Text

Application for Technical Specification Change to Adopt Technical Specifications Task Force (TSTF)-339, Relocate TS Parameters to the COLR Consistent with WCAP-14483, Revision 2
	ML16116A033
Person / Time
Site:	Robinson
Issue date:	04/24/2016
From:	Glover R; Duke Energy Progress
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RNP-RA-/15-0016
	Download: ML16116A033 (36)
	v • d • e

(~ DUKE R. Michael Glover H. B. Robinson Steam Electric Plant Unit 2 ENERGYQP Site Vice President Duke Energy Progress 3581 West Entrance Road Hartsville, SC 29550 0 . 843 857 1704 F: 843 857 1319 Mike. Glovet@d11ke-e11ergy.com 10 CFR 50.90 Serial: RNP-RA/15-0016 APR 2 4 2016 U.S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington, DC 20555-0001 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 I RENEWED LICENSE NO. DPR-23 APPLICATION FOR TECHNICAL SPECIFICATION CHANGE TO ADOPT TECHNICAL SPECIFICATIONS TASK FORCE (TSTF)-339, RELOCATE TS PARAMETERS TO THE COLR CONSISTENT WITH WCAP-14483, REVISION 2

Dear Sir/Madam:

In accordance with the provisions of 10 CFR 50.90 Duke Energy Progress, Inc. is submitting a request for an amendment to the technical specifications (TS) for H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP2) . The proposed amendment would relocate Reactor Coolant System (RCS)-related cycle-specific parameters and core safety limits from the TS to the Core Operating Limits Report (COLR) . provides a description and assessment of the proposed change, the requested confirmation of applicability, and plant-specific verifications. Attachment 2 provides the existing TS pages marked up to show the proposed change. Attachment 3 provides revised (clean) TS pages. provides proposed TS bases changes, for information only.

HBRSEP2 requests approval of the proposed License Amendment by April 30, 2017, with the amendment being implemented within 120 days of issuance.

In accordance with 10 CFR 50.91 , a copy of this application, with attachments, is being provided to the designated South Carolina Official.

Please address any comments or questions regarding this matter to Mr. Scott Connelly, Acting Manager- Nuclear Regulatory Affairs at (843) 857-1569.

I declare under penalty of perjury that the foregoing is true and correct. Executed on A,,.,1 i 'i , 2016.

Sincerely, R-N1,,AaJ.~~

R. Michael Glover Site Vice President

U.S. Nuclear Regulatory Commission Serial: RNP-RA/15-0016 Page2 RMG/jmw Attachments

1. Description and Assessment

2. Proposed Technical Specification Changes

3. Revised Technical Specification Pages

4. Proposed Technical Specification Bases Changes cc: Administrator, NRC, Region II Dennis Galvin, NRC Project Manager, NRR NRC Resident Inspector, HBRSEP2 Ms. S. E. Jenkins, Manager, Infectious and Radioactive Waste Management Section (SC)

U. S. Nuclear Regulatory Commission to Serial: RNP-RA/15-0016 5 Pages (including this cover page}

Description And Assessment of Proposed Changes

U. S. Nuclear Regulatory Commission to Serial: RNP-RA/15-0016 Page 2 of 5

1.0 DESCRIPTION

AND BASIS OF PROPOSED CHANGES 1.1 Description of Proposed Changes The proposed amendment would revise Technical Specification (TS) 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," and associated Bases, by relocating the pressurizer pressure, Reactor Coolant System (RCS) average temperature, and RCS total flow rate values to the Core Operating Limits Report (COLR).

The proposed amendment would relocate TS Figure 2.1.1-1, "Reactor Core Safety Limits," to the COLR, replacing it with more specific requirements regarding the safety limits (i.e. fuel DNB design basis and the fuel centerline melt design basis) conforming with WCAP-14483-A. As discussed in the Safety Evaluation Reports to Westinghouse Topical Report, WCAP-14483-A, it is necessary to relocate TS Figure 2.1.1-1 to the COLR since cycle-dependent changes to parameters upon which TS Figure 2.1.1-1 is based would require a license amendment request to revise the figure.

The amendment would revise TS Table 3.3.1-1, "Reactor Protection System Instrumentation,"

by relocating numerical values pertaining to Overtemperature LlT and Overpower LlT, nominal RCS operating pressure, nominal Tavg, time constants (-c), and constant (K) values to the COLR.

TS 5.6.5, "Core Operating Limits Report (COLR)," will be modified to reflect the above relocations to the COLR.

The proposed changes will allow Duke Energy Progress, Inc. (DEP) the flexibility of enhancing operating and core design margins without the need for cycle-specific license amendment requests. The relocation of these cycle-specific TS values to the COLR will result in a more complete COLR, containing cycle-specific operating conditions and core reload related parameters. The safety and quality of operations at H. B. Robinson Steam Electric Plant, Unit No. 2, (HBRSEP2) will not be compromised by the implementation of this amendment request as TS 5.6.5(c) requires that all applicable limits of the safety analyses be met when generating cycle-specific requirements in the COLR.

