new upper internals.

Subsequent to performing the Exxon fuel analysis, a more conservative examination of the effect of thimble plug removal was found to yield a slightly higher peak clad temperature. The results of this

. examination applied to the Exxon fuel continues to show a substantial margin to the the 22000F Appendix K acceptance criteria and remains bounded in terms of clad temperature by the limiting case of the Cycle 11 transition core.

These peaking factor limits are conservative for Unit 2 Cycle 10 (current cycle in operation). The use of the new upper internals and thimble plug removal assumption is conservative. The smaller water -

volume in the new upper internals makes the new internals the limiting upper internals (See 11/4/85 submittal). The removal of thimble plugs decreases the active core flow making it the limiting case.

While this change may result in some change in the consequences of a previously-analyzed accident or may change in some way a safety -

margin, the results are clearly within all acceptable criteria.

For these reasons, operation of the Prairie Island Nuclear Generating Plant in accordance with this proposed change will not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of

, accident from any accident previously evaluated; or i

j (3) involve a significant reduction in a mar 81 n of safety.

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5. Deletion of the third line segment of the K(z) curve Proposed Channe !

On Figure TS.3.10-3, delete the third line segment and extend the second line segment to the 12 foot level. !

4 Change the associated Bases.

! Pages affected: TS.3.10-9, Figure TS.3.10-5 i

f Reason For Channe

, This change will allow more F q flexibility in the top of the core.

Sianificant Hazards Evaluation

! This change is supported by recent small break IDCA analyses performed l- by Westinghouse on Westinghouse and Exxon fuel assemblies. The Westinghouse fuel analysis is contained in Exhibit F and the Exxon

fuel analysis is contained in Exhibit G. Both analyses show that the small break IDCA peak cladding temperature is approximately 1000 F and j support the removal of the third line segment of the K(z) curve.

! While this revision may change the consequences of a previously-j analyzed accident or may change in some way a safety margin, the results are clearly within all acceptable criteria.

For these rea: ons, operation of the Prairie Island Nuclear Generating Plant in accordance with this proposed change will not: .

l (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or

, (2) create the possibility of a new or different kind of

, accident from any accident previously evaluated; or

(3) involve a significant reduction in a margin of safety.

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P' t Exhibit B Prairie Island Nuclear Generating Plant License Amendment Request Dated January 13, 1986 Proposed Changes Marked Up on Existing Technical Specification Pages [

Exhibit B consists of the existing Technical Specification pages with the proposed changes written on those pages. Existing pages affected-by this change are listed below:

TS-x .

TS.2.1-1 TS.2.1-2 TS.2.1-3 (this page will be deleted)

TS.2.3-6 Figure TS.2.1-1 TS.3.1-17 TS.3.1-18 TS.3.3-1 TS.3.10-1 j TS.3.10-2 TS.3.10-8 -

TS.3.10-9 TS.3.10-10 TS.3.10-11 TS.3.10-13 TS.3.10-17 Figure TS.3.10-5 Figure TS.3.10-7 (this page will be deleted)

Figure TS.3.10-8 (this page will be renumbered to TS.3.10-7) l

t TS-x I REV-73 ?/2 3/05 C

APPENDIX A TECHNICAL SPECIFICA* IONS I

LIST OF FIGURES i

TS FIGURE TITLE 2.1-1 Safety Limits, Reactor Core. Thermal and Hydraulic Two Loop j Operation

, 3.1-1 Unit 1 and Unit 2 Reactor Coolant System Heacup Limitations 3.1.2 Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitations 3.1-3 Effect of Fluence and Copper Content on Shift of RT f"#

Reactor Vessel Steels Exposed to 550 F TemperatureNDT 3.1-4 Fast Neutron Fluence (E >l MeV) as a Function of Full Power Service Life 3.1-5 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity >1.0 uCi/ gram DOSE EQUIVALENT I-131 3.9-1 Prairie Island Nuclear Generating Plant Site Boundary for Liquid Effluents 3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for Gaseous Effluents 3.10-1 ' Required Shutdown Reactivity Vs Reactor Baron Concentration 3.10-2 Control 3:nk Insertion Limits 3.10-3 Insertion Limits 100 Step overlap with One Bottomed Rod

3.10-4 3.10-5 Insertion Limits 100 Step Overlap with One Inoperable Rod 3.10-6 Hot Channel Factor Normalized Operating Envelope Deviation from Target Flux Differenca as a Function of Thermal

} Power

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3.10-% V(Z) as a 7 unction of Core Height 4.4-1 7 Shield Building Design In-Lea' sage Rate 6.1-1 NSP Corporation Organization Relationship to On-Site Operating Organizations 6.1.2 Prairie Island Nuclear Generating Plant Functional Organization for On-Site Operating Group I

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) 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING 2.1 SAFETY LIMIT, REACTOR CORE Applicability Ap,11ies to the limiting combinations of thermal power, l rea ctor coolant system pressure and coolant temperature during operation.

Objjetive To maintain the integrity of the fuel cladding.

Specification i

i 1. The combination of thermal powerlevel, coolant pressure, i and coolant temperature shall not exceed the limits

-shown in Figure TS.2.1-1. The safety limit is exceeded if the point defined by the combination of reactor

, coolant system average temperature and power level is at any time above the appropriate pressure line.

j -

Basis l To maintain the integrity of the fuel . cladding and prevent j .

fission product release, it is necessary to prevent overheating of the cladding under all operating conditions. This is

accomplished by operating the het regions of the core within the nuclease boiling regime of heat transfer wherein the ;

heat transfer coefficient is very large anc the clad surface temocrature is only a few degrees Fahrenheit above the coolant l saturation temperature. The upper boundary of the nucleate l 7~e N-S A/B boiling regime is termed' departure from nucleate boiling (DNB) g,y/,/,i. e used& and at this point there is a sharp reduction of the heat transfe:

h'*^/ g" y de c .. dent, which would result in high clad temperatures and i the poss 'ity of clad failure. DNB is not, however, an

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W A:4-/ 4A/g com observable p amater during reactor operation. Therefore, the

,2 observable par eters; thermal power, reactor coolant tem-

' ; fi M'j 4, g/,,gj,vre perature and pre ure have been related to DNB through the W-3

/se/ DNB correlation. 3 The W-3.DNB correlation' haf been developed to predict the DNB flux and the location of DNB for axially and '

uniform and non-uniform heat flux distributions. The local W4h DNB heat flux ratio, defined as the ratio of the heat flux .

i that would cause DNB at a particular core location to the local i heat flux, is indicative of the margin to DNB. The minimum

value of the DNB ratio, DNBR, during steady state operation, i normal operational transients, and anticipated transients is.

limited to 1.30,. ?. = m ir ef 1

probabia.ity an a 954 confidence hev.X alcorrespond ( to anot that DNB will 95%

occur and is chosen as an appropriate margin to DNB for all

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) The solid curves of Figure TS 2.1-1 represent the loci of points of thermal power, coolant pressure, and coolant average tempera-ture for which either the coolant enthalpy at the core exit is limiting or the DNB ratio is limiting. For the 1685 psig and ;

1985 psig curves, the coolant average enthalpy at the core exit is wqual to saturated water enthalpy below power levels of 914 l and 74% respectively. For the 2235 psig and 2385 psig curves, the coolant average temperature at the core exit is equal to 4

6500F below power levels of 644 and 73% respectively. For all four curves, the DNBR is e7 1 '- 1.3 at higher power levels.

Theareaofsafeoperation}isb su;o tom ese curves.

The plant conditions required o violate the limits in the lower Power range are precluded by the self-actuated safety valves on i

the steam generators. The highest nominal setting of the steam

' generator safety valves is 1129 psig (saturation temperature 5600F). At zero power the difference between primary coolant i and' secondary coolant is zero and at full power it is 500F. The reactor conditions at which steam generator safety valves open 4

is shown as a dashed line on Figure TS.2.1-1.

Except for special tests, power operation with only one loop or j tmerva /". with natural circulation is not allowed. Safety limits for such special tests will be determined as a part of the test procedure.

The curves are :;d en the following nuclear het channel factorsf*T:

8 = -h-Se 'l + 4,-iH1-P) ; and [ = 2."1 As g g -

-s-m Use of these factors results in more conservative safety limits j

l than would result * - - '

from power distribution limits in Specifica-tion TS.3.10. erhir- f :t::: d:: --

--' inclre-f fr er h5t 9 I r:i firterr, N .ft: 1 frrrific: tie: ,

This combination of hot channel factors is higher than that cal-l culated at full power for the range from all control rods fully 1 withdrawn to maximum allowable control rod insertion. The control rod insertion limits are covered by Specification 3.10. Adverse power distribution factors could occur at lower power levels because additional control rods are in the core. However, the control rod insertion limits specified by Figure TS.3.10-1 assure that the DNB ratio is always greater at part power than at full power.

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s TS.2.1-3 The Reactor Control and Protective System is designed to prevent any anticipated combination of transient c itions that would result in a DNB ratio of less than 1.30 .

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S.2.3-6 REV 31 L 1/ H The other reactor trips specified in A.3. above provide additional protection.

