ML20087J953
| ML20087J953 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 08/15/1995 |
| From: | TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | |
| Shared Package | |
| ML20087J931 | List: |
| References | |
| NUDOCS 9508220278 | |
| Download: ML20087J953 (5) | |
Text
,D-e TABLE 2.2-1 (Continued) .
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS .
I FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE
$$ 8. Overpower N-16 sll2% of RTP* $114.5% of RTP*
80 xm 9. Pressurizer Pressure-Low -
o .o a. Unit 1 21880 psig 21863.6 psig '
$$ b. Unit 2 21880 psig 21865.2 psig om
-$SE 10. Pressurizer Pressure-High I$ a. Unit I s2385 psig s2400.8 psig
- b. Unit 2 s2385 psig $2401.4 psig
- 11. Pressurizer Water Level-High $92% of instrument s93.9% of instrument.
span span
- 12. Reactor Coolant Flow-Low -
- a. Unit 1 29 of loop .6% of loop fl ow** fl ow**
- b. Unit 2
~
290% of 1000 a _. of loop a C-t 'n:- r::;c;D C-t '--- r ::;;;t.)
flow
- b flow 'Ee l
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, ; Ias specINd I" w core Oper hs Limih R*P d S " RD i
{11::;-:.^-1.--:::::: ::;; 0,000 ,d l
COMANCHE PEAK - UNIT 1 AND 2 2-6 Unit 1 - Amendment No. 4hM,41 Unit 2 - Amendment Nc 27
, POWER DISTRIBllIION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNS-related parameters shall be maintained within the stated limits: ,
1
- a. Indicated. Reactor Coolant System T, s 592'F i
- b. Indicated Pressurizer Pressure 1 2219 psi I e l c.. Indicated Reactor Coolant System (RCS) Flow kk" ;; ^^ fr WM : T I MZ,;Z ;; ^^ ". 0/.^. : J {
APPLICABILITY: MODE 1. he. Value specified in the. CoM gm (Opera:Wg Limits Repoit (col i With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of l RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE RE00fREMENTS 4.2.5.1 Each of the above parameters shall be verified to be within its limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
~
4.2.5.2 The RCS total flow rate shall be verified to be within its limits at !
least once per 31 days by plant computer indication or measurement of the RCS l albow tap differential pressure transmitters' output voltage.
4.2.5.3 The RCS loop flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months. The channels shall be normalized based on the RCS flow rate determination of Surveillance Requirement 4.2.5.4. 3 4.2.5.4 The RCS total flow rate shall be determined by precision heat balance measurement after each fuel loading and prior to operation above 85% of RATED l THERMAL POWER. The feedwater pressure and temperature, the main steam pres- i sure, and feedwater flow differential pressure instruments shall be calibrated '
within 90 days of performing the calorimetric flow measurement.
- Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.
- Includes a 3.8% flow measurement uncertainty. !
E0MANCHE PEAK - UNITS 1 Ale 2 3/4 2-12 Unit 1 - Amendment No, MW,30 l Unit 2 - Amendment No. 7,16 l i
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= ADMINISTRATIVE CONTROLS m .
v MNTHLY OPERATING REPORTS 'IContinued) '
l shall be submitted on-a monthly' basis to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, with a copy to the :
Regional Administrator of the Regional Office of the NRC, no later than the
$g) _
15th of each month following the calendar month covered by the report. !
g CORE OPERATING LIMITS rep 0RT N
O- Core operating' limits shall'be established and documented in the i 6.9.1.6a}COREOPERATINGLINITSREPORT(COLR)beforeeac Y[ *f m
part of a reload cycle for the following: <
1). Moderator temperature coefficient 80L and E0L limits and 300 ppe sur .
is f veillance limit for Specification 3/4.1.1.3, i g
.g
}g) d 2). Shutdown Rod Insertion Limit for Specification 3/4.1.3.5, g 3). Control Rod Insertion Limits for Specification 3/4.1.3.6, gs (y E 4). AXIAL FLUX OIFFERENCE Limits and target band for Specification 4 l 58 3/* 2 1 - t E
A -e $o 5). Heat Flux Hot Channel Factor, K(Z), W(Z), F/*, and the F/(Z) allowances for Specification 3/4.2.2, jEh 6). Nuclear Enthalpy Rise Hot Channel Factor Limit and the Power Factor
- f. 'g8fN 6 Multiplier for Specification 3/4.2.3.
