Text

Proposed Tech Specs,Ensuring That Adequate Shutdown Margin Be Maintained in Reactor at All Times
	ML17300B115
Person / Time
Site:	Palo Verde
Issue date:	12/02/1987
From:	; ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	;
Shared Package
ML17300B113	List: ML17300B112; ML17300B115;
References
	TAC-66784, NUDOCS 8712090034
	Download: ML17300B115 (206)
	v • d • e

87ia090ose 87i202 PDR ADOCN Oaa00529 ~i t

P PDR ATTACHMENT 1 A. DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE UEST The proposed amendment changes the Shutdown Margin versus Cold Leg Temperature curve as set forth in Technical Specification (T.S.) 3.1.1.2. The change is to the Hot Zero Power endpoint. The change is from 6.0$ 5p to 6.5% 5p .

B. PURPOSE OF THE TECHNICAL SPECIFICATION The purpose of Technical Specification 3.1.1.2 is to ensure that an adequate shutdown margin is maintained in the reactor at all times.

C. NEED FOR THE TECHNICAL SPECIFICATION AMENDMENT Due to the design of Cycle 2, the Cycle 2 moderator temperature reactivity insertion is more adverse than Cycle 1 during a postulated steam line break.

Because of the more adverse cooldown reactivity insertion for Cycle 2, the Shutdown Margin is required to be increased from 6% to 6.5S gp at zero power. The increase in margin is required to maintain the operation of Cycle 2 within the safety analysis.

BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability of consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

A discussion of these standards as they relate to the amendment request follows:

Standard 1--Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the proposed change ensures that the analysis of the most limiting accident, the Steam Line Break event for Cycle 2, is bounded by the reference cycle (Cycle 1) transient analysis. Therefore, there is no increase in the probability or consequences of an accident previously evaluated because operation of Cycle 2 is within the realm of operation, as experienced during Cycle l.

Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because, by increasing the required shutdown margin at zero power, the Cycle 2 transient. analysis is b'ounded by the reference cycle transient analysis.

Requiringa larger shutdown margin does not subject the operation of Cycle 2 to any additional accidents. It restricts the'nit even further in its allowed operation. Therefore, there will be no increase in the possibility of a new or different kind of accident occurring.

Standard 3--Involve a significant reduction in a margin of safety.

The proposed change does not"involve a significant reduction in a margin of safety because the shutdown margin at zero power is being increased to ensure the same margin of safety is maintained for Cycle 2 operation as it was for Cycle 1. The increased shutdown margin ensures that the most limiting event is bounded by the reference cycle transient analysis and thus maintaining margin.

2. The proposed amendment matches the guidance concerning the application of standards for determining whether or not a significant hazards consideration exists (51 FR 7751) by example:

For a nuclear power reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found previously acceptable to the NRC for a previous core at the facility in question are involved. This assumes that no significant changes are made to the acceptable criteria for the Technical Specifications, the analytical methods used to demonstrate conformance with the Technical Specifications and regulations are not significantly changed, and that NRC has previously found such methods acceptable.

ED SAFETY EVALUATION FOR THE AMENDMENT RE VEST The proposed Technical Specification amendment will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. The proposed change does not change or replace equipment or components important to safety.

The change ensures that, during the operation of Cycle 2, the Cycle 2 analysis is bounded by the reference cycle transient analysis. Therefore, there is no increase in the probability of occurrence of the consequences of an accident or malfunction of equipment.

The proposed Technical Specification amendment will not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR. The proposed change ensures that, during the operation of Cycle 2, a shutdown margin of the same magnitude as the margin required 2

0 J

1

during Cycle 1 is maintained. By increasing the margin to 6.5S ,the Cycle 2 analysis is bounded by the reference cycle transient analysis and restricts the Unit even further in its allowed operation. Therefore, there is no increase in the possibility for an accident or malfunction being created.

The proposed Technical Specification amendment will not reduce the margin of safety as defined in the basis for the Technical Specifications. The proposed change ensures that during the operation of Cycle 2, the Cycle 2 analysis is bounded by the reference cycle transient analysis and, therefore, there is no reduction in the margin.

ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental question because operation of PVNGS Unit 2, in accordance with this change, would not:

1. Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the staff's testimony to the Atomic Safety and Licensing Board; or

2. Result in a significant change in effluents or power levels; or

3. Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.

c. MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES Limiting Conditions For Operation And Surveillance Requirements:

t 3/4 1-2a

~

I

'

(500,6.0 6,' I I I I

I I

I I I I I l I I I I I 1 I I I I I I I I I

r I I I

I I II I

I I

5,'EGION I

I I

1 I OF I I

ACCT TABLE--'-----

I I r I I

I LLI I C5 OPERQT ION I I

I I I I C3 (350,3.5)

CL I

3 I I

I I I I I

REGION OF I

I I

I

'UNACCEPTABLE -',

-;

I I OPERATION 2' I I

I I

I I I I

L I I L J I I I I I I I I I I I

I I I I I I I 0'I I I

100 200 300 400 500 600 COLD LEG TEMPERATURE ('F)

FIGURE 3.1- IA SHUTDOWN MARGIN VERSUS COLD LEG TEMPERATURE PALO VERDE UNIT P 3/0 I-2A

e 0

Oi

6EGION I I

I OF

'-------'

I I ACCEP TABLE-I I

I I

OPERATION (350 3 5)--'----

I I

REGION OF I

I

',UNACCEPTABLE I

OPERATION 2'

I I

rI I I I

I I

I J I

I I I

I 0' I 0 lo'0 200 300 ~00 500 600 COLD LEG T":i~iIP RA!URE (:F)

F!GURE 3.I - IA SHUTDOWN ivlARGIN VERSUS COLD LEG TEiAPERATURE PALO VERDE UNIT2. 3/4 I-2A

0 l

~

ATTACHMENT 2 A. DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE UEST The proposed amendment changes the Moderator Temperature Coefficient (MTC)

Figure 3.3-1 as set forth in Technical Specification (T.S.) 3.1.1.3. The changes are two fold. The operating bounds of the MTC are being broadened to accommodate the operation of Cycle 2 and the x axis is being changed to core power level instead of average moderator temperature.

B. PURPOSE OF THE TECHNICAL SPECIFICATION T.S, 3.1.1.3 ensures that the assumptions used in the accident and transient analysis remain valid through each fuel cycle.

C. EED FOR THE TECHNICAL SPECIFICATION AMENDMENT In preparation for future 18 months cycles, the Cycle 2 core physics is such that, a change in the MTC operating band will occur. To accommodate operation throughout Cycle 2, the MTC operating band has become more positive because of the increase in fuel enrichment which requires higher boron concentration at beginning of the cycle. As operation into the cycle proceeds, the MTC will become more negative. In addition, the x axis is to be changed to core power level instead of average moderator temperature. By changing the x axis to core power level, the method of calculating the bounding MTC for the most limiting case becomes simplified. Making the MTC a dependent variable of core power only and not of inlet temperature and core power, as the present curve represents, the calculation of the limiting MTC need only be performed once.

The present method of manipulating MTC requires performing the analyses several times at various average moderator temperatures to be sure of obtaining the most limiting case but, with the new method, MTC can be calculated once and there is assurance that the most limiting case value is obtained. Both graphs are the results of the same set of codes, only the method of manipulating the data is slightly different.

D. BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION

1. The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92 A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

~,

A discussion of these standards as they relate to the amendment request follows:

Standard 1--Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the consequences of any accident, when the unit is operated in the calculated band of the Cycle 2 MTC, is bounded by the reference Cycle (Cycle 1) transient analysis. Therefore, there is no possibility of an accident previously evaluated being increased.

Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated. The results of the analysis performed for Cycle 2, using the proposed MTC band, assures that there will be sufficient margin for the most limiting DBE. By operating within these limits, operation of Cycle 2 will not create any situation where a new or different kind of accident could occur because Cycle 2 analysis results show that Cycle 2 is bounded by the reference cycle analysis.

Standard 3--Involve a significant reduction in a margin of safety.

The proposed change does not involve a significant reduction in a margin of safety because the results for all DBEs affected by the new MTC are bounded by the reference analysis. Therefore, the margin of safety does not change.

2. The proposed amendment matches the guidance concerning the application of standards for determining whether or not a significant hazards consideration exists (51 FR 7751) by example:

(iii) ~ For a- nuclear power reactor,,a change resulting from a nuclear reactor core reloading,'f no fuel' assemblies significantly different from those found previously acceptable to the NRC for a previous core at the facility in question are involved. This assumes that no significant changes are made to the acceptable criteria for the technical specifications, the analytical methods used to demonstrate conformance with the technical specifications and regulations are not significantly changed, and that NRC has previously found such methods acceptable.

SAFETY EVALUATION FOR THE AMENDMENT RE UEST The proposed Technical Specification amendment will not increase the probability of occurrence of the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. The proposed

change does not change or replace equipment or components important to safety.

The proposed change is still bounded by the reference cycle tran'sient analysis and, therefore, the probability of any accident previously evaluated has not changed.

The proposed Technical Specification amendment will not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR. The results of the analysis performed for Cycle 2, using the MTC band as stated in Fig 3.3-1, assure that there is sufficient margin for the most limiting Design Basis Event (DBE). By operating within these limits, operation of Cycle 2 will not create any situation where a new or different kind of accident could occur because Cycle 2 analysis results show that Cycle 2 is bounded by the reference cycle analysis.

The proposed Technical Specification amendment will not reduce the margin of safety as defined in the basis for the technical specifications. The results for all DBEs affected by the new MTC are bounded by the reference analysis.

ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental question because operation of PVNGS Unit 2, in accordance with this change, would not:

Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the staff's testimony to the Atomic Safety and Licensing Board; or I

2. Result in a significant, change in effluents or power levels; or

3. Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.

I MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES Limiting Condition for Operation and Surveillance Requirements:

3/4 1-5

0 4

O o 40.22

+0.22x ~hp/ F (5960F, 0.0hp X

I-Z LU O ALLOWABLE QC Q~

ua LU -1.0 LU U)

CL: <4 D CQ

~Z g

K LU LU -2.0 K

0 LU a TC+ T 0 -3.0 TAVG =

594oF -3.0 x 10 4 hp F) 4.0 8 5ooo r

55o'VERAGE MODERATOR TEMPERATURE, F 6oo'IGURE 3.1-1 ALLOMABLE HTC NODES 1 AND 2 PALO VERDE UNIT 2 CYCLE 1

0 FlGURE 5.l-l 2 MlC MODES l AND ALLOWABLE 2 VERDE UNlT< CYCLE PALO (0/,0. 5) 0.5 (l00/,0.0) 0 U

l~

Ld U

U Ld C) MTC ALLO'IIIABLE

-2.5

~0/,-2.8) 000%,-3.5)

.

loo 80 60 'A&tlll

- 5.5 fpp fgt Qgg A I 20 p p

~,

ATTACHMENT 3 A. DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE UEST The proposed,'mendment 'changes the ', operational pressure band of the pressurizer, as set 'forth in Technical, Specification '(T.S.) '3.2.8 to a tighter operational band. The band is being changed from 1815 psia thru 2370 psia to 2025 psia thru 2300 psia.

B. PURPOSE OF THE TECHNICAL SPECIFICATION T.S. 3.2.8 ensures that the actual value of pies'suriz'er pressure is maintained within the range of values'sed in the safety analyses.

C. NEED FOR THE TECHNICAL SPECIFICATION AMENDMENT To support the Core Protection Calculator (CPC) Improvement Program, the operational pressure band of the pressurizer requires tightening. Potential transients initiated at the extremes of the Cycle 1 pressure range were not analyzed for Cycle 2. Because the calculations were not performed, the CPCs cannot support normal operation outside of the proposed pressurizer pressure band.

D. BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION

1. The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) Involve a significant reduction in a margin of safety.

A discussion of these standards as they relate to the amendment request follows:

Standard 1--Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the change ensures maintaining the safety margin, as required by the reference cycle (Cycle 1) safety analysis or the safety limits as stated in the FSAR. The change restricts normal operation because there are no supporting calculations and related penalty factors for normal operation outside the specified pressure range. The bounds of the safety analysis have not been changed. Therefore, there will be no increase in the

'possibility or consequences of an accident.

0 l'

II 0

Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because the change ensures that the safety margin as required by the reference cycle safety analysis is maintained. Since the operation band is more restrictive in relation to the safety analysis it can be concluded that there will be no increase in the possibility of a new or different kind of accident.

Standard 3--Involve a significant reduction in a margin of safety.

The proposed change does not involve a significant reduction in a margin of safety because the proposed change ensures maintaining the safety margin as required by the reference cycle safety analysis or the safety limits as stated in the FSAR. By reducing the operation band of the pressurizer, initial conditions during an accident are more restricted but, because the bounds of the safety analysis have not changed, the margin of safety has not been reduced.

2. The proposed amendment matches the guidance concerning the application of standards for determining whether or not a significant hazards consideration exists (51 FR 7751) by example:

(iii) For a nuclear power reactor, a change resulting from a nuclear

'-'reactor core reloading, if no fuel assemblies significantly different from those found previously acceptable to the NRC for a previous core at the facility in question are involved. This assumes that no significant changes are made to the acceptable criteria for the Technical Specifications, the analytical methods used to demonstrate conformance with the Technical Specifications and regulations are not significantly changed, and that NRC has previously found such methods acceptable.

SAFETY EVALUATION FOR THE AMENDMENT RE UEST The proposed Technical Specification amendment will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. The proposed change does not change or replace equipment or components important to safety.

The proposed change ensures that the safety margin as required by the reference cycle safety analysis is maintained. The change restricts normal operation because there are no supporting calculations and related penalty factors for normal operation outside the specified pressure range. The bounds of the safety analysis have not been changed.

The proposed Technical Specification amendment will not create the possibility for an accident or malfunction of a different type than any previously

0 0

evaluated in the FSAR. The proposed change ensures that the safety margin as required by the reference cycle safety analysis is maintained. Since the operation band is more restrictive in relation to the safety analysis, it can be concluded that there will be no increase in the possibility of a new or different kind of accident.

The proposed Technical Specification amendment will not reduce the margin of safety as defined in the basis for the Technical Specifications. The proposed change ensures that the safety margin as required by the reference cycle safety analysis is maintained. By reducing the operation band of the pressurizer, initial conditions during an accident are more restricted but, because the bounds of the safety analysis have not changed, the margin of safety has not been reduced.

F. ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental question, because operation of PVNGS Unit 2, in accordance with this change would not:

Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the staff's testimony to the Atomic Safety and Licensing Board; or

2. Result in a significant change in effluents or power levels; or

3. Result in matters not'reviously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.

,G. MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES Limiting Conditions For Operation And Surveillance Requirements:

3/4 2-12

0 Ir

CONTROLLED BY USER POWER 0 I STRIB ION LIMITS i 3/4.2.8 PRESSURIZER PRESSURE LIMITING CONOITION FOR OPERATION ZB~ ~

3.2.8 The psia.

pressurizer pressure shall be maintained between ~

RES psia and

'PPLICABILITY:

MODES 1 and 2".

ACTION:

With the pressurizer pressure outside its above limits, restore the pressure to within its limit wfthin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.8 The pressurizer pressure shall be determined to be within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

"See Special Test Exception 3.l0.5 PALO YEROE " UNIT 2 3i4 2-I.2 gOgyROLLED BY USER

0'

~ I

ATTACHMENT 4 A. DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE UEST The proposed amendment modifies, the, CEA position Technical Specifications (T.S.) 3.1.3.1 and 3.1.3.2 by removing direct. references of the control of insertion of the Part-length Control Element Assemblies (PLCEA) and creates an additional T.S. ;,that addresses the length of time for insertion and the

,

insertion limit of the PLCEA specifically.

B. PURPOSE OF THE TECHNICAL SPECIFICATION 1 bt t The purpose of T.S 3.1.3.1 and 3.1.3.2 is to,'nsure that (1) 'acceptable power distribution limits are maintained, (2) the minimum shutdown margin is maintained, and (3) the potential effects of CEA misalignments are limited to acceptable levels.

C. EED FOR TECHNICAL SPECIFICATION AMENDMENT Creating a separate T.S. for addressing operation of the PLCEA would provide an improvement to the potential consequences of a PLCEA drop or slip initiated from an allowable inserted position. It would also add a more explicit Limiting Condition for Operation to clarify the allowable duration for the PLCEA to remain within the defined ranges of axial position.

BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION

1. The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

A discussion of these standards as they relate to the amendment request follows:

Standard 1--Involve a significant increase in the probability

. or consequences of an accident previously evaluated.

