ML20126H200

Application for Amends to Licenses DPR-42 & DPR-60, Respectively,Revising Action 21 in Table TS 3.5-2B to Incorporate Statement Allowing 6 H to Trip an Inoperable Channel

ML20126H200
Person / Time
Site:	Prairie Island
Issue date:	12/29/1992
From:	Parker T TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20126H167	List: ML20126H162 ML20126H200 ML20126H211
References
NUDOCS 9301050098
Download: ML20126H200 (18)

Text

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UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY' i

PRAIRIE ISLAND NUCLEAR GENERATING PLANT -DOCKET NO. 50 282 50-306 REVISED REQUEST FOR AMENDMENT TO OPERATING LICENSES DPR-42 & DPR-60 REVISION TO LICENSE AMENDMENT REQUEST DATED September 21,.1992-INSTRUMENTATION SPECIFICATION CHANCES Northern States Power Company, a Minnesota corporation, requests authorization; for changes to Appendix A.of the Prairie Island Operating License'as'shown on--

the attachments labeled Attachments 1, 2 and 3. Attachment 1 describes-the proposed changes, reasons-for the changes, and a significant hazards eval-untion. Attachments 2 and 3 are copies of the Prairie Island Technical Specifications incorporating the proposed changes.

This letter contains no restricted or other defense information',

NORTHERN ST S ' ' .R' CO PANY By

/ yhomas W @arker Director--

Fuclear Licensing On thia 29 day of b mAe- / p' fore me a notary public in and for said i County, p'ersonally appeared Thomds' M Parker, Director, Nuclear Licensing, and .

being first duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the contents:thereof, and that to the best of his knowledge, information, and be-lief the statements made in it are true and that it is not interposed for del .

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Attachment 1-

i Revision to Exhibit A' Evaluation of Proposed Changes to the Technical Specifications" License Amendment Request Dated September 21, 1992 4

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Exhibit A Prairic Island Nuclear Generating _ Plant License Amendment kequest Dated September 21, 1992 Revised December. 29 -1992 Evaluation of Proposed changes to the Technical Specifications Appendix A of Operating License _DPR-42 and DPR-60 Pursuant to 10 CFR Part 50, Sections 50.59 and 50.90, the holders of Operating:

Licenses DPR-42 and DPR 60 hereby propose the following changes to Appendix A, Technical Specifications:

Backcround

a. llis tory:

In response to growing concerns of the impact of current testing and-maintanance requirements on plant operation, particularly as_ related to _

instrumentation' systems, the Westinghouse Owners Group (WOG) initiated a -

program to _ develop a justification to be used 'to revise generic and plant specific instrumentation technical specifications. Operating plants experienced many' inadvertent reactor trips and safeguards actuations during performance of instrumentation surveillance, causing unnecessary-transients and challenges _to safety systems. Significant-time and effort on the part.of the operating staff was devoted to performing, reviewing, documenting and tracking the various surveillance activities, which'in many instances seemed unwarranted based on the high reliability of the -

equipment. S16 nificant benefits for operating plants appeared to be achievable through revision of instrumentation test and maintenance-requirements.

In their letter dated February 21, 1985 (Reference 1)..the NRC issued the Safety- Evaluation Report (SER) for WCAP-10271 and Supplement fl. The SERL approved quarterly testing, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.co' place a failed channel in a tripped?

mode, increased Alioved Outage Time (A0T)}for._ test _and testing _In oypass for ' analog channels of the Reactor Protection System (RPS). :The quarterlyJ testing had to be conducted on a staggered basis.

In their letter dated February. 22,- 1989 (Reference 2), the NRC issued thec

-SER for WCAP-10271 Supplement 2 and Supplement 2, Revision 1. /The SER approved quarterly- testing, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to place a failed channel in a tripped g mode, increased Allowed Outage Time for' test and testing in bypass for.

I analog channels of the Engineered Safety Features (ESP) . The. Engineered Safety Features functions approved in the SER were those. presented. in;

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Appendix Al of'the; reference WCAPs. These functions are all: included- in the Westinghouse Standard Technical Specifications. Staggered testing was

- not required for Engineered Safety Features analog ; channels and the l requirement was removed from the Reactor-Protection System' analog l -channels.

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i in their letter dated April 30, 1990 (Reference $), the NRC issued the  !

Supplernental SER (SSER) for WCAP 10271 Supplement 2. and Supplement 2, q Revision 1. The Supplemental SER approved Surveillance Test Interval -!

(STI) and Allowed Outage Time extensions for the Engineered Safety l Features functions that were included in Appendix A2 of WCAP 10271, ,

Supplement 2 Revision 1. The functions approved are associated with the Safety Injection, Steam Line isolation, Main Feedwater Isolation, and i Auxiliary Feedwater Pump Start. signals. The configurations contained in the Appondix A2 are those that are not contained in the Westinghouse ,

Standard Technical Specifications.

