Text

Summary of 990426 Meeting with Util at NRC Headquarters Re Proposed Submittal of Amend to Implement Ists.List of Meeting Participants & Matls Presented by Licensee Encl
	ML20206Q641
Person / Time
Site:	Point Beach
Issue date:	05/12/1999
From:	Wetzel B; NRC (Affiliation Not Assigned)
To:	; NRC (Affiliation Not Assigned)
References
	NUDOCS 9905190185
	Download: ML20206Q641 (36)
	v • d • e

P .,

4 May 12, 1999.

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LICENS' EE: Wisconsin Electric Power Company FACILITY: Point Beach Nuclear Plant, Units 1 and 2

SUBJECT:

MEETING BETWEEN THE WISCONSIN ELECTRIC POWER COMPANY AND THE NRC STAFF CONCERNING IMPROVED STANDARD TECHNICAL SPECIFICATIONS SUBMITTAL -

The NRC staff met with representatives from Wisconsin Electric Power Company at NRC Headquarters on April 26,1999, regarding the proposed submittal of its amendment to implement improved standard technical specifications (ISTS) for Point Beach Units 1 and 2.

Attachment 1 lists the meeting participants. Attachment 2 is a copy of the material presented by i the licensee and discussed at the meeting. .

1 The licensee plans to submit its amendment for ISTS in two parts; Technical Specification

. sections 1.0,2.0 and 3.0 in July 1999, and sections 4.0 and 5.0 in August 1999. The licensee discussed the format and content of their submittal with the staff and asked questions regarding the staff's review process for ISTS. Wisconsin Electric will be requesting staff approval by i October 2000, in order to implement the new Technical Specifications sometime in early 2001, i Original signed'by:

Beth A. Wetzel, Senior Project Manager, Section 1 Project Directorate Ill.

Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301 Attachments: As stated /

cc w/atts: See next page g gy -

l DISTRIBUTION: l Hard Copv E-Mail Docket Files S. Collins /R. Zimmerman (SJC1/RPZ) Bob Dennig (RLD)

PUBLIC B. Sheron (BWS) _

William Beckner (WDB)

PD3-1 r/f - : J. Zwolinski/S. Black (JAZ,/SCB)

B. Wetzel C. Thomas /G. Dick OGC T. Harris '

.ACRS: T. Hiltz (TGH)

C. Harbuck (CCH) -

DOCUMENT NAME:MTS42699.WPD ;

OFFICE PM:LPD3 E' LA:LPD3 (A)SC:LPQ3 6 NAME BWetzel M THarris Odn GDick ff TM u DATE 05////99 05.4/99 054/99^ '

OFFICIAL RECORD COPY 190022 "

9905190185 990512 '

PDR ADOCK 05000266 P PM ;

J

May 12, 1999

, i LICENSEE: Wisconsin Electric Power Company FACILITY: Point Beach Nuclear Plant, Units 1 and 2

SUBJECT:

MEETING BETWEEN THE WISCONSIN ELECTRIC POWER COMPANY AND THE NRC STAFF CONCERNING IMPROVED STANDARD TECHNICAL SPECIFICATIONS SUBMITTAL The NRC staff met with representatives from Wisconsin Electric Power Company at NRC Headquarters on April 26,1999, regarding the proposed submittal of its amendment to implement improved standard technical specifications (ISTS) for Point Beach Units 1 and 2.

Attachment i lists the meeting participants. Attachment 2 is a copy of the material presented by the licensee and discussed at the meeting.

The licensee plans to submit its amendment for ISTS in two parts; Technical Specification sections 1.0,2.0 and 3.0 in July 1999, and sections 4.0 and 5.0 in August 1999. The licensee discussed the format and content of their submittal with the staff and asked questions regarding the staff's review process for ISTS. Wisconsin Electric will be requesting staff approval by October 2000, in order to implement the new Technical Specifications sometime in early 2001.

Original signed by:

. Beth A. Wetzel, Senior Project Manager, Section 1 Project Directorate lli Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301 .

Attachments: As stated cc w/atts: See next page DISTRIBUTION:

Hard Copy E-Mail Docket Files S. Collins /R. Zimmerman (SJC1/RPZ) Bob Dennig (RLD)

PUBLIC B. Sheron (BWS) William Beckner (WDB)

PD3-1 r/f J. Zwolinski/S. Black (JAZ,/SCB)

B. Wetzel C. Thomas /G. Dick OGC T. Harris ACRS T. Hiltz (TGH).

C. Harbuck (CCH)

DOCUMENT NAME:MTS42699.WPD OFFICE PM:LPD3 Ei LA:LPD3 (A)SC:LPQ3 , ,6 NAME BWetzel fA1) THarris Odn GDick J$ M DATE 05/s//99 05M/99 054/99" '

OFFICIAL RECORD COPY 6

"84 g i UNITED STATES

j. j NUCLEAR REGULATORY COMMISSION o $ WASHINGTON, D.C. 20086 0001

- - - - May 12, 1999 LICENSEE: Wisconsin Electric Power Companu FACILITY: Point Beach Nuclear Plant, Units 1 and 2

SUBJECT:

MEETING BETWEEN THE WISCONSIN ELECTRIC POWER COMPANY AND THE NRC STAFF CONCERNING IMPROVED STANDARD TECHNICAL SPECIFICATIONS SUBMITTAL The NRC staff met with representatives from Wisconsin Electric Power Company at NRC Headquarters on April 26,1999, regarding the proposed submittal of its amendment to implement improved standard technical specifications (ISTS) for Point Beach Units 1 and 2.

Attachment i lists the meeting participants. Attachment 2 is a copy of the material presented by the licensee and discussed at the meeting.

The licensee plans to submit its amendment for ISTS in two parts; Technical Specification sections 1.0,2.0 and 3.0 in July 1999, and sections 4.0 and 5.0 in August 1999. The licensee discussed the format and content of their submittal with the staff and asked questions regarding the staff's review process for ISTS. Wisconsin Electric will be requesting staff approval by October 2000, in order to implement the new Technical Specifications sometime in early 2001.

k a. ap' Beth A. Wetzel, Senior Project Manager, Section 1 Project Directorate Ill Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301 Attachments: As stated cc w/atts: See next page

~

, . - . ~

. 1 Mr. Michael B. Sellman Wisconsin Electric Power Company Point Beach Nuclear Plant Units 1 and 2 cc: l Mr. John H. O'Neill, Jr.

Shaw, Pittman, Potts & Trowbridge Ms. Sarah Jenkins 2300 N Street, NW Electric Division Washington, DC 20037-1128 Public Service Commission of Wisconsin P.O. Box 7854 Mr. Richard R. Grigg Madison, Wisconsin 53707-7854 President and Chief Operating Officer Wisconsin Electric Power Company ,

231 West Michigan Street Milwaukee, Wisconsin 53201 j

Mr. Mark E. Reddemann !