The trip initiated by steam /feedwater flow nismatch in coincidence with low steam generator water level is designed for protection from a sudden loss of the. reactor's heat sink. The safety injection signal trips the reactor to decrease the severity of the accident condition. The reactor is tripped when the turbine generator trips above a power level equivalent to the load rejection capacity of the steam dump valves. This reduces the severity of the loss-of

'/are.hedu <u load transient. h ,i .g u ;'

,^!p 4 W Ls a Se positive power range rate trip pr ides protection against rapid flux increases which are characteristic of rod ejection events from any power evel. Specifically, this trip complements the power range nuclear flux hignand low trip to assure that the criteria are net for rod ejection from partial power.

JyIb4 The negative he'ser range rate trip provides p etion satisfy ; all IEEE criteria to asstre that minimum DNBR is maintained e .1 308?or all multiple control rod drop accidents.- Analysis indicates (Section 4.1 3) that in the case of a single rod drop, a return to full power will be indi:: M by the automatic reactor control system in response to a continued full power turbine load demand and it will not result . Thus, automatic inaDNBRoflessthan13(ivelimitsin protection for a single rod drop is not required. Administrat Specification 3.10 require a power reduction if design power distribution limits are exceeded by a single misaligned or dropped rod.

References:

(1) FSAR 14.1.1

. (2) FSAR Page 14-3 (3) FSAR 14.2.6 o/

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(6) FSAR 7.2, 7.3 uM d bdVC' /*I7 4r (7) FSAR (8) FSAR 321 14 1 9 g 7[,,yg M (9) FSAR 14.1.11

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u-4-._ mmes9 at _ .

,ao ::: ...:

.4. . : . . @. . . t. . .f @,

_:. d.::. ..

t:

i!:e:1:

t=l=::!::q=. t=

-:==_:l- : .== =ut_...=j:=:tnel==

= F: . t: p=uj=un:.:u

. ar. j":..:p

~ . ::l=: p.:). .. ~.-

[..:

l .._

.:. . p: ::-t :..:l; n ===v j-

. : ~u r;u.:.4-""

=v --"! t-- -- -l: ..a - ---" - " - --I..a=

= x.::._: -" I . a: " " :- n " tn. =:=!:: * .

ll!!!! @' I!! !llU!ll- ll!:ll}E5lNi!!!: l !!!$li I; I I!! IE : !lE! :ii' i !

520 A-O 20 40 60 80 100 120

% Rated Core Power .

SAFETY LIMITS, REACTOR CORE, THERMAL AND HYDRAULIC WO-LOOP OPERATION

)

FIGURE TS.2.1-1

9 e l l

l l

1 l

TI.3 1-17 l l

REV ~3+ 4/20/7^

{.

l F. MINIMUM CONDITIONS FOR CRITICALIM d/,;4,f d c 3./-/7 Soectitcation f,, f.g,,,, pef spes h5co (5 f ~

1. reactor shall be made critical only at or above the coolan temper and remains 4 4t which for fegative the any following coolantreactivity temperaturecoefficiease p (r"except negative during lov gover ph tes ts):

(a) Moderator terperat ef. t for a reactor loaded with Westinghous e only.

sothernal temperature coefficient for a reacto *her full or partially loaded with Exxon fuel.

2. The reactor shall not be brought to a critical condition until the reactor coolant temperat2re is higher than that defined by the criti-cality limit line shown in Figure'TS.3 1-1.

3.- "..:n -k . reactor coolant temperature is below the min 4a"- _ gYa~ cure as specified in A. .L .1, *he reactor u critical by an amount equal to or r * . tne p _' *==n ivity insertion due to r ant depressurization. -

Basis At the beginning of a fuel cycle the ecderator temperature ccmfficient has its =ost positive or least negative value. As the boren cent.=ntration is reduced throughout the fuel cycle, the moderator temperature > oefficient becemes more negative. ~h; ;;fe:) enely;;; conducted for ,"z.i.ie .el.ul

"*'*" " "--Ti--h^r: . zl ;;;u.;d ; - s;;1:1.; _________. i y .......

fici--r. The isothermal te=perature coefficient is defined as the reactivity change associated with a unit change in the e.oderator and fuel te=peratures. Essentially, the isother=al temperature coefficient is the sum of the moderator and fuel temperature coefficients. This coefficient

'- "a a*"**~ *---#~ ""-*--- *"- -^d=*~^-

is measured

'*"-- ccdirectly

icie-2during tr 2r r"herred

' pr :::::: S :--ir:d tf :;t;;;;;in;;

er predic -f f re'_

__,4-os <e,.w-,*

t-......,.sA.m.,s4-4

er2tu-/ creicient f:: :n er;::in n;;11j d::::

/. g ,. p p4 es feb >~ *

s ,,;f wl/ial pieJahks.

t

TS.3 1-18 i

REY 34 './ Z/79

~^

sJf gade,~d

&ns useih G 1Sc & 0x*cocf "r

N

.de hothrpal l5 o'h1 SpaaWEd S 3 b f l* .

e ~

For extended optimum fuel urne it is necessary to either oad the reactor vith burnable po ons o increase the baron concentration in the reactor coolant system. If the eter approach is emphasized, it is possible that a positive -modeeeeee toeperature coefficient could exist at beginning of cycle (BOC). '- E- '-- =1*h '--- '"-1. r -' e*7 --- ' re re 2rc tr:frr H rrr '-- 2 ;::itir: xdx_:x : , xn_. : n;ffi i r:- 5:::

.t ...s#4,4--. ,a s ---. 4-- e. -- .5i .a-

---1 7- - 7,.a 4 ,. , ch. 4...w re, tr rere ;: :r ::sfitic _ Other conditions, e.g., higher power or partial rod insertion would cause the isothermal coefficient to have a acre negative value. These analyses demonstrate that applicable crite'ria in the NRC Standard Xaview Plan (NUREG 75/087) are met.

Physics maa.aurements and analyses are conducted. during the reload startup test program to (1) verify that the plant will operate within saf ety analyses assumptions and (2) establish operational procedures to ensure saf sty analyses assumptions are met. The 3 1.F.1 requirements are waived during low power physics tests to permit measurement of reactor temperature coefficient and other physics design parameters of interest. Special operating precautions will be taken during these pgyjics tests. In addition, the strong negative Doppler cod.!ficient and the small

(-

integrated Ak/k would limit the magnitude of a power excursion resulting from a reduction of moderator density. ,

The requirement that the reactor is not to be made critical except as specified in figure TS.3 1-1 provides increased assurance that the proper relationsnip between reactor coolant pressnre and temperature will be maintained during system heatup and pressu.iza-lon whenever the reactor vessel is in the nil ductility temperature range'. Heatup to this tempera-ture will be accomplished by operating the reactor coolant pumps and by

'the pressurizar heaters. The pressurizer heater and associated power cables have been sized for continuous operation at full heater power. The shutdown margin in Specification 3 10 precludes the possibility of accidental criticality as a result of an ' crease of moderator temperature or a l decrease of coolant pressure.gg Referencest

(

'A

-, (1) FSAR Figure 3.2-10

, (2) FSAR Table 3 2-1

~

J (V ,

'l e TS.3.3-1

( i REV 51 2/2/"2

, 3.3 ENGINEERED SAFETY FEATURES

\

Applicability

. Applies to the operating status of the engineered safety features.

objective To define those limiting conditions that are necessary for operation of engineered safety features: (1) to remove decay heat from the core in an emergen,cy or normal shutdown situations, and (2) to remove heat from containment in normal operating and emergency situations.

Specifications A. Safety Iniection and Residual Heat Removal Svstems

1. A reactor shall not be made or maintained critical not shall it be heated or maintained above 200*y unless the following conditions are satisfied except as permitted in Specificiation 3.3. A.2.

i, 4

, a. The refueling water tank contains not less than 200,000 4 gallons of water with a boron concentration of at least

,. 1950 ppm.

i l b. Each reactor coolt't-system accumulator shall be opersble j y when reactor cooltat system pressure is greater than 1000 psig.

Operability requires: igio A 20 (1) The isolation valve is open (2) Volume is ter ;;; 125^ ;;d 12"2.^ cubic feet of borated water (3) A minimum boron concentration of 1900 ppm (4) A nitrogen cover pressure of at least 700 psig

c. Two safety injection pumps are operable except that pump control switches in the control room shall meet the require-ments of Section 3.1.G whenever the reactor coolant system temperature is less than MPT.

d. Two residual heat removal pumps are operable.

e. Two residual heat exchangers are operable.

f. Automatic valves, interlocks and piping associated with the above components and required to function during accident

, , , conditions, are operable.

i p; V.arual valves in the above systems that could (if one is improperly positioned) reduce injection flow below that

' assumed for steident analyses, shall be blocked and tagged in the prope' position for injection. RHR system valves, however, may be positioned as necessary to regulate plant v heatup or cooldown rates when the reactor is suberitical.

All changes in valve position shall be under direct admini-strative s.ontrol.

s .

