Text

Proposed Tech Specs,Relocating Shutdown Margin Requirements from TS to COLR
	ML20205Q035
Person / Time
Site:	Prairie Island
Issue date:	04/12/1999
From:	; NORTHERN STATES POWER CO.
To:	;
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ML20205Q023	List: ML20205Q019; ML20205Q026; ML20205Q035;
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	NUDOCS 9904210068
	Download: ML20205Q035 (22)
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EXHIBIT B PRAIRIE ISLAND NUCLEAR GENERATING STATION l .

License Amendment Request dated April 12,1999 l i

l Appendix A, Technical Specification Pages i 1

Marked Up Pages (shaded material to be added, strikethrough material to be removed) 1 Table TS.1-1 1 TS.3.10-1 TS.3.10-5 Figure TS 3.10-1 TS.6.7-3 B.3.8-1 B.3.10-1 l

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TABLE TS.1-1 RE'? 111 9'10'9d TABLE TS.1-1 OPERATIONAL MODES REACTOR

% RATED AVERAGE VESSEL HEAD REACTIVITY THERMAL COOLANT CLOSURE BOLTS MODE TITLE CONDITION POWER TEMPERATURE FULLY TENSIONED 1 POWER OPERATION Critical > 2% NA YES j 2 HOT STANDBY ** Critical 52% NA YES 3 HOT SHUTDOWN ** Suberitical NA 2 350*F YES 4 INTERMEDIATE Subcritical NA < 350*F YES SHUTDOWN ** 2 200*F 5 COLD SHUTDOWN Subcritical NA < 200*F YES 6 REFUELINO NA* NA NA NO Boron concentration of the reactor coolant system and the refueling cavity sufficient to ensure that the more restrictive of the following conditions is mets

a. K,,, S 0.95, os

b. Boron concentration > 2000 ppmy [

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- Prairie Island specific MODE title, not consistent with Standard Technical Specification MODE titles. MODE numbers are consistent with Standard Technical Specification MODE numbers.

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TS.3.10-1

,'o REV-145 3/,28/48 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Applicability Applies to the limits on core fission power distribution and to the limits on control rod operations. 4 Qhjective To assure 1) core subcriticality after reactor trip, 2) acceptable core power distributions during POWER OPERATION, and 3) limited potential reactivity insertions caused by hypothetical control rod ejection.

Specification A. Shutdown Margin

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1. Ecacter Ccciant Syste- Average Temperature >--200 F The SHUTDOWN MARGIN shall bejeentee-than er aal-t~-the a--licable--value shown in Fi-urc TS.3.10-1 '"T E ""M?"TNMP> EdjimedEERi41PR$$

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when in HOT SHUTDOWN, INTERMEDIATE SHUTDOWN

2. Eccctor Ccclent Svatc= Average Temperature f 200 F The-BHUTDOWN-MARGIM-shal-1-bc greater-than or equal tc IL^k/h when in COLD BWUTDOWN-r i

32.With the SHUTDOWN MARGIN less than the applicable limit specified in 3.10.A.1 '

or 3.10.A.2 above, within 15 minutes initiate boration to restore SHUTDOWN MARGIN to within the applicable limit.

B. Power Distribution Limits

1. At all tigs, excJpt during low power PHYSICS TESTING, measured hot channel factors, Yo and F M, as defined below and in the bases, shall meet the following limits:

F"o x 1.03 x 1.05* 5 (F["/ P) x K(Z)

F"M x 1.04** 6 Fd " x [1+ PFDH(1-P)]

where the following definitions apply:

- F[" is the Fo limit at RATED THERMAL POWER specified in the CORE OPERATING LIMITS REPORT.

FM " is the FM limit at RATED THERMAL POWER specified in the CORE OPERATING LIMITS REPORT.

- PFDH is the Power Factor Multiplier for F"M specified in the CORE OPERATING LIMITS REPORT.

K(Z) is a normalized function that limits Fo(z) axially as specified in the CORE OPERATING LIMITS REPORT.

For Unit 1, Cycle 19, when the number of available moveable detector thimbles is greater than or equal to 50% and less than 75% of the total, the 5% measurement uncertainty shall be increased to [5% + (3-T/9)(3%)] where T is the number of available thimbles.

