ML20044G582

Application for Amend to License NPF-42,requesting Rev of TS 3/4.4.9.1,pressure/temp Limits RCS & 3/4.4.9.3, Overpressure Protection Sys

ML20044G582
Person / Time
Site:	Wolf Creek
Issue date:	05/27/1993
From:	Hagan R WOLF CREEK NUCLEAR OPERATING CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20044G583	List: ML20044G582 ML20044G772
References
NA-93-0131, NA-93-131, NUDOCS 9306030346
Download: ML20044G582 (11)

Text

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i s -e W8LF CREEK ' NUCLEAR OPERATING CORPORATION Robert C. Hagan .

Mce President Nuclear Assurance NA 93-0131 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk

• Mail Station Pl-137 Washington, D. C. 20555

Subject:

Docket No. 50-482: Revision to Technical Specifications 3/4.4.9.1 - Pressure / Temperature Limits Reactor Coolant System and 3/4.4.9.3 - Overpressure Protection System Gentlemen: ,

This letter transmits an application for amendment to Facility Operating i License No. NPF-42 for Wolf Creek Generating Station (WCGS). This license _ ,

amendment request proposes revising Technical Specifications 3.4. 9.1, Figare 3.4-2, Figure 3.4-3 and Figure 3.4-4. These revisions are in compliance'with the requirements cf 10 CFR 50, Appendix H and Technical Specification 4.4.9.1.2. The results of the analysis of reactor vessel material Capsule

'Y", removed during the fifth refueling outage, and 'the. effects of the: ,

proposed power rerate for WCGS were considered in developing these revisions.

Attachment I provides a safety evaluation along with a description of the proposed change. Attachment II provides a significant hazards consideration determination and Attachment III provides an _ environmental impact determination. The specific changes to the technical specificatione proposed  :

by this request are prc,vided in Attachment IV.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Kansas State Official. This proposed revision to the WCGS technical specifications will be fully implemented within t 30 days of formal Nuclear Regulatory Commission approval.

6 020080 j 9306030346 930527 Ro. Box 411/ Burkngton. KS 66839 / Phone (316) 364-8831 gh 3, cua,i oppo,,un,ty envoyer wrmcuct ii PDR ADDCK 050004B2 P PDR .h f.

.. NA,93-0131' .1 Page ?'.of 2 ,

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l If you have any questions concerning this ' matter, please - contact me at 1 (316) 364-8831 extension 4553 or Mr. Kevin:J. Moles'at extension 4565.  !

Very.truly yours, 'f

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P bert C. Hagan- . _.

Vice President Nuclear Assurance.' .{

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Attachments I - Change Description.and Safety Evaluation 'i' II - No Significant Hazards Consideration Determination III - Environmental Impact Determinatio.4 l IV - Proposed Technical Specification Changes- .j cc: G. W. Allen (KDHE), w/a l W. D. Johnson (NRC), w/a  !

J. L. Milhoan (NRC), w/a '

G. A. Pick (NRC)., w/a  ;

W. D. Reckley (NRC), w/a r

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STATE OF KANSAS )

) SS COUNTY OF COFFET )

Robert C Hagan, of lawful age, being first duly sworn upon oath says.that he is Vice President Nuclear Assurance of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the. content thereof; that he has executed that same for and on. behalf of said' Corporation with full power and authority to do so;- and that the facts-therein stated are true and correct to the best- of his knowledge, information and belief.

Rober/C.Hagan // .

Vice President [/

Nuclear Assurance SUBSCRIEED and sworn to before me this 27 day of dep,1993.

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a.;r...$:y ( +/a%)/nov Notary Public

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,,- . U B L 1.Gih,i EXP i ration Date // R$

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i Attachment I to NA 93-0131 ,

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Attachment I .

Safety Evaluation .[

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. i Attachmerit I~ to NA 93-0131

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Safety Evaluation i Proposed Channe f This license amendment request proposes to revise the heatup, cooldown and f Cold Overpressure Mitigation System (COMS) Power-Operated Relief Valve (PORV)  ;

setpoint pressure / temperature (P/T) limits as required by 10 CFR 50 Appendix H- l and Technical Specification 4.4.9.1.2. This amendment request proposes a j revision to Figure B 3/4.4-1 to indicate the projected fluence values used in .{

determining the limits for Figures 3.4-2, 3.4-3 and 3.4-4. Figure B 3/4.4-1  ;

was revised using data from WCAP-13365, " Analysis of Capsule Y from the Wolf Creek Nuclear Operating Corporation Wolf Creek Reactor Vessel Radiation Surveillance Program."