TS 3.10-1

' REV Sf, 1^/3/^3 l

, 3.10 CONTROL RCD AND POWER DISTRIBUTION, LIMITS Aeolicability 1 j l

Applies to the lisits on core fission power distribution and to the limits on l control rod operations.

I Obiective - l 1

To assure 1) core subcriticality after. reactor trip, 2) acceptable core power distributions during power operation, and 3) Limited potential reactivity in- .

sortions caused by hypothetical control rod ejection.

Soecifiestion -

A. Shutdown Rasceivity l The shutdown margin with allowance for a stuck control rod assembly shall exceed the applicable value shown in Figure TS.3.10-1 under all steady-state operating conditions, except for physics tests, from zero to full power, including effects of etial power distribution. The shutdown margin as used here is defined as the asount by which the reactor core would be suberitical at hot shadown conditions if all control rod assemblies were tripped, assuming j that the highest worth control rod assembly remained fully withdrawn, and assuming no changes in xenon or baron concentration.

B. Power Distribution Limits

1. At all times, except during law power physics testing, snasured as defined b elow and in the hot bases,< hannel f actors, shall meet the!T'allowing.

and Th,'imits:

T x 1.03 x 1.05 i (2.;-/?) x :( ; ;'; ;- ) (2.30/P) X K(b

,T, AH x 1.04

<*""m ' - -'m1'-

6 a r-x 0 '3(1-P)3 where the following definitions apply:

-fe-) K(2) is the axial dependence function shown in Figure TS.3.10-5.

41>)- is the core height location.

ra r ,_ .u. _..,,_.._....,,___....,_n...

. __; , e- . .u , - u

."..,~~ d -'"~~' '" '~~ ~" ' ~

' ~

[ 'y'h' $_

Q (d) ?'f(r ) i: th e --- ' i - -

- - ---- 4--- >--c- '"--d~- 'a-Ta*

- -f hel :: C:- ; ; ' :1 ch -- i: Fi; :: '"?.2.10 '.

/ .'..,.u..._._

s._..__., w

__,e_ 3, _ . __ e_

-.044 P is theyraction of full power at which the core is operating.

In the T qlimit decernination when PJ< .50, set P = 0.50. ,

(2.21/?) ch:11 he ur-d f er "r--1 ;hruce ::-f lire.

-- .-. _ -- . - .- . - .=. -

e .

TS.3.10-2

% \REV',', 10/3/03

-(ft [ or [ is defined, as the measured F or F respectively, w9th thhsmallest margin or greatest ekcess 0, limit 43-)- 1.03 to the is the engineering measured T' to acenunthot for channel manufacturi f actor, [$g applied tolerance.

4h)- 1.05 is applied to the measured F to account for measurement uncertainty.

44- 1.04 is applied to the measured [# to account for measure-ment uncertainty

2. and [ shall be measured and the target

' Hot channel flux difference determ factors, [Sned, 'ahe,quilibrium conditions according to the following conditions, whichever occurs first:

(a) At least once per 31 effective full-power days in conjunction with the target flux difference determination, or (b) Upon reaching equilibrium conditions after exceeding the reactor power at which target flux difference was last determined, by 10%

or more of raced power. ,

(equil) shall meet the following limit for eh =4dd1= mv4a180%

[!thecore:

o (g So/p') x KUth F N q (equil) x V(2) x 1.03 x 1.051-(2.';1/r; a :(2) 2 !"(If where V(2) is defined Figure 3.10-[ and other terms are defined in 3.10.3.1 above.

3. (a) If either measured hot channel factor exceeds its limit specified in 3.10.3.1, reduce reactor power and the high neut::en flug trip set-point by 12 for each percent that the measured T or r g exceeds the 3.10.3.1 11mit. Then follow 3.10.3.3(c).

(b)

If thethe measured 3.10.3.1 11m1 [9,(equil) take one of exceeds the followingtheactions:

3.10.3.2 limits but not

1. Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months place the reactor in an equilibrium configura-tion for which Specifier.cion 3.10.3.2 is satisfied, or

2. Reduce reactor power and the high neutron ' lux trip setpoint by 1% for sech percent that the measured T'q x 1.05 x V(2) exceeds the ;^. 3"/I; . R( ; 0(equil)

. 2"(2 ; xlimit.

1.03 3 l by s.ss % ,.r.e ud pu c~+ dd A l

me.aswd y e

-m - ,.- - ,-

e ,

TS.3.10-8 RIV 42 12/17/",0

.~.,

3. If one or both of the quadrant power tilt monitors is inoperable, individual upper and lower excore detector calibrated outputs and the calculated power tilt shall be logged every two hours af ter a load change greater than 10* of rated power.

J. DNB Parameters the following DNB related parameters limits shall be maintained during power operation:

a. Beactor Coolant System Tavs 1564 F

b. Pressurizer Pressure 12:20 psia *

c. Reactor Coolant Flow 1178,000 spa With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce thermal power to less than 5*.

of rated thermal power using noi, mal shutdown procedures.

Compliance with a and b. is demonstrated by verifying that each of the parameters is within its limits at least once each 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

! Compliance with c. is demonstrated by verifying that the parameter is within it's limit af ter each refueling cycle.

Bases Throughout the 3.10 Technical Specifications, the terms " rod (s)" and "RCCA(s)"

are synonomous.

Shuedown Reactivity Trip shutdown reactivity is provided consistent with plant safety analyses as sumotions . Oce percent shutdown is adequate except for the steam break analysis, which requires more shutdown reactivity due to the more negative moderator temperature coefficient at end of life (when boron concentration is low) . Figure TS.3.10-1 is drawn accordingly.

Power Distribution Control The specifications of this section provide assurance of fu-l integrity j,. fxrou A/reder- during Condition I (Normal Operations) and II (Incidents of Moderate freousney) events by: (a) maintaining the minimum DNBR in the core fgf gg/,11$,. < >1.30'during normal operation and in short ters transients, and (b)

Timiting the fission gas release, fuel pellet temperature and cladding i

A/chu. 7 b'f6 4f .

Limit not applicable during either a THERMAL POWER ramp increase in excess of (5%) RATED THERMAI. POWER per minute or a THERMAL, POWER step l

increase in excess of (10~) RATED THERMAL POWER.

t i

{

l I

o .

1 a

TS.3.10-9 RES. ,, .. . , . . , -

mechanical properties to within assumed design criteria. In addi: ion, limiting the peak linear power density during Condition 1 events provides assurance that the initial condi:1ons assumed for :he LOCA analyses are met and the ICCS acceptance criteria limit of 2200*F is not exceeded.

Dyring eqeration, :he plant staff compares the measured hot channel factors, i i

F' andf OCA anahs,es(described later) to the limit determined in the. transient and Th I' ':ing i = includ:: r n ::: ---- : , uginnring, '

l ni;;ini:n1 :;n::in:ie:.- 0*he terms on :he right side of the equations in section 3.10.3.1 represent the analytical limits. Those terms on the i

lef: side represent the measured het channel factors corrected for engineer-i 1

ing, calculational, and measurement uncer:ainties.

0 ,, .___. ^*-Broendeu; 'he:

Tis ik Cte..u.; Ten n , i. ;; fin d n :M ----

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

, _ _ .n

- ..s -----

w b- :h: nn:g: ___fn i ::d E h_ ___ _=--i fh1 ;-.11 :: nf ::d:. '"h : --- ' rr rD;; f : naufenning ::',::::::e-en-

- 22

.0 Le Lene-teeee<. :: . Tai: r:1:: i ::: ri:::d iurcher h- '" t iD) , ' " ' '

--d f:---ir-- d ::ribed b:12-_ - T: : d :' -h::: -h r":: f:: :::

j

he K(2) fune:1on shown in Figure TS . . 310-5 is a normalized-uc f n :1on tha:

limits ? -t9t axially, f:: -h::: : ;ien . The K(0) specified for the icwest '

six(6) hee: of the core is arbi:rarily flat since :he lower part of the core is generally not 11=1:ing. Above that region. the K(Z) value is based on large and s=all break LOCA analvses,f 7 m ___ ____m__ ___ 0_ . ..;. t_ =2 2.._,__

_ _ _ . . ' _ . .G. 2 -- ~ =

f::__,,...,,,...u.

i: : ; f .... - - - - i: n:n::d n :::nn :- ::::il- :::: :-c br-9

?=GGAev--

_--..,e m ___,__ _._ ____ ._

........m

- - --,. gr-= . .... ..... . ..,,... . ...._ - _- .m___,___-. . ..___

~ .....

.=-_-...=t ___ __z_ _ _ t____., _ _ _ _. - ,. . .. . . . ,. . . --- =*_,. .._ .,._._ .....-_. .-.

m .e

__?..___ _-__,i__ . *. _ . . _ . _ . _

_ _ _ _ _ _ _ _ _ _ _ _ _ _ . .t_.s__ :.._ _ . . _ . . . _ _

. . . . '7. . . ' ! 7 ' ;. 5-~ ' ?.' Z -- " ~~ ~ ' ' -" ' '-" ~ ~ ' "'- " ' -' _ N -

N , na ,.< w s ef61 cod 70 is the casured Nuclear liet sannel Factor, defined as the naximum local hea: flux'i :::: divided by the average'hea: flux in the core. Hea:

fluxes are derived from measured neu ron fluxes and fuel enrichmen:.