1.2 Basis For Proposed Change NRC Generic Letter 88-16, "Removal of Cycle-Specific Parameter Limits From Technical Specifications," dated October 3, 1988, provides guidance to licensees for the removal of cycle-dependent variables from the TS provided that these values are included in a COLR and are determined with NRG-approved methodologies referenced in the TS. Westinghouse Electric Company (Westinghouse) subsequently developed WCAP-14483, "Generic Methodology for Expanding Core Operating Limits Report," describing how cycle-specific parameters may be relocated to the COLR. WCAP-14483 was accepted for referencing by the NRC on January 19, 1999. The Safety Evaluation Report, contained in the January 19, 1999 NRC letter approving WCAP-14483-A, concludes that additional information contained in the TS may be relocated to the COLR.

U. S. Nuclear Regulatory Commission to Serial: RNP-RA/15-0016 Page 3 of 5 The limits on the parameters which are removed from the TS and added to the COLR must be developed or justified using NRG-approved methodologies. All accident analyses, performed in accordance with these methodologies, must meet the applicable NRG-approved limits of the safety analysis. The removal of parameter limits from the TS and their addition to the COLR does not obviate the requirement to operate within these limits. Furthermore, any changes to those limits must be performed in accordance with TS 5.6.5(c). If any of the applicable limits of the safety analyses are not met, prior NRC approval of the change is required, as is the case for a license amendment request. For more routine modifications, where NRG-approved methodologies and limits of the safety analysis remain applicable, the potentially burdensome and lengthy process of amending the TS may be avoided. The requested changes are essentially administrative in nature; therefore, the required level of safety will be maintained.

The requested changes are based upon NRG-approved Westinghouse Owner's Group (WOG)

Technical Specifications Task Force (TSTF)-339, "Relocated TS Parameters to the COLR Consistent with WCAP-14483, " Revision 2, and Westinghouse WCAP-14483-A. In accordance with these documents, previously approved RCS minimum total flow rates are retained in the TS to preclude the use of lower flow rates without prior NRC approval.

In addition, it has been discovered that there is conflicting information in the Bases of TS 3.4.1.

The Applicable Safety Analysis section of the Bases states:

"The pressurizer pressure limit of 2205 psig and the RCS average temperature limit of 579.4°F correspond to analytical limits used in the safety analyses, with allowance for measurement uncertainty."

The Limiting Conditions for Operation (LCO) section provides the following guidance:

"The LCO numerical values for pressure, temperature and flow rate are given for the measurement location but have not been adjusted for instrument error."

The above conflicting statements were contained in NUREG-1431, Standard Technical Specifications Westinghouse Plants, Revision 1, which served as the basis for HBRSEP2's conversion from Custom Technical Specifications to Improved Standard Technical Specifications. This conflict was subsequently corrected in Revision 2.2 of NUREG-1431 to reflect that instrument uncertainty should be included in the values and this correction continues to be present in the latest revision of NUREG-1431 (Revision 4). DEP has revised the Bases of the Bases of the HBRSEP2 TS to reflect that instrument uncertainty should be included in the LCO and surveillance values consistent with the guidance in the last revision of NUREG-1431 .

The LCO and Surveillance values in TS 3.4.1 for RCS pressure and temperature do not currently account for instrument uncertainty and are therefore non-conservative. The LCO and surveillance values for RCS flow do account for instrument uncertainty. In accordance with the guidance in NRC Administrative Letter 98-10, DEP has implemented administrative controls to utilize conservative limits for RCS pressure and temperature that do account for instrument uncertainty. It is intended to incorporate RCS pressure and temperature LCO and surveillance

U. S. Nuclear Regulatory Commission to Serial: RNP-RN15-0016 Page 4 of 5 values, which do account for instrument uncertainty, into the COLR as part of the implementation of the proposed amendment.

2.0 ASSESSMENT 2.1 Applicability of TSTF-339 DEP has reviewed TSTF-339, Rev. 2, and has concluded that the information in WCAP-14483-A and TSTF-339 is applicable to HBRSEP2 and justify the proposed amendment for the incorporation of the changes to the HBRSEP2 TS.

2.2 Optional Changes and Variations DEP is not proposing any variations or deviations from the TS changes described in TSTF-339, Revision 2.

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Determination As required by 10 CFR 50.91 (a)(1 ), this analysis is provided to demonstrate that the proposed license amendment does not involve a significant hazard.

Conformance of the proposed amendment to the standards for a determination of no significant hazards, as defined in 10 CFR 50.92, is shown in the following:

1) Does the proposed license amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

No. The relocation of RCS-related cycle-specific parameter limits from the TS to the COLR proposed by this amendment request does not result in the alteration of the design, material, or construction standards that were applicable prior to the change. The proposed change will not result in the modification of any system interface that would increase the likelihood of an accident since these events are independent of the proposed change. The proposed amendment will not change, degrade, or prevent actions, or alter any assumptions previously made in evaluating the radiological consequences of an accident described in the Updated Final Safety Analysis Report (UFSAR). Therefore, the proposed amendment does not result in an increase in the probability or consequences of an accident previously evaluated.

2) Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

No. There are no new accident causal mechanisms created as a result of NRC approval of this amendment request. No changes are being made to the facility which would introduce any new accident causal mechanisms. This amendment request does not

U.S. Nuclear Regulatory Commission to Serial: RNP-RA/15-0016 Page 5 of 5 impact any plant systems that are accident initiators. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3) Does the proposed change involve a significant reduction in margin of safety?

No. Implementation of this amendment would not involve a significant reduction in the margin of safety. Previously approved methodologies will continue to be used in the determination of cycle-specific core operating limits that are present in the COLR.

Additionally, previously approved RCS minimum total flow rates for HBRSEP2 are retained in the TS to assure that lower flow rates will not be used without prior NRC approval. Based on the above, it is concluded that the proposed license amendment request does not impact any safety margins and will not result in a reduction in margin with respect to plant safety.

Based on the preceding analysis, it is concluded that the relocation of RCS-related cycle-specific parameter limits from the TS to the COLR does not involve a significant hazards consideration finding as defined in 10 CFR 50.92.

4.0 ENVIRONMENTAL ANALYSIS DEP has concluded that the proposed amendment meets the criteria provided by 10 CFR 51.22(c)(9) for categorical exclusion from the requirement for an Environmental Impact Statement. The proposed amendment does not involve a significant hazards consideration, an increase in the types and amounts of effluents that may be released offsite, or result in an increase in individual or cumulative occupational radiation exposures.

U. S. Nuclear Regulatory Commission to Serial: RNP-RA/15-0016 11 Pages (including this cover page)

Proposed Technical Specification Changes

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest cold leg temperature, and pressurizer ressure shall not exceed the Sbs specified in FigijFB 2.1 .1 1. ( Add INSERT # i here 2.1.2 RCSPressure ~

In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained s 2735 psig.

2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

HBRSEP Unit No. 2 2.0-1 Amendment No. tte-

the COLR; and the following Sls shall not be exceeded:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained?_ l .141 for the HTP correlation and ?_l.17 for the XNB correlation.

2.1.1.2 The peak fuel centerline temperature shall be maintained< ((2790-17.9 x P - 3.2 x B) x 1.8 +

32] °F where P is the maximum weight percent of Gadolinia (%) and B is the maximum pin burnup (GWD/MTU).

SLs 2.0 DO NOT OPERATE L&J IN THIS AREA 0:::

~

L&J 0...

c

~

(.!' 600 LL.I

_J Cl

_J 580 0

u I-V> 560 LL.I

c

~

c 540 V>

u 0:::

HIGH FLUX TA 520 AT 118 'II. OF RATED POWER__....,...

500 0 20 40 60 80 PERCENT OF RATED THERMAL POWER 6

NOTE: BASED ON A MINIMUM RCS FLOW OF 97.3 x 10 lbm/hr Figure 2.1.1-1 (page 1 of 1)

Reactor Core Safety Limits DELETED Figure 2.1.1-1 HBRSEP Unit No. 2 2.0-2 Amendment No. +96-

RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 7)

Reactor Protection System Instrumentation Note 1: Overtemperature b.T The Overtemperature b.T Function Allowable Value shall not exceed the following Nominal Trip Setpoint by more than 2.96% of b.T span.

b.T setpoint Sb.T 0 { K1 - K2 (1 +-r1 S) I (1 +-r 2 S)} (T - T) + K3(P - P) - f(b.I)}

Where: b.To is the indicated lff at RTP, °F.

s is the Laplace transform operator, sec-1 . __.--- [ 1 Tis the measured RCS average temp~°F .

0 T is the reference Tavg at RTP, s 57&.-9°F.

[*)

P is the measured pressurizer pressure, psig /

p' is the nominal RCS operating pressure, <:!: ~ psig . __..,.. [ * )

~[*] k:

K1 S1.1269 -iE-- [* ] K2 =0. 01 ~~ [.] K3 =0.0008Q/psig [.]

-r 1~ sec -r2 s&GS -se-c [* 1 / [.

1

[*J -----:-:.. , [*1 \ It .I ,V f(b.I) = 24!¥e{(qb - qt) - ~ when ~ - qb < ~ RTP it-'

[*] ~ 0%ofRTP when ~ RTPSq t -qb ~ RTP v

~(qt - qb) - ~ when q, - qb > ~.J:ff P i

[*]~ [*] ~ [*]

Where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt+ qb is the total THERMAL POWER in percent RTP.

The values denoted with [ * ] are specified in the COLR.