With the issuance of the SER and the Supplemental SER, the relaxations for'  !

the analog channels of the Reactor Protection Systern and Engineered Safety i Features are now the unme and the special conditions applied to shared  !

analog channels are no longer applicable. 3 i

To facilitate the incorporation of the revised Surveillance Test Intervals and Allowed Outage Times with appropriate ACTION requirements into the Prairie Island Technical Specifications, the Tables applicable to the instrumentation Technical Specifications have been reformatted to be ,

consistent with the forrant of the Westinghouse Standard Technical Specifications and OPERATIONAL MODES have been defined. Where a .

requirement in the Prairie Island Instrumentation Tables is not-included in Standard Technical Specifications Instrurnentation specifications, and -

is adequately covered elsewhere in the Technical Specifications, the.

requirement is requested to be deleted from the Prairie Island Instrumentation Technical Specifications.- Where a surveillance requirement-is not adequately covered elsewhere, even when it is not a requirement in Standard Technical Specifications, it has been retained in Technical Specification Table TS.4.1 1C.

2. liardware Modification.

No plant modifications are required to implement the items requested in i this License Amendment Request. Increased alland outage time and allowed testing in bypass mode will be accomplished with the prw ert plant 'i configuration. At present Prairie Island Nuclear Ceneratirg Plant does not have bypass testing capability for any of the anaP ' ns trumentat. ion 'l associated with the Reactor Protection Systen. or Engi- d Safety Features.

If in the future Prairie Island Nuclear Generating Plant dosa elect to ,

test in bypass,; plant modifications will be required. Any future bypass- "

-testing modification would be accomplished without reliance upon lifted leads or temporary jumpers and would provide bypass status indications to the plant operators in the control room.

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Proposed ChanreH This License Arnenditent Request proposes to revise the prairlo Island Technical i Specifications and associated bases as described below. The specific wording  ;

changes to the Technical Specifications are shown in Exhiaits B and C.

1, Definitions 1.0

n. Added new definit tons to support new Inst.rumentation Specifications.

Definitions are consistent with current industry Standard Technical-  !

Specification Revision 4a.

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o ACTION o OPERATIONAL HODE MODE o STACGERED TEST SASIS o New Table TS.1-1 with OPERATIONAL MODES

b. Repinced definitions for COLD SilUTDOWN, il0T SIIUTDOWN, POWER OPERATION-and REPUELING with new Table TS.1 1 " OPERATIONAL MODES". The proposed HODE table (Table TS.1-1) is consistent with the MODE table in the revised Westinghouse Standard Technical Specifications with the following exceptions:

1) The titles for MODES 2, 3 and 4 are not consistent with tho-Revised Standard Technien1 Specifications. An asterisked statement noting this inconsistency is included in the proposed Table TS.1 1,

2) The RATED TilERMAL p0VER conditions for MODES 1 and 2 are' based on 2n rather than St.

3) The status of the reactor vessel head closure bolts is specified in a separate column rather than as an asterisked statement,

4) The reactivity conditions are specified using the terms " Critical" and "Subcritical" instead of the kerr values _used in the Revised Standard Technical Specifications,

c. Associated with the elimination of the definition of Il0T SilUTDOWN, the current SilUTDOWN MARGIN requirements of Technical Specification 3.10 A "

are being replaced with a new definition for SilUTDOWN MARCIN and new SilVTDOWN MARGIN requirements. The proposed SilVTDOWN NARGIN definition in Section 1,0 and the proposed SilUTDOWN MARGIN requirements in Section -

3.10.A are consistent with the Revised Westinghouse Standard Technical Specif1 cations except that portions.of the Revised Standard Technical.

Specification SilUTDOWN MARGIN definition have been ' incorporated into .

the bases for Section 3.10. A rather. than Section 1.0. In addition, the bases for the shutdown margin requirements of Technical Specification Section 3.10 are being revised to support'the revisions to Section

'3.10.A. A note-is also being incorporated into the Bases for Section 3.10 to clarify that shutdown margin is a function of hot full power boron concentration in Figure TS.3.10 1.

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2. Technical Specification 2.3.A.2.g

a. Technical Specification 2.3.A.2.g and the associated bases are being revised to-clarify that the Teennical Specification required reactor coolant pump bus undervoltage reactor trip is the direct undervoltage reactor trip, not the reactor coolant pump circuit breaker'undervoltage trip which indirectly results in a reactor trip.

3. Technical Specification 3.5 and Table TS.3.5 2 through TS.3.5-6

a. Technical Specification Section 3.5 is revised to refer to new Tables TS.3.5 2A and TS.3.5 2B. parts C and D of specification 3.5 have been replaced by incorporating ACTIONS or notes into the new Tables as '

appropriate.

b. Table TS.3.5-2A replaces old Table TS.3,5-2. The new Table is consistent with the format and content of Standard Technical Specifications Revision 4a and also incorporates the Allowed Outage Times approved in References 1 and 5.

c. Table TS.3.5 2B replaces old Tables TS.3.5 3, TS.3.5 4 (except . ,

Functional Unit 4), and TS.3.5 6. The new Tables are consistent with the format and content of Standard Technical Specifications Revision 4a and also incorporate the Allowed Outage Times' approved in References 2 and 5. j

d. Functional Unit 10 of Table TS.3.5-2 specifies the. requirements for the single loop and two loop : loss of flow reactor trips the two trips are listed separately. In Functional Unit 12 of Tables 3.5-2A and TS.4.1 =