The proposed change does not involve a, significant increase in the probability or consequences of an accident previously evaluated because the proposed change provides additional assurance that adverse axial shapes and rapid local power changes, which affect radial power peaking

0, P" i i

t f

factors and DNB considerations, do not occur as a result of the part length CEA group being positioned in the same axial segment of fuel assemblies for an extended period of time during operation. Because the proposed change will impose more restrictive limits along with surveillance requirements to ensure adherence with the insertion limits, this proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed change provides additional assurance that adverse axial shapes and rapid local power changes, which affect radial power peaking factors and DNB considerations, do not occur as result of the part-length CEA group being positioned in the same axial segment of fuel assemblies for an extended period of time during operation. Because the proposed change will impose more restrictive limits with respect to previously analyzed events, along with surveillance requirements to ensure adherence with the insertion limits, this proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Standard 3--Involve a significant reduction in a margin of safety.

The proposed change does not involve a significant reduction in a margin of safety because the proposed change provides additional assurance that adverse axial shapes and rapid local power changes, which affect radial power peaking factors and DNB considerations, do not occur as a result of the part-length CEA group being positioned in the same axial segment of fuel assemblies for an extended period of time during operation. Because the proposed change will impose more restrictive limits along with surveillance requirements to ensure adherence with the insertion limits, this proposed change does not involve a significant reduction in the margin of safety.

2. The proposed amendment matches the guidance concerning the application of standards for determining, whether or not a significant hazards consideration exists (51 FR 7751) by example:

(ii) A change constitutes an additional limitation, restriction or control not presently included in the Technical Specifications: for example, a more stringe'nt surveillance requirement.

SAFETY EVALUATION FOR THE AMENDMENT RE UEST i li The proposed Technical Specification amendment- will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. The proposed change does not change or replace equipment or components important to safety.

0 ll

The proposed change provides additional assurance that adverse axial shapes and rapid local power changes, which affect radial power peaking factors and DNB considerations, do not occur as a result of the part-length CEA group being positioned in the same axial segment of fuel assemblies for an extended period of time during operation. Because the proposed change will impose more restrictive limits, along with surveillance requirements to ensure adherence with the insertion limits, this proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

The proposed Technical Specification amendment will not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR. The proposed change provides additional assurance that adverse axial shapes and rapid local power changes, which affect radial power peaking factors and DNB considerations, do not occur as a result of the part-length CEA group being positioned in the same axial segment of fuel assemblies for an extended period of time during operation. Because the proposed change will impose more restrictive limits along with surveillance requirements to ensure adherence with the insertion limits, this proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

The proposed Technical Specification amendment will not reduce the margin of safety as defined in the basis for the technical specifications. The proposed change provides additional assurance that adverse axial shapes and rapid local power changes, which affect radial power peaking factors and DNB considerations, do not occur as a result of the part-length CEA group being positioned in the same axial segment of fuel assemblies for an extended period of time during operation. Because the proposed change will impose more restrictive limits, along with surveillance requirements to ensure adherence with the insertion limits, this proposed change does not involve a significant reduction in a margin of safety.

ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental question because operation of PVNGS Unit 2, in accordance with this change, would not:

1. Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the staff's testimony to the Atomic Safety and Licensing Board; or

2. Result in a significant change in effluents or power levels; or

3. Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.

G. MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES Limiting Conditions For Operation,and Surveillance Requirements:

3/4 1-21 XIX 3/4 1-22 IV 3/4 1-23 3/4, 1-1 3/4 1-24 3/4 1-2 3/4 1-25 3/4 10-2 B 3/4 1-6 . 3/4, 10-4 B 3/4 1-7

0 INDEX LIH!TING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIRENENTS SECTION PAGE 3/4. 0 APPLICABILITY. 3/4 O-l 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - ALL CEAs FULLY INSERTED............. 3/4 1-1.

SHUTDOWN MARGIN - KN

" ANY CEA MITHDRAMN........ 3/4 1-2 1

MODERATOR TEMPERATURE COEFFICIENT. 3/4 1-4 MINIMUM TEMPERATURE FOR CRITICALITY. 3/4 1-6 3/4.1. 2 BORATION SYSTEMS FLOW PATHS - SHUTDOWN.. 3/4 1-7 FLOM PATHS - OPERATING....... 3/4 1-S CHARGING PUMPS - SHUTDOWN. 3/4 1-9 CHARGING PUMPS - OPERATING.......... 3/4 1-10 BORATED MATER SOURCES - SHUTDOMN....;. 3/4 1-11 BORATEO MATER SOURCES - OPERATING.. 3/4 1-13 BORON DILUTION ALARMS. 3/4 1-14 3/4.1. 3 MOVABLE CONTROL ASSEMBLIES CEA POSITION..........,......,.....:.... 3/4 1 POSITION INDICATOR CHANNELS - OPERATING. 3/4 1-25 POSITION INDICATOR CHANNELS - SHUTDOWN. 3/4 1-26 CEA DROP TIthE SHUTDOWN CEA INSERTION LIt1IT 3/4 1-28 REGULATING CEA INSERTION LIHITS...... 3/4 1-29 Putts t e.gqyH ~e.A ~osaarlo~ ue>7s 3/0 t-PALO VERDE - UNIT 2 IV AHEHDHEttT tl0. 13

~,i INDEX LIST OF FIGURES PAGE

= "3.'1" lA SHUTDOWN MARGIN VERSUS COLD LEG TEMPERATURE............ 3/4 1-2a

3. 1-1 ALLOWABLE MTC MODES 1 AND 2 3/4 1-5

=.-3. MINIMUM BORATED WATER VOLUMES................;......... 3/4 1"12 1=2-'.1-2A PART LENGTH CEA INSERTION LIMIT VS THERMAL POWER....... 3/4 1-23

3. 1-28 CORE POWER LIMIT AFTER CEA DEVIATION.......... 3/4 1-24 3% 1 3 CEA INSERTION LIMITS VS THERMAL POWER (COLSS IN SERVICE)...;.'..'.. 3/4 1-31

3. 1-4 CEA INSERTION LIMITS VS THERMAL'POWER (COLSS OUT OF SERVICE)...... 3/4 1-32 3.l .Q PAa.T Lt';VcqH 0aA WS<ezlDQ @<IX MS ~seaka~gO~<a.

>/s i-

3. 2-1 DNBR MARGIN OPERATING LIMIT BASED OH COLSS (COLSS IN SERVICE). 3/4 2-6

3. 2-2 DNBR MARGIN OPERATING LIMIT BASED OH CORE PROTECTION CALCULATOR (COLSS OUT QF SERVICE)........ 3/4 2-7

3. 2-3 REACTOR COOLANT COLD LEG TEMPERATURE VS CORE POWER

.LEVEL. 3/4 2"10 3.3 1 DNBR MARGIN OPERATING LIMIT BASED ON COLSS FOR BOTH CEAC'S INOPERABLE.. . . ...... ....... . 3/4 3-10

3. 4-1 DQSE E(UIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY

> 1.0 pCi/GRAM DOSE E(UIVALEHT I"131 ................. 3/4 4-27

3. 4-2 REACTOR COQLAHT SYSTEM PRESSURE TEMPERATURE LIMITATIONS FOR'0 TO 10 YEARS OF FULL POWER OPERATION. 3/4 4-2e

4. 7-1 SAMPLIHG PLAN FOR SNUBBER FUNCTIONAL TEST 3/4 7-26 B 3/4.4-1 NIL-DUCTILITYTRANSITION TEMPERATURE INCREASE AS A FUNCTIOH OF FAST (E > 1 MeV) HEUTRON FLUENCE (550 F IRRADIATION). 8 3/4 4-10

5. 1-1 SITE AHD EXCLUSIOH BOUNDARIES...................,...... 5-2

5. 1-2 LOW POPULATION ZONE ~ ~ ~ ~ ~ \ ~ ~ ~

5. 1-3 GASEOUS RELEASE POINTS .. 5-4

6. 2-1 OFFSITE ORGANIZATION .. 6-3

6. 2-2 ONSITE ORGANIZATION 6-4 PALO VERDE - UHIT 2 XIX AtlEHDMEHT HO. )3

~ l 1

,:.,-.CONTE<<LE>>Y 3/4. 1.3 MOVABLE CONTROL ASSEMBLIES CEA POSITION LIMITING CONDITION FOR OPERATION

3. 1.3. 1 All full-length (shutdown and regulating) CEAs, and all part-length CEAs which are inserted in the core, shall be OPERABLE with each CEA of a given group positioned within 6.6 inches (indicated position) of all other CEAs in its group.

APPLICABILITY: MODES 1* and 2".

ACTION:

With one or more full-length CEAs inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3. 1. 1.g is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With more than one full-length or part-length CEA inoperable or misaligned from any other CEA in its group by more than 19 inches (indicated position), be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With one or more full-length qr part-length CEAs misaligned from any other CEAs in its group by more than 6.6 inches, operation in MODES 1 and 2 may continue provided that core power is reduced in accordance with Figure 3. 1-2 and that within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the misaligned CcA(s) is eit.her:

Restored to OPERABLE status within its above specified alignment requirements, or

2. Declared inoperable and the SHUTDOWN MARGIN requirement os Specification 3. 1. 1. 15 saiissied. After declaring the CEA(s}

inoperable, operation in MODE5 1 and 2 may continue pursuant to the requirements of Specification~ 3. 1. 3. 6Vprovided:

Qsid 8 le 3 7 a} Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the remainder of the CEAs in the group with the inoperable CEA(s) shall be aligned to within 6.6 inches of the inoperable CEA(s) while maintainino the allowable CEA sequence and insertion limits shown on Figures 3. 1-2A,

3. 1-3 and 3. 1-4; .he THERMAL POW=R level shall be restricted pursuant to Specification~3. 1.3.6~during subsequent operation.

S

~See Special Test Exceptions 3. 10.2 and 3. 10.4.

PALO VFRDE - UNIT 2 3/4 1-21 CONTROLLED BY USER

~ '

j f,

,,,,,,t-ggTPOLLED BY USER ACTION: (Continued) b) The SHUTDOWN MARGIN requirement of Specifica ion 3.1. 1.4 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Otherwise, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

d. With one full-length CEA inoperable due to causes other than addressed by ACTION a., above, but within its above specified align-ment requirements, operation in MODES 1 and 2 may continue pursuant

. to the requirements of Specification 3. 1. 3. 6.

e. With one part-length CEA inoperable and inserted in the core, operation may continue provided the alignment of the inoperable part length CEA is maintained within 6. 6 inches (indicated position) of 0

all other part-length CEAs in its group> a~d the CEA is ~oii Iokcl Pg>span+

W>t par

'to 048 l.eyAi>>eiiienls cS SPeelg>ca t'Io~ 3AI ASAP eng er eyon nser > on >mi ts, xcept for

~

su veillance t ting pursu t to Spe ification 4.1.3."., within hours ther:

e

1. Rest e the part ength CE to withi their mits, or d

2. Re uce THERMA POWER to ess than r equal o that fr tion RATED THE AL POWER hich is lowed b part leng CEA group osition u ng Figur 3. 1-2A.

SURVEILLANCE REQUIREMENTS

4. 1.3. 1. 1 The position of each full-length and part-length CEA shall be determined to be within 6.6 inches (indicated position) of all other CEAs in its group at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s except during time intervals when one CEAC is inoperable or when both CEACs are inoperable, then verify the individual CEA positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4. 1.3. 1.2 Each full-length CLA not fully inserted and each part-length CEA which is inserted in the core shall be determined to be OPERABLE by movement of at least 5 inches in any one direction at least once per 31 days.

PALO VERDE - UNIT 2 3/4 1-22 CQNTRGLLED BY USER

f7 1

%7 090-0.0 ACCEPTABLE 0.70 ERAT ION UNACCEP LE OPE ION O.GO 50K WEB LINE w 050 INSER N LIMIT.

o.no Z

0.30 O.20 0.1 150 140 30 120 110 100 90 80 70 Go 50 40 30 20 10 PART LENGTII CEA POSITION, INCUBI'ES VYITI)DRAWN FIGURE 3.1-2h PhRT LENGTII CER INSERTION LIMIT Vs. IIIERMnL POWER

0 CONTROLLED BY USER I

QJ m O

~ C og C

20 (60 MIN, 2')

b I g b 10 0 (

ILJ ~

I I

I

.J. I I ~,I ~

2 0

~O 0 10 20 0 40 50 60 z TIME AFTER E I TION, MINUTES WHEN CORE POWER I REDUCED TO 56'eOF RATED THERMAL POW RPER "THIS LIMIT URQE, FURTHER RED'TION IS NOT REQUIRED FIGURE 3.~-2g g CORE POWER LIMIT AFTER CEA DEVIATION" PALO VERDE - UNIT 2 3/4 1-24 CONTROLLED BY USER

0 FIGURE 3.I.2A CORE POWER LIMIT AFTER CEA DEVIATION C)

I-~

O~

~o LU~ I I

- I I

I I I I

I I I I I LU

~ 20 (60 MIN, 20%)

O~

Q

~l- ~

IO IJJ

~w O I

I I M0 0 I I

I CL I 20 U 0 10 20 30 40 50 60 z TIME AFTER DEVIATION, MINUTES

+WHEN 'CORE POWER IS REDUCED TO 55% OF RATED THERMAL POWER PER THIS LIMIT CURVE, FURTHER REDUCTION IS NOT REQUIRED FIGURE 3.I-2A CORE POWER LIMIT AFTER CEA DEVIATIONS PALO VERDE - UNIT2.'/4 I-

0 l

l

,,,,pggTgOLLED BY USE~

LIMITING CONDITION FOR OPERATION 3.1.3.2 At least two of the following three CEA position indicator channels shall be OPERABLE for each CEA:

a. CEA Reed Switch Position Transmitter (RSPT 1} with the capability of determining the absolute CEA positions within 5.2 inches,

b. CEA Reed Switch Position Transmitter (RSPT 2) with the capability of determining the absolute CEA positions within 5.2 inches, and

c. The CEA pulse counting position indicator channel.

APPLICABILITY: MODES 1 and 2.

ACTIDN:

kith a maximum of one CEA per CEA group having only one of the above required CEA.position indicator channels OPERABLE, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either:

a. Restore the inoperable position indicator channel to OPERABLE status, or

b. Be in at least HOT STANDBY, or J S,I,S,7

c. Position the CEA group(s) with the inoper ble position indicator(s) at its fully withdrawn position while m intaining the requirements of Specifications 3.1.3. 1~~ 3.1.3.6. Operation may then continue provided the CEA group(s) with the inoperable position indicator(s) is maintained fully withdrawn, except during surveillance testing pursuant to the requirements of Specification 4. 1.3. 1.2, and each CEA in the group(s} is verified fully withdrawn at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter by its "Full Out" limit".

SURVEILLANCE Rr UIREHEHTS 4.1.3.2 Each of the above required position indicator channels shall be determined to be OPERABLE by verifying that for the same CEA, the position indicator channels agree within 5. 2 inches of each other at leas~ once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

"CFAs are fully withdrawn (Full Out) when withdrawn to at least 144. 75 inches.

PALO VERDE - UNIT 2 3i4 1-25 goNTROLL,ED BY USER

REACTIYITY CONTROL SYSTEMS PART LENGTH CEA INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.7 The part len EA groups shall. be limited to the insertion limits shown on Figur e . with PLCEA inser tion between the Long Term Steady State Insertion Limit and the Transient Insertion Limit restricted to:

a. < 7 EFPD per 30 EFPD interval, and

b. < 14 EFPD per calender year.

APPLICABILITY: MODE 1 above 20~ THERHAL POWER.

ACTION:

a. With the part length CEA groups inserted beyond the Transient Insertion Limit, except for surveillance testing pursuant to Specification 4. 1.3. 1.2, within two hours, either:

1. Restore the part length CEA group to within the limits, or b.

2.

position using Figure ~

Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the PLCEA group

'l,l -5.

With the part length CEA groups inserted between the Long Term Steady State Insertion Limit and the Transient Insertion Limit for intervals > 7 EFPO per 30 EFPD interval or > 14 EFPO per calendar year, either:

Restore the part length group within the Long Term Steady State Insertion Limits ~ithin two hours, or

2. Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS

4. 1.3.7 The posi ion of the par iength CEA grouo shall be determined to be within the Transient Insertion Limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

"See Special Test Exception53.10.2 oar( 3 lO f

~ .