Site Vice President Point Beach Nuclear Plant Wisconsin Electric Power Company 6610 Nuclear Road '

Two Rivers, Wisconsin 54241 Mr. Ken Duveneck Town Chairman Town of Two Creeks i 13017 State Highway 42 l

Mishicot, Wisconsin 54228 Chairman Public Service Commission of Wisconsin P.O. Box 7854 Madison, Wisconsin 53707-7854 Regional Administrator, Region til U.S. Nuclear Regulatory Commission 801 Warrenville Road l

Lisle, Illinois 60532-4351 Resident inspector's Office L U.S. Nuclear Regulatory Commission 6612 Nuclear Road Two Rivers, Wisconsin 54241 1

ocionwises l

MEETING ATTENDEES Naale Tde Beth A. Wetzel Senior Project Manager, NRC Bill Herrman ITS Implementation Team, WEPCo Curt Castell ITS Project Manager, WEPCo Vito Kaminskas Regulatory Services Manager, WEPCo Jack Gadzala Licensing Manager, WEPCo Craig Harbuck TSB Lead Reviewer, NRC Bob Dennig TSB Section Chief, NRC William Beckner Chief TSB, NRC ra Attachment 1 l

i p

i Topics for Discussion Standard Technical Specifications Conversion Point Beach Nuclear Plant April 26, 1999 '

- Format of the submittal:

WEPC0 does not currently use Wordperfect The Standard Tech Specs and Current Tech Specs are in Word 97 DOCS, JFDs and NSHCs are RTF files output from Access 97 database !

Not planning to submit DOC Summary Tables with initial submittal Number of Paper Copies Required Number of Digital Copies (File by Section: CD-ROMS or 3.5" Disks)

Contents of submittal:

Cross-reference Tables I l

Description of Changes )

Current Tech Spec Mark-up No Significant Hazards Consideration Justification for Deviations NUREG-1431 Mark-up Clean copy of Proposed Technical Specifications Schedule July 1999: 1.0, 2.0 and 3.0 (All LCOs)

August 1999: 4.0 and 5.0 October 2000: NRC Approval for Implementation in 2001 Attachment 2

M INF3RMRTl0N gyty Cross-Reference Report -ITS LCO 3.04.10 l CTS to ITS os. gov.98 -

CTS ITS DOC l l

15.03.01 A04.A FSAR R.01 -

15.03.01 A04.8 LCO 3.04.10 M.02 LCO 3.04.10 M.01 LCO 3.04.10 A.01 15.04.01 T 15.04.01-0211 SR 3.04.10.01 R.01 I SR 3.04.10.01 1A01 SR 3.04.10.01 A.01 15.04.01 T 15.04.010211 (11) N/A LA.01 Page 1 of 1

. .....e.. . . . . -.+e. . . - . - + ~*

FOR INJORMATION DlL.y Cross-Reference Report -ITS LCO 3.04.10 .

ITS to CTS os.Nov-98 ITs cts Doc j l

LCO 3.04.10 15.03.01 A04.B M.02 ]

15.03.01 A04.B M.01 )

15.03.01A04.B A.01 LCO 3.04.10 COND A NEW LO1 j LCO 3.04.10 COND A RA A.1 NEW LO1 l LCO 3.04.10 COND B NEW Loi l LCO ?.04.10 COND B RA B.1 NEW LO1 LCO 3.04.10 COND B RA B.2 NEW Lot LCO 3.04.10 NOTE NEW LO2 SR 3.04.10.01 15.04.01 T 15.04.010211 R.01 !

15.04.01 T 15.04.01-0211 LA.01 l 15.04.01 T 15.04.01-0211 A.01 l

t i

Page 1 of 1

V FOR INF0kMATION DIEY Description of Changes -ITS LCO 3.04.10 .

05-Nov-98 DOC Number DOC Text A.01 In the conversion of Point Beach current Technical Specifications (CTS) to the proposed plant specific improved Technical Specifications (ITS) certain wording preferences or conventions are adopted which do not result in technical changes (either actual or interpretational). Editorial changes, reformatting, and revised numbering are adopted to make the ITS consistent with the Standard Technical Specifications, Westinghouse Plants, NUREG-1431, Revision 1 (i.e.,

improved Standard Technical Specifications (ISTS)).

CTS: ITS:

15.03.01.A04.8 LCO 3.04.10 15.04.01 T 15.04.01-0211 SR 3.04.10.01 LO1 CTS 15.3.1.A.4.b requires both pressurizer safety valves to be operable whenever the reactor is critical, but does not provide any actions if this LCO is not satisfied. Therefore in accordance with CTS 15.3.0.b, the plant is placed in a non-applicable mode in 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months . Proposed ITS 3.4.10, Condition A, is entered whenever a pressurizer safety valve is inoperable. Condition A Actions

, require the res% ration of the valve to an operable status within 15 minutes. The Completion Time of 15 minutes reflects the importance of maintaining the RCS overpressure protection system. An inoperable pressurizer safety valve coincident with an RCS overpressure event could challenge the integrity of the pressure boundary, in the event the pressurizer safety valve cannot be restored within 15 minutes, or both pressurizer safety valves are inoperable, Condition B is entered. Condition B Actions require the plant to be placed in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 4 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . These actions result in placing the plant in a non applicable mode in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is based on operating experience to reach the required plant condition from a full power condition in an orderly manner and without challenging plant systems. Extending the time allowed to place the plant in a non-app!! cable mode from 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is less restrictive. This is acceptable, based on the broader LCO App!!cability adopted as part of ITS 3.4.10, and the increased time required to place the plant in a non-app!' cable mode from full power conditions.

CTS: IT8:

NEW LCO 3.04.10 COND A LCO 3.04.10 COND A RA A.1 LCO 3.04.10 COND B LCO 3.04.10 COND B RA B.1 LCO 3.04.10 COND B RA B.2 Page 1 of 4

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FOR HF0aMni10N ON!.Y DOC Number DOC Text LO2 CTS Specification 15.3.1 A4.b, which requires that both pressurizer safety valves be operable when the reactor is critical, is revised to add ITS LCO 3.4.10 NOTE, which allows the safety valve lift settings to be outside the LCO limits for the purpose of setting the safety valves under ambient (hot) conditions. Because this note allows the pressurizer safety valves to be potentially inoperable in MODE 3 until the safety valves can be tested and set, this change is less restrictive. This change is acceptable because the limitations included in the note (i.e., a maximum of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months allowed following entry into MODE 3) assure that reactor decay heat is significantly reduced below the assumptions in the applicable safety analyses for LCO 3.4.10.

This permits testing and examination of the safety valves at high pressure and temperature near their normal operating range, but only after the valves have had a preliminary cold setting. The cold setting gives assurance that the valves are OPERABLE near their design condition. Only one valve at a time will be removed from service for testing. The 36 hour4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months exception is reasonable based on 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months outage time for each of the valves. The 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months period is derived from operating experience that hot testing can be performed in this time frame.