1A the loss of reactor coolant flow reactor trip is listed as a single item, with no reference to single loop or two loop trips.

c. Functional Unit 15 of Table TS.3.5 2 is deleted since: the control rod misalignment monitor is not associated with the reactor protection system and because Technical Specification Section 3.10 I specifically addresses the actions to be.taken if rod position deviation or quadrant  ;

power tilt monitors are inoperabic.

f. Functional Unit 4 of Table TS.3.5-4 is deleted since this requirement is adequately addressed by revising specification 3.4,C to specifically state that the actuation logic includes the temperature sensors,

g. Technical Specification 3.4.C is revised to include the actuation instrumentation that was previously addressed in Table TS.3.5 4 Functional Unit 4 This-increases tho' time that the temperature sensors may be inoperable to be consistent.with the time that the >

actuated components are allowed to be inoperable,

h. Table TS.3.5-5 is deleted since this requirement is adequately addressed in specifications 3.6.F and.3,6.H.

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4. Technical Specifications 4.1.A, 4.1.D and Table TS.4.1-1

a. Technical Specification 4.1.A is revised to refer to new Tables TS.4.1-1A through TS.4.1-lC.

b. Technical Specification 4.1.D is revised to delete the sentence about APPLICABILITY at all times. APPLICABILITY has been incorporated into the individual new Tables.

c. Tables TS.4.1 1A and TS.4.1-1B replace old Table TS.4.1 1 for Reactor Trip and Engineered Safety Features Surveillance Requirements. Those functions not related to Reactor Trip or Engineered Safety Features have been incorpt. rated into a new Table TS.4.1 1C for miscellaneous instrumentation surveillance requirements. The new Tables are consistent with the format and content of Standard Technical specifications Kevision 4a. In addition, Tables TS.4.1-1A and TS.4.1 1B incorporate the Surveillance frequencies approved in References 1, 2 and S.

d. Specific surveillance requirements for the auxiliary feedwater system actuation instrumentation which are consistent with current requirements in the Prairie Island Technical Specifications or the Standard Technical Specifications have been incorporated into (

Functional Unit 7 in Table TS.4.1 1B.

e. Functional Unit 43 of Table TS.4.1-1 is deleted since the Control Room Ventilation System Chlorine lionitors are no longer required by the Prairie Island Technical Specifications.

5. Change to the bases to insert the necessary wording for referencing the WCAP-10271 and supplements.

6. Editorial Technical Specification Changes -

a. Technical Specification 2.3.B.1 through 2.3.B.5 have added headings of the interlock names for clarity,

b. Deleted footnote "See Specification 4.1.D" at the bottom of Table TS.4.1-2B page 2 of 2 and added a footnote " Required at all times" at the bottom of page 1 of 2 in this table,

c. In addition to the changes to page TS.3.10-1 described above, the term

" power operation" in the objective section of Technical Specification Section 3.10 is being fully capitalized because it is a defined term.

slus t i fi c a t i on increasing the Surveillance Test Interval for the Reactor Protection System and Engineered Safety Features instrumentation minimizes the potential number of inadvertent Engineered Safety Features actuations and reactor trips during surveillance testing. Less frequent surveillance testing has been estimated to result in 0.5 fewer inadvertent reactor trips, per unit, per year. Also, increasing the surveillance interval enhances the operational effectiveness of plant personnel. The amount of time plant personnel spend performing

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i bhtbit A he. s et ts t surveillance testing will be reduced. This allows manpower to be used for.

other tasks such as preventative maintenance. The increased Allowed Outage Time has been shown to result in fewer human factor errors, since more time is allowed to perform an action. l i

WCAP 10271 results show that the reduction in testing and the increase in i testing and maintenance Allowed Outage Times do not adversely affect public health and safety. The proposed revision will reduce the number of  !

inadvertent Engineered Safety features actuations and reactor trips and allow i Prairie Island to better manage resources to maintain the plant. ,

Reformatting the Tables in the instrumentation Technical Specifications to a -

format and content consistent with Revision 4a of the Westinghouse Standard Technical SpecificatiLns ensures implementation of the approved Allowed Outage Timou and Surveillance Test Intervals in a manner consistent with the SERs and the Supplement,al SER of References 1, 2 and 5. By defining ACTION, OPERATIONAL MODE MODE and STAGGERED TEST BASIS, and creating Table TS.1 1 to  ;

define OPERATIONAb MODES, a consistent set of Action Statements, with r Applicable MODE requirements can be established.

Definitions for C0bD SilUTDOWN, POWER OPERATION, REFUELING and 110T SilUTDOWN 4 would be deleted and replaced by the OPERATIONAL MODEJ defined in Table TS.1- i

1. The proposed MODE table (Table TS.1 1) is consistent with the MODE table in the revised Westinghouse Standard Technical Specifications with the -;

following exceptions:

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1) The titles for MODES 2, 3 and 4 are not consistent with the Revised Standard Technical Specifications. The proposed titles for MODES 2, 3 and 4 are "Il0T STANDBY", "110T SilVTDOWN" and "lNTERMEDIATE S!!UTDOWN" respectively. The Standard Technical Specification titles are not being  :

used because the term "Il0T hilUTDOWN" is used throughout the Prairie Island '

Technical Specifications. Use of the Standard Technical Specification title for MODE 3 (ll0T; STANDBY) would involve a significant revision to the Prairie Island Technical Specifications, which is beyond the scope of this amendment request. The title "110T STANDBY" is used for MODE 2 becauso.

plant procedures define the MODE 2 conditions as 110T STANDBY and because the Standard Technical Specification title for MODE 2 is "STARTUP" which could be confused with.the term "STARTUP OPERATIONS" currently defined in the Prairie-Island Technical Specifications.