HBRSEP Unit No. 2 3.3-18 Amendment No. ~

RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 7 of 7)

Reactor Protection System Instrumentation

(

Note 2* Overpower AT The Overpower AT Function Allowable Value shall not exceed the following Nominal Trip Setpoint by more than 3.17t of AT span.

l::iTwpoi111Sl::iTo{KrK,[ -r3 S ]r- K6(T-T')-f(l11)}

1+ -r3 S Where: ATo is the indicated AT at RTP, °F.

s is the Laplace transform operator. sec-1

T is the measured RCS average temperature, °F.

TI ; s the reference Tavg at RTP, s 515-:-9~ [ ] [* 1

~ [* ]

IL' K, s i-:-96 Ks ~ 9-:-921°F for increasing Tavv Ks ~ 9.992771°F when T > T'

(* J,71 r* J 7 &l°F for decreasing Tavv f*l -7 &l°F when T s T'

-r3 ::!!:, g. sec

~[*]

<<AH

a5 defined in Note ~ fer O't'ertem~rature AT 0

HBRSEP Unit No. 2 3.3-19 Amendment No. i9&

INSERT #3 0% of RTP when [ * ] RTP ~qt - qb ~ [ * ] RTP Where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt+ qb is the total THERMAL POWER in percent RTP.

The values denoted with [ * ] are specified in the COLR.

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be .WtlRIA~i-ffl:Rf&.!rs.t:>ecJffe~-..--

below:

a.

b. RCS average temperature s; 67Q.4 ~

c. RCS total ftow rate ~97.3x10' lbm/hr\ ,r-...("""'l,("""'I,"'~~~.-....-...._

APPLICABILITY: MODE 1.

N 0 TE---------------------------------------------

Pressu rizer pressure limit does not apply during:

a. THERMAL POWER ramp > 5% RTP per minute; or

b. THERMAL POWER step> 10% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more RCS DNB A.1 Restore RCS DNB 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months parameters not within parameter(s) to within limits. limit.

B. Required Action and B.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

HBRSEP Unit No. 2 3.4-1 Amendment No. -4*-

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is ::<! 2209 13sig. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.1.2 Verify RCS average temperature is s: 57Q. 4°i;. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.1.3 Verify RCS total flow rate is~ 97.3 x 106 lbm/hr.

SR 3.4.1.4 - - - - - - - - N O T E - - - - - - - - - - -"

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after C::90% RTP.

Verify by precision heat balance that RCS total flow 18 months 6

rate is;:: 97.3 x 10 lbm/hr.

less than or equal to the limit specified in the COLR.

HBRSEP Unit No. 2 3.4-2 Amendment No. tT6'-

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

7. Axial Flux Difference (AFD) limits for Specification 3.2.3; and Boron Concentration limit for Specification 3.9.1.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. The approved version shall be identified in the COLR. These methods are those specifically described in the following documents:

1. Deleted

2. XN-NF-84-73(P), "Exxon Nuclear Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," approved version as specified in the COLR.

3. XN-NF-82-21 (A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations,"

approved version as specified in the COLR.

4. Deleted ~

5. XN-75-32(A), "Computational Procedure for Evaluating Rod Bow,"

approved version as specified in the COLR.

6.

7.

Deleted.

Deleted

,{

8. XN-NF-78-44(A), "Generic Control Rod Ejection Analysis," approved version as specified in the COLR.

9. XN-NF-621{A), "XNB Critical Heat Flux Correlation," approved version as specified in the COLR.

10. Deleted ~

11 . XN-NF-82-06(A). "Qualification of Exxon Nuclear Fuel for Extended Burnup," approved version as specified in the COLR.

12. Deleted

13. Deleted.

(continued)

HBRSEP Unit No. 2 5.0-25 Amendment No .~

INSERT#4

9. Reactor Core Safety Limits Figure for Specification 2.1.1

10. Overtemperature tff and Overpower Lff setpoint parameter values for Specification 3.3.1

11. Reactor Coolant System pressure, temperature and flow Departure from Nucleate Boiling (DNB) limits for Specification 3.4.1

U. S. Nuclear Regulatory Commission to Serial: RNP-RA/15-0016 9 Pages (including this cover page)

Revised Technical Specification Pages

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest cold leg temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained

~ 1.141 for the HTP correlation and ~ 1.17 for the XNB correlation.

2.1.1.2 The peak fuel centerline temperature shall be maintained < [(2790 -

17.9 x P-3.2 x B) x 1.8 + 32] °F where Pis the maximum weight percent of Gadolinia (%)and Bis the maximum pin burnup (GWD/MTU).

2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained ~ 2735 psig.

2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

HBRSEP Unit No. 2 2.0-1 Amendment No.

Sls 2.0 DELETED Figure 2.1.1-1 HBRSEP Unit No. 2 2.0-2 Amendment No.

RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 7)

Reactor Protection System Instrumentation Note 1: Overtemperature IiT The Overtemperature IiT Function Allowable Value shall not exceed the following Nominal Trip Setpoint by more than 2.96% of /iT span.

0 fiTsetpoint S/iTo { K1 - K2 [ (1 +i:1 S) I (1 +i:2 S)] (T-T°) + KJ(P- P )-f(/il)}

Where: /1T0 is the indicated /iT at RTP, °F.

