ML20087L458

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Proposed Tech Specs Removing Several Cycle Specific Parameter Limits & Adding Limits to COLR
ML20087L458
Person / Time
Site: Comanche Peak  Luminant icon.png
Issue date: 08/15/1995
From:
TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:
Shared Package
ML20087L449 List:
References
NUDOCS 9508250262
Download: ML20087L458 (14)
v • d • e


Text

l IEEE SAFETY LIMITS A M LIMITING SAFETY SYSTEM SETTINGS l 1

SECTION

, P. AGE 2.1 SAFETY LIMITS 2.1.1 REACTOR 2.1.2 C0RE................................................ 2-1 REACTOR COOLANT SYSTEM PRESSURE............................. 2-1 44!c"ai :.:-:. = :T : = :T= : = 0 = TT L;":T.....................

] L G:0GE 2.i-T C T  % ACTOR 0000-0ATCT:

LiHii..................... -b J i 2.2 LIMITING SAFETY SYSTEM SETTINGS 1 2.2.1 r 1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0!NTS.......

TABLE 2.2-1 2 2 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP2/-3 SETPOIN BASES L i 4

' SECTION PAGE j 2.1 SAFETY LIMITS -

! 2.'1.1 REACTOR 2.1.2 C0RE................................................ B 2-1 REACTOR COOLANT SYSTEM PRESSURE............................. B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 4

2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0lNTS... B 2-3 4

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COMANCHE PEAK - UNITS 1 AND 2 tii i

9508250262 950915 Unit 1 - Amendment No. 14 PDR ADOCK 05000445 P PDR

2.0" SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

! 2.1 SAFETY LDRITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T,) shall not exceed the limits shown in l

~~ ' ~~'

APPLICABILITY: MODES I and 2. {[+he.C, ore.Operding*

Qmd5 llepo/T (COLR\. ,

AGI1Qli:  ;

l Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STAND 8Y within I hour, and comply with the require-monts of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

AGI1Qti:

l MODES I and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within I hour, and comply with the requirements of Specification 6.7.1.

MODES 3, 4 and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.

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COMANCHE PEAK - UNITS 1 AND 2 2-2 Unit 1 - Amendment No. H. 21 '

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COMANCHE PEAK - UNITS 1 AND 2 2-3 Unit 1 - Amendment No. 14

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2.1

  • SAFETY LIMITS BASES i

2.1.1 REACTOR CORE

! The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is pre-vented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface tempera-i ture is slightly above the coolant saturation temperature.

f Operation above the upper boundary of the nucleate boiling regime could

result in excessive cladding temperatures because of the onset of departure i from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and i~

therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been

! related to DNB. This relation has been developed to predict the DNB heat flux and the location of DNB for axially uniform and non-uniform heat flux distri-

! butions. The local heat flux ratio (DNBR), defined as the ratio of the heat i flux that would cause DNB at a particular core location to the local heat j flux, is indicative of the margin to DNB.

The DNB design basis is that the minimum DNBR of the limiting rod during Con-dition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used. The correlation DNBR limit is established based on i the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence level that DNB will not occur when the l minimum DNBR is at the DNBR limit. In meeting this design basis, uncertain-

ties in plant operating parameters are considered such that the minimum DNBR
for the limiting rod is greater than or equal to the DNBR limit. In addition, margin has been maintained in the design by meeting safety analysis DNBR limits in performing safety analyses.
+

The0 curves U t .. I a : show the loci of points of THERMAL POWER, l

Reactor Coolant System pressure and average temperature below which the calculated DNBR is no less than the safety analysis limit value, or the average enthalpy at the vessel exit is less than the enthalpy of saturated i liquid.

Re.ase Tor Cor4. McTy NNI l

COMANCHE PEAK - UNITS 1 AND 2 8 2-1 Unit 1 - Amendment No. 14

LIMITING SAFETY SYSTEM SETTINGS 4

BASIS Intermediate and Source Ranoe. Neutron Flux l The Intermediate and Source Range', Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a suberitical condi-tion. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. In addition, the Source Range Neutron Flux trip provides similar protection during shutdown operations with the reactor trip breakers closed and the rod control system capable of control rod with-drgwal. The Source Range channels will initiate a Reactor trip at about 10 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will Initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked )

when P-10 becomes active. i Overtemnerature N-16 The Overtemperature N-16 trip provides core protection to prevent DNB for i all combinations of pressure, power, coolant temperature, and axial power dis-tribution, provided that the transient is slow with respect to piping transit delays from the core to the N-16 detectors, and pressure is within the range between the Pressurizer High and Low Pressure trips. The setpoint is auto , .