IO 20 30 40 50 60 TRANSIENT INSERTION LIMIT (TS.O INCHES )

TO zo 80 90 UNACCEPTABLE RESTRICTED OPERATION OPERATION IOO IIO LONG TERM STEADY STATE INSERTION LIMIT I20

( ll2.5 INCHES) 130 I40 I50 oo o CTI o

CD o o CP o

Q) oM o Y1 o o o o o o o o o o o o FRACTION OF RATED THERMAL POWFR FIGURE 3.I-5 PART LENGTH CEA INSERTION LIMIT VS THERMAL POWER PALO VERDE"UNIT 2 3/4 1-33

0 t t

CONTROLLED BY USER SP:r IAL . EST EyC.=P. i "qS 3/4. i O. 2 i~!GDERATGR TEMPERATURE COEr FICIENT. GROUP HETQHT TNSEPTTON AQD LIMITING CONDITION ."GR OPERATION 9i f>9 I7~

3. 10. 2 T mo'derator temperature coefficient, grouo height, inse. tion, and j.

power di ribution limits of Specifications 3. 1. 3, 3. 1. 3. 1, 3. 1. 3. 5,

3. 1. 3. 6, 3. 2. 2, 3. 2. 3, 3. 2. 7, and the Minimum Channels OPERABLE reouirement of i.C.j.(CEA Calculators) of Table 3.3-1 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER is'restricted to the test power plateau which shall not exceed 85% of RATED THERMAL POWER, and

b. The limits oi Specification 3.2. 1 are maintained and determined as speci fico in Soeciiication 4. 10. 2. 2 below.

APPLICABILITY: MODES 1 and 2.

ACTION gl3,9'i:

n any of -he l',mits of Soecification 3.2.1 being exceeded while reouiremen:s oi Soecifications 3.1. j.:, 3. 1.3. 1, 3.1.3.5, 3.1.3.6, 3.2.2,

."." ., 3.2.7, and -he Minimum Channels OPERABLE requirement of I.C. (CEA Calculators) of Table 3.3-1 are suspended, either:

a. Reduce THERMAL POWER suiiiciently io satisfy the reouirements oi Speci>ication 3.2.', or

b. Be in HGT STANDBY wiiihin 6 hours.

SURVEILLANCE REOU R=MENTS >

9> I) 3I7~

4. 10. 2.i The THERMAL GWER shall be determined a least once per hour auring PHYSICS TES:5 in w cn -he requirements of Speciiications 3.i.'.3, 3. 1.3.:,

3. 1. 3. 5 . 3. j. 3. 6, 3. 2. 2, 3. 2. 3, 3. 2. 7, or the Minimum Channels OPERABLE reauire-ment of >. C. 1 (CEA Calcuia.-ors) of Table 3.3-1 are suspended and siiai l be verified to be with'.n he test power plateau.

4. 0.2.2 The linear nea: rate shall be determined 'o be wi.hi n 'he limi:s o>

Specif ic tion 3.2.'y monitoring i: continuously with:he Inc ore Detector Monitor ing System pu. suant to .he reauiremen s of Soeci Ticatio ns -'.2.i.2 and

3. 3. 3 ' o'urina PHY :CS T=STS aoove 2Cio of RATED THERMAL PGW"R in wnicn the recuirements o> Soeci-,ica:ions 3.'.i.3, ~. 1.3.... h h

~...-, c. j. -" 6 3 2.2 I

~

"..2.3, 3.2.7, o. tne Miniimum Channels OPERAB = reouirement of (CEA Calicula ors) "-, Ta=lie ..- are susoenoed.

3,1,3,$J CONTROLLED BY USER

~,

I

'l f

~ i,

~ '

CONTROLLED BY USER SP CIAL TES ='gC=P-;nNS 3/ . 10. - C A ~C ' IOiV. R"'4.1.ATI. G C fn INST< l .OiV M': > s<0 --+C ". CQO'sNT

>

~ S. J,367j LIMITING COsVD I ' 'N ."OR OPERRnTIOiV 3.1Q.4 The reaoirements of:.pecf-factions 3.1.3.', 3.1.:..6 =.",a 3.".6 me! be suspended durinc .he performance of PHYSICS T=STS:o determir e the isothermal temperature coe-:-.'.icient, moceratcr temoerature coeff iclent, a.-,d oc-er coefficient provided he '.mi:s of Speci-.icat on 3.2.1 are mainitained anc aeter~inec as specified in Specification 4.10.4.2 below.

APPLICABILITY: MODES 1 and 2.

~P g I 6 9)7~

ACTION:

With any of he:-:omits of Specification 3.2. 1 be no exceeaea wni le =ne reauirements of Specifications 3.:.3.1, 3.1. 3.6 ana 3.2.= are susoe.".oed. either:

a. Reduce THERMAL POW'ER suificiently to satisfv tne re" ire.-..~~ts

b. "=e i.". -3T STANDBY ithin o hours.

SURVEILLANCE REOL':.REMENTS

4. 10.4.

PHYSICS TESTS 1 The THERMAL POMFR in wnich the reouirements of Specifi "ations 3 and/or 3.2.6 are suspenaed and snail be verifiea to be within :ne =es- powder

.,:

snail oe determined at least once ver r."ur curing

'."- 5p3 j36'7~

plateau.'.

10.-'.2 The guinea. heat ra=e shall be determined o be wi-h-:n .ne I imi 5 of Soecification "=.2. 1 by monitorino . continuously ith t.",e rc=r e

~ e-ec-"r Nonitorino Sys-e... pursuant to he rendu'irements o "pecif-cat',:n n

~ 4 2

~

during PHYSICS TESTS above 2",. of RATED THERMAL PO'ff'ER in wnic.- : .".e of Specifications 3. 1. 3. 1, 3. 1. 3.".and/or 3. 2. o are suspended.

9, I, 3,'7 "CONTPOLL'hD BY USER

~ '

CGNTRQLILEB 8'f USiER REACTIYIn CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES (Continued) and load maneuvering. Analyses are performed based on the expected mode of operation of the NSSS (base load maneuvering, etc.) and from these analyse

, CEA insertions are determined.and a consistent set of radial peaking factors

'defined. 'The Long Term Steady State and Short Term Insertion Limits are deter-mined based upon the assumed mode of oper ation used in the analyses and orovide a means of preserving the assumptions on CEA insertions used. The limits speci-fied serve to limit, the behavior of the radial peaking factors within the bounds determined from analysis. The actions specified serve to limit the extent of radial xenon redistribution effect's to those accommodated in the analyses. The Long and Short Term Insertion Limits of Specifications3. l. 3. 6 are specified for the plant which has been designed for primarily base loaded peration. but which has the ability to accommodate a limited amount of load mane vering.

m* S,t.37'-

The Transient Insertion'imits of Specifications 3. 1.3.6 and the Shutdown CEA Insertion Limits of Specification 3. 1.3.5 ensure"th'at (1) the minimum SHUT-

-DOWN MARGIN is maintained, and (2) the potential effects of a CEA e'ection accident are limited to acceptable levels. Lying-term operation at the Insertion Limits is not, permitted since such operation could have effects Tran-'ient on the core power distribution which could invalidate assumptions used to .deter-mine the behavior of the radial peaking factors.

.,'Rhe PYNGS CPC and COLSS systems are responsible for the safety and monitorin functions, respectively, of the reactor core. COLSS monitors the DNB Power Operating Limit (POL) and various ooerating parameters to help the operator main-tain plant operation within the limiting conditions for operation (LCO). Operat-ing within the LCO guarantees that in the event of an Anticipated Operational Occurrence (AOO), the CPCs will provide a reactor trip in time to prevent un-acceptable fuel damage.

The COLSS reserves the Required Overpower Margin (ROPM) to account for the Loss of Flow (LOF) .transient which is the limiting AOO for the PVNGS plants.

When the COl SS is Out of Service (COOS), the monitoring function is performed via the CPC calculation of DNBR in conjunction with a Technical Specification COOS Limit Line (Figure 3. 2-2) which restricts the reactor power sufficiently'o preserve the ROPM; The reduction of the CEA deviation penalties in accordance with the CEAC (Control Element Assembly Calculator ) sensitivity reduction program has been performed. This task involved setting many of the inward single CEA deviation penalty factors to 1.0. An inward CEA deviation event in effect would not be accompanied by the application of the CEA deviation penalty in either the CPC DNB and LHR (Linear Heat Rate) calculations for those CEAs with the reduced penalty factors. The protection for an inward CEA deviation event is thus accounted for separately.

1% ~ ~

I c

~ i CGNTROLLED BY USER REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES (Continued}

If an inward CEA deviation event occurs, the current CPC algorithm app'.ies two penalty factors to each of the ONB and LHR calculations. The first, a static penalty factor, is applied upon detection of the event. The second, a xenon redistribution oenalty, is apolied linearly as a function of time after the CEA drop. The expected margin degradation for the inward CEA deviation event for which the pena1ty factor has been reduced is accounted for in two ways.

The ROPM reserved in COLSS is used to account for some of the margin degrada-tl on.

a power reduction in accordance with the " rve in V. 4J,e> ~

Fi ure 3.~- is reouired. In aadition, the part length CEA maneuvering is restricted in acta>nance with Figure 3.1+ to justiiy reduction oi -ne Ptg devi ati on penal ty factor s.

The technical soecification permits plant ooeration if both CEACs are considered inoperaol e s or saf ety purposes af-'er:ni s peri oa.

PALO VERDE - i B

" 1 U i

0'

!

l t

1

ATTACHMENT 5 A. DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE UEST The proposed amendment changes the response time of the DNBR -Low Reactor Coolant Pump (RCP) shaft speed trip in Technical Specification (T.S.) 3.3.1, Table 3.3-2. The change is due to redefining the events which take place before the Control Element Assemblies drop into the core. During Cycle 1, the response time of .75 seconds was measured from the time a trip condition existed, such as a loss of power to the RCP motors, to the moment the Control Element Drive Mechanisms (CEDM) coil breakers opened. During Cycle 2 operation, the response time of .3 seconds will be defined from the time a signal is sent down the RCP shaft speed sensor line to the CPCs to the moment the CEDM coil breakers open.

B. PURPOSE OF THE TECHNICAL SPECIFICATION The purpose of T.S. 3.3.1 is to ensure that (1). the associated Engineered Safety Features Actuation action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.

C. NEED FOR THE TECHNICAL SPECIFICATION AMENDMENT During the Cycle 1 startup testing, it was found that the projected Reactor Coolant flow, ratetrip software, housed in the Core Protection Calculators, which monitors the RCP shaft speed and projects what the Reactor Coolant System flow will be in the future, was too sensitive to small deviations in RCP shaft speeds and caused unnecessary trips to the Unit. To correct this problem, the software dealing with the projected flow rate trip was taken out. In its place, trip software, which trips the unit when the RCP shaft speed slows to 95% of its normal speed as did the projected flow rate trip, was installed.

Because of this change, the response time, as defined for the RCP shaft speed trip, has been redefined for Cycle 2 to reflect the purpose of the new trip.

As a result of the redefinition of the response time, the safety analysis for Cycle 2 has taken credit, for the faster time and to ensure that the Unit is operated within the safety analysis, Table 3.3-2 will have to reflect the credited response time as 'it was used in the safety analysis.

D. BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION

1. The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in

'I I

~ i

accordance with, a proposed amendment would'not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from 'any accident previously 'valuated; or (3) Involve a significant reduction in a margin of safety.

A discussion, of these standards, as they 'relate to the, amendment I

request follows:

I j

t Standard 1--Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the changed response time ensures sufficient margin for mitigating the most limiting Design Basis Event (DBE). The Cycle 2 safety analysis results are still bounded by the reference cycle analysis. Therefore, there is no increase in the probability or consequences of an accident previously evaluated.

Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because the change maintains the margin of safety. The redefinition of the response time insures that the results of the Cycle 2 safety analysis will remain within the bounds of the Specified Acceptable Fuel Design Limits (SAFDLs) and, by maintaining the .3 second response time, the Unit will be operated within the realm of the safety analysis. Therefore, the change will not create the possibility of a new or different kind of accident.

Standard 3--Involve a significant reduction in a margin of safety.

The proposed change does not involve a significant reduction in a margin of safety because the change ensures the margin of safety for Cycle 2 is maintained. The analysis results show that there is sufficient margin to mitigate the most limiting DBE and that the results are bounded by the reference cycle. Therefore, no reduction in margin will arise.

2. The proposed amendment matches the guidance concerning the application of standards for determining whether or not a significant hazards consideration exists (51 FR 7751) by example:

(iii) For a nuclear power reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found previously acceptable to the NRC for a previous core at the facility in question are involved. This assumes that no significant changes are made to the acceptable criteria for the Technical Specifications, the analytical methods used to demonstrate conformance with the Technical Specifications and regulations are not significantly changed, and that NRC has previously found such methods acceptable.

ll 0

E. SAFETY EVALUATION FOR THE AMENDMENT RE UEST The proposed Technical Specification amendment will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR, The proposed change does not change 'or'eplace equipment or components which are important to safety. The change reflects the actual response time of the trip circuitry.

The proposed Technical Specification amendment will not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR. The change maintains the margin of safety. The redefinition of the response time insures that the results of the Cycle 2 safety analysis will remain within the bounds of the Specified Acceptable Fuel Design Limits (SAFDLs) and, by maintaining the .3 second response time, the Unit will be operated within the realm of the safety analysis. This does not increase the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR.

The proposed Technical Specification amendment will not reduce the margin of safety as defined in the basis for the Technical Specifications. The change ensures the margin of safety for Cycle 2 is maintained. The analysis results show that there is sufficient margin to mitigate the most limiting DBE and that the results are bounded by the reference cycle. Therefore, no reduction in margin will arise.

F. ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental question because operation of PVNGS Unit 2, in accordance with this change, would not:

1. Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the staff's testimony to the Atomic Safety and Licensing Board; or

2. Result in a significant change in effluents or power levels; or

3. Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.

G. MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES'imiting Conditions For Operation And Surveillance Requirements:

0 II V

l t

~

TABLE 3.3-2 I RL'ACTOR PROTECTIVE INSTRUHENTATION RESPONSE TIHES

(

CD rn R7 FUNCTIONAL UNIT RESPONSE TIHE CD fR I. TRIP GEHFRATION

h. Process

l. Pressurizer Pressure - lligh < 1.15 seconds

2. Presstrrizer Pl essrrre - Low < l. 15 seconds

3. Steam Generator Level - Low < 1. 15 seconds Steam Generator Level - High < 1.15 seconds Steam Generator Pressure - Low < 1.15 seconds Containment Pressrrre - lligh < 1.15 seconds Reactor Coolant Flow Low < 0.58 second Local Power Density - High

a. Neutron Flux Power from Excore Neutron Detectors < 0.75 second*

b. CEA Positiorrs < 1.35 second*"

c. CEA Positions: CEAC Penalty Factor < 0.75 second*"

9. DNOR - Low

a. Neutron Flrrx Power from Excore Neutron Detectors < 0.75 second*

h. CEA Positions < 1.35 second*"

C. Cold Leg Temperature < 0.75 secondNlhr d.

e.

f.

Hot Leg Temperature Primary Coolant Pump Shaft Speed Reactor Coolant Pressure from Pressurizer 0.30 ~~ <

<

0.75 secondNf seconds 0.75 seconds'mt

g. CEA Positions: CEAC Penalty Factor < 0.75 second"*

B 0. Excore Neutron Flux D

o. Variable Overpower Trip < 0.55 second" Logarithmic Power Level - lligh r+

a. Startrrp and Operating < 0.55 second"

h. Shutdown < 0.55 second"

I t

l I

~ '

I

'

ATTACHMENT 6 A. DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE VEST The proposed amendment revises the CEA Insertion Limits as set forth in Technical Specification (T.S.) 3.1.3.6. Operation of the regulating Control Element Assemblies (CEAs) during Cycle 2 will be more limited than in Cycle 1.

The revisions to the curves will maintain the margin of safety and insure that there will be sufficient shutdown margin to handle the most limiting Anticipated Operational Occurrence (AOO) and limiting fault events.

B. PURPOSE OF THE TECHNICAL SPECIFICATION The purpose of T.S. 3.1.3.6 is to ensure that (1) acceptable power distribution limits are maintained, (2) the minimum shutdown margin is maintained, and (3) the potential effects of CEA misalignments are limited to acceptable levels.

C. NEED FOR THE TECHNICAL SPECIFICATION AMENDMENT The proposed changes made to the CEA Insertion Limits are due to the change in the Cycle 2 core physics. Because of the change to the core, the worth of the CEAs has changed and as a result, the effects of the dropped and ejected CEA events change. To ensure that there is sufficient margin to mitigate such events, CEA insertion has to be restricted by the insertion limits set forth in the proposed T.S. 3.1.3.6.

D. BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability of consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident'reviously evaluated; or (3) Involve a significant reduction in a margin of safety.

A discussion of these standards as they relate to the amendment request follows:

Standard 1--Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because by restricting the insertion of the rods to Gp 3 60" withdrawn, margin is

maintained to mitigate the most limiting events, the dropped or ejected rod accidents as they are, described in the FSAR. By complying with the proposed changes during Cycle 2 operation, the Cycle 2 safety analysis results will be bounded by the reference cycle (Cycle 1) safety analysis.