CTS: ITS:

NEW LCO 3.04.10 NOTE LA01 CTS 15.4.1, Table 15.4.1-2, item 11, requires that pressurizer safety valve setpoints be checked at a frequency of"every five years." The frequency is modified by Note (11), which specifies "An approximately equal number of valves shall be tested each refueling outage such that all valves will be tested within a five year period. If any valve fails its tests, an additional number of valves equal to the number originally tested shall be tested. If any of the additional tested valves fall, all remaining valves shall be tested." These details have been moved from the Technical Specification to licensee control as these details are not necessary to describe the actual regulatory requirement. Therefore, proposed ITS SR 3.4.10.1 requires verifying "each pressurizer safety valve is OPERABLE in accordance with the inservice Testing Program", at a frequency of"In accordance with the Inservice Testing Program."

The testing details located in CTS 15.4.1, Table 15.4.1-2, item 11, are not required to be in the ITS to provide adequate protection of public health and safety, as the regulatory requirement (IST Program) is being maintained in the Technical Specifications. Changes to plant procedures and other plant controlled documents are subject to controls imposed by plant administrative procedures, which endorse applicable regulations and standards. Inservice Testing of pressurizer safety valves will continue to be performed in accordance with the IST Program.

CTS: ITS:

15.04.01 T 15.04.010211 SR 3.04.10.01 15.04.01 T 15.04.01-0211 (11) N/A M.01 CTS 15.3.1 A4.b requires both pressurizer safety valves to be operable. Proposed ITS 3.4.10 requires two pressurizer safety valves to be operable with lift settings greater than or equal to 2410 psig and less than or equal to 2560 psig. The pressurizer safety valve settings are not stated in the CTS, but are maintained in ths ITS Program. Stating the safety valve settings in ITS LCO 3.4.10 is more restrictive and is consistent with NUREG 1431.

CTS: ITS:

15.03.01 A04.B LCO 3.04.10 Page 2 of 4

. FOR INFORfl ATION ONI.Y DOC Number DOC Text M.02 CTS 15.3.1 A4.b requires both pressurizer safety valves to be operable whenever the reactor is critical. Propos'ed ITS 3.4.10 requires two pressurizer safety valves to be operable in MODES 1, 2, and 3. In MODES 4 and 5, and MODE 6, with the reactor vessel head on, overpressure protection is provided by operating procedures and by meeting the requirements of LCO 3.4.12, l

" Low Temperature Overpressure Protection (LTOP) System." In MODES 1,2 and 3 operability I of the pressurizer safety valves is required because the combined capacity keeps reactor coolant pressure below 110% of its design value during certain accidents. Expanding the applicability of the LCO to include MODE 3 places additional requirements on plant operation and is therefore more restrictive. This change is necessary to assure the RCS is provided with adequate overpressure protection in all required modes of operation. '

CTS: ITS: '

15.03.01.A.04.B LCO 3.04.10 t i e

i Page 3 of 4

FOR IHORid ATION ONI.Y DOC Number DOC Text R.01 Wisconsin Electric Power Company has utilized the selection criteria provided in the 10 CFR l 50.38.11, and has concluded that the Pressurizer Safety Valve LCO and Surveillances for l conditions below the LTOP enabling Temperature can be relocated to licensee control. The basis for this conclusion is as follows:

The pressurizer safety valves protect the RCS from being pressurized above the RCS pressure Safety Limit. The pressurizer safety valves provide over-pressurization protection during Modes 1,2 and 3 above the low temperature overpressure protection (LTOP) enabling temperature.

The pressurizer safety valves are not assumed to function to mitigate a DBA or transient below the LTOP enabling temperature. Overpressure protection is provided under these conditions by the LTOP requirements.

Comparison to Deterministic Screening Criteria:

1. Pressurizer safety valves are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.

2. Pressurizer safety valves are not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient.

3. Pressurizer safety valves are not part of a primary success path in the mitigation of a DBA or transient in Mode 4 below the LTOP arming temperature and Mode 5 or 6.

4. The loss of pressurizer safety valves (< 200*F) was found to be a non significant risk '

contributor to core damage frequency and offsite releases. Wisconsin Electric Power Company has reviewed this evaluation and considers it applicable to Point Beach Station. The Point Beach IPE does not cover plant conditions with RCS temperatures less than 200 *F, and so it presents no conflict with the generic analysis.

Conclusion:

Since the screening criterla have not been satisfied, the Pressurizer Safety Valvo LCO and Surveillances for conditions below the LTOP arming Temperature may be relocated to other plant controlled documents outside the Technical Specifications.

CTS: ITS:

15.03.01.A.04.A FSAR 15.04.01 T 15.04.01-0211 SR 3.04.10.01 Page 4 of 4

't

""""" ""'V F O R I N "w' "nResiausi nvieRemwe rodp (x). M =>== 3 4 to

,,,,1 ,,3 i (d) Residual Heat Removal Loop (B)* i U)" If the conditions of specification (1) above cannot be met, corrective action to retum a second decay heat removal method to operable status as soon as possible shall be initiated immediately.

(3) If no decay heat removal method is in operation, except as permitted by (4) !

below, all operations causing an increase in the reactor decay heat load or a i reduction in reactor coolant system boron concentration shall be suspended. !

Corrective actions to retum a decay heat removal method to operation shall be !

initiated immediately.

(4) At least one of the above decay heat removal methods shall be in operation.

(a) All reactor coolant pumps and residual heat removal pumps may be deenergized for up to 1 bour in any 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided:

:

~~No operations are permitted that would cause dilution ofreactor J~~

~~(1) '~~

coolant system boron concentration, and (2) Core outlet temperature is maintained at least 10 F below j saturation temperature, j l b. Reactor Coolant Temperature Less Than 140"F : ~~j (1) Both residual heat removal loops shall be operable except as permitted in items j (3)or(4) below. -~~

~~l (2) If no residual heat removal loop is in operation, all operations causing'an i increase in the reactor decay heat load or a reduction in reactor coolant system ,~~

boron concentration shall be suspended. Corrective actions to retum a decay heat removal method to operation shall be initiated immediately. j (3) One residual heat removal loop may be out of service when the reactor vessel l

~~head is removed and the refueling cavity flooded. !~~

~~(4) One of the two residual heat removal loops may be temporarily out of service I to meet surveillance requirementsj g ;~~