1 s is the Laplace transform operator, sec- .

Tis the measured RCS average temperature, °F.

0 T is the reference Tavg at RTP, s [ * ]°F.

P is the measured pressurizer pressure, psig p' is the nominal RCS operating pressure, ;:: [ * ] psig K1 s [ *] K2 =[ * ]/°F K3 =[ * ]/psig i: 1 ;:: [ *] sec i: 2 s [*]sec f(/11) = [

1{(qb - q,) - [ * ]} when q, - qb < [ * ] RTP 0% of RTP when [ * ] RTP s qt - qb s [ * ] RTP

[

1{( q, - qb) - [ * ]} when q, - qb > [ * ] RTP Where q, and qb are percent RTP in the upper and lower halves of the core, respectively, and q, + qb is the total THERMAL POWER in percent RTP.

The values denoted with [ * ] are specified in the COLR.

HBRSEP Unit No. 2 3.3-18 Amendment No.

RPS Instrumentation 3.3.1 Table 3.3.1-1(page7of7)

Reactor Protection System Instrumentation Note 2: Overpower t::.T The Overpower l1T Function Allowable Value shall not exceed the following Nominal Trip Setpoint by more than 3.17% of f1T span.

f1 Tsetpoint st::. Ta { K4 - Ks [-c3 SI (1 +-c 3 S)] T - KB(T - T) - f(l::.I)}

Where: l1T0 is the indicated f1T at RTP, °F.

1 s is the Laplace transform operator, sec- .

Tis the measured RCS average temperature, °F.

T' is the reference T avg at RTP, s [ * ] °F.

l<.tS[*] K5 ;;::: [ * ]/°F for increasing Tavg Ks;::: [ * ]/°F when T > T'

[ * ]/°F for decreasing T avg [ * ]/°F when T s T'

't3 ;;::: [ *] sec f(f11) = [ * ] {(qb - q,) - [ * ]} when q, - qb < [ * ] RTP 0% of RTP when [ * ] RTP s q1- qb s [ * ] RTP

[ * ] {(q, - qb) - [ * ]} when % - qb > [ * ] RTP Where q1and qb are percent RTP in the upper and lower halves of the core, respectively, and q1+ qb is the total THERMAL POWER in percent RTP.

The values denoted with [ * ] are specified in the COLR.

HBRSEP Unit No. 2 3.3-19 Amendment No.

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below:

a. Pressurizer pressure is greater than or equal to the limit specified in the COLR;

b. RCS average temperature is less than or equal to the limit specified in the COLR; and 6

c. RCS total flow rate 2: 97 .3 x 10 lbm/hr and greater than or equal to the limit specified in the COLR.

APPLICABILITY: MODE 1.

NOTE----------------------------------------------

Pressurizer pressure limit does not apply during:

a. THERMAL POWER ramp > 5% RTP per minute; or

b. THERMAL POWER step> 10% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more RCS DNB A.1 Restore RCS DNB 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months parameters not within parameter(s) to within limits. limit.

B. Required Action and B.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

HBRSEP Unit No. 2 3.4-1 Amendment No.

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1 .1 Verify pressurizer pressure is greater than or equal to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months the limit specified in the COLR.

SR 3.4.1 .2 Verify RCS average temperature is less than or equal 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to the limit specified in the COLR.

6 SR 3.4.1.3 Verify RCS total flow rate is;:: 97.3 x 10 lbm/hr and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months greater than or equal to the limit specified in the COLR.

SR 3.4.1.4 ---------------------------NOTE---------------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after

90% RTP.

Verify by precision heat balance that RCS total flow 18 months 6

rate is;:: 97.3 x 10 lbm/hr and greater than or equal to the limit specified in the COLR.

HBRSEP Unit No. 2 3.4-2 Amendment No.

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

7. Axial Flux Difference (AFD) limits for Specification 3.2.3; and

8. Boron Concentration limit for Specification 3.9.1 .

9. Reactor Core Safety Limits Figure for Specification 2.1.1

10. Overtemperature !:!,.T and Overpower!:!,.T setpoint parameter values for Specification 3.3.1

11. Reactor Coolant System pressure, temperature and flow Departure from Nucleate Boiling (DNB) limits for Specification 3.4.1

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. The approved version shall be identified in the COLR. These methods are those specifically described in the following documents:

1. Deleted

2. XN-NF-84-73(P), "Exxon Nuclear Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," approved version as specified in the COLR.

3. XN-NF-82-21 (A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations,"

approved version as specified in the COLR.

4. Deleted

5. XN-75-32(A), "Computational Procedure for Evaluating Rod Bow,"

approved version as specified in the COLR.

6. Deleted.

7. Deleted

8. XN-NF-78-44(A), "Generic Control Rod Ejection Analysis. approved version as specified in the COLR.

9. XN-NF-621(A}, "XNB Critical Heat Flux Correlation," approved version as specified in the COLR.

10. Deleted (continued)

HBRSEP Unit No. 2 5.0-25 Amendment No.