es g ,
6.9.1.6b The following analytical methods used to determine the core
- 4 Difference [ Constant Axial Offset Control).) s COMANCHE PEAK - UNITS 1 AND 2 6-20 Unit 1 - Amendment No. 6d4,34 Unit 2 - Amendment No. 20 7 'i- AmlINI1TRATIVE CONTROLS' ? , CORE OPERATING LIMITS REPORT '(Continued) 5). WCAP-10216-P-A, Revision IA, " RELAXATION OF CONSTANT AXIAL 0FFSET . CONTROL F JProprieta. SURVEILLANCEfor TECHNICAL SPECIFICATION," 3.2.2 - Heat Flux February 1994 (W y } ry)..(Methodology Specification Hot Channel Factor (W(z) surveillance requirements for F Methoiology).) c 6). WCAP-10079-P-A, "NOTRUMP, A N00AL TRANSIENT SMALL BREAK AND GENERAL. ,ars NETWORK CODE," August 1985, (M Proprietary). 7). WCAP-10054-P-A, "WESTINGHOUS SfjALQ BREAK ECCS EVALUATION MODEL USING THE NOTRUMP CODE", August 19 ,(WProprietary). ls 8). WCAP-11145-P-A, " WESTINGHOUSE LL BREAK LOCA Ecr EV LUATION MODEL GENERIC STUDY WITH THE NOTRL#tP CODE", October 1980,(M oprietary). b 9). RXE-90-006-P, " Power Distribution Control Analysis and Overtemperature-N-16 and Overpower N-16 Trip Setpoint Methodology," February 1991. cv (Methodology-for Specification 3.2.1 - Axial Flux Difference, 3.2.2 - d _ _ Heat Flux Hot Channel Factoy.) ] 10). RXE-88-102-P, "TUE-1 Departure from Nucleate Boiling Correlation",- January 1989. 6 11). RXE-88-102-P, Sup. 1, "TUE-1 DN6 Correlation - Supplement 1", j l December 1990. , ) 12). RXE-89-002, "VIPRE-01 Core Thennal-Hydraulic Analysis Methods for l La Comanche Peak Steam Electric Station Licensing Applications", June i ci g 1989. + j hh , w 13). RXE-91-001, " Transient Analysis Methods for Comanche Peak Steam . Electric Station Licensing Applications", February 1991.#-) j ~- om t; p- 14). RXE-91-002, " Reactivity Anomaly Events Methodology", May 1991. 8 (Methodology for Specification 3.1.1.3 - Moderator Temperature 25 Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit 3.1.3.6 - '60C Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Not Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot . A' Channel Factor.) dikycr 15). RXE-90-007, "Large Break Loss of Coolant Accident Analysis Methodology", December 1990. a j .$ C 16). TXX-88306, " Steam Generator Tube Rupture Analysis", March 15, 1988. 17). RXE-91-005, " Methodology for Reactor Core Response to Steamline Break Events," May, 1991. k' al 4H 7 32 2E I (L. COMANCHE PEAK - UNITS 1 AND 2 6-21 Unit 1 - Amendment No.1,5,10,l".21,2*,34 Unit 2 - Amendment No. 0,7,10,20 e- 4 @ M n ,k on A. Vy.F
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-L r I 9 A , b .i 'a-T' L .l ' L .l It 4 ^ , _.5 l .) .k 'i ENCLOSURE 1 TO TXX-95216 f I GENERIC LETTER 88 16 REMOVAL OF' CYCLE SPECIFIC PARAMETER i i LIMITS FROM TECHNICAL SPECIFICATIONS i ? I ) i P f 'f s i i + I i l 'l i H I f a m-7. w;. . /# . UNITED STATES .!' 'o' NUCLEAR REGULATORY COMMISSION , '[ wasunseron o.c.seous ; 9, i %,..... QCT 0 4 m ! + TO ALL POWER REACTOR LICENSEES AND APPLICANTS :
SUBJECT:
REMOVAL 0F 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).
l The ,
processing of changes to TS that are developed using an NRC-approved method- i ology is an unnecessary burden on licensee and NRC resources. A lead plant 1 proposal for an alternative that eliminates the need for a license amendment- !
to update the cycle-specific parameter limits each fuel cycle was~ submitted i for the Oconee plant with the endorsement of the Babcock and Wilcox Owners 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. ,
Generally, the methodology for determining cycle specific parameter' limits is !
documented in an NRC-approved Topical Report or in a plant-specific submittal.
As a consequence, the NRC review of proposed changes to TS for these limits i is primarily limited to confirmation that the updated limits are calculated -
using an NRC-approved methodology and consistent with all applicable limits 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 l continued trending of these limits without the' necessity of prior NRC review l and approval.
Licensees and applicants are encouraged to propose changes to TS that are consistent with the guidance provided in the enclosure. Conforming amendments will be expeditiously reviewed by the NRC Project Manager for the facility. ;
Proposed amendments that deviate from this guidance will require a longer, more :
detailed review. Please contact the Project Manager if you have questions on- !
this matter. :
Sincerely, l
-i i
~!