~~4. Pressurizer Safety Valves '~~
la. At least one pressunzer safety valve shall be operable whenever the reactor head is onl the vesecl.
lLco 3.4.10 l b. lBoth pressurizer safety valves shall be operabicHwhenever the reactor is criticalp Q
h 4^.dd Actions A & B l l Add LCO 3.4.10 NOTE k
' Mechanical design provisions of the residual heat removal system afford the necessary flexibility to allow an operable residual heat removal loop to consist of the RHR pump from one loop coupled with the RHR heat exchanger from the other loop. Electrical design )
provisions of the residual heat removal system afford the necessary flexibility to allow the f normal or emergency power source to be inoperable or tied together when the reactor coolant j temperatureisless than 200*F F < s Leo 3.4.6 > !
i Unit 1 - Amendment 149 15.3.1-2 August 16,1994 l Unit 2 - Amendment 153 )
k - . - - ~ , - .,. .~__. . . - . m f
FOR IN F 0 B A TAELT Q15.4.1 4 ihp(Contmued)
Y C Pd 3 Inst Freauency l 7. Spent Fuel Pit a) Boron Concentration Monthly +--4 See LCOs 3.7.15 b) Water Level and 3.7.16 >
Verification Weekly l 8. Secondary Coolant Gross Beta gamma Weekly
~~l ,~~
Activity or gamma l isotopic analysis : < See LCO 3.4.16 >
Iodine concentration Weekly when gross Beta-gamma activity equals or exceeds r-1.0 pCi/gWl
~~19. Control Rods a) Rod drop times of all Each refueling or fulllength rods~~
after maintenance that could afTect proper functioning *
< See LCO 3.1.5 ' : b) Rodworth measurement Following each refueling shutdown prior to commencing power operation 110 Control Rod Partial movement of Every 2 weeks 00 I all rods I
~~11. Pressurizer Safety Valves Set point l Every five years ""l b~~
l 12. Main Steam Safety Valves Set Point Every five years "Ul N See LCO 3.7.1 >
l 13. Containment Isolation Trip Functioning Each refueling shutdown l N See LCO 3.3.2 >
l14. Retueling System Interlocks functioning Each refueling shutdown l N See LCO 3.9.3 >
j 15. Scrylce Water System Functioning Each refueling shutdown l N See LCO 3.7.8 >
l16. Primary System Leakage Evaluate Monthly "' l N See LCO 3.4.13 >
l17. Diesel Fuel Supply fuelinventory Daily L N See LCO 3. 8.3 >
~~18. Deleted~~
~~19. Deleted l 20. Boric Acid System Storage Tank and Daily"Ul piping temperatures 2 temperature required : < See LCO 3.5.2 >~~
by Table 153.21 Unit 1 - Amendment No.176 Unit 2 - AmendmentNo.180 Page 2 of 5 August 6,1997
. E O R W F 0 R M Mg,N,[,J, i 0 N ,.,<c,,,,,,,,,
C,T,'3 , j (30. Pressurizer Heaters Verify that 100 KW of Quarterly l
. lheaters are available. I See LCO 3.4.9 >
131. CVC5 Charging Pumps Verify rability Quarterly l See LCO 3.5.2 >
l 32. Potential Dilution in Verify operability of Prior to placing plantini lProaress Alarm alarm.1 Icold shutdown. I 133. Core Power Distribution Perform power distribu- Monthly I"\ See LCO 3. 3. 9 >
" ""E8 8"
~~incore detector system < see LCos 3.2.1 and 3.2.2 >~~
to confirm hot channel factors.I 13 4. Shutdown Martin Perform shutdown margin Daily " l {
lcalculation l N ,,,ggo 3,1,y > )
t
[1) Required only during periods of power operation. I M 8ee LCO 3.4.16 >
[2) Q determination will be started when the gross activity analysis of a filtered sample indicates ;t10pCi/cc l and will be redetermined if the primary coolant gross radioactivity of a filtered sample increases by l more than 10uci/cc. I (3) Drop test shall be conducted at rated reactor coolant flow. Rods shall be dropped under both cold and hot l condition, but cold drop tests need not be timed. l: see LCo 3.1.5 >
(4) Drop tests will be conducted in the hot condition for rods on which maintenance was performed.l (5) As accessible without disassembly of rotor. l: < 8ee LCO 3.3.2 >
(6) Not required during periods of refueling shutdown. : < See LCos: 3.3.2, 3.4.16, (7) At least once per week during periods of refueling siutdown. l 3.5.4, 3.7.18, and 3.4.13 >
(8) At least three times per week (with maximum time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months between samples) during periods of l Irefbeling shutdown. l: < see Lco 3.4.1s >
l(9) Not required during periods of cold or refueling shutdown, but must be performed prior to exceeding 200'T if it has l Inot been performed during the previous surve llance per od.l: < 8ee LCOs 3.3.1, 3.6.3 >
l(10) Sample to be taken after a minimum of 2 EFP3 and 20 c sys power operation since the reactor was last suberitical I ,
Ifor 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months orlonger.k < see LCo 3.4.16 > l l(11) An approximately equal number of valves shall be tested each refueling outage such that all valves will be tested within a five year period. If any valve falls its tests, an additional number of valves equal to the number originally tested shall be tested. If any of the additional tested valves fall, all remaining valves shall be tested.1 l(12) The specified buses shall be determined energized in the requiied manner at least once per shift bY Verifying correct l Istatic transfer switch alignment and indicated voltage on the buses.le--< See LCOs 3. 8. 9 and 3. 8.10 >
l(13) Not required if the block valve is shut to isolate a PORV that is inoperable for reasons other than excessive seat l Ieakage.l l
(14) Only applicable when the overpressure mitigation system is in service.l See LCO 3.4.11 >
(15) Required to be performed only if conditions will be established, as defined in Specification 15.3.15, where the PORVs are used for low temperature overprenure protection. The test must be performed prior to establishing these conditions J
< see LCO 3.4.12 >
M Unit 1 - Amendment No.171 Unit 2 - Amendment No.175 Page 4 of 5 January 16,1997
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FOR lt4FORNiAT10N ONLY No Significant Hazards Considerations -ITS LCO 3.04.10 24-Feb-98 NaHC Number NSHC Text A in accordance with the criteria set forth in 10 CFR 50.92, PBNP has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided in support of this conclusion. ,
~~1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?~~
The proposed change involves reformatting and rewording of the current Technical Specifications. The reformatting and rewording process involves no technical changes to existing requirements. As such, this change is administrative in nature and does not impact initiators of analyzed events or assumed mitigation of accident or transient events. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previous'y evaluated.
~~2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?~~
The proposed change does not necessitate a physical alteration of the plant (no new or different type of equipment will be installed) or changes in parameters goveming normal plant operation. The proposed change will not impose any different requirements. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. .
~~3. Does this change involve a significant reduction in a margin of safety?~~
l The proposed change will not significantly reduce the margin of safety because it has no impact on any safety analysis assumptions. This change is administrative in nature. As such, there is no technical change to the requirements and therefore, there is no significant reduction in the margin of safety.
Page 1 of 6
FOR IMGRMATIDM ORY NSHC Number NSHC Text LO1 In accordance with the criteria set forth in 10 CFR 50.92, PBNP has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided in support of this conclusion.
~~1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?~~
The proposed change does not involve any physical alteration of plant systems, structures or components, changes in parameters goveming normal plant operation, or methods of operation. This change adopts ISTS LCO 3.4.10 Actions A and B. These actions result in extending time allowed to place the plant in a mode in which the requirement does not apply from 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is based on operating experience to reach the required plant condition from a full power condition in an orderly rnanner and without challenging plant systems. This relaxation is acceptable, based on the broader LCO Applicability adopted as part o' - 3 3.4.10, and the increased time required to place the plant in a non-applicable mode fro . * .I power conditions. The Completion Time is consistent with the time allowed by ITS LCO 3.0.3 to bring the plant to a Hot Shutdown condition from full power operation. Therefore, this change does not involve an increase in the probability or consequences of an accident previously evaluated.