An asterisked statement, noting this inconsistency, is included in the proposed Tablo TS.1 1.

- 2) The RATED TllERMAL POWER-conditions for MODES 1 and 2 are based on 2%

rather than the St specified in the Revised Standard Technical Specifications. The 2% RATED TilERMAL POWER conditions for MODES 1 and 2 is. consistent with current Prairie. Island Technical Specification'and procedural definitions. Changing the. MODES 1 and 2 RATED TilERMAL POWER condition to 5% would provide no improvement in plant safety and could increase the possibility of human error, 1

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3) The status of the reactor vessel head closure bolts is specified in a separate column t o climinate additional asterisked notes f rom the tablo and to more cicarly define reactor head bolt requirements and the dif ference between the C01.D SilUTDOWN and REFUEi. LNG MODES.

4) The ll0T SilVTDOWN reactivity requirements, in the current definition, are based on the shutdown margin requisements in Table TS.3.10 1. Because of this past history of MODES being associated with .hutdown margin, the reactivity conditions in the proposed MODE table are specified using the terma "Gritical* and "Subcritical" instead of the k gr values used in the Rnvised Standard Technical Specificationa, The terms "Gritical" and "Suberitical" meet the intent of the k.tr conditions speoffied in the Revised Standard Technical Specifications.  !

POWER OPERATION AND Go!.D SilUTDOWN, as defined in new Table TS.1-1, corrospond to the definitions being deleted and therefore do not represent changes, ll0T STANDBY AND INTEIUiEDIATE SilVTDOWN and the conditions they represent are not currently delined in the Prairlo Island Technical Speciflentions.

lloweve r , the proposed MODES they represent (MODES 2 and 5) are consistent (cy. cept; as discussed above) with the Standard Technical Specification MODES.

The definition of Il0T SilUTDOWN in the now OPERATIONAL MODES table differs from the current definition in two respects, the reactivity conditions and the average reactor coolant temperature.

In the current definitions of Il0T SliUTDOWN, reactivity requirements are based on the SilVTDOWN MARGIN requirements in Table TS.3.101. The definition of Il0T SilVTDOWN por new Table TS.1 1 specifies reactivity conditions equivalent to those in the Westinghouse Standard Technical Specifications 4 Associated with '

these proposed changes, the current SilUTDOWN MARGIN requirements of Technical Spectilcation'3.10.A are being replaced with a new definition for SliUTDOWN MARGIN and new SlIUTDOWN MARGIN requirements, This change to' Specification 3,10,A is being made to ensure there is no confusion between the ll0T SilUTDOWN reactivity conditions and the requirements of Table TS,3,101 and to ensure that shutdown margin requirements exist for all plant modes.

The bases for the shutdown margin requirements of Technical Specification Section 3.10 are revised to support the revisions to Section 3,10.A and to f urther clarify the shutdown margin requirement:s. A noto is also being incorporated into the Bancs for Section 3.10 to clarify that shutdown margin la a function of hot full power baron concentration in Figure TS.3.10 1.

The new SilUTDOWN MARGIN requirement.s in Sect. ion 3.10. A' and t ho- new SilUTDOWN MARGIN definition in Section 1.0 aro consistent with the revised Westinghouse Standard Technical Specifications except that portions of the revised Standard Technical Specification SliUTDOWN MARGIN definition have been incorporated into-the bases for Section 3.10.A rather than Section 1.0 The portions of the Revised Standard Technical Specification SIIUTDOWN MARGIN definition'placed in the Section 3.10 Bases involve guidance on how to calculate Sil0TDOWN MARGIN when'in MODES 1 or 2 or with rods not fully inserted and as such do not specifically define SilUTDOWN MARCIN.

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l These changes t o the def inition of 110T SituTDOWN, to Specification 3.10.A and the bases to Section 3.10 will expand shutdown margin requirements to the INTEPJiEDIATE and COLD SilVTDOWN conditions, but will have no affect on the actual shutdown margin limits in Figure TS.3.10.1.

The adoption of Standard Technical Specif ication OPEPATIO!'AL MODES will result.

In the temperature condit ions f or ll0T SilUTDOWN (MODE 3) changing f rom a specific temperature (547'F) to a range of temperatures (2350*F). The addition of SilVTDOWN MAEGIN requirements for temperatutes below 547'r, as discussed above, will ensure that adequat e SilVTDOWN MARCIN will be innintained for the full HOT SilVTDOWN (MODE 3) temperature range.

The definition of REFUELING currently includes a 140'F temperature condition.