matically varied with: (1) coolant temperature to correct for temperature I induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the cold leg temperature detec-tors, (2) pressurizer pressure, and (3) axial power distribution. With no mal axial r distribution this Reactor trip limit is always below the N . If axial peaks are greater than design, as inc cated by the difference between top and bottom power range nuclear 3.,

detectors, the Reactor trip is automatically reduced-r- m r r: nnMx 5 !x:: :.: ;^ l Overnower N-16 t.odof Cote M48$ Y biJS 00W l The Overpower N-16 trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature trip, and provides a backup to the High Neutron Flux trip. The Overpower N-16 trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releases."

Pressurizer Pressure In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pres-l sure trip thus limiting the pressure range in which reactor operation is per-mitted. The Low Setpoint trip protects against low pressure which could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure.

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COMANCHE PEAK - UNITS 1 AND 2 B 2-5

__m_____ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ __ ._

, @ STRATIVE CONTROLS -

, MONTHLY OPERATING REPORTS (Continued) i i

shall be submitted on a monthly basis to the U.S. Nuclear Regulatory l Comeission, Document Control Desk, Washington, D.C. 20555, with a copy to the l Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING LINITS REPORT 4- 6.9.1.6a Core operating limits shall be established and documented in the j ,,, CORE OPERATING LINITS REPORT (COLR) before each reload cycle or any remaining

g. part of a reload cycle for the following:
  • "I 1). Moderator temperature coefficient 80L and EOL limits and 300 ppe sur-i el $ veillance limit for Specification 3/4.1.1.3, y 2). Shutdown Rod Insertion Limit for Specification 3/4.1.3.5, )

, ,y 3). Control Rod Insertion Limits for Specification 3/4.1.3.6, l E3 4). AXIAL FLUX DIFFERENCE Limits and target band for Specification O* $ 3/4.2.1., i i

4- 5). Heat Flux Hot Channel Factor, K(Z), W(Z)., F,", and the F.C(Z) d# allowances for Specification 3/4.2.2, at

  • 6). Nuclear Enthalpy Rise Hot Channel 4ctor '.imit and the Power Factor 4 Multiplier for Specification 3/4..c .

M 6.9.1.6b The following analytical methods wed to determine the core operating limits are for Units I and 2, unless otherwise stated, and shall be those previously approved by the NRC in:

1). WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY,"

July 1985 (M Proprietary). (Methodology for Specifications 3.1.1.3 -

Modarator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 - Control Bank Insertion Limits, 1.2.1 - Axial Flux g Difference, 3.2.2 - Heat Flux Hot Channel FactorW, 3.2.3 - Nuclear g Enthalpy Rise Hot Channel Factog.)

  • 2). W 4P-8385, " POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES -

TOPICAL REPORT,' September 1974 (W Proprietary). (Methodology for t$- Specification 3.2.1 - Axial Flux Difference (Constant Axial Offset g , Control).)

W .L.

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j 3). T. N. Anderson to K. Kniel (Chief of Core Performance Branch, NRC g "d -

January 31, 1980--

Attachment:

Operation and Safety Analysis Aspects i rs of an Improved Load Follow Package. (Methodology for Specification x  !

b 3.2.1 - Axial Flux Difference (Constant Axial Offset Control).) I 1 j- d 4). NUREG-0800, Standard Review Plan, U.S. Nuclear Regulatory Commission,

"" Section 4.3, Nuclear Design, July 1981. Branch Technical Position '

CPS 4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Rev. 2, July 1981. (Methodology for Specification 3.2.1 - Axial Flux Difference (Constant Axial Offset Control].)

COMANCHE PEAK - UNITS 1 AND 2 6-20 Unit 1 - Amendment No. M4. 34 Unit 2 - Amendment No. 20 l

a-

ADMINISTRATIVE CONTROLS

' CORE OPERATING LINITS REPORT (Continued) i 5). WCAP-10216-P-A, Revision IA " RELAXATION 0F CONSTANT AXIAL OFFSET l CONTROL F. SURVEILLANCE TECHNICAL SPEC!FICATION," February 1994 (M i Proprietary). (Methodology fgr Specification 3.2.2 - Meat Flux Hot j Channel Factor (W(z) surveillance requirements for F. Metho1 ology).)