This then ensures that the Cycle 2 operation will experience the same probability of consequences of an accident. The proposed change is made to ensure that Cycle 2 safety analysis is bounded by the reference cycle (Cycle 1) safety analysis. Therefore, there is 'o change in the probability or 'consequences of an accident occurring.

Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed change is more limiting than the reference cycle insertion limits. By restricting the insertion limits, there become fewer opportunities for the Unit to experience accidents. Since the change is more conservative a new or different kind of accident will not be created.

Standard 3--Involve a significant reduction in a margin of safety.

The proposed change does not involve a significant reduction in a margin of safety because the proposed change is being made to maintain Cycle 2 margin of safety and sufficient shutdown margin for the most limiting Anticipated Operational Occurrence (AOO) and limiting fault event.

Therefore, the reduction of safety margin does not arise.

2. The proposed amendment matches the guidance concerning the application of standards for determining whether or not a significant hazards consideration exists (51 FR 7751) by example:

(iii) For a nuclear power reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found previously acceptable to the NRC for a previous core at the facility in question are involved. This assumes that no significant changes are made to the acceptable criteria for the Technical Specifications, the analytical methods used to demonstrate conformance with the Technical Specifications and regulations are not significantly changed, and that NRC has previously found such methods acceptable.

SAFETY EVALUATION FOR THE ENDMENT RE UEST The proposed Technical Specification amendment will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. The proposed change is not a change or replace equipment or components important to safety.

Therefore, there is no increase in the probability of occurrence or the consequences of an accident occurring.

The proposed Technical Specification amendment will not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR. The proposed change places limits on the insertion of the CEAs such that the results from any accident occurring, while within the bounds set by T.S. Figure 3.1-3 and 3.1-4, will have the same consequences as those determined for the reference cycle. Thus, the proposed change is a result of maintaining the Cycle 2 safety analysis results within the reference cycle bounds and no new or different kinds of accidents will be created.

The proposed Technical Specification amendment will not reduce the margin of safety's defined in the basis for the Technical Specifications, The proposed change is being made to maintain Cycle 2 margin of safety and sufficient shutdown margin for the most limiting AOO and limiting fault events.

Therefore, the reduction of safety margin does not arise.

F. ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATIO The proposed change request does not involve an unreviewed environmental question because operation of PVNGS Unit 2, in accordance with this change, would not:

Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the staff's testimony to the Atomic Safety and Licensing Board; or

2. Result in a significant change in effluents or power levels; or E

'I 3 ~ Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.

MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES Limiting Conditions For Operation And Surveillance Requirements:

3/4 1-31 3/4 1-32

0.90 W O m 0 0.80 ~ Vl C/l m R9 o 0.70 o~

0.60 TRANSIE INSERTION LIMIT C/l 0.50 m~M MJ 0.40 5 mm C U mJ Kl D

CD O 0.30 J C/l mD a

c CI 0.20 D O.IO 0.00 5 3 I 150 l20 90 60 30 0 l50 l20 90 60 30 0 l50 I20 90 60 30 0 2

I50'20 90 60 30 0 I50 l20 90 60 30 0 CEA WITHORAWAL - INCHES

m 0.90 m

0.

A- 0.~0 Qro 0 0.60 Cl H Ch

~G) ~

Rl .0.50 Ag HO tlat Pg

~H A IHSERTIgtf L Pt< O 0.40 HW Hg g.

g Vl-

0. 30 lQ~ g$ g o OZO Q5 0.10 0.00 tA FVI 150 120 90 )0 QO 0 l'SO 120 ')0 6 30 0 l. l20 0 60 30 0 50 I l 90 60 30 0 l50 l 0 90 60 0 l

CEA'IT)IDRAWAL - INC l S FIGURE 3.1-3 CEA INSERTION LIMITS VS THERHAL POWER (COLSS IN SERVICE)

~,

0

1.00

(

Pl C3 0.90 XI I D K I/I Pl 0.80 C/l W 3> O Vl

+C m o~ I/I Pl 0.70 n C7 m z~'l O

CO I I/I I/l Z 0.60 &CO I O

I I I/I g C) O IO

+0 I/I 27 g -o 5o. I

-I~ 'Z

(/I W C) 27 I/l I m O4O m mm O m 27 I Pl I U n

Pl I VI I

0 I C) 0.30 I Z~

I- m II .I .

o O I Ko Ql 0.20

( O m

O.IO 0.00 5 3 I l50 l20 90 60 30 0 I50 l20 90 60 30 0 l50 l20 90 60 30 0 4 2 150 l20 90 60 30 0 l50 l20 90 60 30 0 CEA WITHDRAWAL"INCHES

(

m 1.00 C7 m 0.90 /cn g

~Q.

0.80 I~

M lR 0.70 QH

.g 0 A H CO 0

0.60 C osl

~

i=i o l/l h>

~

C s, 0.50 C)

HO v) Ch& A TRANSI NSE RT 10 .IMITQ 0.40 Q~o

~4 HP~

o 4' Ln it~ 0 30 C Cjl fi Hfo 0.20 0.10 O.na

')

I 150 l20 90 60 ln 0 lSO I20 90 6O 3O 0 150 l 90 60 '30 0 150 l2l) 9t) 60 30'; 150 l 20 90 60 30 0 CEA WITIIDRAWAL - INCIIES FIGURE 3.1-4 CEA INSERTION LIHITS VS THERHAL POWER (COLSS OUT OF SERVICE)

I

ATTACHMENT 7 A. DESCRIPTION OF THE PROPOSED CHANGE The existing PVNGS Unit 1 Technical Specifications provide an allowance for entering penalty factors into the Core Protection Calculators (CPCs) to compensate for Resistance Temperature Detector (RTD) response times greater than 8 seconds (but less than or equal to 13 seconds). These CPC penalty factors are provided in Technical Specification Table 3.3-2a and are supported by the Cycle 1 safety analyses. However,, the Cycle 2'afety analyses will not support these CPC penalty factors. Therefore, Table 3.3-2a must be deleted and Table 3.3-2 must be revised to remove this CPC penalty factor allowance.

B. PURPOSE OF THE TECHNICAL SPECIFICATION Technical Specification Table 3.3-2 (and associated Table 3.3-2a) provide the allowable response times for instrumentation used in the PVNGS reactor protective system. By ensuring that the reactor protective instrumentation meets these response time requirements, the assumptions used in the safety analyses are complied with and the associated protective action (i.e., reactor trip) is received within the time frame allowed by the safety analyses.

The RTDs that are the subject of this proposed Technical Specification change measure the Reactor Coolant System (RCS) hot and cold leg temperatures. The temperature measurements are provided as an input to the CPCs for use in the DNBR calculation. Each CPC channel receives temperature inputs from both RCS hot legs and from two diametrically opposed RCS cold legs.

C NEED FOR THE TECHNICAL SPECIFICATION AMENDMENT This Technical Specification change is necessary in order to ensure that the Cycle 2 safety analyses assumptions are complied with during Unit 1, Cycle 2 operations. The Cycle 2 safety analyses assume a maximum RTD response time of 8 seconds and do not include an allowance to enter CPC penalty factors to compensate for RTD response times greater than 8 seconds. Therefore, there should not be any allowances in the Technical Specifications for using the CPC penalty factors. For this reason, Technical Specification Table 3.3-2a should be deleted and Table 3.3-2 should be revised to remove the penalty factor allowances.

D. BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION

1. The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability of consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

l!

0 t tl 0

A discussion of these standards as they relate to the amendment request follows:

Standard 1 -- Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed Technical Specification change will not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change involves revising Table 3.3-2 and deleting Table 3.3-2a to remove the allowance which provides for CPC penalty factors to compensate for RTD response times greater than 8 seconds. The subject RTDs measure the RCS hot and cold leg temperatures, and provide an input to the associated CPC channel for use in the CPC DNBR calculation. The response times of these RTDs has no impact on the probability of occurrence of any of the accidents that depend on a CPC low DNBR reactor trip.

This revision to Table 3.3-2 and the deletion of Table 3.3-2a will ensure that the consequences of the analyzed accidents will be no worse than evaluated for the Cycle 2 safety analyses. The existing Cycle 1 safety analyses support the use of CPC penalty factors to compensate for RTD response times slower than 8 seconds. The Cycle 2 safety analyses do not support the use of the CPC penalty factors. Thus, during Cycle 2, any RTD response times greater than 8 seconds will be unacceptable and the use of Table 3.3-2a will not be supported by the Cycle 2 safety analyses. Therefore, Table 3.3-2a should be deleted and Table 3.3-2 should be revised to assure that operation of PVNGS Unit 1 is in accordance with the Cycle 2 safety analyses.

Standard 2 -- Create the possibility of a new or different kind of accident from any accident previously analyzed.

This proposed Technical Specification change will not create the possibility of a new or different kind of accident from any accident previously analyzed.

This proposed change, to delete the Technical Specification allowance for degraded RTD response times, does not affect the operation of the RTDs or the associated CPC channels. With the change, if a RTD, response time is greater than 8 seconds, the associated CPC channel must, be declared inoperable until repairs and/or retest are successfully completed.

Standard 3 -- Involve a significant reduction in a margin of safety.

This proposed Technical Specification change will not involve a significant reduction in a margin of safety. The 'asis for the existing Technical Specification Table 3.3-2a is the Cycle 1 safety analysis which analyzed the cases where the RTD response times were greater than 8 seconds but less than 13 seconds. For Cycle 2, there will not be an analysis to support the CPC penalty factors for degraded RTD response times. Therefore, Table 3.3-2a must be deleted since it will have no supporting basis, during Cycle 2.

The Commission has provided guidance concerning the application of the Standards for determining whether a significant hazards consideration exists by providing certain examples (51 FR 7751) of amendments that are considered least likely to involve a significant hazards consideration. This proposed amendment matches example (ii) in that it is a change that constitutes an additional

0 0

N 0

limitation, restriction or control not presently included in the Technical Specifications. Specifically, this proposed Technical Specification change constitutes an additional limitation because the allowance for RTD response times greater than 8 seconds has been deleted. Thus, if a RTD response time is measured greater than 8 seconds, then that channel of the CPCs must be declared inoperable until repairs and/or retest are satisfactorily completed.

SAFETY EVALUATION FOR THE PROPOSED CHANGE This proposed Technical Specification change will not increase the probability of occurrence of an accident previously evaluated in the FSAR. The subject RTDs measure the RCS hot and cold leg temperatures and provide an input to the CPCs for use in the CPC DNBR calculations. The response times of these RTDs have no effect on the probability of occurrence of any of the accidents that rely on a CPC low DNBR trip.

This proposed Technical Specification change will not increase the consequences of any accidents previously evaluated in the FSAR. The existing Cycle 1 safety analyses assure a RTD response time of no greater than 8 seconds. Additional analysis was performed for Cycle 1 to justify the application of CPC penalty factors if the measured RTD response times are greater than 8 seconds but no more than 13 seconds. This additional analysis supported the provisions contained in Technical Specification Tables 3.3-2 and 3.3-2a to apply CPC penalty factors to compensate for degraded RTD response times. The Cycle 2 safety analyses also assumed a maximum RTD response time of 8 seconds. However, no additional analysis was performed for Cycle 2 to support RTD response times greater than 8 seconds. Therefore, the Cycle 2 safety analyses do not support Table 3.3-2a and it must be deleted to ensure operation of PVNGS Unit 1 within the Cycle 2 safety analyses. Therefore, this Technical Specification change will ensure that the consequences of any accidents will be no greater than that of the Cycle 2 safety analyses.

This proposed Technical Specification change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

This proposed change, to delete the Technical Specifications allowance for degraded RTD response times, does not affect the operation of the RTDs or the associated CPC channels. With the change, if a RTD response time is greater than 8 seconds, the associated CPC channel must be declared inoperable until repairs and/or retest are successfully completed.

This Technical Specification change will not reduce the margin of safety as defined in the basis for any Technical Specifications. The basis for the existing Table 3.3-2a is the Cycle 1 safety analyses which analyzed the cases where the RTD response times were greater than 8 seconds but less than 13 seconds. For Cycle 2, there is no longer an analysis to support the CPC penalty factors for degraded RTD response times. Thus, Table 3.3-2a must be deleted since it will have no basis during Cycle 2.

0 F. ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental questionbecause operation of PVNGS Unit 2, in accordance with this change, would not:

1. Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the Staff's testimony to the Atomic Safety and Licensing Board .

(ASLB), Supplements to the FES, Environmental Impact Appraisals, or in any decisions of the ASLB; or

2. Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental'mpact.

I G. MARKED-UP TECHNICAL SPECIFICATION CHANGES PAGES Enclosed are revised pages 3/4 3-12; 3/4 3-13 of the PVNGS Unit 2 Technical Specifications'

TAOLE 3.3-2 t :iued)

I REACTOR PROTECTIVE IHSTRUHEHTATIOH RESPOHSE TIMES C)

(

m FUHCTIOHAL UHIT RESPONSE TIME C) m C. Core Protection Calculator -'System

1. CEA Calculators Hot Applicable

2. Core Protection Calculators Hot Applicable

0. Supplementary Protection System Pressurizer Pressure - Iligh < 1 15 second II. RPS LOOIC A. Matrix Logic Hot Applicable

0. Initiation Logic Hot Applicable III. RPS ACTUATIOH nEVICES A. Reactor Trip Breakers i Hot Appl cable

8. Manual Trip Hot Applicable hJ pe~~ 77~<

of the neutron Heutron detectors are exempt from response time testing. The

~

flux signal portion of the channel shall be measured from the detector output or from the input. of first electronic. component in channel.

AA @<5~~

.Respen~M~ shall be measured from the output of the sensor. Acceptable CEA sensor response shall be demonstrated by compliance with Specification 3. 1.3.4.

IThe pulse transmitters measuring pump speed are exempt from response time testing. The

'I 8 h,lf b d f th p I h p I p t: 8 R-.&WM rr~W I '

h 11 b d f tb tP t f lb I t.

response time shall be measured at least once per months. The measured I lt8b (sensor). RTD 1 p tl fib 8 tdfb hllb I tb d.

~h~PC addressab)e-constants-given-in-Tabl e-3-.3-2a-shaH-be-.made-4o-accommodat~

-current-va4 ue~~he-RTD-t4me-cons%a exceeds the .value-corresponding-to-the-penal-ties-algren (s) 8 ll-l I R 8-I R.-~

bl tH.-P -I PP 888~~I EItNAespense-tive shall be measured from the output, of. the pre'ssure transmitter. The transmit.ter response time shall be leis than or "equal to 0. 7 second.

~ '

I

CONTROLLE~ BY USER TABLE 3. 3-2a INCREASES IN BERRO, BERR2. AND BERR4 VERSUS RTD DELAY TIM S BERRO ERR2 BERR4 RTD DELAY TIME INCREASE INCREASE INCREASE

(~) () (~) (5) t < 8 0 sec 0 0 8.0 sec < x < 10.0'ec 2.5 2.0 1.0 10.0 sec < x < 13.0 se 6.0 4.0 6.0 NOTE: BERRY increases are not cumulative. For, example, the time

~ant changes from the range of 8.0 < t < 10.0 sec to t e~ange 10.0 < x < 13.0, the BERRO increase from its original (x < 8.0 st@-

.value is 6.0 not 2.5 + 6.0.

PALO VERDE - UNIT 2 CGRTRQLLED BY USER

'

ATTACHMENT 8 A. DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE UEST The proposed amendment changes references to the calculated Departure from Nucleate Boiling Ratio (DNBR) from 1.231 to 1.24 as set forth in Technical Specification (T.S) 2.1.1.1, Table 2.2-1, Basis 2.1.1, and Basis 2.2.1. The amendment also deletes references to the calculation of additional rod bow penalties if the rod bow penalty incorporated into the DNBR limit is not sufficient for any part of the cycle. The low pressurizer pressure floor is also changed from 1861 to 1860 because of the changed DNBR value.

B. PURPOSE OF THE TECHNICAL SPECIFICATION The purpose of T.S. 2.1.1 is to prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large,, and the cladding surface temperature is slightly above the coolant saturation temperature.

C. NEED FOR THE TECHNICAL SPECIFICATION AMENDMENT During Cycle 1 operation, the rod bow penalty factor was applied to the DNBR in increments. This method provided a means for not penalizing the operational margin unnecessarily during the cycle. As the fuel assemblies approach higher burnup the advantage of the Cycle 1 method no longer exists.

The application of a rod bow penalty factor large enough to provide protection throughout the cycle is now more advantageous. This can be accomplished because the physics of the Cycle 2 core is such that, by applying a rod bow penalty factor of 1.75% Minimum DNBR (MDNBR) to the DNBR limit, there will be sufficient margin to compensate for the effects of rod bow caused by those bundles with burnups of less than 30,000 MWD/MTU. For those bundles with burnups of greater than 30 GWD/MTU, there is sufficient margin from other factors to offset the small increase in the rod bow penalty.