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements (continued)

11. XN-NF-82-06(A), "Qualification of Exxon Nuclear Fuel for Extended Burnup," approved version as specified in the COLR.

12. Deleted

13. Deleted.

(continued)

HBRSEP Unit No. 2 5.0-25a Amendment No.

U. S. Nuclear Regulatory Commission to Serial: RNP-RA/15-0016 9 Pages (including this cover page)

Proposed Technical Specification Bases Changes

Reactor Core SLs B 2.1.1 BASES ~

~

APPLICABLE predict the ~-ffttx and the location of DNB for axially SAFETY ANALYSES uniform and non-uniform heat flux distributions. The local c tinued) -?? QN8 heat flux ratio, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB. The minimum DNB ratio, or DNBR, during normal operational and anticipated transients, is restricted to the safety limit. A DNBR at the safety limit corresponds to a 95% probability, at a 95%

confidence level, that DNB will not occur, and is chosen as an appropriate margin to DNB for all operating conditions. The DNBR safety limit is a conservative design value which is used as a basis for setting core safety limits. Based on rod bundle tests, no fuel damage is expected at this DNBR or greater. For the standard mixing vane fuel, the Siemens Power Corporation XNB correlation has a DNBR safety limit of 1.17 (Ref. 2) and for the high thermal performance fuel the Siemens HTP correlation has a DNBR safety limit of 1.141 (Ref. 3). The safe~ liFRit Gl:IF\'es pre11iaea iA Fig1:1re a.1.1 1 remaiR \'alia 1:1siAf:J U:ie SieFRens MTP saFFelatiaR.

1 The Reactor Trip System setpoints specified in Limiting Condition for Operations (LCO) 3.3.1, in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System (RCS) temperature, pressurizer pressure, flow, core power distribution, and THERMAL POWER level that would result in a departure from nucleate boiling ratio (DNBR) of less than the DNBR limit and preclude the existence of flow instabilities.

~

/

r Automatic enforcement of these reactor core SLs is provided by the fellewiAg f1:1Aetians:

The fuel centerline temperature limit is -i3. OveFteFA~erattJre 6T trip, appropriate operation of the RPS and the main steam safety valves a function of weight percent of Gadolinia b. Ova1 power lff b ip, and pin burnup as presented in 6. Pewer Range ~'el:ltren Fl1:u< trip; enel Reference 5 and

d. Main steam safety vah1ss.

approved for use at RNP per Reference M!il"ltaining the D~48R abe't'E: the limit eAstires that tl=~e average enthalpy

6. in the het leg is less than er eei1:1al te tl::ie entl::ialpy of eahnatea li~1:1ia ana alse ensttFes tl=iat the AT (continued)

HBRSEP Unit No. 2 B 2.0-3 Revision No. ~

AA'leAdFFlent Ne. 178

Reactor Core SLs B 2.1.1 BASES APPLICABLE meest1Fed ey iAStftlffieAtetieA, 1:1sed iA tl:ie RPS desigA as a SAFETY ANALYSES R=teas1:1Fe ef eere pewer, is profJortieAal to eore power.

(continued)

The SLs represent a design requirement for establishing the RPS trip setpoints identified previously. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) safety limits Limits," or the assumed initial conditions of the safety analyses (as figure indicated in the Updated Final Safety Analysis Report (UFSAR), Ref. 4) provide more restrictive limits to ensure that the SL a e xceeded.

SAFETY LIMITS The G1:1rves provided in

e o of THERMAL POWER, RCS pressure, and reactor vessel inlet temperature for which the minimum DNBR is not less than the safety analyses limit, that fuel centerline temperature remains below melting, that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, or that the core exit quality is within the limits defined by the DNBR correlation. Fig1:1re 2.1.1 1 st:iows tt:le allewable pe*..,ier le*Jel EieGreasing v.*itl=l ineFOasiAg roaster 'Jessel iAlet teff113ei:at1:1Fe at seleeteEi 13ress1:1ri~er PLACE press1:1res fer eeAstant fie\*,* (i.e., tl=tree lee13 ei;1eratien, minim1:1m flew INSERT ~ ~ lbff'!Jl:ir). Tl=te area where elaa integri~ is ass1:1rea is belew

2 HERE

> tl=lese lines. Tl=le teff'lfJOFat1:1re liff'lits at law fJOwer are eensiaerably ff'!ere Gonserv-ative ~an weula be FO"f l:lirea if ~ey were basoet eA tl=te miniff'lt:JFR allewable 0~18 ratie, but ar:o set te pFeeluae b1:1lk sailing at tl=te vessel e>Eit.