V(2) is an.. axially dependen: func:Lon applied :o che equilibrium measured F'N to bound ?"O's :ha: could be measured ar non-sou111brium condi: ions. This function is based on power dis:ribu: ion con:rol analyses :ha: evaluated :he effec: of burnable poisons, rod posi: ion, axial effee:s, and xenon wor:h.

e F"",

Incineerine Hea: Flux Het Channel Fsetor. is defined as the allowance on heat flux required fer =anufac:uring :olerances. The engineering fac:or j

allows for local variations in enrichment, pellet densi:y and dia=e:er, ,

surface area of :he fuel rod and eccen:ricity of the gap between pelle: and j clad. Cembined sta:is:1cally :he ne: affec: is a fac:or of 1.03 :o be applied to fuel rod surface hea: flux. '

}

Y

)

1 -- -, --. . . _ . _ . . - _ . , - - - - . . .

e .

TS.3.10-10 REV i: *0/D/10.

The 1.05 multiplier accounts for uncertainties associated with measure-ment of the power distribution with the moveable incore detectors and the use of those asseurements to establish the assembly local power gy .g g distribution.

Ndelear Enthalev Rise Hot Channel Factor, is defined as the ratio

8"

/ /4,/Af

the Integral of linear power along tne rod with the highest integrated I di4nc/4 power to the average rod power.1 T is based on an integral and f'

- a usea as such in the DNB calcula tons. 1.ocal heat fluxes are

' cp,4 # gg

,jf.

ob ined by using hot channel and adjacent channel explicit power shape ich take into account variations in horizational (x y) power

- f ,/

'i 7 shanes thqghout the core. Thus the horizontal power shape at the pgint of 4x' heat flux is not necessarily directly related to

@,fy,: ode AH*

In the specified limit fF there is an 8 percent allowance for un-certainties which means normal cperation of the core is expected to N

result in fl.55/1.08. Th logic behind the larger uncertainty in this cas is that:

(a) abnormal y perturbations in the radia over shape (e.g. ref misalignment' af fect Tg, in most cases without nece rilyaffectingFf, (b) the operator has a direct influence on F th h movement of rods, and cag limit it' to the desired value, ubile he no direct control overjkB and, (c) an error in the predictions for radial power shape, which v be dqtected during startup physics tests can be connensated for T by tighter axial control, but compensation for (g is less 1 rkadilyavailable.

, y l When a measurement of [ is taken, n :r ' :::1-error must be allowed for and 4 percenk is the appropria allowance for a full core uap taken with the movable incore detector flux ma pi s em.

m. .w < mt- <

Measurements of the hot channel f actors are requi. rec ~as rt of startup

- physics tests, at least once each effective full power month of operation, and whenever abnormal power distribution conditions require a reduction of core power to a level based on measured hot channel f actors. The incore map taken follovir.r initial loading provides confirmation of the basic nuclear design bases including proper fuel loading patterns. The periodic monthly incore mapping provides additional assurance that the nuclear design bases remais inviolate and identify operational anomalies which would otherwise affect these bases.

For normal operation, it is not necessary to measure these quantities.

Instead it has been determined that, provided certain conditions are observed, the hot channel f actor limits will be met; these conditions are as follows:

! 1. Control rods in a single bank move together with no individual rod insertion differing by more than 15

}

/

- - - - - r n om- r-- , , , , -aw - n. g y my r-- - - - - , e-- -

TS.3.10-11 REV 00 10/0/00 inches from the bank demand 'p' o sition. An accidental misalignment limit of 13 steps precludes a rod misalign-ment greater than 15 inches with consideration of maximum ;

instrumentation error.

2. Control rod banks are sequenced with overlapping banks as described in Technical Specification 3.10. ,

3. The control bank insertion limits are not violated.

4 Axial power distribution control procedures, which are given in terms of flux difference control and . control bank insertion limits are observed. Flux difference refers to the difference in signals between the top and bottom halves of two-st.ction excore neutron detectors. The flux difference is a measure of the axial offset which is defined as the difference in normalized power between the top and bottom halves of the core.

The permitted relaxation in T N N and T allows for radial power AR shape changes witti rod insertion to the insertion limits. It has been determined that provided the above conditions 1 through I. are obse-ved, y these hot channel f actor limits are met. In specification 3.10, F arbitrarily limited for F < 0.5 (except for low power physics testk).is The procedures for axial power distribution control referred to above are designed to minimize the effects of xenon redistribution on the axial power distribution during load-follow maneuvers. Basically i

control of flux difference is required to limit the difference between the current value of Flux Difference ( *AI) and a reference valua which corresponds to the full power equilibrium value of Axial Offset -

(Axial offset = *AI/ fractional power). The reference value of flux difference

8 varies with power level and burnup but expressed as axial set it varies only with burnup.

60 /'M' 1 The echnical specifications on power distribution control assure that the r T .0.p app;. tu? ; _1:;; cf 2."'? W:: Fi p :: ?? ?.1^-5 '

.0 7 is not exceeded and xenon di'stributions are not developed i

which at a later time, would cause greater local power peaking even though the flux difference is then within the licits specified by the procedure.

The target (or reference) value of flux difference is determined as follows: At any time that equilibrium xenon conditions have been established, the indicated flux difference is noted with the full length rod control rod bank more than 190 steps withdrawn (i.e. ,-

normal full power operating position appropriate for the time in life, usually withdrawn farther as burnup proceeds). -This value, divided by the fraction of full power at which the core was operating is the full power value of the target flux difference. Values for all other core power levels are obtained by multiplying the full power value by the fractional power. Since the indicated equilibrium was noted, no allowances for excore detector error are necessary and indicated

} deviation of :5 percent: AI are permitted from the indicated reference value. Figure TS.3.10-6 shows the allowed deviation from the target flux difference as the function of thermal power.

l l

. o l 1

1

- ^ TS.3.10-13 m

i [ hrs REV i: 10/ " /20 resulting from operation withic ne target band. he consequences of being outside the +5% target b d but within the Figure TS.3.10-6 limit I for power levels between 50% d 90 has been evaluated and determined to result in' acceptable * 'd ::1;;;. Therefoze, while the deviation exists the power level iskisited to 90 percent or lower depending on the indicated axial flux difference In all cases the +5 percent target band is the Limiting Condition fo- Operation. Only when the target band is violated do the limits under Figura TS.3.10-6 apply.

controlled If, for any reason, the indicated axial flux difference is not within the 2 percent 5 band for as long a period as one hour, then menon distributions may be significantly changed and operation at 50 percent is required to protect against potentially more severe consequences of some accidents.

As discussed above, the essence of the procedure is to maintain tLe xenon distribution in e.he core as close to the equilibrium full power condition as possible. 'his is accomplished by using the baron system to position the full length control rods o produce the r uired i ic flux l

difference. g g g,g g 6 g 4.c For Conditio II events th d from overpower and a minimum DNBR of 1.30 by an automatic protection system. Compliance with operating procedures is assumed as a precondition for Condition II trasients, however, operator error and ecuipment malfunctions are separately assumed to lead to the cause of the transients considered.

Quadrant Power Tilt Limits Quadrant power tilt limits are based on the following considerations. Fre-quant power tilts are not anticipated during normal operation since this phenomenon is caused by some aavsmsetric perturbation, e.g. rod misalignment, x-y xenon transient, or inlet temperature mismatch. A dropped or misaligned rod will easily be detected by the Rod Position Indication System or core instrumentation per Specification 3.10.7, and core limits protected per .

Specification 3.10.E. A quadrant tilt by'some other wtans (x y xenon tran-sient, etc.) would not appear instantaneously, but would build up over

' several hours and the quadrant tilt limits are set to protect against this situation. They also serve as a backup protection against the dropped or misaligned rod.

Operational experience shows that normal power tilts are less than 1.01.

Thus, sufficient time is available to recognize the presence of a tilt and correct the cause before a severe tilt could build up. During stiart-up aad power escalation, however, a large tilt could be initiated.

Therefore, the Technical Specification has bet.n written so as to prevent escalation above 50 percent power if a large tilt is present.

)

,e----,-m -

. TS.3.10-17 RIV i; ' /17/"^

If the rod position deviation monitor and quadrant power tilt monitor (s) are inoperable, the overpower reactor trip setpoint is reduced (and also power) to ensure that adequate core protection is provided in the event that unsatis-factory conditions arise that could affect radial power distribution.

Increased surveillance is required, if the quadrant power tilt monitors are in operable and a load change occurs, in order to confirm satisfactory power distribution behavior. The automatic alarm functions related to quadrant power tilt must be considered incapable of alerting the operator to unsatis-factory power distribution conditions.

DNB Parameters The RCS flow rate, T and Pressurizer Pressure requirements are based on trans ient analyses a!aEm,ptions. The flow rate shall be verified by calorimetric fl o'v data and/or elbow taps. Elbow taps are used in the reactor coolant system as an instrument device that indicates the status of the reactor coolant flow.