As a result of the DNBR change, a reevaluation of the safety analysis was performed to determine if the low pressurizer pressure floor for the DNBR-low trip would change. The low DNBR trip provides protection in the event of an increase in heat removal by the secondary system and subsequent cooldown of the reactor coolant. The analysis has shown that a pressurizer pressure of 1860 instead of 1861 will ensure that, if a reactor trip occurs on Low-DNBR, the plant will not reach the Specified Acceptable Fuel Design Limits (SAFDLs).

t I

~ '

D BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION

1. The Commission has provided standards for determining wh ether a significanthazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the po'ssibility of a new or"different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.,

t A discussion of these standards as'hey relate'o the'mendment request follows:

Standard 1--Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the proposed change incorporates the reference cycle (Cycle 1) approved fuel rod bow penalty factor into the DNBR limit for fuel assembly burnups of up to 30,000 MWD/MTU. For those assemblies which will reach burnups of greater than 30,000 MWD/MTU in Cycle 2, there is sufficient available margin, due to lower radial power peaks, to offset any increase in the rod bow penalty. Thus, the probability or consequences of an accident occurring during Cycle 2 is the same as the reference cycle.

Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed change incorporates the reference cycle approved fuel rod bow penalty factor into the DNBR limit for fuel assembly burnups of up to 30,000 MWD/MTU. For those assemblies which will reach burnups of greater than 30,000 MWD/MTU in Cycle 2, there is sufficient available margin, due to lower 'adial power peaks, to offset any increase in the rod bow penalty. Therefore, the possibility of a new or different kind of accident will not increase.

Standard 3--Involve a significant reduction in a margin of safety.

The proposed change does not involve a significant reduction in a margin of safety because the proposed change incorporates the reference cycle approved fuel rod bow penalty factor in the DNBR limit for fuel assembly burnups of up to 30,000 MWD/MTU. For those assemblies which will reach burnups of greater than 30,000 MWD/MTU in Cycle 2, there is sufficient available margin, due to lower radial power peaks, to offset any increase in the rod bow penalty. Therefore, there is no reduction in the margin of safety.

~ i

2. The proposed amendment matches the guidance concerning the application of standards for determining whether or not a significant hazards consideration exists (51 FR 7751) by example:

For a nuclear power reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found previously acceptable to the NRC for a previous core at the facility in question are involved. This assumes that no significant changes are made to the acceptable criteria for the Technical Specifications, the analytical methods used to demonstrate conformance with the Technical Specifications and regulations are not significantly changed, and that'RC has previously found such methods acceptable.

E. SAFETY EVALUATION FOR THE AMENDMENT RE UEST The proposed Technical Specification amendment will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. The proposed change does not change or replace any equipment or components important to safety. The proposed change changes the DNBR margin by incorporating the reference cycle approved fuel rod bow penalty for a burnup of up to 30,000 MWD/MTU. Assemblies which will reach a burnup of greater than 30,000 MWD/MTU in Cycle 2, will not contribute a large enough rod bow penalty to require a larger penalty factor to be applied to the DNBR limit. The reference cycle safety analysis has incorporated into the analysis results. The effects of the higher burnups and, therefore, the DNBR for Cycle 2 is bounded by the reference cycle.

The proposed Technical Specification amendment will not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR. The proposed change is bounded by the reference cycle safety analysis because the effects of higher burnups on the fuel rod bow penalty factor were incorporated into the analysis. Therefore, the possibility of a new or different kind of accident stays the same.

The proposed Technical Specification amendment will not reduce the margin of safety as defined in the basis for the technical specifications. The proposed change is bounded by the reference cycle safety analysis because the effects of higher burnups on the fuel rod bow penalty factor were incorporated into the analysis. Therefore, the margin of safety stays the same.

F. ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental question, because operation of PVNGS Unit 2, in accordance with this change, would not:

1. Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the staff's testimony to the Atomic Safety and Licensing Board; or

2. Result in a significant change in effluents or power levels; or

3. Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.

G. MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES Limiting Conditions For Operation And Surveillance Requirements:

2-1 B 2-5 2-3 B 2-6 2-5 B 2-1 B 2-2

CONTROLLED BY USER 2.0 SAFETY LIMITS ANO LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS

2. 1. 1 REACTOR CORE DNBR

2. 1. 1. 1: The calculated DNBR of the reactor core shall be maintained gr eater than or equal to ~H..l 2.ct APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the calculated DNBR of the reactor has decreased to less than Q. . BM; be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specifi-cation 6.7. 1.

PEAK LINEAR HEAT RATE

2. 1. 1.2 The peak linear heat rate (adjusted for fuel rod dynamics) of the fuel shall be maintained less than or equal to 21 kw/ft.

APPLICABILITY: MODES 1 and 2.

ACTION:

whenever the peak linear heat rate (adjusted for fuel rod dynamics) of the fuel has exceeded 21 kM/ft, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requi rements of Speci fi cati on 6. 7. l.

REACTOR COOLANT SYSTEM PRESSURE

2. 1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

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

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 275O psia, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7. 1.

MODES 3, 4, and 5:

Mhenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to wi hin i:s limit wi hin 5 minu-es, and comply with the requirements of Specification 6.7. 1.

PALO YEROE - UNIT 2 2-1 CONTROLLED BY USER

1 I

TABLE 2.2-1 REACTOR PROTECTIVE INSTRUHENTATION TRIP SETPOINT LIHITS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES I. TRIP GENERATION A. Process

1. Pressurizer Pressure - High < 2303 psia < 2388 psia

2. Pressurizer Pressure - Low > 1837 psia (2) > 1822 psia (2)

3. Steam Generator Level - Low > 44.2X (4) . > 43.7X (4)

Steam Generator Level - lligh < 91.0X (9) < 91.5X (9)

5. Steam Generator Pressure - Low > 919 psia (3) > 912 psia (3)

6. Containment Pressure - Iligh < 3.0 psig < 3.2 psig

7. Reactor Coolant Flow - Low

a. Rate < 0.115 psi/sec (6)(7) < 0.118 psi/sec (6)(7)

b. Floor > 11.9 psid (6)(7) > 11.7 psid(6)(7)

c. Band < 10.0 psid (6)(7) < 10.2 psid (6)(7)

0. Local Power Density - lligh < 21.0 kW/ft (5) < 21.0 I(W/ft (5)

9. DNOR - Low > k-.PK(5) > a-aaa (5) i a'I .1 <'I B. Excore Neutron Flux I. Variable Overpower Trip

a. Rate < 10.6X/min of RATED < ll. OX/min of RATED TIIERHAL POWER (8) TIIERHAL POWER (8)

b. Cei ling < 110.0X of RATED < 111.0X of RATED THERHAL POWER (8) TIIERHAL POWER (8)

c. Band < 9 8X of RATED < 10.0X of RATED TIIERHAL POWER (8) TIIERHAL POWER (8)

I TABLE 2. 2-1 (Conti nued) r REACTOR PROTECTIVE INSTRlNENTATION TRIP SETPOIRT LIHITS TABLE NOTATIONS (1) Trip may be manually bypassed above 10- X of RATED CAROL PMR; bypass shall be automatically ~vied when THERMAL PSKR is less than or equal to 10-~X of RATED THERMAL PtWER.

(2) In HODES 3-4, value say be decreased aanually, to a in$ em of 100 psia, as pressurizer pressure is reduced, provided the Nargin between the pres-surizer pressure and this value is maintained at lasa than or:equal to 400 psi; the setpoint shall be increased autoeatically as pressurizer pressure 'is increased until the trip setpoirrt is ron:hed. Trip uay be aranual+y bypassed below 400 psia; bypass shall be a4uaetically removed whenever pressurizer pressure is greater than or equal to 500 psia.

(3) In HODES 3-4, value say be decrea'sed aanually as stae generator pressure

. is reduced, provided. the margin between the steae generator pressure and this value is maintained at less than or equal to 2CQ psi; the setpoint shall be increased autceatically as steam generator pressure is increased until the trip setpoint is reached.

(4) X of the distance between steam generator upper and"lower level wide range instrument norxles.

(5) As stored within the Core Protection Calculator (CPC). Calculation of the trip setpo$ nt inc'te5es eaaasurooent, calculatiorN1 and processor uncer tainti es, Trip aury be nanually bypassed below 1" of RATED THERMAL POSER; bypass shall be autoaatically reaoved when THERMAL P&ER is greater than or equal to R of RATED THERNhL POWDER; approved DNBR liarit is 1.231 which includes a portial rod bow pena compe tion. If the fuel burnup exceeds that for ich an incr! rod bow penal s required, the DNBR limit shall be acf$ usted. is case a DNBR trip setp of 1.231 is allowecf provided that the ference is com-pensated by an inc e in the CPC addressable cons BERRl as follows:

- RB where BERR1'ld1 is the unc sated value o bo~ penalty ir, 'X QN, B <s the fuel rod bow pena RR1; RB is the fuel rod in ~ DHBR already accounted fo n the DNBR limit; POL is the paver Qperatin <mit; and d (~ PD (X DNBR} is the absolute value of the most adverse t ivative

~ ~ith respect to DHBR.

8<<8 ~ PALO VERDE U~T. ~ 2>>5

>>>> I fl>> l ~>>ld'- - >>0 9'JLi <<n nT

~ i~ ~ ~ -- ~ .

II' CONTROLLED BY USER 2.1 and 2.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES 2;1.1 REACTOR CORE The restrictions of these safety limits prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the 'reactor coolant. Overheating of the fuel cladding is prevented by (1) restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature, and (2) maintaining the dynamically adjusted peak linear heat rate of the fuel at or less than 21 kM/ft which will not cause fuel centerline melting in any fuel rod.

First, by operating within the nucleate boiling regime of heat transfer, the heat transfer coefficient is large enough so that the maximum clad surface temperature is only'lightly greater than the coolant saturation temperature.

The upper boundary of the nucleate boiling regime is termed "departure from nucleate boi ling" (DNB). At this point, there is a sharp reduction of the heat transfer coefficient, which would result in higher cladding temperatures and the possibility of cladding failure.

'orrelations predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB ratio (DNBR}, defined a the ratio of the predicted ONB heat flux at a particular core loc actual heat flux at that location, is indicative of the minimum value of ONBR during normal operatio esi gn b

'o

'nti ...The ci pated e

operational occurrences is limited to M~

of CE-1 CHF correlation and engineering facto based u statistical combination ertainties and is established

. t 75 > ~

as a Safety Limit. The DNBR limit of on ONBR.

~~ includes a rod bow corn ensation of burnups which exceed that for which an od bow penalty is required, the e In this case the e( eke 8 ONBR trip setpoint of owe if the re

'

crease is c an increase, of the addressable. constant BERR1.

Second, operation with a peak linear heat rate below that which would cause fuel centerline melting maintains fuel rod and cia'dding integrity.

Above this peak linear heat rate level (i.e., with some melting in the center),

fuel rod integrity would be maintained only if the design and operating conditions are appropriate throughout the life of the fuel rods. Volume changes which accompany the solid to liquid phase change are significant and require accommodation. Another consideration involves the redis ribu ion of the fuel which depends on the extent of the melting and the physical state of the fuel rod at the time of melting. Because of the above factors, the steady state value of the peak linear heat rate which would not cause fuel centerline melting is established as a Safety Limit. To account ,or fuel rod dynamics (lags}, the directly indicated linear heat rate is dynamically adjusted by the CPC program.

PALO VERDE - UNIT 2 B 2-1 CCINTROLLED BY USER

0 0

i CGRTRGLLED BY USER .

BASES Limiting Safety System Settings for the Low DNBR, High Local Power Density, High Logarithmic Power Level, Low Pressurizer Pressure and High Linear Power Level trips, and Limiting Conditions for Operation on DNBR and kw/ft margin are specified such that there is a high degree of confidence that the specified acceptable fuel design limits are not exceeded during normal operation and design basis anticipated operational occurrences.

2. 1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the

'containment atmosphere.

The Reactor Coolant System components are designed to Section III, l974 Edition, Summer 1975 Addendum, of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110K (2750 psia) of

, design pressure. The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code 'requirements.

The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity. prior to initial operation.

2.2.1 REACTOR TRIP SETPOINTS Reactor Trip Setpoints,specified'in Table-2.'2-1 are the value's-wt

.r'he which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and Reactor Coolant System are prevented from exce'eding their Safety Limits dur ing normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint end the Allowable Value is equal to or less than the drift allowance assumed 'for each trip in the safety analyses.

l,ag The DNBR - Low and Local Power Density High are digital~y generated trip setpoints based on Safety Limits, of . and 21 kwlft, respectively.

,Since these trips are digitally generated by the Core Protection Calculators,

.the. trip values are not subject to drifts common to trips generated by analog type equipment. The Allowable Values for these trips are therefore the same as the Trip Setpoints.

To maintain the margins of safety assumed in the safety analyses, the calculations of the trip variables for the ONBR - Low and Local Power Oensity-High trips include the measurement, calculational and processor uncertainties and dynamic allowances as defined in CESSAR System 80 applicable system descriptions and safety analyses.

PALO VERDE - UNIT 2 B 2-2

~GgTRGLLED BY USER

~ i BASES Local Power Oensit - Hi h (Continued)

a. Nuclear flux power and axial power distribution from the excore flux monitoring system;

b. " Radial peaking factors from the position measurement for the CEAs;

c. Delta T power from reactor'coolant temperatures and coolant flow measurements.

The local power density (LPO), the trip variable, calculated by the CPC incorporates uncertainties and dynamic compensation routines. These uncer-tainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core peak LPD is sufficiently less than the fuel design limit such that the increase in actual core peak LPD after the trip will not result .

in a violation of the Peak Linear. Heat Rate'afety Limit. CPC uncertainties related to peak LPD are the same types used for DNBR calculation. Dynamic compensation for peak LPD is provided for the effects of core fuel centerline temperature delays (relative to changes in power density), sensor time delays, and protection system equipment time delays.

ONBR " Low I8i,O The ONBR -;Low trip-is provided Co prevent the D R in the limiting

-"

'coolant channel in the core from exceeding the fuel esign limit in the event

< of design bases anticipated operational occurrences The DNBR - Low trip incorporates a low pressurizer pressure floor of psia. At this pressure a ONBR - Low trip will automatically occur. The DNBR is calculated in the CPC utilizing the following information:

a. Nuclear flux power and axial power distribution from the excore neutron flux monitoring system;

b. Reactor Coolant System pressure from pressurizer pressure measurement; C. Differential temperature (Delta T) power from reactor coolant temperature and coolant flow measurements;

d. Radial p'caking factors from the position measurement for the CEAs;

e. Reactor coolant mass flow rate from reactor coolant pump speed; Core inlet temperature from reactor coolant cold leg temperature measurements.

~ ~ ~

PALO VERDE - UNET 2 B 2-5

0, 0

,li

SAFETY LIMITS ANO LIMITING SAFETY SYSTEMS SETTINGS BASES

- (Continued)

DNBR Low I,~4 The DNBR; %he trip variable, calcul ted by the CPC incorporates various uncer-tainties and dynamic compensation outines to assure a trip is initiated prior to violation of fuel design limit . These uncertainties and dynamic compensa-tion routines ensure that a reac or trip occurs when the calculated core ONBR is sufficiently greater than . such that the decrease in calculated core ONBR after the trip wi 1] not result in a violation of the DNBR Safety Limit.

CPC uncertainties related to DNBR cover CPC input measurement uncertainties, algorithm modelling uncertainties, and computer equipment processing uncertainties. Dynamic compensation is provided in the CPC calculations for the effects of coolant transport delays, core heat flux delays (relative to changes in core power), sensor time delays, and protection system equipment time delays.*

I The DNBR algorithm used in the CPC is valid only within the limits indicated below and operation outside of these limits will"result in a CPC initiated trip.

Parameter Limitin Value a ~ RCS Cold Leg Temperature-Low > 470 F b RCS Cold:Leg Temperature-High <;610 F C. Axial Shape Index-Positive 'Not more positive than + 0.5

d. Axial Shape Index-Negative Not more negative .than - 0.5

e. Pressurizer Pressure-Low sia

f. Press'urizer. Pressure-High < 2388 psl a I 840

g. Integrated Radial Peaking Factor-Low > 1.28

h. Integrated Radial Peaking Factor-High < 4.28 equality Margin-Low . > 0 Steam Gene~ato~ Level - Hi h The Steam Generator Level - High trip is provided to protect the turbine from excessive moisture carry over. Since the turbine is automatically tripped when the reactor is tripped, this trip provides a reliable means for providing protection to the turbine from excesssive moisture carryover. This trip's setpoint does-not correspond to a safety limit, and provides protection in the event of excess feedwater flow. The setpoint is identica! to the main steam isolation setpoint. Its functional capab'ility at the specified trip setting enhances the overall reliability of the reactor protection system.