For Unit 1. Cycle 19, when the number of available moveable detector thimbles is greater than or equal to 50% and less than 75% of the total, the 4% measurement uncertainty shall be increased to [4% + (3-T/9)(2%)] where T is the number of available thimbles.

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. TS.3.10-5 l REV 02 1413490 l l

l 3.10.C.2. If the QUADRANT POWER TILT RATIO exceeds 1.02 but is less than j 1.07 for a sustained period of more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . or if such a tilt recurs intermittently, the reactor shall be brought to the HOT SHUTDOWN condition. Subsequent operation below 50% of rating, for testing, shall be permitted.

3. Except for PHYSICS TESTS if the QUADRANT POWER TILT RATIO exceeds 1.07. the reactor shall be brought to the HOT SHUTDOWN condition.

Subsequent operation below 50% of rating, for testing, shall be permitted.

4. If the core is operating above 85% power with one excore nuclear channel inoperable, then the core quadrant power balance shall be determined daily and after a 10% power change using either 2 movable detectors or 4 core thermocouples per quadrant. per Specification 3.11.

D. Rod Insertion Limits

1. The shutdown rods shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT when the reactor i.s critical or approaching criticality.

2. When the reactor is critical or approaching criticality. *he .

control banks shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT.

3. Insertion limits do not apply during PHYSICS TESTS or during periodic exercise of individual rods...The shutdown margin _ehewn.

4n41gure T4r3,404 must be maintained except for low power PHYSICS TESTING. For this test the reactor may be critical with all but one high worth full-length control rod inserted for a period not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per year provided a rod drop test is run on the high worth ,

full-length rod prior to this particular low power PHYSICS TEST. I i !

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,. FIGURE TS 3.10-1 l REV 91 10/27/89 1

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equired Shutdown Margin vs Reactor Boron Concentrat'on

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NLI \ I I/ I N \ I / l N \! i / l

^ 1600

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3 ** I NN A I I g \ NI / j i \l M 3 l\l / I I i i ! ! I i l l l l 1 I Vi l I I I I i l I i

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0 0 200 400 600 800 1000 1200 Boron Concentration (ppm)

FIGURE TS. .10-1

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.N TS.6.7-3 RE" 122 l'2d'95 6.7.A.4. Radioactive Effluent Report The Radioactive Effluent Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant. The material provided shall be consistent with the objectives outlined in the ODCM and in conformance with 10CFR50.36a and 10CFR50, Appendix I, Section IV.B.1.

6.7.A.5. Annual Summaries of Meteorological Data An annual summary of meteorological data shall be submitted for the previous calendar year in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability at the request of the Commission.

6.7.A.6. Core Operating Limits Report

a. Core operating limits shall be established and documented in the I CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

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1. Heat Flux Hot Channel Factor Limit ( Fg"* ) , Nuclear Enthalpy Rise Hot Channel Factor Limit ( FA/") , PFDH, K(Z) and V(Z)

(Specifications 3.10.B.1, 3.10.B.2 and 3.10.B.3) l

2. Axial Flux Difference Limits and Target Band j (Specifications 3.10.B.4 through 3.10.B. 9)

3. Shutdown and Control Bank Insertion Limits l (Specification 3.10.D) !

4. Reactor Coolant System Flow Limit j (Specification 3.10.J)

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b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

NSPNAD-8101-A, " Qualification of Reactor Physics Methods for Application to PI Units" (latest approved version)

NSPNAD-8102-A, " Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units" (latest approved version)

WCAP-9272-P-A, " Westinghouse Reload Safety Evaluation Methodology", July, 1985

B.3.8-1 RE" 119 '/2/95 3.8 REFUELING AND FUEL HANDLING Bases Core alteration containment isolation specifications are provided to minimize releases following a fuel handling accident (FHA). Allowing both airlock doors open during core alterations will facilitate evacuation of containment following a FHA and help maintain the seals in good working order. The FHA does not cause containment pressurization, however, with an assumed single failure the operating purge system supply fan is assumed to continue supplying air to containment. To maintain post-FHA releases well within the limits of 10CFR100, only the inservice purge system is allowed to be operating during core alterations. Two containment fan coil unit fans are required to operate in the high speed mode following a fuel handling accident in containment to assure that radioactive material in i containment is well mixed and any releases will leave containment at i a lower concentration over the duration of the accident. The provision that one door is OPERABLE and under procedural control will ensure that at least one door will be closed in within 30 minutes as required, thus assuring radioactive releases are well within the limits of 10CFR100.