This temperature is not associated with any plant safety analysis and uns eliminated from the OPERATIONAL MODE Table in the Revised Westinghouse Standard Technical Spectitcations. Therefore, the 140'F temperaturo limitation for the REFUELING MODE was not included in the proposed OPERATIONAL MODE Table The propose REFUELING MODE definition is consistent with the Revised Westinghouse Standard Technical Specifications.

Technical Specification 2.3.A.2.g currently lists the reactor coolant pump circuit breaker undervoltage trip setpoint as a protective i ns t rtunentat ion setting for reactor trip. While the trip of a reactor coolant ptunp breaker as the result of this undervoltage instrumentation would indirectly result in a reactor trip because of the opening of the breaker, this is not the undervoltage trip utilized in the analysis of the loss of flow accident in the Updated Safety Analysis Report. Per Section 14.4.8.1 of the Updated Safety Analysis Report, the direct. reactor coolant pump bus undervoltage reactor trip is used in the analysis of the reactor coolant system flow coastdown event.

Because the direct reactor coolant pump bus undervoltage reactor trip is utilized in the plant safety analysis, Technical Specification 2.3.A.2.g is _

being revised to refer to that undervoltage trip rather than the reactor coolant pun.p breaker undervoltage trip currently referenced, The loss of reactor coolant flow trip is being listed as a single item, with no reference to the number of loops, in Functional Unit 12 of Tables 3.5 2A and TS.4.1 1 A, because the P-7 and P-8 interlocks , which enable the single loop and two loop loss of flow trips, have the same setpoint (>10% power).

Because the single loop and two loop loss of flow trips are enabled at the same power level, and because these two trips utiltre the same flow instrumentation, there is no need to list both trips separately. A reactor coolant system flow channel failure would affect both the singic loop and the two loop loss of flow trips, and the actions taken in response to the failure would be the same. Listing only a single loss of reactor coolant flow functional unit in Table TS.3.5 2A will make the response to a loss of a reactor coolant flow channel less confusing and will simplify the Technical Specifications.

Functional Unit 15 of Table TS.3.5-2 is being deleted since the control rod misalignment monitor is not associated with the reactor protection system and because Technical Specification Section 3.10.1 specifically addresses the actions to be taken if rod position deviation or quadrant power tilt monitors are inoperable.

tshn44 A r... e or is Funct ional Unit 4 of Table TS.3.5 4 is not proposed to be incorporated into Table TS.3.5 2B as it is more appropriate to add it to Lechnical Specification 3.4.C. This increases t.he time the temperature sensors may be inoperable and makes the action to be taken consistent with actions for the actuation logic and actuated components. In the current Technical Specification, if the temperature sensor is inoperabic, the plant must be taken to 110T SHUTDOWN and -

then to COLD SilUTDOWN if the minimuni conditions are not met in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The actions to be taken if either the actuation logic or actuated dampers are inoperabic is only to close the associated dampers. Provided the datepers are closed, plant shutdown or MODE changes are not required. The proposed change is justified because there is only one temperature sensor per channel and because the change makes the action for the sensors consistent with the actions for the actuation logic that takes its input f rom the sensors, and the -

action-for the actuated componento These channels are not in the inutrumentation Standard Technical Specifications.

The Instrument Operating Conditions contained in the current Table TS.3.5-5 are deleted as these are-adequately covered in the normal operability determination for the vent 11ation systeres in Technical Specifications 3.6.F and 3. 6.11. This does not represent a change in requirements. These channels are not in the instrumentation Standard Technical Specifications.

Adding headings to Technical Specification 2.3.B.1 through 2.3.B.S.is proposed for clarity and does not represent a change in requirements.

Deleting the footnote "See Specification 4.1.0" at the bottom of Table TS.4.1-2B page 2 of 2 and adding a footnote " Required at all times" at the bottom of page 1 of 2 in the same table does not represent a change in requirements.

The original footnote on page 2 of 2 only applied to a function on page 1 of

2. Specification 4.1.D required the asterisked items in this tabic t o be operable at all times. Changing the footnote as proposed therefore is only makes the existing requirement more readily apparent to the operators.

The surveillance requirements -for the auxiliary feedwater system actuation instrumentation were incorporated into Table TS.4.1-1B for consistency with the Standard Technical Specifications. Those surveillance requirements for the auxiliary feedwater actuation instrumentation which are not consistent with the Standard Technical Specification requirements are consistent with current requirements in Prairie Island Technical Specification Section 4.8.

Functional Unit 43 of Table TS.4.1 1 is deleted since the Control Room Ventilation System Chlorine Monitors are no longer required by the Prairie Island Technical Specifications. The Control Room Ventilation System Chlorine Monitors were climinated from the Prairie Island Technical Specifications by Licenno Amendments DPR 42/102 and DPR 60/95 dated September 29, 1992. The chlorino monitor surveillance requirements were mistakenly nor. included in the 4 License Amendment-request which requested the-climination of the chlorine monitors.

The changes to the bases are consistent with the NRC requirements included in the References 1, 2 and 5 and only add the applicable references for the revised Allowed Outage Times and Surveillance Test Intervals.