This amendment request also proposes a change to Bases page B 3/4.4-7 which deletes the word *either" so that this paragraph now indicates that only j '

- Regulatory Guide 1.99, Revision 2 was used. This change was an oversight in an earlier Technical Specification Change Request and is being changed . to f clarify this section, j

Reason for Proposed Change Technical Specification 4.4.9.1.2 requires that the heatup and cooldown .!

pressure / temperature (P/T) limit curves shown in Figures 3.4-2 and 3.4-3 and.

the COMS PORV setpoint limit curve shown in Figure 3.4-4 be re-evaluated and, )

if necessary, updated in the technical specifications, on the basis of the i results of surveillance testing of irradiated vessel material samples. )'

Capsule "Y" of the Wolf Creek Generating Station (WCGS) reactor vessel surveillance program specimens was withdrawn from the reactor during the fifth .. ;

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refueling outage. The results of testing and evaluation of the specimens from Capsule "Y were provided in WCAP-13365, " Analysis of Capsule Y.from the Wolf i' Creek Nuclear Operating Corporation Wolf Creek Reactor Vessel Radiation 4

Surveillance Program" which was submitted to the NRC on November 5, 1992 by  ;

letter ET 92-0216. j I

Evaluation  ;

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The heatup and cooldown P/T limit curves, and the COMS PORV setpoint limit l curve define the range of acceptable operation for the Reactor Coolant System' (RCS) in order to protect the vessel against non-ductile ' f ailure. These j limits ensure that the margin of protection against non-ductile failure is  ;

maintained in accordance with 10 CFR 50, Appendix G, requirements. This is  ;

accomplished by limiting the maximum allowable pressure ' at low temperatures. {

The fluence projections for an increased power level from 3411 megawatts.

thermal (MWth) to 3565 MWth have been factored into the development of the j limit curves.

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Attachment I to NA 93-0131 Page 3 of 4 The heatup and cooldown limit curves were generated by Westinghouse using the  ;

information from the testing and analysis of Capsule 'Y" specimens. These  ;

limit curves were calculated in accordance with 10 CFR 50, Appendix G and ASME l 4

Code Section III Appendix G requirements. The basis for the revised heatup and cooldown limit curves for the reactor pressure vessel are:

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a. the lower ~shell plate, R2508-3, is the limiting material for heatup and cooldown curves applicable up to 13.6 Effective Full Power Years (EFPY),

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b. lower calculated RTNDT (reference nil-ductility temperature) values I of 89*F and 79*F at 1/4 and 3/4 wall thickness locations respectively, as referenced from the inner wetted surface of the i reactor vessel, since surveillance capsule data was used to determine the limiting Adjusted Reference Temperature 'of the l beltline materials, and i

c. the revised RTNDT values were calculated using Regulatory Guide 1.99 >

Revision 2. t l

The revised PORV setpoint limits for the COMS were derived using the same l methodology employed in. the development of the current COMS PORV setpoints.

The COMS PORV setpoint limit curve (Figure 3.4-4) is determined based on the  !

revised heatup and cooldown limit curves, heatup and .cooldown rate i applicability limits, and the analysis results of limiting Low Temperature ,

Over-Pressure (LTOP) transients. The limiting LTOP mechanisms analyzed for j WCGS under water solid conditions were:

a. FOR LIMITING MASS ADDITION LTOP MECHANISM t

Operation of one Centrifugal Charging Pump (CCP) with instrument' air  ;

failure resulting in the flow control valve in the letdown line i

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failing closed (letdown isolation) and the flow control valve in the charging line failing open (maximum charging flow), and' l I

b. FOR LIMITING HEAT ADDITION LTOP MECHANISM i l

Inadvertent start-up of a reactor coolant pump with a maximum 50'F l temperature mismatch between the RCS and the hotter steam.  ;

generators. l l

These analyses using the LOFTRAN computer code take into consideration l pressure overshoot and undershoot beyond the PORV open and close setpoints. .l vhich can occur as a result of time delays in signal processing and valve .;

stroke times. The maximum expected pressure overshoot and undershoot -l calculated from the limiting mass input and heat input transients, in conjunction with the Appendix G pressure limits and reactor coolant pump No. 1 seal pressure limit, are utilized in the selection of the pressure setpoints for the PORV. It should be noted that the mass injection rate assumed in the design basis mass input transient include an allowance of 100 gpm. This allowance has been added to envelop the maximum flow possible by the operational configuration that uses the Positive Displacement Pump for charging with one Centrifugal Charging Pump (CCP) remaining operable during shutdown modes.

1 Attachment I to NA 93-0131 ,

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The pressure difference from the wide range pressure transmitter to the i t

beltline region of the reactor vessel, has been properly accounted for.in the determination of the maximum allowed PORV setpoint. Specifically, the  !

calculated maximum allowable PORV setpoint (throughout the entire temperature  ;

range for which COMS is enabled) is reduced by an amount equivalent to the applicable differential pressure when the reactor coolant.. pumps are in  ;

operation. The actual open. setpoints of the PORVs fa the COMS are less than the maximum allowed PORV setpoints and are normally staggered so that only one j valve would open during any cold overpressure transient. Operation with the setpoints of both valves within the limits of Figure 3.4-4 ensures that the Appendix G limits will not be exceeded for the events analyzed. 1 The design and analysis of COMS satisfies single failure criteria (failure of one of two PORVs available for over-pressure relief) by assuming that only one ,