PALO VERDE - UNIT 2 B 2-6 CGNTRGLLED iBY USE~

0

~ ~

e

Attachment 9 A. DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE UEST Reactor Coolant System (RCS) total flow

'he proposed amendment changes the rate as set forth ig Technical Specification (T.S.) 3.2.5 from gregter than or equal to 164.0 x 10 ibm/hr to greater than or equal to 155.8 x 10 ibm/hr.

B. PURPOSE OF THE TECHNICAL SPECIFICATION The purpose of T.S. 3.2.5 ensures that the actual RCS total flow rate is maintained at or above the minimum value used in the safety analysis.

C. NEED FOR THE TECHNICAL SPECIFICATION AMENDMENT T.S. 3.2.5 is being changed to eliminate an ambiguity in where instrument uncertainty is to be included when comparing measured RCS flow rate against the RCS flow rate used in the safety analysis. As currently worded, actual total RCS f)ow rate is to be compared against the 100% design flow value of 164.0 x 10 ibm/hr. The term "actual" implies that the RCS flow rate determined by the Reactor Coolant Pump (RCP) delta-pressure method is to be corrected for pressure transmitter uncertainty. The uncertainty amounts to a maximum of 4% of flow for transmitters within their calibration period. The corrected flow rate is then compared to 164.0 x 10 ibm/hr. The RCS flow ratg used in the safety analysis, however, is 95% of the d~sign flow or 155.8 x 10 ibm/hr. The 100$ design flow rate of 164.0 x 10 ibm/hr conservatively accommodated the maximum instrument uncertainty of 4%, removing the need to correct for instrument uncertainty. The T.S. basis states that the specification is provided to ensure that the actual total RCS flow rate is maintained at or above the minimum value used in the safety analysis. This T.S. change will remove the ambiguity and permit any changes in instrument uncertainty to be handled procedurally rather than requiring additional T.S.

changes.

D. BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION

1. The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

A discussion of these standards as they relate to the amendment request follows:

Standard 1--Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the value of 155.8 x 10 ibm/hr for minimum RCS flow rate is the value used in the reference 'cycle (Cycle 1), safety analysis. Therefore, the probability or consequences of an accident is the same for Cycle 2 as it is for the reference cycle.

Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because the same value was used for both the reference cycle and Cycle 2 safety analysis.

Therefore there is no possibility of creating a new or different kind of accident with the reduced RCS total flow.

Standard 3--Involve a significant reduction in a margin of safety.

The proposed ,change does not involve a significant reduction in the margin of safety because, no changes have been made to the safety analysis. ,The proposed value in the T.S. is the value used in both the reference cycle" and Cycle 2 safety analysis. Therefore, the margin of safety is the same for Cycle 2 as it is for the reference cycle.

2. The proposed amendment matches the guidance concerning the application of standards for determining whether o', not 'a significant hazards consideration exists (51 FR 7751) by example:

(iii) For a nuclear power reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found previously acceptable to the NRC for a previous core at the facility in question are involved. This assumes that no significant changes are made to the acceptable criteria for the technical specifications, the analytical methods used to demonstrate conformance with the technical specifications and regulations are not significantly changed, and that NRC has previously found such methods acceptable.

SAFETY EVALUATION FOR THE AMENDMENT RE UEST The proposed Technical Specification amendment will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. The proposed change does not change or replace equipment or components important to safety.

The safety analysis for the proposed change is the same as the reference cycle and, therefore, the probability of occurrence or the consequences of an accident is the same.

0 r 61 f

1 E

'1

The proposed technical specification amendment will not create the possibility for an accident or malfunction of a, different type than any previously evaluated in the FSAR. The Cycle' safety analysis for the proposed change uses the same value for RCS minimum flowrate as for the reference cycle and therefore, the possibility for an accident is the same.

The proposed Technical Specification amendment will not reduce the margin of safety as defined in the bases for the technical specifications. No changes have been made to the safety analysis. The proposed value in, the T.S. is the value used'. in both the reference cycle and Cycle 2 safe'ty analysis. Therefore, there is no reduction in the margin'of safety.

F. ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental question because operation of PVNGS Unit 2, in accordance with this change, would not:

Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the staff's testimony to the Atomic Safety and Licensing Board; or

2. Result in a significant change in effluents or power levels; or

3. Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.

G. MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES Limiting Condition for Operation and Surveillance Requirements:

3/4 2-8 B 3/4 2-4

l~

}J I

0

CONTROLLED BY USER POWER DISTRIBUTION LIMITS 3/4.2.5 RCS FLOW RATE LIMITING COND IT ION FOR OPERATION 3.2.5 The. actual Reactor Coolant System total flow rate shall be greater than

/1 t~s 8 MODE

~io'PPLICABILITY:

l.

ACTION: .

With the actual Reactor Coolant System total flow rate determined to be less than the above limit, reduce THERMAL POWER to less than 5X of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURYEILLANCE REOUIREMENTS 4.2.5 The actual Reactor Coolant System total flow rate shall be determined to be greater than or equal to its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

PALO VERDE - UNIT 2 3/4 2-8 CONTROLLED BY USER

~,

CQRTRGLLED BY USER POWER 'DISTRIBUTION LIMITS BASES 3/4. 2. 5 RCS. FLOW RATE This specificat'ion is provided to ensure that the actual RCS total flow qQh. rate is. ma'intained at er above the minimum\ value used in the safety analyses.

3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE This specification is provided to ensure that the actual value of reactor coolant 'cold leg temperature 'is, maintained within the range of values used in the.safety analyses.

3/4.2.7 AXIAL SHAPE INDEX This, specification is provided to ensure that the actual value of the core average AXIAL SHAPE INDEX is maintained within the range of values used in the

.

safety analyses.

3/4. 2. 8 PRESSURIZER PRESSURE r

This specification is provided to ensure that the actual value of pressurizer pressure is maintained within the r ange of values used in the safety analyses.

T4~ >i6+u~ valw ~pe< i~+he ~chef 0~a'lq ',s i~ 'I % 0(-+4<<.

des>q~ 9lo~~Q ( la4.OX<e lb~(hn) o~ iM P Xiii it m/hi .~i aJ'Aea c ~~~<~, <, + J 1 ~, P~c~ht<r w< f ib G4nhi(4<cd.

Ye.c~u.inc.6. v<~

PALO VERDE - U,. 2

ATTACHMENT 10 A. DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE VEST The proposed amendment changes the Linear Heat Rate (LHR) limit as defined in Technical Specification (T.S.) 3.2.1 from 14.0 kw/ft to 13.5 kw/ft. The change also provides information for the appropriate methods of monitoring LHR and formats the T.S. with regard to human factors.

B. PURPOSE OF THE TECHNICAL SPECIFICATION The purpose of T.S. 3.2.1 is to limit Linear Heat Rate which will ensure that, in the event of a Loss of Coolant Accident (LOCA), the peak temperature of the fuel cladding will not exceed 2200'F.

C. EED FOR THE TECHNICAL SPECIFICATION AMENDMENT In support of the Unit 1 reload, the reanalysis of the Safety Analyses resulted in a change in the Linear Heat Rate limit to ensure the peak fuel clad temperature is not exceeded. The change in the LHR is, in part, due to the change in the method of performing the safety analysis. As part of the analysis, penalties are applied to compensate for increased power peaking caused by the densification of small interpellet gaps. These penalties are called Augmentation Factors and were not used for the Cycle 2 analysis. This method change has been approved by the NRC in "Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 104 to Facility Operating License No. DPR-53, Baltimore Gas and Electric Company, Calvert Cliffs Nuclear Power Plant Unit No. 1, Docket No. 50-317". Other factors contributing to the change in LHR are from increased fuel enrichment and the core loading pattern.,

In addition to changing the references to LHR, the amendment also delineates how LHR is to be monitored. By "providing more detail of the monitoring of LHR, assurance is provided that the LHR will be maintained below the specified limit. The amendment, also changes the format of the ACTION statement in such a w'ay as to facilitate assessment of the, actions required be exceeded.

if the limit should D. BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION

1. The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

~ '

A discussion of these standards as they relate to the amendment request follows:

I Standard 1--Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the safety analysis of the proposed change is bounded by the safety limits set forth by 10 CFR 50.46. Changing the LHR limit will ensure that there is sufficient margin for the most limiting Design Basis Event (DBE). The change is also more conservative than the value used in Cycle 1. The format changes to the LCO and Action statements further define and clarify the actions required to be taken to ensure maintaining the LHR below the limit. Therefore, there will be no increase in the probability or consequences of an accident.

Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because the safety analysis results of the proposed change are bounded by the safety limits set forth by 10 CFR 50.46. The proposed change to the LHR is more conservative than the LHR allowed by Cycle 1, thus reducing the consequences of an event but not creating any new or different accidents.

The format modification changes the presentation of information within the T.S. but does not delete required actions and adds additional restrictions. Therefore, there will be no increase in the possibility of a new or different kind of accident.

Standard 3--Involve a significant reduction in a margin of safety.

The proposed change does not involve a significant reduction in a margin of safety because the safety analysis results of the proposed change are bounded by the safety limits set forth by 10 CFR 50.46. Changing the LHR limit will maintain sufficient margin for the most limiting DBE.

Therefore, there will be no reduction in the safety margin.

The proposed amendment matches the guidance concerning the application of standards for determining whether or not a significant hazards consideration exists (51 FR 7751) by examples:

A purely administrative change to Technical Specifications: for example, a change to achieve consistency throughout the Technical Specifications, correction of an error or a change in nomenclature.

and

0 t

h t

)

t I

H

M (iii) For a nuclear power reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found, previously, acceptable to the NRC for a previous core at the facility in question are involved. This assumes that no significant changes are made to the acceptable criteria for the Technical Specifications, the analytical methods used to demonstrate conformance with the Technical Specifications and regulations are not significantly changed, and that NRC has previously found such methods acceptable.

SAFETY EVALUATION FOR THE AMENDMENT RE UEST The proposed Technical Specification amendment will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. The safety analysis results of the proposed change are bounded by the safety limits set forth by 10 CFR 50.46 and do not change or replace equipment or components which are important to safety.

The proposed Technical Specification amendment will not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR. The safety analysis results of the proposed change are bounded by the safety limits set forth by 10 CFR 50.46. The proposed change to the LHR is more conservative than the LHR allowed by the reference cycle (Cycle 1), thus reducing the consequences of an event but not creating any new or different accident or malfunction; The format modification changes the presentation of information within the T.S., but does not delete required actions and adds additional restrictions.

The proposed Technical Specification amendment will not reduce the margin of safety as defined in the basis for the technical specifications. The safety analysis results of the proposed change are bounded by the safety limits set forth by 10 CFR 50.46. Changing the LHR limit for Cycle 2 will maintain sufficient margin for the most limiting DBE, thus maintaining the margin of safety.

ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental question because operation of PVNGS Unit 2, in accordance with this change, would not:

1. Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the staff's testimony to the Atomic Safety and Licensing Board; or

2. Result in a significant change in effluents or power levels; or

3. Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.

G. MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES Limiting Conditions For Operation And Surveillance Requirements:

3/4 2-1 B 3/4 2-1

0 1 1 K

I 0

i>sist M sac ~ rise'

~ r Wi

/Sr ~ ri ~ ~ rr ~ r ~

L M.>:HG COHO;Ti"H F"R 0. =<<"T".'QN l>~1- a 4 1 35

.2.1 The linear heat ra.e -' - '." kM/ S4<<tt be w~i4n<'npQ Q ss>s Q f$ f4 e Q<> ll <>win >( n<4'~~<>45 >S n pp l 'c>>bfe 0, R<>,>'ntCj>n<>>>> COLS>5 cc~p p.~m Ipse ft 46 r~~<<t W We COlI"<>+l'a(c,ul~te4 SS catC>st<>~4 pa><I<<r ep< r<>,f'<rig l ~1 b<<sip 0vl ( 'r><<<ir> k p<"s ( w4 e><<CQ-SS >'s in stir> c4> '1 ~ II<

~ ~

Ra.<'<><ai<<g reap p>>>>, k r<>fF m<~>n IM j <haft <nq <<ny ~ peroÃe CPC Pol I'"" C C I>,annot (~pe>n (+C

= I zhove 20' c Q ir

>

O~O OR pA i c J I ~

<<v '~).

.'CHAL PG4 AC .ON:

<* gin+ ~<>>a4>>>n+d I ~

Mith :he 1inear heat rate as i ndi ca ec hy ~

n> Sq r i ro r ~ ~

r r rr>>

~

/ +<< ~ er

~ \ 'VV I

~ r ~

~ ~ V 'V Sl> ~ 4 i p ass

~ saa ~

), COLSS, c tc t ied cor pv e PYc~ ck g

~4'4'~ 5 c 't I >s<>>l OR<>A On l><<par 4>. >> i roHC'z c>r p~ay cpu<>>g>nq Peak, (In<.tv h~i rg~ o><gs>h >< j>>s>>>~ Mg'>ssq>>>>y optrahft << c,6>> <'><r f

(<kw Cc>l.55 I'z c>~ if srrv>c<r)I 1:- minutes inii ate correc:ive ac:Ion:o recvee:."e linear neat rate wl n ii