~~2. Does the change create the possibility of a new or different Idnd of accident from any accident previously evaluated?~~
The proposed change does not involve any physical alteration of plant systems, structures or components, nor does it alter parameters goveming normal plant operation. The proposed change does not introduce a new mode of operation or alter the method of normal plant operation Therefore, the possibility of a new or different kind of accident from any accident previously evaluated is not created.
~~3. Does this change involve a significant reduction in a margin of safety?~~
There are no margins of safety related to safety analyses that are dependent upon the proposed change. The requirements will continue to assure that limiting conditions for pressurizer safety valves are property maintained. Therefore, this change does not involve a reduction in a margin of safety.
Page 2 of 6
. m . . .mem em~ ~ *~~
FOR 11 FORMAT 101 DNLv NSHC Number NSHC Text l
LO2 in accordance with the criteria set forth in 10 CFR 50.92, PBNP has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided in support of this conclusion.
~~1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?~~
The proposed change does not involve any physical alteration of plant systems, structures or components, changes in parameters goveming normal plant operation, or toethods of operation. This change adopts LCO 3.4.10 NOTE, allowing the pressurizer safety valve lift settings to be outside the LCO limits for the purpose of setting the safety valves under '
ambient (hot) conditions. Only one valve at a time will be removed from service for testing.
This NOTE allows a maximum of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months following entry into MODE 3 for the exception, provided a preliminary cold setting was made prior to heatup. This assures that reactor decay heat is significantly below the assumptions in the appilcable safety analyses for LCO 3.4.10. Therefore, this change does not involve an increase in the probability or consequences of an accident previously evaluated.
~~2. Does the change create the possibility of a new or different idnd of accident from any accident previously evaluated?~~
The proposed change does not involve any physical alteration of plant systems, structures or components, nor does it alter parameters goveming normal plant operation. The proposed change does not introduce a new mode of operation or alter the method of normal plant operation. Therefore, the possibility of a new or different idnd of accident from any accident previously evaluated is not created.
~~3. Does this change involve a significant reduction in a margin of safety?~~
There are no margins of safety related to safety analyses that are dependent upon the proposed change. The requirements will continue to assure that limiting conditions for refueling are property maintained. Therefore, this change does not involve a reduction in a margin of safety.
Page 3 of 6
?0R INF0ftMAT10N ONLY NSHC Number NSHC Text LA in accordance with the criteria set forth in 10 CFR 50.92, PBNP has evaluated this proposed Technical Specifications change and detemiined it does not represent a significant hazards consideration. The following is provided in support of this conclusion.
~~1. Doss the change involve a significant increase in the probability or consequences of an accident previously evaluated?~~
The proposed change relocates requirements from the Technical Specifications to the Bases, FSAR, TRM, or other plant controlled documents. The Bases, FSAR, TRM will be maintained using the pravisions of 10 CFR 50.59. In addition to 10 CFR 50.5gprovisions, the Technical Specifications Bases are subject to the change process in the Administrative Controls l Chapter of the ITS. The FSAR is subject to the change provisions of 10 CFR 50.71(e), and the plant procedures and other plant controlled documents are subject to controls imposed by i plant administrative procedures, which endorse applicable regulations and standards. Since any changes to the Bases, FSAR, TRM or other plant controlled documents will be evaluated in accordance with the requirements of the Bases Control Program in Chapter 5.0 of the ITS, 10 CFR 50.59, or plant administrative processes, no increase in the probability or consequences of an accident previously evaluated will be allowed. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
~~2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?~~
The proposed change does not necessitate a physical alteration of the plant (no new or different type of equipment will be installed) or changes in parameters goveming normal plant operation. The proposed change will not impose any different requirements and adequate control of the information will be maintained. Thus, this change does not create the possibility i of a new or different kind of accident from any accident previously evaluated. !
~~3. Does this change involve a significant reduction in a margin of safety?~~
The proposed change will not reduce a margin of safety because it has no impact on any safety analysis assumptions. In addition, the requirements to be transposed from the Technical Specifications to the Bases, FSAR, TRM or other plant controlled documents are as they currently exist. Since any future changes to the requirements in the Bases, FSAR, ,
TRM, or other plant controlled documents will be evaluated in accordance with the !
requirements of 10 CFR 50.59, the applicable plant process no reduction in a margin* of safety 1 will be allowed. l l
Based in 10 CFR 50.g2, the existing requirement for NRC review and approval of revisions, to these requirements proposed for relocation, does not have a specific margin of safety upon !
which to evaluate. However, since the proposed change achieves consistency with the '
Standard Technical Specifications, Westinghouse Plants, NUREG-1431, approved by the NRC staff, revising the Technical Specifications to reflect the approved level of detail ensures no significant reduction in the margin of safety.
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l Page 4 of 6 l t
I 1
FOR INFORMAT10N ONI.Y NSHC Number NSHC Text l
M in accordance with the criteria set forth in 10 CFR 50.92, PBNP has evaluated this proposed i Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided in support of this conclusion, l
~~1. Does the change involve a significant increase in the probability or consequences of an !~~
accident previously evaluated?
The proposed change provides more stringent requirements for the Point Beach Technical Specifications. These more stringent requirements are not assumed to be initiators of analyzed events and will not alter assumptions relative to mitigation of accident or transient events. These more stringent requirements are imposed to ensure process variables, structures, systems and components are maintained consistent with the safety analyses and licensing basis. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
~~2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?~~
The proposed change does not necessitate a physical alteration of the plar:t (no new or different type of equipment will be installed) or changes in parameters goveming normal plant operation. The proposed change does impose different requirements. However, these changes are consistent with assumptions made in the safety analysis and licensing basis.
Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
~~3. Does this change involve a significant reduction in a margin of safety?~~