The equipment and general procedures to be utilized during refueling are discussed in the FSAR. Detailed instructions, the precautions specified above, and the design of the fuel handling equipment incorporating built-in interlocks and safety features, provide assurance that no incident could occur during CORE ALTERATIONS that would result in a hazard to public health and safety (Reference 1) .

Whenever changes are not being made in core geometry, one flux monitor is sufficient. This permits maintenance of the instrumenta-tion. Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. The residual heat removal pump is used to maintain a uniform boron concentration.

Under rodded and unrodded conditions, the Ku, of the reactor must be less than or equal to 0.95 and the boron concentration must be greater than or equal to 2000 ppm.

Per c c ecks o ref ing water boron concentration naure t at proper shutdown margin is maintained. 3.8.A.1.h allows the control room operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.

No movement of fuel in the reactor is permitted until the reactor has been suberitical for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months to permit decay of the fission products in the fuel. The delay time is consistent with the fuel handling accident analysis (Reference 2).

Fuel will not be inserted into a spent fuel cask unless a minimum boron concentration of 1600 ppm is present. The 1800 ppm will ensure that ku, for the spent fuel cask, including statistical uncertainties, will be less than or equal to 0.95 for all postulated arrangements of fuel within the cask.

The number of recently discharged assemblies in Pool No. I has been limited to 45 to provide assurance that in the event of loss of pool cooling capability, at least eight hours are available under worst case conditions to make repairs until the onset of boiling.

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.'s B.3.10-1 RE" 111 9/10/9' 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Bases Throughout the 3.10 Technical Specifications, the terms " rod (s)" and "RCCA(s) " are synonymous.

A. Shutdown Margin A sufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made suberitical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition.

? JTOO"P "?_"f!F requir----te u2ry threughout cere life 2r fu ction of fuel depletion re2eter reelznt ryrte- herer corre-tratier --d re20ter coel?-t 2"erzge te per:ture. The ert rertrictive re dition errurr 2t ^-d Of life --A ir ereci2ted "ith 2 pertul?ted rte 2r line bre2' 2 C r i d e"_ t -"A rerelting uncentrelle d rearter reel 2nt ryrte reclde"- In the ---lyrir of *hir 2crid_ert 2 -iri u- ? " " = "*"fIF (rhe'c'- in Figure T?.2.10-1 2r furetier Of equilibrium het full pe"er berer correntratic ' ir required a

te centrcl the re22ivity tr- rient- rrerdingly, the ?F"~~ = "^"fIF __

requir^ rnte 2re b2 red upe- t'ir lf-iting re-dition --d ?re renrictent "ith plant r2fet; ---lyrir scr" ptienr. "It' re rter recl2nt ryrter 2"cr2;e t- perature lerr -- 000*F the re?rti"it; tr2nciente rerulting fre 2 pertul?ted rte 2~ line bre2' ccelde"n 2re -!ni 21 --d 2 1? Ah

FMt" Do= F_^ "r IF p rev! d e r adequete pretection In POWER OPERATION and HOT STANDBY, with k,,, 2 1, SHUTDOWN MARGIN is ensured by complying with the rod insertion limitations in Specification 3.10 D. In HOT SHUTDOWN, INTERMEDIATE SHUTDOWN and COLD SHUTDOWN, the SHUTDOWN MARGIN requirements in Specification 3.10 A are applicable to provide sufficient negative reactivity to meet the assumptions of the safety analyses discussed above. For REFUELING, the shutdown reactivity requirements are specified in Table TS.1-1.