Tl=le safe~ liff'lit Gl:IF\'OS given in Fig1:1Fe 2.1.1 1 aFe fer eenstant flew eenditiens. n1ese s1:1rves webllef net be at>rlisable in eases *M:lere tetal reaoter sealant flew is less tl=lan 97.3 K ~ leff'l/Rr. Tl=le evalblatien ef SYGR an event webllet t:ie baseet 1:113en the analysis 13resented iA Seetien 15.a ef the yi;::sAR Tt:ie SL is l=ligher than the limit eale1:1latea wl=len the N<ial Flbl>E OiUeFense (Ali'D) is within tt:ie limits ef the F 1{Al) f.l:inetieA of tl=le evertem13eratYre AT reaster tri13. When tl=lo AFD is not 'Nitl=lin tl=le teleranse, tl=lo AFO oftest on tl=le e*,*ertemperatl:1Fe anet everpeweF AT roaster tFifJS will reablse tl=lo setpeints ta 13re*;iae 13reteetien eensistent *.*o1itl=I tl=le roaster sere Sls (Ref. 4) .

(continued)

HBRSEP Unit No. 2 B 2.0-4 Revision No.~

Amendment No. 178

INSERT #2 The reactor core Sls are established to preclude violation of the following fuel design criteria :

a.. There must be at least a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB; and

b. There must be at least a 95% probability at a 95% confidence level that the hot fuel pellet in the core does not experience centerline fuel melting.

The reactor core Sls are used to define the various RPS functions such that the above criteria are satisfied during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs}. To ensure that the RPS precludes the violation of the above criteria, additional criteria are applied to the Overtemperature and Overpower Lff reactor trip functions. That is, it must be demonstrated that the average enthalpy in the hot leg is less than or equal to the saturation enthalpy and the core exit quality is within the limits defined by the DNBR correlation. Appropriate functioning of the RPS ensures that for variations in the THERMAL POWER, RCS Pressure, RCS average temperature, RCS flow rate, and 61 that the reactor core Sls will be satisfied during steady state operation, normal

Reactor Core SLs B 2.1 .1 BASES APPLICABILITY SL 2.1.1 only applies in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The main steam safety valves and automatic protection actions serve to prevent RCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the unit into MODE 3. Setpoints for the reactor trip functions are specified in LCO 3.3.1, "Reactor Protection System (RPS) Instrumentation." In MODES 3, 4, 5, and 6, Applicability is not required since the reactor is not generating significant THERMAL POWER.

SAFETY LIMIT If SL 2.1.1 is violated, the requirement to restore VIOLATIONS compliance and go to MODE 3 places the unit in a safe condition and in a MODE in which this SL is not applicable.

The allowed Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable, and reduces the probability of fuel damage.

REFERENCES 1. 10 CFR 50, Proposed Appendix A, 32FR10213, July 11, 1967.

2. XN-NF-621 (P)(A) Revision 1, "Exxon Nuclear DNB Correlation PWR Fuel Designs," Exxon Nuclear Company, September 1983.

3. EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel."

4. UFSAR, Sections 3.1, 4.4, 7.2, and 15.0.

5. XN-NF-79-56(P)(A) Revision 1, "Gadolinia Fuel Properties for LWR Safety Evaluation.

6. XN-NF-85-92(P)(A) , "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results.

HBRSEP Unit No. 2 B 2.0-5 Revision No. Q, 3 Amendment No. 178

RPS Instrumentation B 3.3.1 BASES APPLICABLE5. Overtemperature )T (continued)

SAFETY ANALYSES, LCO, axial power distribution - f(til}, the Trip and APPLICABILITY Setpoint is varied to account for imbalances in the axial power distribution as detected by the NIS upper and lower power range detectors. If axial peaks are greater than the design limit, as indicated by the difference between the upper and lower NIS power range detectors, the Trip Setpoint is reduced in accordance with Note 1 of Table 3.3.1-1 .

Dynamic compensation is included for system piping delays from the core to the temperature measurement system and RTD response time.

The Overtemperature 8 T trip Function is calculated for each loop as described in Note 1 of Table 3.3.1-1 . Trip occurs if Overtemperature tiT is indicated in two loops. The function (1 +r1s)/(1 + r 2s); is generated by the lead-lag controller for Tavg dynamic compensation and f(til) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument res onse durin lant startu test ~

The shape of the every % that (Efo et~ eMGeeds 17%, the O*.tertempei:at1:1re AT f{t.I)penalty is setpoint is red1:1oed by 2.4% and for every °lo that (E11 etiJ e>eoeeds described in the 12%, tt:ie 01Jertm~:iperat1:1re AT setpoiRt is red1:1see by 2.4%. Note Core Operating that this Function also provides a signal to generate a turbine Limits Report (COLR) . runback prior to reaching the Trip Setpoint. A turbine runback will reduce turbine power and reactor power. A reduction in power will normally alleviate the Overtemperature lff condition and may prevent a reactor trip.

The LCO requires all three channels of the Overtemperature tiT trip Function to be OPERABLE. Note that the Overtemperature 6 T Function receives input from channels shared with other RPS Functions. Failures that affect multiple Functions require entry into the Conditions applicable to all affected Functions.