) 6). WCAP-10079-P-A, "NOTRUMP, A N00AL TRANSIENT SMALL BREAK AND GENERAL .

j g; NETWORK CODE," August 1985, (M Proprietary). I

- J'

! 7). WCAP-100b4-P-A, " WESTINGHOUSE BREAK ECCS EVALUATION MODEL USING l THE NOTRUNP CODE", August 198 1,(MProprietary). i 8). WCAP-11145-P-A, "WESTINGH0USE LL BREAK LOCA ECl UATION MODEL

] $e 64 t GENERICSTUDYWITHTHENOTRUMPCODE", October 198)v,(Mf

, *j 9). RXE-90-006-P, " Power Distribution Control Analysis and Overtemperature N-16 and Overpower N-16 Trip Setpoint Methodology," February 1991.  !

d d (Methodology for Specification 3.2.1 - Axial Flux Difference, 3.2.2 -

  • D> Heat Flux Hot Channel Facto 3.)

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+- 10). RXE-88-102-P, "TUE-1 Departure from Nucleate Boiling Correlation",

January 1989.

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11) . RXE-88-102-P, 'a i

., "TUE-1 DNB Correlation - Supplement 1",

l December 1990.

12). RXE-89-002, "VIPRE-01 Core Thermal-Hydraulic Analysis Methods for i

Comanche Peak Steam Electric Station Licensing Applications", June 1989.

13). RXE-91-001, " Transient Analysis Methods for Comanche Peak Steam

Electric Station Licensing Applications", February 1991.

14). RXE-91-002, " Reactivity Anoisaly Events Methodology", May 1991.

! (Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit 3.1.3.6 -

control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 -

Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot

! Channel Factor.)

15). RXE-90-007, "Large Break Loss of Coolant Accident Analysis Methodology", December 1990.

16). TXX-88306, " Steam Generator Tube Rupture Analyr.is", March 15, 1988.

17). RXE-91-005, " Methodology for Reactor Core Resr,onse to Steamline Break Events," May, 1991'.

f COMANCHE PEAK - UNITS 1 AND 2 6-21 Unit 1 - Amendment No. 1,5,10,10,21.2",34 Unit 2 - Amendment No. i,7,10,20 i

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.1 ENCLOSURE 1 TO TXX 95213 l l

GENERIC LETTER 88 16 REMOVAL OF CYCLE SPECIFIC PARAMETER  !

I LIMITS FROM TECHNICAL SPECIFICATIONS i I

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umHTED STATES I  ! o NUCLEAR REGULATORY COMMISSION

, {. waewueeron, p. c. noses

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s TO ALL POWER REACTOR LICENSEES AND APPLICANTS l

SUBJECT:

REMOVAL OF CYCLE-SPECIFIC PARAMETER LIMITS FROM TECHNICAL SPECIFICATIONS (GENERIC LETTER 88-16)  :

l License amendments are generally required each fuel cycle to update the values

! of cycle specific parameter limits in Technical Specifications (TS). The ;

4 processing of changes to TS that are developed using an NRC approved method-  ;

a ology is an unnecessary burden on licensee and NRC resources. A lead plant i proposal for an alternative that eliminctes the need for a license amendment  ;

to update the cy:le-specific parameter limits each fuel cycle was submitted i j for the Oconee plant with the endorsement of the Babcock and Wilcox Owners  !

1 Group. On the basis of the NRC review and approval of that proposal, the en-  ;

closed guidance for the preparation of a license amendment request for this  ;

. alternative was developed by the NRC staff.

4-

! Generally, the methodology for determining cycle-specific parameter limits is

. documented in an NRC-approved Topical Report or in a plant-specific submittal. i 1 As a consequence, the NRC review of proposed changes to TS for these limits

! is primarily limited to confirmation that the updated limits are calculated .

! using an NRC-approved methodology and consistent with all applicable limits i i of the safety analysis. These changes also allow the NRC staff to trend the  ;

values of these limits relative to past experience. This alternative allows  ;

i continued trending of these limits without the neccssity of prior NRC review j j and approval.  !