When in POWER OPERATION and HOT STANDBY, SHUTDOWN MARGIN is determined assuming the fuel and moderator temperatures are at the nominal zero power

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EXHIBIT C PRAIRIE ISLAND NUCLEAR GENERATING STATION License Amendment Request dated April 12,1999 l

l Appendix A, Technical Specification Pages t l \

Revised Pages j t

Table TS.1-1 l

TS.3.10-1 TS.3.10-5 l TS.6.7-3 B.3.8-1 B.3.10-1 i

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l TABLE TS.1-1 OPERATIONAL MODES REACTOR

% RATED AVERAGE VESSEL HEAD REACTIVITY THERMAL COOLANT CLOSURE BOLTS MODE TITLE CONDITION POWER TEMPERATURE FULLY TENSIONED 1 POWER OPERATION Critical > 2% NA YES 2 HOT STANDBY ** Critical 5 2% NA YES 3 IIOT SHUTDOWN ** Suberitical NA 2 350*F YES 4 INTERMEDIATE Subcritical NA < 350*F YES SHUTDOWN ** 2 200'F 5 COLD SHUTDONN Subcritical NA < 200 F YES 6 REFUELING NA* NA NA NO Boron concentration of the reactor coolant system and the refueling cavity sufficient to ensure that the more restrictive of the following conditions is met:

a. K.,, 3, 0.95,

b. Boron concentration > 2000 ppm, or

c. Shutdown Margin as specified in the Core Operating Limits Report.

- Prairie Island specific MODE title, not consistent with Standard Technical Specification MODE titles. MODE numbers are consistent with Standard Technical Specification MODE numbers.

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

gbjective 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.

Specification l A. Shutdown Mitgin

1. The SHUTDOWN MARGIN shall be maintained within the limits specified in the ,

Core Operating Limits Report when in HOT SHUTDOWN, INTERMEDIATE SHUTDOWN ;

and COLD SHUTDOWN.

2. With the SHUTDOWN MARGIN less than the applicable limit specified in 3.10.A.1 i above, within 15 minutes initiate boration to restore SHUTDOWN MARGIN to I within the applicable limit. !

B. Power Distribution Limits f

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1. At all timps, excgpt during low power PHYSICS TESTING, measured hot channel l factors, F a and F m, as defined below and in the bases, shall meet the I following limits:

F*o x 1.03 x 1.05 * $ (Fa*"/ P) x K(Z)

F"a x 1.04** 5 Fu*" x [1+ PFDH(1-P)] l t

I where the following definitions apply:

- Fo*" is the Fa limit at RATED THERMAL POWER specified in the CORE OPERATING !

LIMITS REPORT.

- Fa*" is the Fu limit at RATED THERMAL POWER specified in the CORE OPERATING LIMITS REPORT.

- PFDH is the Power Factor Multiplier for F"M specified in the CORE OPERATING LIMITS REPORT. .

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- K(Z) is a normalized function that limits Fo(z) axially as specified in the CORE OPERATING LIMITS REPORT.

For Unit 1. Cycle 19, when the number of available moveable detector thimbles is greater than or equal to 50% and less than 75% of the total, the 5% measurement i uncertainty shall be increased to [5% + (3-T/9)(3%)] where T is the number of available thimbles.

For Unit 1 Cycle 19, when the number of available moveable detector thimbles is greater than or equal to 50% and less than 75% of the total, the 4% measurement uncertainty shall be increased to [4% + (3-T/9)(2%)] where T is the number of available thimbles, l

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TS.3.10-5 3.10.C.2. If the QUADRANT POWER TILT RATIO exceeds 1.02 but is less than 1.07 for a sustained period of more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . or if such a tilt recurs intermittently the reactor shall be brought to the HOT SHUTDOWN condition. Subsequent operation below 50% of rating, for testing shal: be permitted.

3. Except for PHYSICS TESTS if the QUADRANT POWER TILT RATIO exceeds 1.07. the reactor shall be brought to the HOT SHUTDOWN condition.

Subsequent operation below 50% of rating, for testing shall be permitted.

4. If the core is operating above 85% power with one excore nuclear channel inoperable, then the core quadrant power balance shall be determined daily and after a 10% power change using either 2 movable detectors or 4 core thermocouples per quadrant, per Specification 3.11.

D. Rod Insertion Limita

1. The shutdown rods shall be limited in physical insertion as specified in the-CORE OPERATING LIMITS REPORT when the reactor is critical or approaching criticality.

2. When the reactor is critical or approaching criticality. the control banks shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT.