In MODE 1 or 2, the Overtemperature 8 T trip must be OPERABLE to prevent DNB. In MODE 3, 4, 5, or 6, this (continued)

HBRSEP Unit No. 2 B 3.3-14 Revision No . .Q-

The shape of the RPS Instrumentation f (lU)penalty is B 3.3.1 described in the Core Operating BASES Limits Report (COLR) .

APPLICABLE 6. Overpower LH (continued)

SAFETY ANALYSES, LCO, constant utilized in the rate-lag controller for Tavg*

and APPLICABILITY Note that this Function also provides a signal to generate a turbine runback prior to reaching the Allowable Value. A turbine runback will reduce turbine power and reactor power. A reduction in power will normally alleviate the Overpower !>.T condition and may prevent a reactor trip.

The LCO requires three channels of the Overpower !>.T trip Function to be OPERABLE. Note that the Overpower D.T trip Function receives input from channels shared with other RPS Functions. Failures that affect multiple Functions require entry into the Conditions applicable to all affected Functions.

In MODE 1 or 2, the Overpower D.T trip Function must be OPERABLE. These are the only times that enough heat is generated in the fuel to be concerned about the heat generation rates and overheating of the fuel. In MODE 3, 4, 5, or 6, this trip Function does not have to be OPERABLE because the reactor is not operating and there is insufficient heat production to be concerned about fuel overheating and fuel damage.

7. Pressurizer Pressure The same sensors provide input to the Pressurizer Pressure - High and - Low trips and the Overtemperature

[ff trip.

a. Pressurizer Pressure - Low The Pressurizer Pressure - Low trip Function ensures that protection is provided against violating the DNBR limit due to low pressure.

The LCO requires three channels of Pressurizer Pressure - Low to be OPERABLE.

In MODE 1, when DNB is a major concern, the Pressurizer Pressure - Low trip must be OPERABLE. This trip Function is automatically enabled on increasing power by the P-7 interlock (NIS power range P-10 or turbine impulse pressure greater (continued}

HBRSEP Unit No. 2 B 3.3-16 Revision No.-&

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABLE result in meeting the DNBR criterion ef ~ 1.17 fer tl:le SAFETY ANALYSES StaRelare Mi><iRg VaRe ~el , BREI > 1.141 fer tf:le ~igA Tt:leFFRal .r-(continued) PerfeFff:laRse t.:lel (~ef. 2) . This is the acceptance limit for the RCS DNB parameters. Changes to the unit that could impact these parameters must be assessed for their impact on the DNBR criteria. The transients analyzed for include loss of coolant flow events and dropped or stuck rod events. A key assumption for the analysis of these events is that the core power distribution is within the limits of LCO 3.1.6, "Control Bank Insertion Limits"; LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)"; and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)."

The RCS DNB parameters satisfy Criterion 2 of the NRC Policy Statement.

LCO This LCO specifies limits on the monitored process variables - pressurizer pressure, RCS average temperature, and RCS total flow rate - to ensure the core operates within the limits assumed in the safety analyse Operating within these limits will result in meeting the DNBR crite

n in the event of a DNB limited transient.

sed on tibrate the

~rRN'ftllr~S?!Wure, and flow rate are een adjusted for APP ICABILITY In MODE 1, the limits on pressuri s re, RCS coolant average temperature, and RCS flow rate must be maintained during steady sta operation in order to ensure DNBR criteria will be met in the event of a unplanned toss of forced coolant flow or other DNB limited transient. In all The variables are contained in the COLR to provide operating and analysis flexibility from cycle to cycle. However, the minimum RCS flow is retained in the TS LCO. (continued)

HBRSEP Unit No. 2 B 3.4-2 Revision No. -9; AmeAdment No. 176

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABILITY other MODES, the power level is low enough that DNB is not a (continued) concern.

A Note has been added to indicate the limit on pressurizer pressure is not applicable during short term operational transients such as a THERMAL POWER ramp increase > 5% RTP per minute or a THERMAL POWER step increase> 10% RTP. These conditions represent short term perturbations where actions to control pressure variations might be counterproductive. Also, since they represent transients initiated from power levels < 100% RTP, an increased DNBR margin exists to offset the temporary pressure variations.

AnoU:ier set of liFRits on ONB relate~ paraFReters is provided in SL 2.1 .1, "Reactor Core SLs." Those liFRits re less restrictive than the limits of this LCO, but violation of a Safety um* SL) merits a stricter, more severe Required Action. Should a violation f this LCO occur, the operator must check whether or n tan SL ma hav been exceeded .

The conditions which define the DNBR limit ACTIONS RCS pressure and RCS average temperature are controllable and measurable parameters. With one or both of these parameters not within LCO limits, action must be taken to restore parameter(s).

RCS total flow rate is not a controllable parameter and is not expected to vary during steady state operation . If the indicated RCS total flow rate is below the LCO limit, power must be reduced, as required by Required Action 8.1, to restore DNB margin and eliminate the potential for violation of the accident analysis bounds.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based on plant operating experience.

(continued)

HBRSEP Unit No. 2 B 3.4-3 Revision No.