8810050058
- p Denn s M. Crutchfiel
- Acting Associate Di ctor for Projects RECE D Office of Nuclear Reactor Regulation l l
Enclosure:
As stated OCT 211988 ;
WILLIAM G. COUNSIL
~
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-l
+
x . 1 Gener.ic; Letter 88-.16i Enclosure'-
GUIDANCE FOR TECHNICAL SPECIFICATION CHANGES FOR CYCLE-SPECIFIC PARAMETCR LIMITS i INTRODUCTION 4
~A number of Technical Specifications (TS) address _ limits. associated with ;
reactor _ physics parameters that generally change'with each reload core, requir-ing the processing of changes to TS to update these limits each fuel' cycle. M If these limits are developed using an NRC-approved methodology, the' license !
amendment process is an unnecessary burden on the licensee.and-the NRC. An t alternative to including the values of these cycle-specific parameters-in in-dividual specifications is'provided and is responsive to industry and NRC !
efforts on improvements in TS.
This enclosure provides guidance for the preparation of a license amendment i request to modify TS that have~ cycle-specific parameter limits. An acceptable !
alternative to specifying the values of cycle-specific parameter limits in TS l was developed on the basis of the review and approval of a lead p'lant proposal for this change to the TS for the Oconee units. The implementation of this ,
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 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 consistent with all applicable limits of the safe- !
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- .i specific parameters shall be maintained within the limits provided in the l defined formal report.
- In the evaluation of this alternative, the NRC staff concluded that it is 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 i that are addressed in the Final Safety Analysis Report (FSAR). Additionally, '
it was concluded that a formal report should be submitted to NRC with the '
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 required. ;
' 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- !
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 :
i
. - . - + -, , , , . , . - - - - , , _ - . - - -. .-. - . - - . - , . -
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-tions including any mid-cycle revisions to such parameter values.
The following items show the changes to the TS for this alternative. A defined 1 formal report, " Core Operating Limits Report" (the name used as an example for l the title for this report), shall be added to the Definitions section of the TS, as follows.
[ CORE] OPERATING LIMITS REPORT
- 1. xX The [ 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 l for each reload cycle in accordance with Specification 6.9.X. Plant I operation within these operating limits is addressed in individual i specifications.
A new administrative reporting requirement shall be added to existing reporting requirements, as follows.
- CORE] OPERATING LIMITS REPORT
- 6.9.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 NRC letter and date). The [ core) operating limits shall be determined so that all applicable 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 revisicns or supplements thereto, shall be provided 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.)
r 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.
k i
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- s. , c., 1
., a Enclosure ii LIST OF RECENTLY ISSUED GENERIC LETTERS !
Generic- Date 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 1 APPLICANTS -i 88-14 INSTRUMENT AIR SUPPLY 08/08/88 ALL HOLDERS OF SYSTEM PROBLEMS AFFECTING h OPERATING LICENSES -
SAFETY-RELATED EQUIPMENT OR CONSTRUCTION- 1 PERMITS FOR NUCLEAR- !
POWER REACTORS !
88-13 OPERATOR LICENSING 08/08/88 ALL POWER REACTOR EXAMINATIONS !
LICENSEES AND APPLICANTS FOR I AN OPERATING LICENSE.-
88-12 REMOVAL OF FIRE PROTECTION 08/02/88 ALL POWER REACTOR REQUIREMENTS.FROM TECHNICAL LICENSEES AND- !
SPECIFICATIONS APPLICANTS j
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 t HOLDERS-0F PART 95- !
APPROVALS '
88-09 PILOT TESTING OF FUNDAMENTALS 05/17/88 ALL LICENSEES OF ALL EXAMINATION l BOILING WATER REACTORS .
AND APPLICANTS FOR A !
BOILING WATER. REACTOR i OPERATOR'S LICENSE. i UNDER 10 CFR PART 55
(
88-08 l MAIL SENT OR DELIVERED TO
~
05/03/88 ALL LICENSEES FOR POWER THE OFFICE OF NUCLEAR REACTOR q REGULATION AND NON-POWER REACTORS AND HOLDERS OF CONSTRUCTION PERMITS FOR NUCLEAR' POWER REACTORS 88-07' MODIFIED ENFORCEMENT POLICY 04/07/88 RELATING TO 10 CFR E0.49, ALL POWER REACTOR LICENSEES AND
" ENVIRONMENTAL QUALIFICATION APPLICANTS OF ELECTRICAL EQUIPMENT IMPORTANT TO SAFETY FOR NUCLEAR POWER PLANTS"
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