3. Insertion limite do not apply during PHYSICS TESTS or during periodic exercise of individual rods. The shutdown margin specified in the Core Operating Limits Report must be maintained except for low power PHYSICS TESTING. For this test the reactor may be critical with all but one high worth full-length control rod inserted for a period not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per year provided a rod drop test is run on the high worth full-length rod prior to this particular low power PHYSICS TEST.

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TS.6.7-3 6.7.A.4. Radioactive Effluent Report The Radioactive Effluent Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 of e each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant. The material provided shall be consistent with the objectives outlined in the ODCM and in conformance with 10CFR50.36a and 10CFR50, Appendix I, Section IV.B.1.

6.7.A.5. Annual Summaries of Meteorological Data An annual summary of meteorological data shall be submitted for the previous calendar year in the form of joint frequency distributions 4 of wind speed, wind direction, and atmospheric stability at the request of the Commission.

6.7.A.6. Core Operating Limits Report

a. Core operating limits shall be established and documented in the ,

CORE OPERATING LIMITS REPORT before each reload cycle or any i remaining part of a reload cycle for the following:

1. Heat Flux Hot Channel Factor Limit ( Fe*" ) , Nuclear Enthalpy Rise Hot Channel Factor Limit ( FA,"") , PFDH, K(Z) and V(Z)

(Specifications 3.10.B.1, 3.10.B.2 ano 3.10.B.3)

2. Axial Flux Difference Limits and Target Band (Specifications 3.10.B.4 through 3.10.B.9)

3. Shutdown and Control Bank Insertion Limits i (Specification 3.10.D) l

4. Reactor Coolant System Flow Limit (Specification 3.10.J)

5. Shutdown Margin (Table TS.1-1 and Specifications 3.10 A.1 and 3.10.D.3)

b. The analytical methods used to determine the core operating limits shall be t ose previously reviewed and approved by the j NRC, specifically those described in the following documents: 1 NSPNAD-8101-A, "Qualificatien of Reactor Physics Methods for Application to PI Units" (latest approved version)

NSPNAD-8102-A, " Prairie Island Nuclear Power Plant Reload l Safety Evaluation Methods for Application to PI Units" l (latest approved version)

WCAP-9272-P-A, " Westinghouse Reload Safety Evaluation j Methodology", July, 1985 l

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B.3.8-1 I*

3.8 REFUELING AMD FUEL HANDLING l

i Bases i

Core alteration containment isolation specifications are provided to minimize releases following a fuel handling accident (FHA). Allowing both airlock doors open during core alterations will facilitate i evacuation of containment following a FHA and help maintain the seals in good working order. The FHA does not cause containment I pressurization, however, with an assumed single failure the operating ;

purge system supply fan is assumed to continue supplying air to I containment. To maintain post-FHA releases well within the limits of '

10CFR100, only the inservice purge system is allowed to be operating during core alterations. Two containment fan coil unit fans are required to operate in the high speed mcde following a fuel handling l accident in containment to assure that radioactive material in containment is well mixed and any releases will leava containment at a lower concentration over the duration of the accident. The provision that one door is OPERABLE and under procedural control will ensure that at least one door will be closed in within 30 minutes as required, thus assuring radioactive releases are well within the limits of 10CFR100.

l The equipment and general procedures to be utilized durino sfueling are discussed in the FSAR. Detailed instructions, the ; . - ations specified above, and the design of the fu.1 handling egt .ent incorporating built-in interlocks and safety features, p.avide assurance that no incident could occur during CORE ALTERATIONS that j would result in a hazard to public health and safety (Reference 1). #

l Whenever changes are not being made in core geometry, one flux monitor is sufficient. This permits maintenance of the instrumenta-tion. Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. The residual heat removal pump is used to maintain a uniform boron concentration.

Under rodded and unrodded conditions, the K.,, of the reactor must be less than or equal to 0.95 and the boron concentration must be greater than or equal to 2000 ppm. Also, the Shutdown Margin shall be maintained within the limits specified in the Core Operating Limits Report. Periodic checks of refueling water boron concentration insure that proper shutdown margin is maintained. 3.8.'.1.h allows the control room operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.

I No movement of fuel in the reactor is permitted until the reactor has i been suberitical for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months to permit decay of the fission products in the fuel. The delay time is consistent with the j fuel handling accident analysis (Reference 2).