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f Exhibit A f ras e 10 cf 16 5 EA.fety Evaluatlon In WCAp.10271 and its supplements, the Westinghouse Owners Group evaluated the

, impact of the proposed Surveillance Test Interval and Allowed Outage Time changes on core damage frequency and public risk. The NRC staff concluded in  !

its evaluation (Reference 2) of the Westinghouse owners Group evaluation that ,

an overall upper bound of the core damage frequency increase due to the proposed Surveillance Test Interval / Allowed Outage Time changes is less than 6 percent for Westinghouse Pressurir.ed Water Reactors (PVR) plants. The NRC Staff also concluded that actual core dama6e frequency increases for individual plants are expect ed to be substantially less than 6 percent. The NRC Staff considered this core damage frequency increase to be small compared to the range of uncertainty in the core damage frequency analyses and therefore acceptabic, 1

Additionally the NRC Staff concluded that a staggered test strategy need not-he implemented for Engineered Safety Features analog channel testing and is no

, longer required for Reactor protection System analog channel testing. This conclusion was based on the small relative contribution of the analog channels -

to Reactor Protection System / Engineered Safety features unavailability, process parameter signal diversity and normal operational testing sequencing.

The proposed changes in Surveillance Test Intervals and Allowed Outage Times are consistent with NRC Safety Evaluation Reports dated February 21, 1985 (Reference 1), February 22, 1989 (Reference 2), and April 30, 1990 (Reference *

5) regarding WCAp.10271, WCAp.10271 Supplement 1, WCAp.10271 Supplement 2, and WCAp.10271 Supplement 2 Revision 1 (References 1, 2 and 5). The changes to make the Tables, the MODES and the ACTIONS in the Standard Technical Specification format are consistent with the assumptions used in the analyses of WCAp 10271 and the supplements. The SERs and the Supplemental SER therefore apply to the format and content changes as proposed. Where a Functional Unit in the current prairie Island Technical Specifications is not included in the Standard Technical Specifications, it is retained in the new Table 4.1 1c. In the few cases where an existing Functional Unit has not been included in one of the new Tables,-the requirement is retained elsewhere in the Technical Specifications and the safety function is maintained. The only-exceptions are the surveillance requirements for the control Room Ventilation System Chlorine Monitors which are no longer required by the Prairle Island Technical Specifications and have been removed from service.

The NRC Staff has stated tnat approval of the changes approved _in their SERs is contingent upon confirmation that certain conditions are met, Although the

_ Safety Evaluation Reports of References 2 and 5 apply to Engineered Safety Features instrumentation, conditions given in the Reference 1 SER for the Reactor Protection System instrwnentation also apply to Engineered Safety

- Featurca where appropriate. The prairic Island response to - these conditions

-is provided below.

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i hhibit A Pagellof16

1. Reactor Protection System SER Conditions:

a. SER Condition NRC Staff stated in the Reactor Protection System SER (Reference 1, page 10) that approval of an increase in Surveillance i Test Interval for the analog channel operational tests from once per month to once per quarter is contingent on performance of the testing on a staggered test basis. In the Engineered Safety Features SLR (Reference 2, page 4 of enclosure 1) this requirernent was removed.

Response - This SER Condition is not a concern for Prairie Island as the changes proposed in this IAR implement Reactor Protection System and Engineered Safety Features at the same time. As the increase in Surveillance Test Interval for the analog channel operational tests from once per month to once per quarter with the contingency to perform the testing on a staggered test basis was not implemented for Reactor 1 i Protection Systern functions, it is not necessary to remove this j requirement,

b. SER Condition - NRC Staff stated in the Reactor Protection System SER (Referenco 1, page 10) that approval of iterns related to extending Surveillance Test Intervals is contingent on procedures-being in place i to require evaluation of failures for cormoon cause and to require additional testing if neccusary.

Response - Prairic Island has implemented procedures and procedural '

steps to evaluate failures for common cause and require additional testing as necessary in accordance with the Westinghouse owners Group position given in " Westinghouse Owners Group Guidelines for Preparing Submittals Requesting Revision of Reactor Protection System Technical Specification, Revision 1". These guidelines were reviewed and approved by NRC Staff.

c. SER Condition NRC Staff stated in the Reactor Protection System SER (Reference 1, page 10) that for channels which provide _ dual inputs to other safety related systems such as Engineered Safety Features,- the approval of items that extend Surveillance Test Intervals and Allowed Outage Times apply only to the Reactor Protection System function.

Response The Engineered Safety Features SER has been issued (References 2 and 5). The extensions approved for the Engineered Safety Features analog Channels are the- same as the Reactor Protection System and so this SER Condition is not a concern for Prairic Island.

d. SER Condition - NRC Staff stated in the' Reactor Protection System SER (Reference 1, page 10) that approval of channel testing in a bypassed condition is contingent on the capability of the Reactor: Protection System design to allow such testing without lifting Icads or installing temporary jumpers.

Response - At present Prairie Island does not have bypass testing

capability for any of the analog instrumentation associated with the Reactor Protection Syatem or Engineered ~ Safety Features with the exception of the source range and intermediate range. reactor trips.