The imposition of more stringent requirements will provide for, increased testing, increasing the scope of the specification to include additional plant equipment, increasing the applicability of the specification, providing additional actions, decreasing restoration times, imposing new surveillances, and/or decreasing surveillance intervals. These changes are consistent with the safety analysis and licensing basis. Therefore, this change does not involve a reduction in a margin of safety.
Page 5 of 6
~
FOR INF0iiniATlaN Diny NSHC Number NSHC Text R in accordance with the criteria set forth in 10 CFR 50.92, PBNP has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided in support of this conclusion.
~~1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?~~
The proposed change relocates requirements and surveillancS for structures, systems, components or variables which did not meet the criteria for inclusion in Technical Specifications as identified in the Application of Selection Criteria to the Point Beach Technical Specifications. The affected structures, systems, components or variables are not assumed to be initiators of analyzed events and are not assumed to mitigate accident or transient events. The requirements and surveillances for these affected structures, systems, components or variables will be relocated from the Technical Specifications to en appropriate administratively controlled document and maintained pursuant to 10CFR 50.59. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
1
~~2. Does the change create the possibility of a new or different kind of accident from any I accident previously evaluated?~~
The proposed change does not necessitate a physical alteration of the plant (no new or different type of equipment will be installed) or change in parameters goveming normal plant operation. The proposed change will not impose any different requirements and adequate control of information will be maintained. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
~~3. Does this change involve a significant reduction in a margin of safety?~~
The proposed change will not reduce a margin of safety because it has no impact on any i safety analysis assumptions. In addition, the affected requirement will be relocated to an owner controlled document for which future changes will be evaluated pursuant to the requirements of 10CFR 50.59. Therefore, this change does not involve a significant reduction in a margin of safety.
Page 6of 6
i FOR INFORM ATl01 ONi.Y JustifiCatlOn For DeviatlOns -ITS LCO 3.04.10 05-Nov-98 JFD Number JFD Text 01 The brackets have been removed and the proper plant specific information has been provided.
in some Iristances, even though the information was designated as being site specific information in the LCO (bracketad), the corresponding Bases information was not bracketed.
These cases are self evident, corresponding to the bracketed information in the LCO, and the have had the appropriate site specific information provided.
ITS: NUREG: i B 3.04.10 B 3.04.10 LCO 3.04.10 LCO 3.04.10 LCO 3.04.10 NOTE LCO 3.04.10 NOTE .
1 02 ITS Specification 3.4.10 is modified to reflect a safety valve operability setpoint tolerance of +/- l 3% to allow for drift, in accordance with Section ill of the ASME Boiler and Pressure Vessel '
Code.
ITS: NUREG:
B 3.04.10 B 3.04.10 LCO 3.04.10 LCO 3.04.10 1
03 LCO 3.4.10 is modified to reflect Point Beach LTOP enabling temperature of 355 F. Therefore the Applicability statement regarding MODE 4 is deleted. Additionally, in order to simplify implementation of this specification, the applicability of LCO 3.4.10 is expanded to include all of MODE 3, i.e., greater than or equal to 355 F. This results in an overlap in the requirements for LCO 3.4.10 and LCO 3.4.12 (LTOP). This change is also reflected in LCO 3.4.10, Condition B, Required Action B.2, which will require placing the unit in MODE 4 to exit the applicability of the LCO.
ITS: NUREG: j B 3.04.10 B 3.04.10 LCO 3.04.10 LCO 3.04.10 LCO 3.04.10 COND B RA B.2 LCO 3.04.10 COND B RA B.2 LCO 3.04.10 NOTE LCO 3.04.10 NOTE 04 The actual numerical values for an LTOP enabling temperature are replaced with a reference to the temperature specified in the PTLR. The LTOP enabling temperature will then be calculated and controlled by the licensee in accordance with the topical reports identified in the PTLR.
ITS: NUREG:
B 3.04.10 B 3.04.10 Page 1 of 2
FOR INF0f! MAIL 0N ONLY JFD Number JFD Text 05 Consistent with the range specified in PBNP calculation 98-0096, as tested lift setting of the pressurizer safety valves (+2.67% / -1.78%), SR 3.4.10.1 is modified to specify a pressurizer safety valve lift setting of greater than or equal to 2440.71 psig and less than or equal to 2551.25 psig.
ITS: NUREG:
B 3.04.10 B 3.04.10 SR 3.04.10.01 SR 3.04.10.01 Page 2 of 2
. FOR IUORMAT10N ON!.y ""*"""' " **' N!!
3.4 REACTOR COOLANT SYSTEM (RCS) .
3.4.10 Pressurizer Safety Valves 1
,/YMS i LCO 3.4.10 [Three] pressurizer safety valves shall be OPERABLE with lift settings 2 [2460)lpsig and s [2510]lpsig.
R >"
410 56o APPLICABILITY: MODES 1. 2. and 341 I"0DE ' "ith :l' 9CS 01d "
100 temccr:ture: ' E2751*c] ,
mm a l
................. .... ......N0TE-------.----....--..--...... l The lift settings t'e,not required to be within the LC0 limits during MODEE'3 ene-4lfor the purpose of setting the pressurizer safety valves under ambient (hot) conditions.
This exception is allowed for 54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months following entry into MODE 3 provided a preliminary old setting was made prior to heatup.
36 : E use ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pressurizer safety A.1 Restore valve to 15 minutes valve inoperable. OPERABLE status.
B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. MD QR B.2 Be in MODE 4 It 1th" tr"l 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months RCS :Old le; Twoler ?.crel temper:tur^' ga pressurizer safety c [2751*e as valves inoperable.
WOG STS 3.4-21 Rev 1. 04/07/95
FOR IMFORMAIl0N ONLy ***'5'd8 SURVEILLANCE RE0VIREMENTS l SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify each pressurizer safety valve is In accordance OPERABLE in accordance with the Inservice with the Testing Program. Fol10wina testing, lift Inservice settings shall belwithin i 1%I Testing Program l
2 2440.71 psig and 52551.25 psig a
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.WOG STS 3.4-22 Rev 1, 04/07/95
c .
Pressurizer Safety Valves FOR INFORMATION ONLY S 3 4 2o B 3.4 REACTOR COOLANT SYSTEM (RCS)
B-3.4.10 Pressurizer Safety Valves BASES BACKGROUND The pressurizer safety valves provide, in conjunction with the Reactor Protection System, overpressure protection for the RCS. The pressurizer safety valves are totally enclosed pop type, spring loaded, self actuated valves with backpressure compensation. The safety valves are designed l to prevent the system oressure from exceeding the system i Safety Limit (SL),((2735] psig, which is 110% of the design pressure.
Because the safety valves are totally enclosed and self 288,000 actuating. they are considered indeoendent components. The relief capacity for each valve,T380.000]llb/hr, is based on postulated overpressure transien; conditions resulting from a complete loss of steam flow to the turbine. This event results in the maximum surge rate into the pressurizer. l which specifies the minimum relief capacity'for the safety '
valves. The discharge flow from the pressurizer safety
. valves is directed to the pressurizer relief tank. This discharge flow is indicated by an increase in temperature i downstream of the pressurizer safety valves or increase in the pressurizer relief tan (emperatureorlevel, e l Overpressure protection equired in MODES 1, 2, 3, 4, the LTOP enabling temperature and 5; however in MODE specified in the PH.R temperatures sT275]'FL andwithtigQL5 cneand or more MODE RCS coldthe 6 with lea , MODES I
4andl reactor vessel head on, over ressure protection is provided by operating procedures and y meeting the requirements of LCO 3.4.12,'" Low Temperature Overpressure Protection (LTOP)