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' Licensees and applicants are encouraged to propose changes to TS that are "

consistent with the guidance provided in the enclosure. Conforming amendments i

! will be expeditiously reviewed by the NRC Project Manager for the facility.

! Proposed amendments that deviate from this guidance will require a longer, more t

, detailed review. Please contact the. Project Manager if you have questions on this matter.

i Sincerely, t

1 1

8810050058

  • p l Dennis M. Crutchfiel ,

I DU p RL D n

U Acting Associate Di ctor for Projects  ;

Office of Nuclear Reactor Regulation  :

Enclosure:

As stated 00T 211988  :

e

. WILLIAM G. COUNSIL 9 Tim n5551 y j

4 Generi,c Letter 88- 16 Enclosure i

! GUIDANCE FOR TECHNICAL SPECIFICATION CHANGES  :

FOR CYCLE-SPECIFIC PARAMETER LIMITS INTRODUCTION  ;

A number of Technical Specifications (TS) address limits associated with  !

i reactor physics parameters that gencrally change with each reload core, requ1r- -

ing the processing of changes to TS to update these limits each fuel cycle. '

If these limits are developed using an NRC-approved methodology, the license f 1

amendment process is an unnecessary burden on the licensee and the NRC. An  ;

. alternative to including the values of these cycle-specifi parameters in in- t

dividual specifications is provided and is responsive to industry and NRC l

! efforts on improvements in TS. j

- This enclosure provides guidance for the preparation of a license amendment 1 request to modify TS that have cycle-specific parameter limits. An acceptable

. alternative to specifying the values of cycle-specific parameter limits in TS j was developed on the basis of the review and approval of a lead plant propusal  !

j for this change to the TS for the Oconee units. The implementation of this J alternative will result in a resource savings for the licensees and the NRC by ,

! eliminating the majority of license amendment requests on changes in values of

cycle-specific parameters in TS.

DISCUSSION l

l This alternative consists of three separate actions to modify the plant's TS:

(1) the addition of the definition of a named formal report that includes the

[ values of cycle-specific parameter limits that have been established using an NRC-approved methodology and consist,at with all applicable limits of the safe-l ty analysis, (2) the addition of an administrative reporting requirement to sub-

mit the formal report on cycle-specific parameter limits to the Commission for ,

information, and (3) the modification of individual TS to note that cycle- t

! specific parameters shall be maintained within the limits provided in the

! defined formal report. .

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In the evaluation of this alternative, the NRC staff concluded that it is '

l essential to safety that the plant is operated within the bounds of cycle-specific parameter limits and that a requirement to maintain the plant within ,

the appropriate bounds must be retained in the TS. However, the specific
values of these limits may be modified by licensees, without affecting nuclear

safety, provided that these changes are determined using an NRC-approved method-ology and consistent with all applicable limits of the plant safety analysis '

that are addressed in the Final Safety Analysis Report (FSAR). Additionally, j

, it was concluded that a formal report should be submitted to NRC with the i values of these limits. This will allow continued trending of this information, i

, even though prior NRC approval of the changes to these limits would not be 1 required. i The current method of controlling reactor physics parameters to assure conform-ance to 10 CFR 50.36 is to specify the specific value(s) determined to be with- l in specified acceptance criteria (usually the limits of the safety analyses) ~

using an approved calculation methodology. The alternative contained in this guidance controls the values of cycle-specific parameters and assures conform-  !

ance to-10 CFR 50.36, which calls for specifying the lowest functional  !

1

,, Generic Letter 88- 16 Enclosure performance levels acceptable for continued safe operation, by specifying the calculation methodology and acceptance criteria. This permits operation at any specific value determined by the licensee, using the specified methodology, to be within the acceptance criteria. The Core Operating Limits Report will docu-ment the specific values of parameter limits resulting from licensee's calcula- ,

l tions including any mid-cycle revisions to such parameter values.

l The following items show the changes to the TS for this alternative. A defined formal report, " Core Operating Limits Report" (the name used as an example for  !

the title for this report), shall be added to the Definitions section of the '

TS, as follows.

[CORF1 OPERATING LIMITS REPORT 1.XX l he [ CORE] OPERATING LIMITS REPORT is the unit-specific document that provides [ core] operating limits for the current operating reload cycle. These cycle-specific [ core] operating limits shall be determined for each reload cycle in accordance with Specification 6.9.X. Plant operation within these operating limits is addressed in individual specifications.