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txhibit A res. n ette If in the future Prairie Island does elect to test other channels in bypass, plant inodifications will be required. Any future bypass testing modification would be accomplished without reliance upon lifted leads or temporary jumpers and will provide bypass status indications to the plant operators in the control room,

e. SER Condition NRC Staff stated in the Reactor Protection Systern SER (Reference 1, page 9) that acceptance was contingent on confirination that the instrurnent setpoint methodology includes sufficient margin to offset the drif t anticipated as a result of less frequent surveillance.

Response - Prairie Island impicmented a prograta to evaluate setpoint drif t in accordance with the Westinghouse Owners Group position given in the " Westinghouse Owners Group Guidelines for Preparing Submittals Requesting Revision of Reactor Protection System Tec'.nical Specification, Revision 1". These guidelines were reviewed and approved by NRC Staff.

Prairie Island has deterinined that the values used in the setpoint methodology properly account for drift due to extended Surveillance Test Intervals.

2. Engineered Safety Features SER Conditions;

a. SER Condition - NRC Staff stated in the Engineered Safety Features SER (Reference 2, Table 1 of enclosure 1) that the licensee must confirrn the applicability of the generic analyses to the plant.

Response - The generic analyses used in WCAP 10271 and Supplements is  ;

applicable to. Prairie Island. Prairie Island uses the'Foxboro ll Line Process Control System and the Westinghouse Relay _ Protection System for both the Engiieored Safety Features and Reactor Protection System.

Both of rhose systems were specifically modelled in the-generic  !

analyses. The Engineered Safety Features Functional. Units implemented at Prairie Island are all addressed by the generic analyses.

b. SER Condition - NRC Staff stated in the Engineered Safety Features SER (Reference 2, Table 1 of enclosure 1) that the licensee must confirm that any increase in instrument drift due to the extended Surveillance Test Intervals is properly _ accounted for in the sotpoint calculation methodology.

Response - Same as Reactor Protection System SER Condition e. above.

The changes being made to the definition of Il0T SIIUTDOWN separate shutdown margin from the. plant mode definitions and make the definition of Il0T SilUTDOWN more consistent with the Standard Technical- Specifications.

Because these proposed changes are consistent uith the guidance in the Standard Technical Specifications and because the proposed changes to Section 3.10. A will ensure that shutdown margin will be maintained' during Il0T SilUTDOWN conditions, there will be no reduction in plant safety.

~ ~ ~ -

i Exaba A i r s.nor te .

The change to Specification 3.10.A and associated bases will ensure there is no confusion between the HOT SHUTDOWN reactivity conditions and the requirements of Table TS.3.10 1. The changes to Specification 3.10.A and the associated bases will expand shutdown margin requirements to the INTERMEDIATE and COLD SHUTDOWN conditions consistent with the Westinghouse Standard Technical Specifications, and thus will result in more restrictive Technical Specification requirements. The proposed changes will have no affect on the actual shutdown margin limits in Figure TS.3.10.1. Because the proposed shutdown margin requirements are consistent with the guidance in the Westinghouse Standard Technical-Specifications and are more restrictive than the current Technical ,

Specification requirements, they will not reduce the plants margin of safety.

Technical Specification 2.3.A.2.g currently lists the reactor coolant pump circuit breaker undervoltage trip setpoint as a protective instrumentation setting for reactor trip. While the trip of a reactor coolant pump breaker as the result of this undervoltage instrumentation would indirectly result in a reactor trip because of the opening of the breaker,. .

this is not the undervoltage trip utilized in the analysis of the loss of flow accident in the Updated Safety Analysis Report. The. proposed change to Section 2.3.A.2.g will correct the Technical Specifications to reference the reactor coolant pump bus undervoltage trip utilized in the plant safety analysis and will thus help ensure that protective function-is maintained operable.

In conclusion, Northern States Power believes there is reasonable assurance that the health and safety of the public will not be adversely affected by the proposed Technical Specification changes.

Determination of Strnificant Hazards Considerations The proposed changes to the Operating License have been evaluated to determine whether they constitute a significant hazards consideration as required by 10 CFR Part 50, Section 50.91 using the standards provided in Section 50.92.

This analysis is provided below:

1. The proposed amendment will not involve a significant increase in the nrobability or consequences of an accident previous 1v evaluated.

The determination that the results of the proposed change are within all acceptable criteria have been established in the SERs prepared for WCAP-10271, WCAP-10271 Supplement 1, WCAP-10271 Supplement 2 and >

WCAP-10271 Supplement 2, Revision 1 issued by References 1, 2 and 5.

Implementation of the proposed changes is expected to' result in an acceptable increase-in total Reactor Protection and Engineered Safety

~

Features Systems yearly unavailability. This increase, which is primarily-due. to less frequent surveillance, results in a increase. of similar--

magnitude'in the probability of an Anticipated Transient Without Scram (ATWS) and in the probability of core melt resulting from an ATWS and also results in a small increase.in core demage: frequency-(CD) due to Engineered Safety Features unavailability.

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Exhibit a Page 14 0f 16 Implementation of the proposed changes is expected to result in a significant reduction in the probability of core melt from inadvertent reactor trips. This is a result of a reduction in the number of inadvertent reactor trips (0.5 fewer inadvertent reactor trips por unit per year) occurring during testing of Reactor Protection System instrumentation. This reduction is primarily attribt able to less frequent surveillance.