System."
The jM.i safety The upper and lower pressure limits are based on the i 1%
valve setpointis
3% for tolerance requirement (Ref.1) for lifting pressures above OPERABILITY;bowever, 1000 psig. The lift setting is for the ambient conditions the valves are reset to associated with MODES 1, 2, and 3. This requires either
+2.67%/ l.78% during that the valves be set hot or that a correlation between hotl surveillance to allow for and cold settinas be established.1 drift and account forthe ambient conditions The pressurizer safety valves are part of the primary associated with MODES 1, success path and mitigate the effects of postulated 2 ands. accidents. . OPERABILITY of the safety valves ensures that the RCS pressure will be limited to 110% of design pressure.
WOG STS B 3.4.10-45 ,
Rev 1, 04/07/95 a . _ _ _ _ . . . _ . _ _ . _ . . _ . _
i Pressurizer Safety Valves B 3 4 to
,3 . FOR INFORMATlON ONLY BASES
' BACKGROUND- The consequences of exceeding the American Society of (continued) Mechanical Engineers (ASME) pressure limit (Ref. 1) could include damage to RCS components. increased leakage, or a requirement to perform additional stress analyses prior to i resumption of reactor operation.
APPLICABLE All accident and safety analyses in the FSAR (Ref. 2) that !
SAFETY ANALYSES require safety valve actuation assume operation oflthreel pressurizer safety-valves to limit increases in RCS N pressure. The overpressure orotection analysis (Ref. 3) is Itwo l l also based on operation ofl[threelfsafety valves. Accidents "
that could result in overpressurization if not properly terminated include: ll
~~a. Uncontrolled rod withdrawal from full power;~~
~~b. Loss of reactor coolant flow:~~
~~c. Loss of external electrical load;~~
~~d. Loss of normal feedwater:~~
~~e. Loss of all AC power to station auxiliaries: and~~
~~f. Locked rotor.~~
Detailed analyses of the above transients are contained in Reference 2. Safety valve actuation is required in events c, d, and e (above) to limit the pressure increase.
Compliance with this LCO is consistent with the design bases and accident analyses assumptions.
Pressurizer safety valves satisfy Criterion 3 of the NRC Policy Statement, tWO %
LCO The'l[three)) pressurizer safety valves are set to open at the RCS design pressure (2500 psia), and within the ASME specified tolerance, to avoid exceeding the maximum design pressure SL. to maintain accident analyses assumptions, and to comply with ASME reauirements. The upper and lowerI pressure tolerance limits are baset on the i 11 toleranceI requirements (Ref.1) for lifting prestures-above 1000 psig l The pressurizer safety valve setpoint is i 3% for OPERABILITY; however, the valves are reset to +2.67% / -1.78% during surveillance to allow for drift.
E E ;
1 WOG STS B 3.4.10-46 Rev 1. 04/07/95
BASES Y 05$ f LC0 The limit protected by this Specification is the reacter
-(continued) coolant pressure boundary (RCPB) SL of 110% of design pressure. Ino)erability of one or more valves could result in exceeding t1e SL if a transient were to occur. The consequences of exceeding the ASME pressure limit could include damage to one or more RCS components. increased leakage. or additional stress analysis being required prior to resumption of reactor operation.
H APPLICABILITY In MODFS 1. 2. and 3. End 00rt10r: c' uggg a 35 ye gge (79p l
!: rec tem:0r:ture. ' OPERABILITY of((three]Lvalves is Itwo @
required because the combined capacity 1s required to keep reactor coolant pressure below 110% of its design value during certain accidents. MODE 3 bnd oortions of MODE 4 are conservatively included, although the listed accidents may not require the safety valves for protection.
The LCO is not applicable in MODE 4 M er 9 RCE cc!d ice
@ $0mper:ture: :re c [275]" lor in MODE 5 because LTOP is provided. Overpressure protection is not required in MODE 6 with reactor vessel head detensioned.
The Note allows entry into MODE 31and-4 lwith the lift settings otitside the LCO limits. This permits testing and examination of the safety valves at high pressure and temperature near their normal operating range, but only after the valves have had a preliminary cold setting. The cold setting gives assurance that the valves are OPERABLE near their design condition. Only one valve at a time vill be removed from service for testing. The [L54_]fhour m eotion is based on 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months outage time for each of the h
$ @wo l--4 :threell valves. The 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months period is derived from operating experience that hot testing can be performei in Es aa this timeframe.
ACTIONS L,1 l With one pressurizer safety valve inoperable, restoration l must take place within 15 minutes. The Completion Time of 15 minutes reflects the importance of maintaining the RCS Overpressure Protection System. An inoperable safety valve WOG STS B 3.4.10-47 Rev 1. 04/07/95
o ,
""'"""" S "
F 0 R I IN y ATION ONLi' S 3 4 to BASES ACTIONS L1 (continued) coincident with an RCS overpressure event could challenge the integrity of tha pressure boundary.
B.1 and B 2 I If the Required Action of A.1 cannot1be met within the . l required Completion Time or if two ler-aer41 pressurizer safety valves are inoperable, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months anc to MODE 4 With := 9CS : ld 10c h 4tc cer:turc: c [2:5]'clwithin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Q I Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full !
power conditions in an orderly manner and without !
challenging plant systems. With any RCS cold leg ee Lu wung temperatures at or belowl275]*FL overpressure protection is "mPm A 8PMM provided by the LTOP System. The change from MODE 1, 2
" ** M or 3 to MODE 4 reduces the RCS energy (core power and l pressure), lowers the potential for large pressurizer l insurges, and thereby removes the need for overpressure protection by [threellpressurizer safety valves.
N '
tWO '
SURVEILLANCE SR_ 3.4 10.1 REQUIREMENTS SRs are specified in the Inservice Testing Program.
Pressurizer safety valves are to be tested in accordance am t with the requirements of Section XI of the ASME Code Ep (Ref. 4), which provides the activities and Frequencies necessary to satisfy the SRs. No additional requirements "
are specified. 3 The pressurizer safety valve setpoint is
~~Tj for OPERABILITY: however, the valves are reset ;o 1%Iduring the Surveillance to allow for drift.~~
~~REFERENCES 1. ASME, Boiler and Pressure Vessel Code,Section III.~~
~~2. FSAR, Chapter 15 14 WOG STS B 3.4.10-48 Rev 1, 04/07/95~~
.. ....c.,% , . ~ - ..._._-%.._,_y %g. . . , . _ , . . . _ - . .
FDP,1)!FORfli AT10Ji Olity """*"""*"S**s!T!!
BASES l
REFERENCES 3. WCAP-7769. Rev. 1. June 1972*
(continued)
~~4. ASME. Boiler and Pressure Vessel Code.Section XI.~~
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l WOG STS B 3.4.10-49 Rev 1. 04/07/95 l
l
' Pressurizer Safety Valves
, 0R li4FGiifd A..Ii0 M 0?4LY 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Pressurizer Safety Valves LCO 2.4.10 Two pressurizar safety valves shall be OPERABLE with lift settings 2 2410 psig and s 2560 psig.
APPLICABILITY: MODES 1. 2, and 3.
............................N0TE------------..............-.
The lift settings are not required to be within the LCO limits during MODE 3 for the purpose of setting the pressurizer safety valves under ambient (hot) conditions.
TNs exception is allowed for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months following entry into MOT.E 3 provided a preliminary cold setting was made prior to ;
he6 tup.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pressurizer safety A.1 Restore valve to 15 minutes valve inoperable. OPERABLE status.
B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND .
DR B.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Two pressurizer safety valves inoperable.
Point Beach 3.4-20 Draft, 11/19/98