he .limits ana ei-ner:
Co Res.ore he linear hea: I1 i"ill ts Wl i.i i il hcur, cr
h. Re+ "a Ti-:=RMAL Pic<<o 4<re rg <<< I
'eon r>s'il POMr.<<wi.:hin:he next 6 hou. s.
sUo'I= ' "HC- r,-uU <<= H i"
<.2.1.1 The provisions.c. spec -,',cation -'..0.4 are nc. avail 1 cah I e.
<. 2.1.
TH-"DOMAL 2 The linear heat ".=-
PO'4".<<1s 2"ove 20 2 5: rI he cete.... nec to. he w . >1 '.
'
l1 $ w>1en t><1 isle Ccrc power cistr;""u: cr, wi-1 ne Core Ooera: nc Limi= cl."e"."is"~ sys:e.-..'
(COLsc) o ~, wl .i isse C'nl cs 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ta: :1e linear hea Dens ts s ~
i:y l r
> ~ i cnannel A: least ence oer
, is '
<>P s4>n jig [>~>t....
2 2 >'S siIC ica ac .s rs Ot'<<n
'~"'-
2 I r ONe.
~
aC rua I at a In" PMAL PG'~'""" level less:nan cr e" a7 -c:,",e. -ore.=cw.er .co'e".2=.-."nc 11' t ha > nc'ar h<'a>
~
rrr <sr c
~,
0,
CQRTRQLLED BY USER 3/4. 2 POWER DISTRIBUTION L:MITS BASES 3/4.2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature o'f the fuel cladding will not exceed 2200 F.
Either of the two core power distribution monitoring systems, the Core Oper ating Limit Supervisory System (COLSS) and the Local Power Oensity channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits. The COLSS performs this function by continuously monitoring the core power distribution and calculating a core power operating corresponding to the allowable peak linear heat rate. Reactor operation of-~
'imit at or below this calculated power level assures that the limits kM/ft are not exceeded. l3,5 The COLSS calculated core power ana the COLSS calculated core power
.operating limits based on linear heat rate are continuously monitored and displayed to the operator. A COLSS alarm is annunciated in the event that the core power exceeds the core power opera-ing limit. This provides adequate margin to the linear beat rate operating limit for normal steady-state opera-.
tion., Normal reactor'ower transients or equipment failures .whichdo not require a reactor trip may result in this core power operating limit being exceeded. In the event this occurs, CO'S alarms, will be annunciated. If the event which causes the COLSS limit to be exceedea results in conditions which approach the core safety limits, a reac:or trip will be initiated by the Reactor Protective Instrumentation. The COLSS calculation of the linear heat rate includes appropriate. penalty'actors which provide', with a 95/95 probability/
confidence level, that the maximum linear heat rate calculated by COLSS is with respect, to the actual maximum linea~ heat rate existing in 'onservative the core. These penalty factors are determined from the uncertainties associated with planar radial peaking measurement, engineering heat flux uncertainty, axial densification, software algorithm modelling, computer processing, rod bow, and core power measurement.
Parameters required to maintain the operating limit power level based on linear heat rate, margin to DNB, and total core power are also monitored by the CPCs Therefore, in the event that the COLSS is by utilizing c a l. T bove listed uncertainty and penalty factors plus those 'associated with th CPC s".artup test acceptance criteria are also inc uded in the CPCs.
A.nq oq~4lc li~e~~ h ~0 ~
-
PAL0 vERDE UQQ gTPQLLPP Bg Upped
t ATTACHMENT 11 A'. DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE VEST f
The proposed amendment will'evise Technical'pecifications (T.S) 3.2.4, 3.3.1, Bases 3.1.3.1/3.1.3.2 and Bases 3.2.4, The changes are as follows:
T.S. 3.2.4-(1) Replaces the T.S. with a new format which addresses the specific conditions for monitoring DNBR with or without COLSS and/or the CEACs, (2) delineates by a new format what ACTIONS should be taken, (3) removes reference to the DNBR Penalty Factor table used in T.S. 4.2.4.4, and (4) replaces the present graph figures 3.2-1 and 3.2-2 of the DNBR limits with graph figures 3.2-1, 3.2-2 and 3.2-2a addressing DNBR operating limits for the conditions mentioned in (1) above.
T.S. 3.3.1-(1) Removes references to the, operation of the reactor with both CEACs inoperable and with or without COLSS inservice, and (2) deletes the graph of DNBR margin operating limit, Figure 3.3-1, based on COLSS for both CEACs inoperable. These changes are a result of being incorporated into the proposed T.S. 3.2.4 Bases 3.1.3.1/3.1.3.2-(l) Removes references to Cycle 1 specific information, and (2) modifies Bases due to T.S. 3.2.4 changes.
Bases 3.2.4-Modifies Bases due to the T.S. 3.2.4 changes.
These changes are due, in part, to ensuring operation of Cycle 2 within the approved safety analysis and to improving the Technical Specifications from a human factors point of 0iew.
B.'URPOSE OF THE TECHNICAL SPECIFICATION The purpose of T.S. 3.2.4 is to ensure the limitation of DNBR, as a function of AXIAL SHAPE INDEX, will be within the conservative envelope of operating conditions consistent with the safety analysis assumptions and which have been analytically demonstrated adequate to maintain an acceptable minimum DNBR throughout all anticipated operational occurrences., Operation of the core with a DNBR at or above this limit provides assurance that an acceptable minimum DNBR will be maintained in the event of a loss of flow transient.
The purpose Safety Features Actuation action and/or reactor trip 'ill of T.S. 3.3.1 is to ensure that (1) the associated Engineered be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.
fl 0
I'l 0
C. NEED FOR THE TECHNICAL SPECIFICATION AMENDMENT The proposed changes are due to (1) ensuring operation of the reactor within approved safety analysis for Cycle 2 by modifying the T AS. graphs, (2) increasing operator reliability by placing DNBR operating'limits in one place, and (3) eliminating superfluous information to reduce confusion and the possibility of misuse. (i.e., eliminating the Table in T.S. 4.2.4.4)
D. BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would no't: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2),Create the possibility of a new or different kind of accident'from any accident'reviously evaluated; or (3) Involve a significant reduction in a margin of safety.
A discussion of these standards as they relate to the amendment request follows:
Standard 1--Involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change to the graphs of T.S. 3.2.4 does not involve a significant increase in the probability or consequences of an accident previously evaluated because the Cycle 2 safety analyses have shown that when COLSS is in service and at least one CEAC is operable, Specification 3.2.4a provides enough margin to DNB to accommodate the limiting Anticipated Operational Occurrence (AOO) without violating the Specified Acceptable Fuel Design Limits (SAFDL). For the case when neither CEAC is operable but COLSS is in service, the CPCs assume a preset CEA configuration because they can not obtain the required CEA position information to ensure that the SAFDL or DNBR will not be violated during an AOO. Thus, as a result of the reevaluation of the limiting AOOs for Cycle 2, Specification 3.2.4.b requires that core power be reduced to a value, (based on Figure 3.2-1) less than the current COLSS calculated power operating limit. This ensures the limiting AOO will not result in a violation of SAFDLs. The proposed revision to Figure 3.2-2 accounts for the situation when COLSS is out-of-service but at least one CEAC is operable. In this case, the Cycle 2 safety analysis has shown that, by maintaining the CPC calculated DNBR above the value shown in the figure, the limiting AOO will not result in a violation of the SAFDLs. When COLSS is out of service and both CEACs are inoperable, there must be additional margin to DNB set aside in the CPCs to ensure they can mitigate the consequences of the limiting AOO. A reevaluation of the limiting transients performed as part of the Cycle 2 safety analysis has shown that, by maintaining the CPC calculated DNBR above the limits shown in the proposed Figure 3.2-2a, there is sufficient thermal margin to ensure that the limiting AOO will not result in a violation of the SAFDLs. Therefore, the proposed change will not significantly increase the probability or consequences of any accident previously evaluated.
I The proposed change to the format of T.S. 3.2.4 and 3.3.1 does not involve a significant increase in the probability or consequences of an accident previously evaluated because consolidation of the DNBR operating limits within one Technical Specification will increase the operator's ability to ensure proper operation of the reactor. The proposed format change still contains the same Limiting Conditions for Operations (LCO),
ACTIONS and surveillance requirements as the original Technical Specifications. Therefore, the change will not significantly increase the probability or consequences of any accident previously evaluated.
The proposed change to eliminate the DNBR penalty factors table of T.S.
4.2.4.4 does not involve a significant increase in the probability or consequences of an accident previously evaluated because the penalty is an allowance for rod bow and has been incorporated into the DNBR value for Cycle 2. This can be done because the burnup of the reactor core in Cycle 2 will reach the value for applying the maximum rod bow penalty and the table will no longer be needed (see Attachment 12). Therefore, the change will not significantly increase the probability or consequences of any accident previously evaluated.
Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change to the graphs of T.S. 3.2.4 will not create the possibility of a new or different kind of accident from any accident previously evaluated because operation of the reactor within the limits as set forth in the graphs ensures that the reactor will not exceed, the SAFDLs as defined for the reference cycle (Cycle 1) during Cycle 2.
Therefore, the possibility of a new or different kind of accident from any accident previously evaluated will not be created.
The proposed change to the format of T.S. 3.2.4,and 3.3.1 will not create the possibility of a new or'ifferent kind of accident from any accident previously evaluated because the proposed change reduces the possibility of human error by consolidating closely related allowable operations into a single entity and by clearly identifying each allowable operation. The contents of the proposed T.S, are the same as those of T.S. 3.2.4 and 3.3.1, thus, the only change is in regard to the human factors element.
Therefore, by keeping the same contents but arranging them so as to reduce human error, the proposed change will not create the possibility of a new or different kind of accident not previously evaluated.
The proposed change to eliminate the DNBR penalty factors table of T.S.
4,2.4.4 will not create the possibility of a new or different kind of accident from any accident previously evaluated because the possibility of misusing the table is eliminated.
'l ir
Standard 3--Involve a significant reduction in f
k a margin of safety.
The proposed change to the graphs of'.S. 3.2.4 does not involve a significant reduction in a margin of safety because the change is to ensure that there will always be sufficient margin to DNBR such that the CPCs can mitigate the consequences of violating the SAFDLs. Figures 3.2-1, 3.3-2, and 3.2-2a represent a conservative envelope of operating conditions for the CPCs and COLSS which is consistent with Cycle 2 safety analysis assumptions. This band of operating conditions has been analytically demonstrated to maintain an acceptable minimum DNBR throughout all AOOs ~ Therefore, the proposed change does not reduce the margin of safety.
The proposed change to the format of T.S. 3.2.4 and 3.3.1 does not involve a significnat reduction in a margin of safety because the contents of the Technical Specifications have remained the same, only a rearrangment of information has taken place. Therefore, the proposed change does not reduce the margin of safety.
The proposed change to eliminate the DNBR penalty factors table of T.S.
4.2.4.4 does not involve a significant reduction in a margin of safety because the maximum rod bow penalty factor has been applied to the DNBR value for Cycle 2 and, therefore, the table is no longer needed and the margin of safety has been maintained for Cycle 2.
2. The proposed amendment matches the guidance concerning the application of standards for determining whether or not a significant hazards consideration exists (51 FR 7751) by examples:
(i) A purely administrative change to Technical Specification: for example, a change to achieve consistency throughout the Technical Specifications, in correction of an error, or a change in nomenclature.
and For a nuclear power reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found previously acceptable to the NRC for a previous core at the facility in question are involved. This assumes that no significant changes are made to the acceptable criteria for the Technical Specifications, the analytical methods used to demonstrate conformance with the Technical Specifications and regulations are not significantly changed, and that NRC has previously found such methods acceptable.
SAFETY EVALUATION FO THE AMENDMENT RE UEST The proposed Technical Specification amendment will not increase the probability of occurrence or the consequences of an accident or a malfunction of equipment important to safety previously evaluated in the FSAR. The proposed change to the graphs of T.S. 3 '.4 ensures that the reactor will be operated within a conservative envelope of operating conditions, consistent with the safety analysis, during Cycle 2, thus ensuring no increase in the probability of occurrence or the consequences of an accident or malfunction.
0 The changes to the format of T.S. 3.2 ' will increase the operator's ability to ensure correct operation of the reactor by consolidating related operation requirements into one Technical Specification. Because the change does not change the LCO, ACTIONS or surveillance requirements only the manner of presentation, no increase in the probability of occurrence or the consequences of an accident or malfunction will be experienced. The proposed change to eliminate the DNBR rod 'bow penalty factors table of T.S. 4.2.4.4 reduces confusion since the table is no longer needed. Because the maximum rod bow penalty factor has been incorporated into the Cycle 2 DNBR value no increase in the probability of occurrence or the consequences of an accident or malfunction will be incurred when the table has been deleted.
The proposed Technical Specification amendment will not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR. The proposed changes to the graphs of T.S. 3.2.4 ensure the operation of the reactor, during Cycle 2 operation, to be within the same limits as for Cycle 1 ~ Therefore, the possibility for an accident or malfunction of a different type will not be created. The proposed changes to the format of T.S. 3.2.4 do not change the LCOs, ACTIONS or surveillance requirements of the T.S., only the manner of presentation, thus the change does not create the possibility of an accident or malfunction of a different kind to occur. The proposed change to eliminate the rod bow penalty factors of T.S.
4.2.4.4. removes information no longer needed or necessary. A maximum rod bow penalty has been applied to the DNBR value, therefore, the change will not create the possibility for an accident or malfunction of a different kind to occur.
The proposed Technical Specification amendment will not reduce the margin of safety as defined in the basis for the Technical Specifications. The proposed changes either ensure sufficient margin will be maintained or do not change LCOs, actions or surveillance requirements required to maintain the margin of safety. Therefore, the margin of safety is not reduced.
ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental question because operation of PVNGS Unit 2, in accordance with this change, would not:
1. Result in a significant ,increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the staff's testimony to the Atomic Safety and Licensing Board; or
2. Result in a significant change in effluents or power levels; or
3. Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.
t I
f.
I MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES Limiting Conditions For Operation And Surveillance Requirements:
XIX 3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-7a 3/4 3-7 3/4 3-8 3/4 3-9 3/4 3-10 B 3/4 2-3 B 3/4 1-6
INDEX LIST OF FIGURES PAGE 3.1-1A SHUTDOWN MARGIN VERSUS COLD LEG TEMPERATURE............ 3/4 1-2a
3. 1-1 ALLOWABLE MTC MODES 1 AHD 2........;. 3/4 1-5
3. 1-2 MINIMUM BORATED WATER VOLUMES 3/4 1-12
3. 1-2A PART LENGTH CEA INSERTION LIMIT VS THERMAL POWER....... 3/4 1-23 3/4 1-24
~
3. 1-2B CORE POWER LIMIT AFTER CEA DEVIATION...
3. 1-3 CEA INSERTION LIMITS VS THERMAL POWER (COLSS IN SERVICE)............................. 3/4 1-31
3. 1-4 CEA INSERTION LIMITS VS THERMAL, POWER (COLSS OUT OF SERVICE). 3/4 1-32 Py ~cr Qllgwan,c~ 708 %07lQ
3. 2-1 ONER NARRNI OFERATI LINIT
~ ~ ~ ~ ~ ~ ~ I~ ~ ~ ~ ~ ~ ~ ~ 3/4 2-6 C.c.Rc'~ lNotS aabm
3. 2-2 LEE OF RERUIOEl................. C,iHi7 t.the I/O II
~Dhole g, gl cbhc, opgg.gg~)
3. 2-3 REACTOR COOLANT COI D LEG TEMPERATURE VS CORE POWER LEVEL . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \ \ ~ ~ 0 ~ J ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 2-10
3. 4-1 I DOSE EQUIVALENT "131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY
> 1.0 pCi/GRAN DOSE EQUIVALENT I-131................... 3/4 4-27
3. 4-2 REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITATIONS FOR 0 TO 10 YEARS QF FULL POWER OPERATION. 3/4 4"29
4. 7-1 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST.............. 3/4 7-26 B 3/4.4-1 HIL-DUCTILITYTRANSITION TEMPERATURE INCREASE AS A FUNCTION OF FAST (E ) 1 MeV) NEUTRON FLUENCE (550 F IRRADIATION) ...........-..-- ~ .. ~ -....<<.. B 3/4 4-10
5. 1-1 SITE AND EXCLUSIOH BOUNDARIES ....... ..........- 5-2
5. 1" 2 LOW POPULATION ZONE 5-3
5. 1-3 GASEOUS RELEASE POINTS.... 5-4
6. 2-1 OFFS ITE ORGANIZATION 6-3
6. 2-2 ONSITE ORGANIZATION . 6-4 PALO VERDE - UNIT 2 XIX AMENOMEHT HO.
Z~ Ca~a 0M~ 0P ggPgl(.g i>t4$ ll. I lH/7 l lUlK
(
0 0