The reduction in inadvertent core melt frequency is sufficiently large to counter the increase in ATWS core melt probability resulting in an overall P reduction in total core melt probability.

The values determined by the Westinghouse Owners Group and presented in-the VCAP for the increase in core damage frequency were verified by i Brookhaven National Laboratory (BNL) as part of an audit and sensitivity analyses for the NRC Staff. Based on the small value of the increase compared to the range of uncertainty in the core damage frequency,Lthe increase is considered acceptable. ,

The changes of an editorial nature, including the change to Standard -

Technical Specification format for the instrumentation Technical ^

Specifications and mode definitions, have no impact on the severity or consequences of an accident previously evaluated. .

The proposed changes do not result in an increase in the severity or consequences of an accident previously evaluated. Implementation of-the proposed changes affects the probability of failure of the Reactor Protection System'and Engineered Safety Features but does not alter the manner in which protection is afforded nor the manner in which limiting criteria are established.

2. The proposed amendment will not create the possibility of a new or different kind of accident from any accident nreviousiv analyzed.

- The proposed changes do not involve hardware changes and do not result in-

- a change in the manner in which the Reactor Protection System and Engineered Safety Features provide plant protection. No change is being made which alters the functioning of the Reactor-Protection-System or Engineered Safety Features. Rather the likelihood or. probability of the Reactor Protection System or Engineered Safety Features functioning properly is affected as described above. . Therefore the proposed changes do not' create the possibility of a new or different kind of accident from any accident-previously evaluated. _

.s

- TheSchanges of an editorial nature, including the change.to Standard Technical Specification format:for the instrumentation'Te'chnical ,

Specifications-and mode definitions _does not create-the possibility of a new or different kind of accident from any.- previously evaluated,-

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J Exhibit. A res. sht is i

3. The proposed amendment will not involve a significant reduction in the  !

margin of safety. l l

The proposed changea do not alter the manner in which safety limits, limiting safety system setpoints or limiting conditions for operation are determined. The impact of reduced testing other than as addressed above i is to allow a longer time interval over which instrument uncertainties (e.g., drift) may act. Experience has shown that the initial uncertainty assumptions are valid for reduced testing.

Implementation of the proposed changes is expected to result in en overall improvement in safety by:

a. Less frequent testing will result in less inadvertent reactor trips and .I actuation of Engineered Safety Features components, i

b. liigher quality repairs leading to improved equipment reliability due to j longer repair times.  ;

c. Improvements in the effectiveness of the operating staff in monitoring and controlling plant operation. This is due to less frequent distraction of the operator and shift supervisor to attend to instrumentation testing.

l The changes of an editorial nature, including the change to Standard Technical Specification format for the instrumentation Technical Specifications and mode definitions does not lead to a reduction in any margin of safety.

Based on the evaluation described above, and pursuant to 10 CFR Part 50, ,

Section 50,91, Northern States Power company has determined that operation of the Prairie _ Island Nuclear Generating Plant in'accordance with the proposed._

license amendment request does not involve any significant hazards

~

considerations as defined by NRC regulations in 10 CFR Part 50, Section 50.92.

Environmental Assessment Northern States Power has evaluated the proposed changes and determined that: P

- 1. The changes do not involve a significant hazards consideration, f 24 The changes do not involve'a significant change in the_ types or 5 significant increase in the amounts of any effluents that may be released

• offsite, or- ,

3. The changes do not involve a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed changes meet the eligibility criterion forn 4

- categorical exclusion set forth in 10 CFR Part.51~, Section 51.22(c)(9).

Therefore, pursuant to 10 CFR Part 51',-Section 51.22(b), an environmental assessment;of the proposed changes is'not required. .

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Enttelt, A Pate16ef16 liefsrenceri.

1. Letter f rom C. O. Thomas (NRC) to J . J . Sheppard (WOG) dated February 21, 1985 " Safety Evaluation by the Office of Nuclear Reactcr .

Regulation WCAP-10271, Evaluation of Surveillance Frequercles and Out of Service Times for the Reactor Protection Instrumentation S,' stem".

2. Letter from Charles E. Rossi (NRC) to Roger A. Newton (WOG) doted .

February 22, 1989 " Safety Evaluation by the Office of Nucletr Reactor '

Regulation Review of Westinghouse Report WCAP-10271 Supplement 2 and WCAP-10271 Supplement 2, Revision 1 on Evaluation of Surveillance Frequencies and out of Service Times for the Engineered Safety ,

Features". i

3. WCAP-10271 Supplement 1 P A, " Evaluation of Surveillance Frequencies and out of Service Times for the Reactor Protection Instrumentation System", May 1986.  ;

4. WCAP-10271-P-A Supplement 2, Revision 1 " Evaluation of Surveillance Frequencies and Out of Service Times for the Engineered Safety Features", May 1989.

5. Letter Charles E. Rossi (NRC) to Gerard T. Goering (WOG) dated April -

30, 1990 (NRC Supplemental Safety Evaluation for WCAP-10271 Supplement t 2, Revision 1).

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