FOR N!0RUAT10N dry 'l'd8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify each pressurizer safety valve is In accordance OPERABLE in accordance with the Inservice with the l Testing Program. Following testing, lift Inservice I settings shall be 2 2440.71 psig and Testing Program s: 2551.25 psig '
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d, Point Beach 3.4-21 Draft, 11/19/98 4
w ..-..-,.e,+...wm.,--. - - - + em-.A- -,.w., ...., ., ., = .-
F0R INF0Rh1AT10N 0NL'I "ra$'"rr Saf*tt M B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.10 Pressurizer Safety Valves BASES BACKGROUND- The pressurizer safety valves provide, in conjunction with the Reactor Protection System, overpressure protection for the RCS. The pressur1zer safety valves are totally enclosed pop type, spring loaded, self actuated valves with backpressure compensation. The safety valves are designed to prevent the system pressure from exceeding the system Safety Limit (SL) 2734 psig, which is 110% of the design pressure.
Because the safety valves are totally enclosed and self actuating, they are considered independant components. The relief capacity for each valve. 288.000 lb/hr. is based on postulated overpressure transient conditions resulting from a complete loss of steam flow to the turbine. This event results in the maximum surge rate into the pressurizer, which specifies the minimum relief capacity for the safety valves. The discharge flow from the pressurizer safety valves is directed to the pressurizer relief tenk. This discharge flow is indicated by an increase in temperature downstream of the pressurizer safety valves or increase in the pressurizer relief tank temperature or level.
Overpressure protection is required in MODES 1, 2, 3. 4 and 5: however, in MODE 3 with one or more RCS cold leg temperatures s: the LTOP enabling temperature specified in the PTLR. and MODES 4 and 5 and MODE 6 with the reactor vessel head on, overpressure protection is provided by op'erating procedures and by meeting the requirements of LCO 3.4.12. " Low Temperature Overpressure Protection (LTOP)
System."
The pressurizer safety valve setpoint is
~~3% for OPERABILITY: however, the valves are reset to +2.67%/-1.78%~~
during surveillance to allow for drift and account for the ambient conditions associated with MODES 1. 2 and 3.
The pressurizer safety valves are part of the primary success path and mitigate the effects of postulated accidents. OPERABILITY of the safety valves ensures that the RCS pressure will be limited to 110% of design pressure.
Point Beach B 3.4.10-1 Draft. 11/19/98
(
n ,
1
" Pressurizer Safety Valves
. . FOR INF0UMTION ONLY B 3.4.10 BASES BACKGROUND (continued)
The consequences of exceeding the American Society of Mechanical Engineers (ASME) pressure limit (Ref. 1) could include damage to RCS components. increased leakage, or a requirement to perform additional stress analyses prior to l resumption of reactor operation.
l l
APPLICABLE All accident and safety analyses in the FSAR (Ref. 2) that l SAFETY ANALYSES require safety valve actuation assume operation of two ;
pressurizer safety valves to limit increases in RCS l pressure. The overpressure protection analysis (Ref. 3) is also based on operation of two safety valves. Accidents that could result in overpressurization if not properly ;
terminated include: !
~~a. Uncontrolled rod withdrawal from full power:~~
~~b .. Loss of reactor coolant flow:~~
~~c. Loss of external electrical load:~~
~~d. Loss of normal feedwater:~~
~~e. Loss of all AC power to station auxiliaries; and~~
~~f. Locked rotor.~~
Detailed analyses of the above transients are contained in Reference 2. Safety valve actuation is required in events c. d. and e (above) to limit the pressure increase.
Compliance with this LC0 is consistent with the design bases and accident analyses assumptions.
Pressurizer safety valves satisfy Criterion 3 of the NRC Policy Statement.
LCO The two pressurizer safety valves are set to open at the RCS design pressure (2500 psia), and within the ASME specified tolerance, to avoid exceeding the maximum design pressure SL'. to maintain accident analyses assumptions, and to comply with ASME requirements. The pressurizer safety valve setpoint is i 31 for OPERABILITY: however, the valves are reset to +2.671/-1.78% during surveillance to allow for drift. The limit protected by this Specification is the Point Beach B 3.4.10-2 Draft. 11/19/98 1
""*"" " U U
. .- . FOR INTOiMAIl0K ONLY BASES
. LC0 (continued) l reactor coolant pressure boundary (RCPB) SL of 110% of design pressure. Ino>erability of one or more valves could
, result in exceeding t1e SL if a transient were to occur.
The consequences of exceeding the ASME pressure limit could include damage to one or more RCS components, increased L leakage, or additional stress analysis being required prior I to resumption of reactor operation.
APPLICABILITY In MODES 1. 2, and 3, OPERABILITY of two valves is required because the combined capacity is recuired to keep reactor coolant pressure below 110% of its cesign value during certain accidents. MODE 3 and portions of MODE 4 are conservatively included, although the listed accidents may not require the safety valves for protection.
The LC0 is not applicable in MODE 4 or in MODE 5 because LTOP is provided. Overpressure protection is not required in MODE 6 with reactor vessel head detensioned.
The Note allows entry into MODE 3 with the lift settings outside the LC0 limits. This permits testing and examination of the safety valves at high pressure and ;
temperature near their normal operating range, but only i after the valves have had a preliminary cold setting. The J cold setting gives assurance that the valves are OPERABLE near their design condition. Only one valve at a time will be removed from service for testing. The 36 hour4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months exception is based on 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months outage time for each of the two valves. i The 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months period is derived from operating experience that !
hot testing can be performed in this timeframe. l ACTIONS L1 With one pressurizer safety valve inoperable, restoration must take place within 15 minutes. The Completion Time of l 15 minutes reflects the importance of maintaining the RCS !
Overpressure Protection System. An inoperable safety valve l
coincident with an RCS overpressure event could challenge l
the integrity of the pressure boundary.
l l
, Point Beach B 3.4.10-3 Draft 11/19/98 l
l
.. . ,--,....,..--?- . - , . , - . , ,
.- i Pressurizer Safety Vahfes
.- - ' 42 FOR IUD PdiATION 011.Y ,
BASES .
ACTIONS (continued) l B.1 and B.2 If the Required Action of A.1 cannot be met within the j required Completion Time or if two pressurizer safety valves are inoperable, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the ;
plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without cht.11enging plant systems. With any RCS cold leg temperatures at or below the LTOP enabling temperature specified in the PTLR, overpressure protection is provided by the LTOP System. The change from MODE 1, 2, or 3 to MODE 4 reduces the RCS energy (core power and pressure), lowers the potential for large pressurizer insurges, and thereby removes the need for overpressure protection by two pressurizer safety valves.
I SURVEILLANCE SR 3.4.10.1 REQUIREMENTS SRs are specified in the Inservice Testing Program. l Pressurizer safety valves are to be tested in accordance with the requirements of Section XI of the ASME Code (Ref. 4), which provides the activities and Frequencies necessary to satisfy the SRs, No additional requirements are specified.
The pressurizer safety valve setpoint is
~~31 for OPERABILITY: howe /er, the valves are reset to'+2.67%/-1.78%~~
during the Surveillance to allow for drift.
REFERENCES 1. ASME, Boiler and Pressure Vessel Code Section III.
~~2. FSAR, Chapter 14.~~
~~3. WCAP-7769. Rev. 1. June 1972.~~
4 .' ASME, Boiler and Pressure Vessel Code,Section XI.
Point Beach- B 3.4.10-4 Draft, 11/19/98
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