Fuel will not be inserted into a spent fuel cask unless a minimum boron I concentration of 1800 ppm is present. The 1800 ppm will ensure that k ,,

for the spent fuel cask, including statistical uncertainties, will be less than or equal to 0.95 for all postulated arrangements of fuel within the cask. l The number of recently discharged asso,ablies in Pool No. 1 has been limited to 45 to provide assurance that in the event of loss of pool cooling capability, at least eight hours are available under worst case conditions to make repairs until the onset of boiling.

. B.3.10-1 Q

la 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Bases Throughout the 3.10 Technical Specifications, the terms " rod (s)" and "RCCA (s) d are synonymous.

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A. Shutdown Margin A sufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown, anticipated operational occurrences (AOOs), and accidents. As such, the SHUTDOWN MARGIN defines the degree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod cluster assembly of highest reactivity worth remains fully withdrawn. The primary safety analyses that rely on the SHUTDOWN MARGIN limits are the boron dilution and Main Steam Line Break analyses.

If the SHUTDOWN MARGIN requirements are not met, boration must be initiated promptly. A Completion Time of 15 minutes is adequate for an operator to f correctly align and start the required systems and components, and the i probability of an accident occurring during this time is very low. It is assumed that boration will be continued until the SHUTDOWN MARGIN requirements are met.

l l In POWER OPERATION and HOT STANDBY, with k,n 2 1, SHUTDOWN MARGIN is l ensured by complying with the rod insertion limitations in Specification l 3.10.D. In HOT SHUTDOWN, INTERMEDIATE SHUTDOWN and COLD SHUTDOWN, the l SHUTDOWd MARGIN requirements in Specification 3.10.A are applicable to provide sufficient negative reactivity to meet the assumptions of the l safety analyses discussed above. For REFUELING, the shutdown reactivity j requirements are specified in Table TS.1-1.

l l When in POWER OPERATION and HOT STANDBY, SHUTDOWN MARGIN is determined l assuming the fuel and moderator temperatures are at the nominal zero power design temperature of 547'F.

With any rod cluster control assembly not capable of being fully inserted, the reactivity worth of the rod cluster control assembly must be accounted for in the determination of SHUTDOWN MARGIN.

B. Power Distribution Control The specific &tJons 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 of greater than or equal to 1.30 for Exxen fuel and 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 mechanical propertier to within assumed design criteria. ' The ECCS analysis was performed in accordance with SECY 83-472. calculation at the 95% probability level was performed as well as one - alation with l

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l EXHIBIT D i l

PRAIRIE ISLAND NUCLEAR GENERATING STATION License Amendment Request dated April 12,1999 l

1 Sample Cote Operating Limits Report j l

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Ccre Operating Limits Report Unit 1, Cycle 20 Preliminkry Shutdown Marain Reauirements Minimum Shutdown Margin requirements are shown in Table 2.

Reference Technical Specification Sections: Table TS.1-1 and Specifications 3.10.A and 3.10.D.3.

LOCA The small and Large Break LOCA analyses performed for this cycle are valid for Fo # 2.80 and Fo # 2.40, respectively. The Fo limit for the Large Break LOCA analysis is more limiting than the Fo limit for the Small Break LOCA analysis. The Small Break LOCA analysis incorporates the K(z) methodology. However, since the Small Break LOCA is less limiting than the Large Break LOCA analysis, no K(z) penalty needs to be j applied to calculations of most limiting Fovalues. Thus for the equation in Technical l Specification 3.10.8, K(z) is equal to 1. K(z) is shown graphically in Figure 1. l I

Transient Power Distribution Penalty for Fg -V(z)

~ Table 1 summarizes the bounding V(z) values for the middle 80% of the core for Prairie Island Unit 1 Cycle 19. The V(z) penalty takes the form of straight lines connecting data points determined as a function of core height. A particular V(z) curve is valid over a given exposure range and equilibrium Axial Offset (AO) range as noted in Table 1. ,

The V(z) penalty for each exposure and AO range is shown graphically in Figures 2a - )

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e Table 2 Minimum Required Shutdown Margin Number of Charging Pumps in Service (2)

Plant Conditions 0 or 1 Pump 2 Pumps 3 Pumps

! Modes 1 and 2 2% 2% 2%

l (1) (1) (1) i Mode 3, T.v. > 520 F 2% 2% 2%

l Mode 3, Tav. < 520 F 2.5% 2.5% 2.5% l

> 350 F Mode 4 2.5% 6% 8%

l Mode 5 2.5% 6% 8%

Mode 6, ARI 5% 6% 8%

Mode 6, ARO 5% 7% 9.5%

(1) For Modes 1 and 2, minimum shutdown margin requirements are provided by the Rod insertion Limits.