The basic function of this device is to provide L2 formation as to whether or not a reduction in flow rate has occurred. If a reduction in flow rate is indicated below the specification value indicated, shutdown is required to investigate adequacy of core cooling during operation.

7: 're re;i r- -i th ' i;' ' r rr ; , t' e

.;7 . c: .:-. ..-a..... g....s. ..a..... --'_r _ pi

. .. -- c' 4'i. ci'-

r --a... -" -- --' '" # ' d -

....wist.. -w...

_.._u:..a u....... .se.... .2.... , . . . s e ': . : . . . N . . . ... . . . . . : 2-.. .-> u--

~~

y...s.i.. C__;Q ' . .i_i . .... ~ . ~AE ._. ._.}

w ~ ; c A A" nAA 24v-- ~ ~

i

. . l TICURE TS.3.10-5 l RIV $5 10 /2/33 1.2 (6.0,1.0) _

(

1.0 m

~

/

m (/$.0,0.S:

3_

'L 0.8 _

n --

[. U K(20 _

'^

5 0.6 - =.. -

=

. 4 0.4 }12

,. OQ.

l 0.2 -

0.0 0 2 4 6' 8 10 12 Core Height (f:)

g H02 CHANNEL TACTOR NORMALIZED l 0?ERATn;G ENVELOPE T:! ! E. 22 -

Q f

- ,-- , , - - -- - . . . - , - - , - -,--wn-.m,-w - - ~ - . - -

FIGURE TS.3.10-7 REV 67 12/28/83

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Exhibit C Prairie Island Nuclear Generating Plant License Amendment Request Dated January 13, 1986 Revised Technical Specification Pages Exhibit C consists of the proposed Technical Specification pages with the changes shown in Exhibit B incorporated. The proposed pages are listed below:

TS-x TS.2.1-1 TS.2.1-2

TS.2.3-6 Figure TS.2.1-1 TS.3.1-17 TS.3.1-18 TS.3.3-1 TS.3.10-1 TS.3.10-2 TS.3.10-8 TS.3.10-9 TS.3.10-10 -

TS.3.10-11 TS.3.10-13 TS.3.10-17 Figure TS.3.10-5 Figure TS.3.10-7 **

TS.2.1-3 will be deleted with this change.

4

- Exinting Figure TS.3.10-7 will be deleted by this change. Existing Figure TS.3.10-8 will be renumbered to TS.3.10-7 i

i

r

a .

TS-x REV APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES TS FIGURE 2.1-1 Safety Limits, Reactor Core Thermal and Hydraulic Two Loop Operation 3.1-1 Unit 1 and Unit 2 Reactor Coolant System Heatup Limitations 3.1-2 Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitations 3.1-3 Effect of Fluence and Copper Content on Shift of RTNDT ' #

Reactor Vessel Steels Exposed to 550*F Temperature 3.1-4 Fast Neutron Fluence (E >l MeV) as a Function of Full Power Service Life 3.1-5 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary l

Coolant Specific Activity >1.0 uC1/ gram DOSE EQUIVALENT I-131 3.9-1 Prairie Island Nuclear Generating Plant Site Boundary for Liquid Effluents 3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for Gas Effluents 3.10-1 Required Shutdown Reactivity Vs Reactor Boron Concentration 3.10-2 Control Bank Insertion Limits 3.10-3 Insertion Limits 100 Step Overlap with One Bottomed Rod 3.10-4 Insertion Limits 100 Step Overlap with One Inoperable Rod 3.10-5 Hot Channel Factor Normalized Operating Envelope 3.10-6 Deviation from Target Flux Difference as a Function of Thermal Power 3.10-7 V(Z) as a Function of Core Height l 4.4-1 Shield Building Design In-Leakage Rate 6.1-1 NSP Corporation Organization Relationship to On-Site Operating Organizatient 6.1-2 Prairie Island Nuclear Generating Plant Functional Organization for On-Site Operating Group

TS.2.1-1 REV 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING 2.1 SAFITY LIMIT, REACTOR CORE Applicability Applies to the limiting combinations of thermal power, reactor coolant systes press.re and coolant temperature during operation.

Objeer3ve To maintain the integrity of the fuel cladding.

Specification

1. The combination of thermal power leve'1, coolant pressure, and coolant temperature shall'not exceed the limits shown in Figure TS.2.1-1. The safety limit is exceeded if the point defined by the combination of reactor coolant system average temperature and power level is at any time above the appropri :e pressure line.

Basis To maintain the integrity of the fuel cladding and prevent fission product release, it is necessary to prevent overheating of the cladding under all operating ceiditions. This is accomplished by operating the hot regions of the cc within the nucleate boiling regime of heat transfer wherein the heat transfer coefficient is very large anu the clad surface temperature is only a few degrees Fahrenheit above the coolant saturation temperature. The upper boundary of the nuc1'eate boiling regime is termed departure from nucleate boiling (DNB) and at

, this point there is a sharp reduction of the heat transfer coefficient, which would result in high clad temperatures and the possibility of clad failure. DNB is not, however, an observable parameter during reactor operation. Therefore, the observable parameters; thermal power, reactor coolant temperature and pressure have been related to DNB through the W-3 and WRB-1 DNB correlations. The W-3 DNB correlation is used for Exxon Nuclear fuel. The WRB-1 DNB correlation is used for Westinghouse fuel. The W-3 and WRB-1 DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The minimum value of the DNB rat $o, DNBR, during steady state operation, normal oper'.tional transients, and anticipated transients is limited to 1.30 for the Exxon Nuclear fuel and to 1.17 for the Westinghouse fuel. l These limits correspond to a 95% probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

t TS.2.1-2 REV The solid curves of Figure TS 2.1-1 represent the loci of points of thermal power, coolant pressure, and coolant average temperature for which either the coolant enthalpy at the core exit is limiting or the DNB ratio is limiting. For the 1685 psig and 1985 psig curves, the coolant average enthalpy at the core exit is equal to saturated water enthalpy below power levels of 91% and 74% respectively. For the 2235 psig and 2385 psig curves, the coolant average temperature at the core exit is equal to 650*F below power levels of 64% and 73%

respectively. For all four curves, the DNBR is limiting at higher power l levels. The area of safe operation is below these curves.

The plant conditions required to violate the limits in the lower power range are precluded by the self-actuated safety valves on the steam generators. The highest nominal setting of the steam generator safety valves is 1129 psig (saturation temperature 560*F). At zero power the difference between primary coolant and secondary coolant is zero and at full power it is 50*F. The reactor conditions at which steam generator safety valves open is shown as a dashed line on Figure TS.2.1-1.

Except for special tests, power operation with only one loop or with natural circulation is not allowed. Safety limits for such special tests will be determined as a part of tlw test procedure.

The curves are conservative for the following nuclear hot chaanel factors:

F = .6 1 + 0.3(1-P)] ; and F = 2.30 q

aH Use of these factors results in more conservative safety limits than would result from power distribution limits in Specification TS.3.10.

l This combination of hot channel factors is higher than ths; calculated at full power for the range from all control rods fully wfchdrawn to naximum allowable control rod insertion. The control rod insertion limits are covered by Specification 3.10. Adverse power distribution factors could occur at lower power levels because additional control rods are in the core. However, the control rod insartion limits specified by Figure TS.3.10-1 assure that the DNB ratio is always greater at part power than at full power.

The Reactor Control and Protective System is dasige.ed to prevent any anticipated combination of transient conditier0 trac would result in a DNB ratio of less than 1.30 for Exxon Nuclear fuel and less than 1.17 l for Westinghouse fuel.

TS.2.3-6 REV The other reactor trips specified in A.3. above provide additional protection.

The trip initiated by steam /feedwater flow mismatch in coincidence with low steam generator water level is designed for protection from a sudden loss of the reactor's heat sink. The safety injection signal trips the reactor to decrease the severity of the accident condition. The reactor is tripped when the turbine generator trips above a power level equivalent to the load rejection capacity of the steam dump valves. This reduces the severity of the loss-of-load transient.

The positive power range rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power level. Specifically, this trip compliments the power range nuclear flux high and low trip to assure that the criteria are met for rod ejection from partial power.

The negative power range rate trip provides protection satisfying all IEEE criteria to assure that minimum DNBR is maintained above 1.30 for Exxon Nuclear fuel and above 1.17 for Westinghouse fuel for all multiple control rod drop accidents. Analysis indicates (Section 14.1.3) that in the case of a single rod drop, a return to full power will be initiated by the automatic reactor control I system in response to a continued full power turbine load demand and it will not result in a DNBR of less than 1.30 for Exxon Nuclear fuel and 1.17 for Westinghouse fuel. Thus, automatic protection for a single rod drop is not required. Admini-strative limits in Specification 3.10 require a power reduction if design power distribution limits are exceeded by a single misaligned or dropped rod.