PORV is available to provide pressure relief. Each of the two PORVs is 3 capable of providing 100Z of the pressure relief necessary to mitigate  ;

overpressure transients. l The major effect of adopting the revised P/T limit curves and the COMS PORV setpoint limit curve is re-defining of the P/T range for acceptable operation [

at low temperatures. The revised range for acceptable operation compensates  !

for in-service radiation induced embrittlement of the WCGS reactor pressure .

vessel in a conservative manner. The revised P/T limit curves,also account '!

for a requirement that the minimum metal temperature of the closure head j

flange and vessel flange regions must be at least 120*F higher than the.  ;

limiting RTNOT for these regions when the pressure exceeds 20% of the ,

I preservice hydrostatic test pressure. The initial RTNDT of 20*F occurs in both the closure head flange and the vessel flange of the Wolf Creek reactor  ;

vessel, so the minimum allowable temperature of . these regions is 140*F. These  ;

limits are shown in Figures 3.4-2 and 3.4-3 whenever applicable.

i The revised PORV setpoint limit curve also provides a basis f or updating the  ;

COMS Setpoint Program. Since the PORV setpoint limit. curve is more limiting i in the lower temperature range than the curve it is intended to replace, the l l breakpoints of the function generator in the current COMS Setpoint Program will be reset within the limits of Figure 3.4-4 to assure the operability of i

l the COMS. The operability of the COMS ensures that the RCS pressure will be maintained within acceptable limits following a design basis overpressure ,

transient which might occur during low temperature, water-solid operation when the Residual Heat Removal System relief valves are isolated from the RCS.

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Attachment II to NA 93-0131 Page 1 of 2 Attachment II No Significant Hazards Consideration Determination

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i Attachment II to NA 93-0131 J Page 2 of 2 No Significant Hazards Consideration Determination i

The proposed amendment has been reviewed per the standards provided in 10 CFR 50.92. Each standard is discussed separately below. .

Standard I - Involves a Significant Increase in the Probability or .

Consequences of an Accident Previously Evaluated {

Incorporating the revised heatup, cooldown pressure / temperature limit curves and the COMS PORV setpoint limit curve into Wolf Creek Generating Station (WCGS) technical specifications does not affect the probability or consequences of an accident previously evaluated.

t The revised limit curves are calculated using the most limiting RTNDT for the reactor vessel components and include a radiation induced shift corresponding i to the end of the period for which the curves are generated. The changes do '

not affect the basis, initial conditions, initiating events, chronology, or availability / operability of safety related equipment required to ' mitigate transients and accidents analyzed for WCGS.

Standard II - Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated Adopting the revised limit curves redefines the range of acceptable operation for the Reactor Coolant System. This redefinition is a result of the 1 analysis of reactor vessel surveillance specimens removed from the reactor in. I a continuing surveillance program which monitors the . effects of neutron irradiation on the WCGS reactor pressure vessel materials under . actual j operating conditions. Incorporating these revised curves does not create the possibility of an accident of a different type from any previously evaluated  !

for WCGS.

Standard III - Involve a Significant Reduction in the Margin of Safety The revision of these limit curves is not a design change and continues to j maintain the margin of safety required for prevention of non-ductile failure  ;

of the WCGS reactor pressure vessel during low temperature . operation as required by 10 CFR 50, Appendices G and H. The revised curves primarily affect RCS operation below 350*F by limiting the available pressure / temperature window for heatup and cooldown. The revised limit curves ,

compensate for the in-service radiation induced embrittlement of the reactor ,

vessel and accounts for the requirement that the closure flange region temperature must exceed the nil-ductility temperature by at least 120 F when .;

pressure exceeds 20% of the preservice hydrostatic test pressure. l 1

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Attachment.III to.NA 93-0131 Page 1 of. 2 ,

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i Attachment III Environmental Impact Determination I

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i Attachment III to NA 93-0131 Page 2 of 2 ,

Environmental linpact Determination This amendment-request meets the criteria specified in 10 CFR 51.22(c)(9) as specified below:  ;

(i) the amendment involves no significant hazards consideration .j

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As demonstrated in Attachment II, the proposed changes do not involve- any-significant hazards considerations.

i (ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

't The proposed changes do not involve generation or release of effluents from ,

the plant. The changes . impact surveillance - requirements for reactor power- ,

distribution . used to assure the operation of the plant within its safety-  !

design basis. Therefore, the proposed changes will have no effect.on normal ~ ,

plant effluents and there will be no change in the types or amounts of any >

effluents released offsite.

t (iii) there is no significant increase in individual or cumulative occupational radiatica exposure ,

t The proposed changes to 'urveillance requirements will have no effect on

, l general levels of radiation present in the plant; nor will: additional.

quantities of radioactive materials be generated as a result of the proposed  ;

changes. Therefore, there will be no increase in individual . or cumulative 6 occupational radiation exposure associated with this proposed change. ;j Based on the above it is concluded that there will be no impact on the  ;

environment resulting from this change and the change meets the criteria ,

specified in 10 CFR 51.22 for a categorical exclusion from the requirements of 5 10 CFR. 51.21 relative to specific environmental assessment by the Commission. ,

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