(2) Charging pump (s) in service only pertains to steady state operations, it does not include transitory operations. For example, operations such as starting a second l

charging pump in order to secure the operating pump would fall under the one pump in service column.

Reviewers Note:

The values in the above table are provided for example purposes only. Actual values for entry into the COLR will be determined using approved methodology.

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g e EXHIBIT E i

PRAIRIE ISLAND NUCLEAR GENERATING STATION I

, License Amendment Request dated April 12,1999 l

Appendix A, Technical Specification Page l

l Marked Up Page From License Amendment 141 (shaded material to be added, strikethrough material to be removed)

TS.6.0-13 l

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TS.6.0-13 C. Radioactive Effluent Reparl The Radioactive Effluent Report covering the operation of the plant l during the previous calendar year shall be submitted by May 15 of l each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant. The material provided shall be consistent with the i

objectives outlined in the ODCM and in conformance with 10CFR50.36a I and 10CFR50, Appendix I, Section IV.B.1. l D. Monthly Operatine Reports Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the pressurizer power operated relief valves or pressurizer safety valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

E. Core Operating Limits Reoort (COLR)

1. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and l shall be documented in the COLR for the following:

a. Heat Flux Hot Channel Factor Limit (F/*') , Naclear Enthalpy Rise Hot Channel Factor Limit (F6/#' ), PFDH, K(Z) and V(Z) l (Specifications 3.10.B.1, 3.10.B.2 and 3.10.B.3)

b. Axial Flux Difference Limits and Target Band

! (Specifications 3.10.B.4 through 3.10.B.9)

c. Shutdown and Control Bank Insertion Limits (Specification 3.10.D) l d. Reactor Coolant System Flow Limit (Specification 3.10.J) l e.

2. The analytical methods used to determine the core operating limits j shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

NSPNAD-8101-A, " Qualification of Reactor Physics Methods for Application to PI Units" (latest approved version)

NSPNAD-8102-A, " Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units"(latest approved version)

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EXHIBIT F PRAIRIE ISLAND NUCLEAR GENERATING STATION s

License Amendment Request dated April 12,1999 Appendix A, Technical Specification Page Revised Page From License Amendment 141 TS.6.0-13 3

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TS.6.0-13

. , C. Radioactive Effluent Renort The Radioactive Effluent Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant. The material provided shall be consistent with the l objectives outlined in the ODCM and in conformance with 10CFR50.36a {

and 10CFR50. Appendix I.Section IV.B.1. I D. Monthiv Operating Recorts y I

Routine reports of operating statistics and shutdown experience, j including documentation of all challenges to the pressurizer power i operated relief valves or pressurizer safety valves, shall be submitted on a monthly basis no later than the 15th of each month i

following the calendar month covered by the report.

E. Core Operating Limits Reoort (COLR) l 1. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle. and shall be documented in the COLR for the following: j

a. Heat Flux Hot Channel Factor Limit (F/*') Nuclear Enthalpy i

Rise Hot Channel Factor Limit (Fo/*' ) . PFDH, K(Z) and V(Z) '

' Specifications 3.10.B.1. 3.10.B.2 and 3.10.B.3)

b. Axial Flux Difference Limits and Target Band (Specifications 3.10.B.4 through 3.10.B.9)

c. Shutdown and Control Bank Insertion Limits (Specification 3.10.D)

d. Reactor Coolant System Flow Limit (Specification 3.10.J)

e. Shutdown Margin (Table TS.1-1 and Specifications 3.10.A.1 and 3.10.D.3)

2. The analytical methods used to determine the core operation limits shall te those previously reviewed and approved by the NRC, specifically those described in the following documents:

NSPNAD-8101-A. " Qualification of. Reactor Physics Methods for Application to PI Units" (latest approved version)

NSPNAD-8102*A. " Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units"(latest approved version)

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ML20205Q035

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