References:

(1) FSAR 14.1.1 (2) FSAR Page 14-3 (3) FSAR 14.2.6 (4) FSAR 14.3.1 (5) FSAR 14.1.2 (6) FSAR 7.2, 7.3

(~) FSAR 3.2.1 (8) FSAR 14.1.9 (9) FSAR 14.1.11 l

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SAFETY LIMITS, RIACTOR CORE, THERMAL AND HYDRAC IC i TWO-LOOP OPERATION TIGURE TS.I.1-1 .

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TS.3.1-18 REV For extended optimum fuel burnup it is necessary to either load.the i

reactor with burnable poisons or increase the boron concentration in the reactor coolant system. If the latter approach is emphasized, it is possible that a positive isothermal temperature coeffieient could exist at beginning of cycle (BOC). Safety analyses verify the accep-tability of the isothermal temperature coefficient for limits specified in 3.1.F.1. Other conditions, e.g., higher power or partial rod insertion would cause the isothermal coefficient to have a more negative value. These analyses demonstrate that applicable criteria in the NRC Standard Review Plan (NUREG 75/087) are met.

Physics measurements and analyses are conducted during the reload startup test program to (1) verify that the plant will operate within safety analyses assumptions and (2) establish operational procedures to i ensure safety analyses assumptions are met. The 3.1.F.1 requirements are waived during low power physics tests to permit measurement of reactor temperature coefficient and other physics design parameters of

! interest. Special operating precautions will be taken during these (3)

{ physics tests. In addition, the strong negative Doppler coefficient and the small integrated ak/k would limit the magnitude of a power i excursion resulting from a reduction of moderator density.

! The requirement th'at the reactor is not to be made ciritical except as l' specified in Figure TS.3.1-1 provides increased assurance that the proper relationship between reactor coolant pressure and temperature will be maintained during system heatup and pressurization whenever the reactor vessel is in the nil ductility temperature range. Heatup to this temperature will be accomplished by operating the reactor coolant pumps and by pressurizer heaters. The pressurizer heater and associ-ated power cables have been sized for continuous operation at full heater power. The shutdown margin in Specification 3.10 precludes the possibility of accidental criticality as a result of an moderatortemperatureoradecreaseofcoolantpressure.{3yreaseof l

l l

l

References:

(1) FSAR Figure 3.2-10 (2) FSAR Table 3.2-1

TS.3.3-1 REV 3.3 ENGINEERED SAFETY FEATURES Applicability Applies to the operating status of the engineered safety features. ,

Objective l

To define those limiting conditions that are necessary for operation of 4

engineered safety features: (1) to remove decay heat from the core in an emergency or normal shutdown situations, and (2) to remove heat from -

containment in normal operating and emergency situations.

Specifications i

A. Safety Injection and Residual Heat Removal Systems t

1. A reactor shall not be made or maintained critical nor shall it

be heated or maintained above 200*F unless the follswing condi-

tions are satisfied except as permitted in Specification 3.3.A.2.

a. The refueling water tank contains not less than 200,000 gallons of water with a boron concentration of at least 1950 i ppa.

, b. Each reactor coolant system accumulator shall be operable-when reactor coolant system pressure is greater than 1000 psig.

4 Operibility requires:

(1) The isolation valve is open (2) Volume is 1270 2 20 cubic feet.of borated water l (3) A minimum boron concentration of 1900 ppm (4) A nitrogen cover pressure of at least 700 psig

(

c. Two safety injection pumps are operable except that pump J control switches in the control room shall meet the require-i ments of Section 3.1.G whenever the reactor. coolant system l temperature is less than MPT.

I 4

d. Two residual heat removal pumps are operable. ;

e. Two residual heat exchangers are operable,

f. Automatic valves, interlocks and piping associated with the above components and required to function during accident conditions, are operable.

. g. Manual valves in the above systems that could (if one is

improperly positioned) reduce injection flow below that 1

assumed for accident analyses, shall be blocked and tagged in the proper position for injection. RHR system valves, however, may be positioned as necessary to regulate plant heatup or

{ cooldown rates when the reactor is suberitical. All changes in

. valve position shall be under direct; administrative control. I i-

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

e .

TS.3.10-1 REV 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Applicability Applies to the limit.: on core fission power distribution and to the limits on control rod operations.

Objective To assure 1) core suberiticality after reactor trip, 2) acceptable core power distributions during power operation, and 3) limited potential reactivity insertions caused by hypothetical control rod ejection.

I Specification A. Shutdown Reactivity The shutdown margin with allowance for a stuck control rod assembly shall exceed the applicable value shown in Figure TS.3.10-1 under all steady-state operating conditions, except for physics tests, from zero to full power, including effects of axial power distribution. The shutdown margin as used here is defined as the amount by which the reactor core would be suberitical at hot shutdown conditions if all control rod assemblies were tripped, assuming that the highest worth control rod assembly remained fully withdrawn, and assuming no changes in xenon or boron concentration.

B. Power Distribution Limits

1. At all times, except dpring lgw power physics testing, measured and F hotchannelfactors,F"hollowinglimits:g bases, shall meet the , as defined below and in the

[ x 1.03 x 1.05 1(2.30/P) x K(Z) d[3 x 1.04 11.60 x [1+ 0.3(1-P)]

where the following definitions apply:

- K(Z) is the axial dependence function shown in Figure TS.3.10-5.

- Z is the core height location.

-Pisthefractiono[fratedpoweratwhichthecoreis operating. In the limit determination when P 1 50, set P = 0.50.

O e TS.3.10-2 REV or [ is defined as the measured F or F respectively, l

- [kth w thh smallest margin or greatest ekcess bk limit.

app i

- 1.03 is tb engineeringhotchannelfactor,[ha,nce.liedtothe to account for manufacturing tole measured

- 1.05 is applied to the measured [q to account for measurement I uncertainty.

- 1.04 is applied to the measured F to account for measurement l uncertainty.

2.

Hotchannelfactors,[9ned,atequilibriumconditionsaccordingandFfH,sh flux difference determ to the following conditions, whichever occurs first:

(a) At least once per 31 effective full-power days in conjunction with the target flux difference determination, or (b) Upon reaching equilibrium conditions af ter exceeding the reactor power at which target flux differan;e was last

~

determined, by 10% or more of rated power.

(equil) shall meet the following limit for the middle axial 80%

[hthecore:

o N (equil) x V(Z) x 1.03 x 1.05 1(2.30/P) x K(Z) where V(Z) is defined Figure 3.10-7 and other terms are l defined in 3.10.B.1 above.

3. (a) If either measured hot channel factor exceeds its limit j

specified in 3.10.B.1, reduce reactor power and the high neutron flux trip setpoint by 1% for each percent that the mgasuredF"0the3.10.B.1 limit.or Then by 3.33%

followfor3.10.B.3(c).

each percent that the measured f]H exceed (b) the If the measured 3.10.B.1 lim 1 [9,(equil) taks one of exceeds the followingtheactiens:

3.10.B.2 limits but not

1. Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months place the reactor in an equilibrium configuration for which Specification 3.10.B.2 is satis-fied, or i

2. Reduce reactor power and the high neutron flux trip sqtpoint by 1% for each percent that the measured f q(equil) x 1.03 x 1.05 x V(Z) exceeds the limit. l 1

TS.3.1-17 REV F. MINIMUM CONDITIONS FOR CRITICALITY Specification

1. Isothermal Temperature Coefficient (ITC)

a. When the reactor is critical, the isothermal temperature coefficient shall be less than 5 pcm/*F with all rods withdrawn, except during low power physics teste and as specified in 3.1.F.1.b and c.

b. When the reactor is above 70 percent rated thermal power with all rods withdrawn, the isothermal temperature coefficient shall be 4 negative, except as specified in 3.1.F.1.c.

1 c. If the limits of 3.1.F.1.a or b cannot be met, Power Operation may continue provided the following actions are taken:

(1) Establish and maintain control rod withdrawal limits sufficient to restore the ITC to less than the'ltaits specified in Specification 3.1.F.1.a and b above within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months gor be in HOT SHUTD0'4N within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdrawal limits shall be in addition to the insertion limits of Figure TS.3.10-2.

(2) Maintain the control rods within the withdrawal limits established above until a subsequent calculation verifies that the ITC has been restored to within its limit for the all rods withdrawn condition.

(3) Submit a special report to the Commission within 30 days, describing the value of the measured ITC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the ITC to within its limit for the all rods withdrawn condition.

2. The reactor shall not be brought to a critical condition until the j . reactor coolant temperature is higher than that defined by the criti-cality limit line shown in Figure TS.3.1-1. ;

4 I

Basis At the beginning of a fuel cycle the moderator temperature coefficient has i its most positive or least negative value. As the boron concentration is

! reduced throughout the fuel cycle, the moderator temperature coefficient I

becomes more negative. The isothermal temperature coefficient is defined as the reactivity change associated with a unit. change in the moderator and fuel 'emperatures.

t Essentially, the isothermal temperature coefficient is the sum of the moderator and fuel temperature coefficients. This co-efficient'is measured directly during low power physics tests in order to verify analytical predictions.