A new administrative reporting requirement shall be added to existing reporting requirements, as follows.

CORE] OPERATING LIMITS REPORT J6.9.X] [ Core] operating limits shall be established and documented in the

[ CORE] OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle. (If desired, the individual specifications that address [ core] operating limits may be referenced.) The analytical methods used to determine the [ core] operating limits shall be those previously re-viewed and approved by NRC in [ identify the Topical Report (s) by number, title, and date, or identify the staff's safety evaluation report for a plant-specific methodology by NaC letter rM date). The [ core] operating limits shall be determined se that all appi, cable limits (e.g., fuel therm-al mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits)

< of the safety analysis are met. The [ CORE] OPERATING LIMITS REPORT, in-cluding any mid-cycle revisions or supplements thereto, shall be provided i upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

Individual specifications shall be revised to state that the values of cycle-specific parameters shall be maintained within the limits identified in the defined formal report. Typical modifications for individual specifications are as follows.

The regulating rods shall be positioned within the acceptable operating range for regulating rod position provided in the [ CORE] OPERATING LIMITS REPORT. (Used where the operating limit covers a range of acceptable operation, typically defined by a curve.)

The [ cycle-specific parameter] shall be within the limit provided in the

[ CORE] OPERATING LIMITS REPORT. (Used where the operating limit has a single point value.)

' Generic Letter 88- 16 Enclosure

SUMMARY

The alternative to including the values of cycle-specific parameter limits in' individual specifications includes (1) the addition of a new defined term for the formal report that provides the cycle-specific parameter limits, (2) the addition of its associated reporting requirement to the Administrative Controls section of the TS, and (3) the modification of individual specifications to re-place these limits with a reference to the defined formal report for the values of these limits. With this alternative, reload license amendments for the sole purpose of updating cycle-specific parameter limits will be unnecessary.

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Enclosure LIST OF RECENTLY ISSUED GENERIC LETTERS Generic Oate of Letter No. Subject Issuance Issued To 88-15 ELECTRIC POWER SYSTEMS - 09/12/88 ALL POWER REACTOR INADEQUATE CONTROL OVER LICENSEES AND DESIGN PROCESSES APPLICANTS 88-14 INSTRUMENT AIR SUPPLY 08/08/88 ALL HOLDERS OF SYSTEM PROBLEMS AFFECTING OPERATING LICENSES SAFETY-RELATED EQUIPMENT OR CONSTRUCTION PERMITS FOR NUCLEAR POWER REACTORS 88-13 OPERATOR LICENSING 08/08/88 ALL POWER REACTOR EXAMINATIONS LICENSEES AND APPLICANTS FOR AN OPERATING LICENSE.

88-12 REMOVAL OF FIRE PROTECTION 08/02/88 ALL POWER REACTOR REQUIREMENTS FROM TECHNICAL LICENSEES AND SPECIFICATIONS APPLICANTS 88-11 NRC POSITION ON RADIATION 07/12/88 ALL LICENSEES OF EMBRITTLEMENT OF REACTOR OPERATING REACTORS VESSEL MATERIALS AND ITS AND HOLDERS OF

IMPACT ON PLANT OPERATIONS CONSTRUCTION PERMITS 88-10 PURCHASE OF GSA APPROVED 07/01/88 ALL POWER REACTOR SECURITY CONTAINERS LICENSEES AND HOLDERS OF PART 95-APPROVALS I

88-09 PILOT TESTING 0F FUNDAMENTALS 05/17/88 ALL LICENSEES OF ALL EXAMINATION BOILING WATER REACTORS

  • AND APPLICANTS FOR A BOILING WATER REACTOR OPERATOR'S LICENSE UNDER 10 CFR PART 55 88-08 MAIL SENT OR DELIVERED TO 05/03/88 ALL LICENSEES FOR POWER THE OFFICE OF NUCLEAR REACTOP AND NON-POWER REACTORS REGULATION AND HOLDERS OF CONSTRUCTION PERMITS FOR NUCLEAR POWER REACTORS 88-07 MODIFIED ENFORCEMENT POLICY 04/07/88 ALL POWER REACTOR RELATING TO 10 CFR 50.49, LICENSEES AND

" ENVIRONMENTAL QUALIFICATION APPLICANTS OF ELECTRICAL EQUIPMENT IMPORTANT TO SAFETY FOR NUCLEAR POWER PLANTS"

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