, e + ,. , , - , , -, , - - - - . ,w- r- --v~-- - ,, ,

l TS.3.10-8 REV ;

3. If one or both of the quadrant power tilt sonitors is inoperable, individual upper _and lower excore detectos calibrated outputs and the calculated power tilt shall be logged every_two hours after a load change greater than 10% of rated power.

i J. DNB Parameters The following DNB related parsmeters limits shall be maintained j during power operation

! a. Reactor Coolant System Tavg 1564*F

b. Pressurizer Pressure >2220 psia *

, i i

i

c. Reactor Coolant Flow >178,000 gpm With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce thermal power to less than 5% of rated thermal power using normal shutdown 3

procedures.

1 Compliance with a. and b. is demonstrated by verifying that each of the parameters is within its limits at least once each 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Compliance with c. is demonstrated by verifying that the parameter is within its. limit after each refueling cycle. -

Bases Throughout the 3.10 Technical Specifications, the terms " rod (s)" and "RCCA(s)" are synonomous.

Shutdown Reactivity l Trip shutdown reactivity is provided consistent with plant safety analyses assumptions. One percent shutdown is adequate except for

.the steam break analysis, which requires more shutdown reactivity due to the more negative moderator temperature coefficient at end of life (when boron concentration is low). Figure TS.3.10-1 is drawn accordingly.

!, Power Distribution Control The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operations) and II (Incidents of Moderate frequency) events by: (a) maintaining the minimum DNBR

, in the core 21.30 for Exxon Nuclear fuel and 3.1.17 for Westinghouse fuel during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding

Limit not applicable during either a THERMAL' POWER ramp increase in excess of (5%) RATED THERMAL POWER per minute or a THERMAL POWER step increase in excess of (10%) RATED ~HERMAL POWER.

I TS.3.10-9. i REV mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density-during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.

4 During opegation,Nthe plant staff compares the measured hot channe.

factors, F and F ) to the limits determined in the transient hnd LOCK a,na(described lyses. The termslater on the right side of the j equations in Section 3.10.B.1 represent the analytical limits. Those terms on the left side represent the measured hot channel factors corrected for engineering, calculational, and measurement uncertainties.

is the measured Nuclear Hot Channel Factor, defined as the maximum (localheatfluxonthesurfaceofafuelroddividedbytheaverage heat flux in the core. Heat fluxes are derived from measured neutron fluxes and fuel enrichment. -

The K(Z) function shown in Figure TS.3.10-5~1s a normalized function that limits F axially. The K(Z) specified for the lowest six (6) feet l o

of the core is abritrarily flat since the lower part of the core is_

generally not limiting. Above that region, the K(Z)_value is based on large and small break LOCA analyses.

4 V(Z) is angaxially depgndent function applied to the equilibrium measuredFgtoboundE'O's that could be measured at non-equilibrium conditions. This function is based on power distribution control analyses that evaluated the effect of burnable poisons, rod position, axial effects, and xenon worth, r

F~, Engineering Heat Flux Hot Channel Factor, is defined as the allow-I akceonheatfluxrequiredformanufacturingtolerances. The engi-neering factor allows for local variations in enrichment, pellet density and diameter, surface area of the fuel rod and eccentricity of the gap between pellet and clad. Combined statistically the net effect l 1s a factor of 1.03 to be applied to fuel rod surface heat flux.

l 'The 1.05 multiplier accounts for uncertainties associated with measure-ment of the power distribution with the moveable incore detectors and the use of those measurements to establish the assembly local power distribution.

Ff(equil)isthemeasuredlimitingF obtained at equilibrium conditions dQring target flux determination.

F , Nuclear Enthalpv Rise Hot Channel Factor, is defined as the ratio

! of the integral of linear power along the rod with the highest integrated power to the average rod power.

i i

TS.3.10-10 REV When a measurement of F H is taken, measurement error must be allowed for and 4 percent is the appropriate allowance for a full core map taken with the movable incore detector flux mapping system.

Measurements of the hot channel factors are required as pect cf startsp physics tests, at least once each effective full power mouth of operation, and whenever abnormal power distribution conditions require a .eductian of cere power to a level based on measured hot channel factors. The incore map taken following initial loading provides confirmation of the basic nuclear design bases including proper fuel loading patterns. The periodic monthly incore mapping provides additional assurance that the nuclear design bases remain inviolate and identify operational anomalies which would otherwise affect these bases.

For normal operation, it is not necessary to measure these quantities.

Instead it has been determined that, provided certain conditions are observed, the hot channel factor limits will be met; these conditions are as follows:

1. Control rods in a single bank move together.with no individual rod insertion differing by more than 15 inches from the bank demand position. An accidental misalignment limit of 13 steps precludes a rod misalign-ment greater than 15 inches with consideration of maximum instrumentation error.

2. Control rod banks are sequenced with overlapping banks as described in Technical Specification 3.10.

3. The control bank insertion limits are not violated.

4. Axial power distribution control procedures, which are given in terms of flux difference control and control bank inser-tion limits are observed.' Flux difference refers to the difference in signals between the top and bottom halves of two-section excore neutron detectors. The flux difference is a measure of the axial affect which is defined as the dif-ference is a measure of the axial offset which is defined as the difference in normalized power between the top and bottom halves of the core.

f

g ,

TS.3.10-11 REV N and FN allows-for radial power #

i The permitted relaxation in F AH '

shapechangeswithrodinsertiontotheinsertionlimits. It has been determined that provided the above conditions 1 through 4 are -

observed,tysehotchannelfactorlimitsaremet. In specifica-tio'n 3.10, F is arbitrarily limited for P <0.5 (except for lov ,

i powerphysichtests). ,

N

The procedures for axial power distribution control referred to above are designed to minimize the effects of xenon redistribution on the axial power distribution during load-follow maneuvers.

1 Basically control of flux difference is required to limit the

! difference between the current value of Flux Difference (61) and I

a reference value which corresponds to the full power equilibrium value of Axial Offset (Axial Offset = AI/ fractional power). The

)'

reference value of flux dit'ference varies with power level and burnup but expressed as axial offset it varies only with burnup.

3 The technical specifications on power distribution control' assure that the F limit is not exceeded and xenon distributions are not l developedOhichatalatertime,wouldcausegreaterlocalpower peaking even though the flux difference is then within the limits-specified by the procedure.

The target (or reference) value of flux difference is determined as j follows: At any time that equilibrium xenon conditions have been-established, the indicated flux difference is noted with the full-length rod control rod bank more than 190 steps withdrawn (i.e., ..

normal full power operating position appropriate for the time in life, usually withdrawn f arther as burnup proceeds). This value, '

divided by the fraction of full power at which the core was oper-b ,

ating is the full power value of the target flux difference. Values for all other core power levels are obtained by multiplying the j full power value by the fractional power. Since the. indicated ,

l equilibrium was noted, no allow.ances for excore detector error are -

necessary and indicated deviation of 25 percent AI are permitted ,

from the indicated reference value. Figure TS.3.10-6 shows the &

allowed deviation from the target flux difference as the function of thermal power.

i 1

s

+v-l l

L j ,

1

'y %

\

6

c.. .

TS.3.10-13 REV resulting from operation within the target band. The consequences of being outside the 15% target band but.within the rigure TS.3.10-6 limit

for pewer-levels between 50% and 90% has been_ evaluated and determined to result in acceptable peaking factors. Therefore, while the deviation l exists the power level is limited to 90 percent or lower depending on the indicated axial flux difference. In all cases the 25 percent target band is the Limiting Condition for Operation. Only when the target band is violated do the limits under Figure TS.3.10-6 apply..

If, for any reason, the indicated axial flux difference is not controlled within the 5 percent. band for as long a period as one hour, then xenon distributions may be significantly changed and operation at 50 percent is required to protect against potentially more severe consequences of some accidents.

l s -

As discussed above, the essen,ce of the procedure is to maintain the xenon s

distribution in the core as close to the equilibrium full power condition as,possible. This is accomplished by using the boron system to position the full length control rods to produce the required indicated flux difference.

For Condition II event's t'he core is protected from overpower and a minimum -

DNBR of 1.30 for Exxon fuel and 1.17 for Westinghouse fuel by an automatic '

l protection system. Compliance with operating procedures is assumed as a precondition for Condition II transients, however, operator error and equip-ment malfunctions are separately assumed to lead to the cause of the transients considered.

Quadrant Power Tilt Limits s s.

)

Quadrant power tilt limits are_ based on the following considerations. Fre-quent power tilts are not anticipated during normal operation since this phenomenon is caused by some. asymmetric perturbation, e.g. rod misalignment, x-y xenon transient, or inlet temperature mismatch. A dropped or misaligned rod will easily be detected by the Rod Position Indication System or core instrumentation per Specification 3.10.F. and core limits protected per Specification 3.10.E.

A quadrant tilt by some other means (x-y xenon tran-sient, etc.) would not appear instantaneously, but would build up;over several hours and the quadrant tilt limits are set to protect against this situation. They also serve as a backup prottetion against the dropped o'r misaligned rod.

Operational experience shows that normal power. tilts are 1ess than 1.01.

Thus, sufficient time is available to recognize the presence of a, tilt and correct up and-powet the cause before a severe tilt could build up. During start-escalation, however, a large tilt could be initiated.

Therefore, the Technical Specification has been written so as to prevent ~

escalation above 50 percent power if a large tilt is present.

A TS.3.10-13 REV t

resulting from operation within the target band. The consequences of being outside the 15% target band but within the rigure TS.3.10-6 limit for nower levels between 50% and 90% has been evaluated and determined to r'sult e in acceptable peaking factors. Therefore, while the deviation l exists the power level is limited to 90 percent or lower depending on the indicated axial flux difference. In all cases the 25 percent target band is the Limiting Condition for Operation. Only when the carget band is violated do the limits under Figure TS.3.10-6 apply.

If, for any reason, the indicated axial flux difference is not controlled within the 25 percent band for as long a period as one hour, then xenon distributions may be significantly changed and operation at 50 percent is required to protect against potentially more severe consequences of soma accidents.

As discussed above, the essence of the procedure is to maintain the xenon distribution in the core as close to the equilibrium full power condition as,possible. This is accomplished by using the horon system to position the full length control rods to produce the reqdtred indicated flux difference. \s For Condition II events the core is protected from overpower and a minimum DNBR of 1.30 for Exxon fuel and 1.17 for Westinghouse fuel by an automatic l protection system. Compliance with operating procedures is assumed as a precondition for Condition II transients, however, operator error and equip-ment malfunctions are separately assumed to lead to the cause of the transients considered, g 6 x Quadrant Power Tilt Limits N Quadrant power tilt limits are based on the following considerations. Fre-quent power tilts are not anticipated during nogma) operation since this phenomenon is caused by some asymmetric-perturbarion, e.g. rod misalignment, x-y xenon transient, or inlet temperature mismatcu. A dropped or misaligned rod will easily be detected by the Rod Position Indication System or core instrumentation per Specification 3.10.F and core , limits protected per Specification 3.10.E. A quadraut tilt by some other means (x-y xenon tran-sient, etc.) would not appear instantaneously, but would build up over several hours and the quadrant tilt limits are set to protect against this situation. They also serve as a backup protection against the dropped or misaligned rod.

Operational experience shows that normal power tilts are less than 1.01.

Thus, sufficient time is available to recognize the presence of a tilt

~

and correct the cause before a severe tilt could build up. During start-up and power escalation, however, a large tilt could be initiated.

Therefore, the Technical Specification has been written so as to prevent escalation above 50 percent power if a large tilt is present.

,f

~>

[ ,

L_.

4 -t TS.3.10-17 REV If the rod position deviation monitor and quadrant power tilt monitor (s) are inoperable, the overpower reactor trip setpoint is reduced (and also power) to ensure that adequate core protection is provided in the event that unsatisfactory conditions arise that could affect radial power distribution.

Increased surveillance is required, if the quadrant power tilt monitors are inoperable and a load change occurs, in order to confirm satisfac-tory power distribution behavior. The automatic alarm functions related to quadrant power tilt must be considered incapable of alerting the operator to unsatisfactory power distribution conditions.

DNB Parameters The RCS flow rate, T and Pressurizer Pressure requirements are based on transient analyses #Es,sumptions. The flow rate shall be verified by calorimetric flow data and/or elbow taps. Elbow taps are used in the reactor coolant system as an instrument device that indicates the status of the reactor coolant flow. The basic function of this device is to provide information as to whether or not a reduction in flow rate has occurred. If a reduction in flow rate is indicated below the specifica-tion value indicated, shutdown is required to investigate adequacy of core cooling during operation.

. U

FIGURE TS.3.10-5 ,

REV l l

l I

1.2 l

(6.0,1.0) 1.0 i

(12.0.0.92) 0.8 K(Z) 0.6 0.4 0.2 0.0 2 4 6 8 10 12 0

Core Height (ft)

HOT CHANNEL FACTOR NORMALIZED OPERATING ENVELOPE

FIGURE TS.3.10-7 REV 1.20 1.18 1.16 ,

(11.25,1.15) 1.14 [

1.12 /

/\

1.10 (9.25,1,11)

V(Z) 1.08 1.06 1.04 1.02 1.00 0 2 4 6 8 10 12 Core Height (Ft)

V(Z) as a Function of Core Height

EXHIBIT D PRAIRIE ISLAND NUCLEAR GENERATING PLANT License Amendment Request Dated January 13, 1986 PRAIRIE ISLAND UNIT 1 WESTINGHOUSE OFA TRANSITION RELOADS Information addressing eleven plant specific items given on page 3 of NRC cover letter for NRC SER on WCAP-9500 (May 22, 1981)

1. For tbase plants usfeg the Improved Thermal Design Procedure (ITDP), the conditions listed in the safety evaluation must be addressed and satisfied.

Response: This item is not applicable to the Prairie Island Unit I since the ITDP methodology is not used.

2. A discussion in the Basis of Technical Specifications of any generic or plant-specific margins that have been used to offset the reduction in DNBR due to rod bowing.

Response: The effects of fuel rod bowing is included explicitly in the transient analysis. A rod bow penalty is applied to the calculated MDNBR for each transient before comparison to acceptance criteria. No generic or plant specific margins have been used to offset rod bow DNBR reductions.

3. A declaration in the Technical Specifications that prohibits N-1 loop operation unless adequately justified in the plant-specific analysis.

Response: Current Technical Specifications require an automatic reactor trip with less than 90% flow in either loop below 10% power.

4. Frequency and description af rod worth tests that would detect gross losses of reactivity worth from boron-containing control rods.

Response: This item is not applicable to Prairie Island Unit I since boron-containing control rods are not used.

5. Confirmation that the predicted cladding collapse time exceeds the expected lifetime of the fuel.

Response: The Prairie Island Unit 1 Westinghouse OFA fuel utilizing the approved Westingh 7 pye fuel performance and model {g) design approved clad flattening model . The OFA fuel is designed so that the calculated fuel rod clad flattening time is greater than the maximum planned fuel irradiation time in the reactor.

(a) Miller, J V. " Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations," WCAP-8720 (Proprietary) and WCAP-8785 (Non-Proprietary), October 1976.

(b) George , R A. (et, al.), " Revised Clad Flattening Model," WCA1-0377 (Proprietary) and WCAP-8381 (Ncn-Proprietary), July 1974.

. . l

6. Supplemental ECCS calculations using NRC-supplied LOCA cladding models, until a generic resolution of this issue is obtained.

Response: This item is not applicable since a generic resolution of NRC concsrns on clad ballooning and assembly flow blockage has been obtained.

The Prairie-Island LOCA analyses of the Westinghouse OFA fuel use approved models which incorporate the generic resolution.

7. A determination that the appropriate seismic and LOCA forces are bounded by the cases considered in WCAP-9401 or additional analyses.

Response: The impact of LOCA and seismic forces on the integrity of 0FAs !

has been analyzed for a homogeneous OFA core and mixed cores of 0FAs and Exxon fuel assemblies. Results confirm the integrity of the 0FAs since ;

the grids do not crush and the stresses in the OFA components are within acceptable limits. The analyses used the generic methods of Reference i (a) which received NRC approval via Reference (b). )

(a) Letter from E P Rahe (Westinghouse) to J R Miller (NRC) dated March !

19, 1982, NS-EPR-2573,

Subject:

WCAP-9500 and WCAP-9401/9402 NRC SER Mixed Core Compatability Items.

(b) Letter from C 0 Thomas (NRC) to E P Rahe (Westinghouse), dated November 12, 1982,

Subject:

Supplemental Acceptance Number 1 for i Referencing of Licensing Topical Report WCAP-9500-A. I

8. A description of plans for on-line fuel system monitoring:

Response: Prairie Island Unit I does not have an on-line fuel system monitor other than gross fuel failure detection with the letdown line liquid process menitors. The current method of fuel monitoring by collecting periodic primary coolant samples to assess fuel integrity is sufficient.

l

9. A description of plans for post-irradiation poolside surveillance of fuel. l l

Response: No special surveillance requirements are necessary since Prairie Island Unit 1 is not a lead plant for using the 14x14 0FAs. The Point Beach units are the lead plants utilizing standard 14x14 0FAs. Therefore, only the normal visual surveillance of representative sample of irradiated 0FAs is planned during refueling shutdowns of Prairie Island Unit 1.

10. For transients analyzed to determine fuel failure, DNBR as function of time (NUREG-1.70 requirement).

Response: The DNBR as a function of time is calculated and the results are published in the Final Reload Design Report (Reload Safety Evaluation) for each cycle. The results of this evaluation are contained in Exhibir H.

11. Initial fuel conditions (i.e., stored energy or centerline temperature) utilized in the transient and accident analyses (as per NUREG-1.70 requirements).

Response: The initial fuel temperatures used in the NSP transient. analyses were supplied by the fuel vendor (Westinghouse) using the PAD Code (WCAP 8720). The initial feel temperatures used in LOCA aralyses performed by Westinghouse used the PAD. Code (WCAP 8720 Addn 2).

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