POWER CGNTRGLLED BY USER DISTRIBUTION LIMITS 3/4. 2. 4 DN MARGIN LIMITING ONDITION FOR OPERATION 3.2.4 The DNBR ar'gin shall be maintained by operating within the Re n of Acceptable.0perat n of Figure 3.2-1 or 3.2"2, as applicable, or in ccordance with the requireme s of, Action 6 of Table 3.3-1. t APPLICABILITY: MODE abo e 20K of RATED THERMAL POWER.
With operation outside of the region of acceptable operati , as dicated by either (1) the COLSS calculated core power exceeding the LSS lculated core power operating limit based op DNBR,; or (2) when the CO S is ot being used, any OPERABLE Low DNBR 'channel below<the DNBR limit, wi in minutes initiate corrective action to restore ei'tger the DNBR core pow r op rating limit or the DNBR tn within the limits anl~eithen:
~
a. Restore the DNBR core power operating li it r DNBR to within its limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or
b. Reduce THERMAL POWER to less~than or q 1 to.20K of RATED THERMAL POWER within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; SURVEILLANCE RE UIREMENTS
.4.2.4.3 'he provisions 'of 'Speci-ficatio "4..4 'are% no't applicable.'
4.2.4.2 The DNBR shall be determined to' w'ithin itts limits when THERMAL POWER is above 20Ã of RATED THERMAL OWgd by continuously monitoring the core power distribution with the C e )crating>Limit~Supervisory System (COLSS) or, with the COL'SS out of ser ice, by verifying, at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months that the DNBR margin, a ind cated on all~OPERABLE DNBR margin
channels, is within the limit ow on Figure 3.2-2.
4.2.4.3 At least once per '31 day , the COLSS Margintglarm>,shall be verified to actuate at a THERMAL POWER 1 el less than or equal to the core power operating limit based on D BR.
4.2.4.4 The following D R qr equivalent penalty factors shal<l be verified to ~ ~
be included in the COLS and~CPC DNBR; calculations at lea'st once per 31 EFPD.
<cwo~
Bur nu MTU DNBR Pena t (I)"
0-10~ 0.5 10~%0 1.0 20-30 2.0 0-.40 e 3.5 40-50 5.5 "The penal for/each batch will be determined from the batch's max> m burnup assembly and +plied to the batch's maximum radial power peak assemb . A single et penalty for COLSS and CPC will be determined from the penalties associ ted with each batch accounting for the offsetting margins due to the lower radial power peaks in the higher burnup batches.
PALO YERQE - UNIT 2 3/4 2-5 ZGNTRGLLED BY USER
POWER DISTRIBUTION LIMITS
, 3/4. 2. 4 OHBR MARGIN LIHITIHG CONDITIOH FOR OPERATION 3.2.4 The OHBR margin shall be maintained by one of the following methods:
a. Maintaining COLSS calculated core power less than or equal to COLSS calculated core power operating limit based on DHBR (when COLSS is in service, and either one or both CEACs are operable); or
b. Maintaining COLSS calculated core power less than or equal to COLSS calculated core po~er operating limit based on DHBR decreased by @e ~ltEnBra~m S~ 3~ Figs~>.~-j is operable); or (when COLSS is in service and neither CiAC c Operating within the region of acceptable operation of Figure 3.2-2 using any operable CPC channel (when COLSS is out of service and either one or both CEACs are operable); or
d. Operating within the region of acceptable operation of Figure 3.2-jg using any operable CPC channel (when COLSS is out of service and ~
neither C EAC is operab1 e).
APPLICABILITY: HQOE'1 above 20~ of RATED THERMAL POWER.
ACTION:
With the DHBR not being maintained:
1. As indicated by COLSS calculated core power exceeding the appropriate COLSS calculated power operating limit; or
2. With COLSS out of service, operation outside the region of acceptable operation of Figure 3. 2-2 or ' 3.2-g, as applicable;
~ ~
.M within 15 minutes inititate correc.ive ac.ion to 'increase the DNBR to within the limits and either:
a. Restore the OHBR to within its limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or
~~~At
'b.
SURVEI LLANC
~~ <~SS REOUIREMEHTS Wchg ar w
g~) FB
+o ZCg of hO~N S.
R~ Rdh Ce
~m~L'P~g ~s442~
4.2.4.1 The provisions of Specification 4.0.4 are not applicable.
4.2.4.2 The ONBR shall be determined to be within its limits when.THERMAL POWFR is above 20 of RATiD THER."NL POWER by continuously 'monitoring he'ore "-
distribution with the Core Operating Limi Supervisory Sys em (COLSS) "'ower or, with the COLSS out of service, by verifying a leas. once per-2 hours tha-
, the ONBR, as indicamu on any OpERABLE ONBR cnanne!, is within the'licit'sho~n on Figure 3.2-2 or Figure 3.2$ .
BRss 4.2.4.3 At leas once per 31 days, 'the COLSS Margin'A1am "5'ha?l~ "verified:: '.= '=
j g'/S
.-to actuate at a THERMAL POWER level less than or e'qua'l Xa h'e'ore'power operating limit based on OHBR.
~LWt& - VNI 2-
J 0
CGNTRGLLED BY USER 100 REGlON QF O ACCEPTABLE OPER'ATION 80
.SQ U
'Z I w bEil 60 EI O g K 0 I
I, QK 0 EGIO OF LV CC Lit 40 , UNACCEPTABLE OP ERAT lON o ~
~
CO cn ~
Pg I g b
~z0 b 20 0
0 20 ao 6o 80'"'x 100 PERCENT OF RATED THERMAL POPPER, FIGURE 3.2"1 OMBRE RGIN OP RATING LIMIT BASEO ON COLSS (COLSS IN SERVICE)
/
PALO VERDf - UNiT 2 3/4 2"6 CGNTRGLLED BY USER
Oi COLSS DNBR POWER OPERATING LIMIT REDUCTION (i! OF RATED THERMAL POWER)
PV Ol n
Ol C)
Ul Sl U
C)
Cl1 O oM C3 Pl CJl Wo FIGURE 3.2-1 COLSS DNBR POWER OPERATING LIMIT ALLOWANCE FOR BOTH CEAC'S INOPERABLE PALO VERDE - UNIT Z. 3 /4 2-6
0 k
CONTRQLLED BY USER 0.60 I I
.::: RE l. NOF
'::: ACCEPTABLE II
(
I ~ ~
~ \
0.55 j'.'PERATION~+
I ~ ~ ~ ~
~ ~ ~ 4 1 ~
g '(:,.05, 0.51):.. :(2S,OS1)
I ~
I I
I 0.50 I
~ 1 ~
U ~ ~ I I
I
~ ~ ~ ~ ~
t I:..: (>0, 0.46) g 0.45
~ ~
l..REGlON~EF Z >, UNACCEPTABL'E z
':
')~OPEJAT1ON 0.40 I
s
(-.30, 038): ~ ':: ' z ~
'L I
~:'::"I
~ ~
0.35 i
~ ~ 'I I
I/
0.30
~ 0.3 ~
.2. / .0.1 0,0 0.1 0.2 4
0.3 CORE AVERAGE AS1 SEE SECTlON 32.7 FOR THE ASI OPERA'TING LIMlTS FIGURE 3.2"2'NBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE)
PAI.O VERDE - UNIT 2 3/4 2-7 CONTROLLED BY USER
COLSS OUT OF SERVICE DNBR LIMIT LINE 2.1 ACCEPTABLE 2.8 OPERATION MINIMUM 1 CEAC OPERABLE
(.1,1.85) (.2,1.85)
(-.2,1.75) 1.7 UNACCEPTABLE OPERATION 1.6 1.5
-8.3 -8.2 -8.1 8.1 - 8.3 CORE AVERAGE ASI FIGURE 3.2-2 DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE, CEAC'S OPERABLE)
PALO VERDE - UNIT 2. 3 /4 2-7
~ j 0'
COI SS OUT OF SERVICE DNBR LIMIT LINE 2.4 ACCEPTABLE OPERATION I
2.3 CEACs INOPERABLE ( 85 2 38) (.2,2.38) 2.2 X
Z 21 (-.2,2.13)
CL UNACCEPTABLE OPERATION 2.8 1.9
-8.2 -8.1 8.1 8.2 8.:.
. CORE AVERAGE ASI F IGURE 3.2-2a DNBR MARGIN OPERATING LIMIT BASEO QN CORE PROTECTION CALCULATOR',
(COLSS OUT OF SERVICE,CEACs INOPERABLE)
PALO VEROE-UNIT 2 3/4 2-7a
CGNTRGLLED BY USER REACTOR PROTECTIVE INSTRUMENTATION ACTION STATEMENTS
. 3. 'team Generator Pressure .- Steam Generator Pressure - Low
~ I Low Steam Generator Level 1-Low (ESF)
Steam Generator Level 2-Low (ESF)
4. Steam Generator Leve'. - Low Steam Generator Level - Low (RPS)
(Mide Range) Steam Generator Level 1-Low (ESF)
Steam Generator Level 2-Low (ESF)
5. Core Protection Calculator Local Power Density - High (RPS)
DNBR - Low (RPS)
STARTUP and/or POMER OPERATION may continue until the performance of the next required CHANNEL FUNCTIONAL TEST. Subsequent STARTUP restored to and/or POMER OPERATION OPERABLE may continue if one channel is status and the provisions. of ACTION 2 are satisfied.
ACTION 4 Mith the number of channels, OPERABLE one less than required by the Minimum Channels OPERABLE requirement, suspend all operations positive reactivity changes. 'nvolving ACTION 5 Mith the number of channel,s OPERABLE one less than required by the minimum Channels OPERABLE .requirement, STARTUP and/or POMER OPERATION may continue provided,.the reactor trip breaker..
~ f ~
'. of the inoperable channel is placed in'the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ,- otherwise, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, the trip breaker associated with the inoperable channel may be closed for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for surveillance testing per Specification 4.3. 1. 1.
ACTION 6 a. Mith 'one CEAC inoperable, operation may continue for up to 7 days provided that at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , each CEA is verified to be within 6.6 inches (indicated position) of all other CEAs in its group. After 7 days, operation may continue provided that the conditions of Action Item-6.b m~ are met.
'
b. Mith both CEACs inoperable , operation may continue provided that:
P. 'in our:
'
a) Opera is res icted to t limits s wn in Fi e 3.3-1. he DNBR m in requir by ecificati 3.2.e is placed by is restricti when bot EAC's are noperabl
') and COL The S
near cificati is in He o
3.2. 1 ation.
Rate Mar i
'equired aintained.
c) The Re or Power tback Syst is place out of service.
PALO VERDE - UNIT 2 3/4 3-7 gGNTROLLEB BY USER
I
~ i
CONTRQ]LE9, BY USER REACTOR PROTECTIVE INSTRUMENTATION 5~0 ACTION STATEMENTS
2. Within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:
a) All full-length and part-length CEA groups are 4
.
withdrawn to and subsequently maintained at the
~ III "Full Out" position, except during surveillance P testing pursuant to the requirements of Specifica-tion 4. 1.3. 1.2 or for control when CEA group 5 may be inserted no further than 127.5 inches withdrawn.
b) The "RSPT/CEAC Inoperable" addressable constant in the CPCs is set to be indicated that both CEAC'.s are inoperable.
c) The Control Element Drive Mechanism Control System (CEDMCS) is placed in and subsequently maintained in the "Standby" mode except during CEA group 5 'motion permitted by a) above, when c~ the CEDMCS may be operated in either the "Manual Group" or "Manual Individual" mode.
3. At least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , all full-length and part-
'enqth CEAs are verified fully withdrawn except during surveillance testing pursuant to Specification
...;:., 4.1.3. 1.2 or .during insertion sf..CEA .group -5 .'as ...
permitted by 2.a) above, then.'verify at least once '*
" per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months that the inserted CEAs are aligned within 6.6 inches (indicated position) of all other CEAs in its roup.
4. Followin a C A misalignment with both AC s inoper e and COLSS i operation, operation may cont ue provided th within 1 hour:
T power is red ed to 85K of the pre-misali ed ower but nee ot be reduced to less th of RATED THERt POMER. This power res ction replaces the powe restriction of Specif ion 3.1.3.1,
. Figur . 1-2B, otherwise Sp 'cation 3. 1.3. 1 remains app cabj e.
C. With oth CEACs inoperab and COLSS out-of-servic o ation may'contin provided that:
Mithin 1 h~o a) d existing CPC value of e CPC addressable constant "BERR1" is mu pled by 1. 19 and the resulting value i~s -entered. into the CPCs.
b) The Reactor P dr Cutback System is placed out of service c) The CO out of service Limit Line n Fig-ure 3.2-2 of Specification 3.2. is not appli-cable to this mode of opera VERDE - UNIT 2 3/4 3-8
'ALO CQiNTROLLED BY USER
0 CONTRQLLED BY USER REACTOR PROTECTIVE INSTRUMENTATION ACTION STATEMENTS .
'\
2. Within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:
a) All full length and part length CEA groups are withdrawn to and subsequently maintained at the "Full Out" position, except during surveillance testing pursuant to the requirements of Specifi-cation 4. 1.3. 1.2 or for control when CEA group 5 may be inserted no further than 127.5 inches withdrawn.
b) The "RSPT/CEAC Inoperable" addressable constant in the CPCs is set to be indicated that both CEAC's are inoperable..
c) The Control Element Drive Mechanism Control System (CEDMCS) is placed ih and subsequently maintained in the "Standby" mode except during b,i~4.h, CEA group 5 motion permitted by a) .above, when the CEDMCS may be operated in either the "Manual
'. Group" or "Manual Individual" mode.
- At least once>er 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,-:all-full-'-length 'and length CEAs are verified fully withdrawn except part during surveillance testing pursuant to Specifica-tion 4. 1.3. 1.2 or during insertion of CEA group 5 as permitted by 2.a) above, then verify at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months that the inserted CEAs are aligned inches (indicated position) of all other CEAs in within'.6 its group.
. 4. Following a CEA misalignment with both CEAC's and COLSS" inoperable, operation may continue provided that within 1 hour:
The power is reduced to 85K of 4he pre-misaligned power but need not be reduced to less than 50K of RATED THERMAL POWER. This power restriction replaces the power restriction of Specification 3. 1.3. 1, Figure. 3. 1-2B, otherwise Specification 3. 1.3. 1 remains applicable.
ACTION 7 - With three or more auto restarts, excluding periodic auto restarts (Code 30 and Code 33), of one non-bypassed calculator during a 12-hour interval, demons'trate calculator OPERABILITY by performing a CHANNEL FUNCTIONAL TEST within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
ACTION 8 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore an inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open an affected reactor trip breaker within the next hour.
PALO VERDE - UNIT 2 3/4 3-9
CQNTRGLLED BY USiER 140
~ ~ ~ ~
I I
I~
i: I .
j
>
1
~
I': .I.
I: I
~ c I ~.
120 ~
(100, 118.7)
~
t
'A 11 .7 '95
~ ~ ~ ~ I zCI '.
z ~ oX l 0 100.
W r
O
'
0 REGION OFy ~ ~
CQ R j
OPERATION
~ ~
Pg 80 (79.4, 79.4) -."
U
'- a z I
~ ~ ~
K 4
0 o 60
~ ~ ~
q ~
Q QJ ~
O O
~ ~ ~, ~
REGION OF 40 'NACGEPTAB E "
OPERAT ON O
~ ~
20 20 40 60 80 100 PERCENT OF RATED THERMAL POWER FIGURE 3. 3-1 DNBR MARGIN OPERATING LIMIT BASED ON COLSS FOR BdTH CEACs INOPERABLE PALO VERDE " UNIT 2 3/4 3-10
~,
POWER DISTRIBUTION LIMITS BASES AZIMUTHAL POWER TILT - T (Continued) o t,.lt t.lt is, the ratio of the power at a core location in -.he presence of a tilt to the power at that location with no tilt.
The AZIMUTHAL POWER TILT allowance used in the CPCs is defined as t.".e value of CPC addressable constant TR-1.0.
3/4.2.4 DNBR MARGIN The limitation on DNBR as a function of AXIAL.SHAPE INDEX represents a conservative envelope of operating conditions consistent with the safety analy-sis assumptions and which have been analytically demonstrated adequate t" main-tain an acceptable minimum ONBR throughout all anticipated one. ational oc"ur-rences, 0 era= on of the core with a ONBR at or above this limit provides assurance that an acceot-able minimum ONBR will be maintained in the event of a loss of flow -.rans',en',.
Either of the two coi'e power distribution monitorina systems, .ne C"re Operating Limit Supervisory System (COLSS) and the ONBR cnannels in '-ne are Protection Calculators (CPCs), provide adequate monitorina of =ne core ocwer distribution and are capable of verifying that the ONBR aoes not viola=e .'-s limits. The COLSS performs this function by continuously moni:orino one =ore
...... power di stribution'and .calculating-a-core.aoperating+imi t--corresponaina co ~he - -,-
a Pi,g 'llowable minimum ONBR.
1 The C"LSS
'alculation of core power operating limit based on ONBR incluces apprcpri te
.
penalty factors which provide, with a 95/95 probability/conficence level, :hai the core power limits calculated by COLSS (based on the minimus DNBR L;m..'=) is conservative with respect to the actual core power limit. These penalty .ac Grs are determined from the uncertainties associated with planar r dial peak;.".g measurement, engineei.ing heat flux, state parameter measuremer.-., so==ware algorithm modelling, computer processing, rod bow, and core power measuremlent.
8,2.% ~~b, 3,2.,z.~
Parameters required to maintain the margin to ONB and total core po~er are also monitored by the CPCs. Therefore, in the event .hat =he CGLSS .:s not being used, operation within the limits of Figuru~~ can be maintainec by utilizing a predetermined DNBR as a function of AXIAL SHAPE INDEX and by monitoring the CPC trip channels. The above listed uncer',ainty and penalty factors are 'also included in the CPCs which assume a minimum core power of 20..
'..
of RATED THERMAL POWER. The 20Ã RATED THERMAL POWER threshold is due to the neutron flux detector system being ~i@accurate be'low 20/. core power. Core noise level at low power is too lar)q to obtain usable detector r eadinqs. Na
~ ' ~q bcc.e wcI~a. i~ eke.c.ecsa -c. d. c-ac I'can RI fNe. DNBR penalty factor) e c Ic aIIn to accommodate the effects of rod bow. The amount of rod bow in eacn assembly is dependent, upon the average burnup experienced by tnat assembly. Fuel assemblies that incur higher average burnup will experience a greater magnitude of rod bow. Conversely, lower burnup assemblies will experierce less roo bow%~~<<q<
..
co.)c,~4 ~ 'the penalty for each batch required to compensate for rod bow is determined from a batch's maximum average assembly burnup applied to the batch's maximum inte-grated planar-radial power peak. A single net penalty for COLSS and CPC is then determined from the penalties associated with each batch, accounting for 'the off-setting margins due to the lower radial power peaks in the higner burnup batches.
PALO VERDE - UN B, 2.
ii CQNTRQLLED BY USiER REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES (Continued) and load maneuvering. Analyses are performed based on the expected mode of operation of the NSSS (base load maneuvering, etc.) and from these analvse CEA insertions are determined and a consistent set of radial peaking factors defined. The Long Term Steady State and Short Term Insertion Limits are deter-mined based upon the assumed mode of operation used in the analyses and provide a means of preserving the assumptions on CEA insertions used. The limits speci-fied serve to limit the behavior of the radial peaking factors within the bounds determined from analysis. The actions specified serve to limit the extent of radial xenon redistribution effects to those accommodated in the analyses. The Long and Short Term Insertion Limits of Specification 3. 1.3.6 are specified for the plant which has been designed for primarily base loaded ooeration. but which has the ability to accommodate a limited amount of 1'oad maneuvering.
The Transient Insertion Limits of Specification 3. 1. 3. 6 and the Shutdown CEA Insertion Limits of Specification 3. 1.3.5 ensu~e 'that (1) the. minimum SHUT-00WN MARGIN is maintained, and (2) the potential effects of a CEA ejection accident are limited to acceptable levels. LAg-term operation at the Tran-
'sient Insertion Limits is not permitted since suc'h operation could have effects on the core power distribution which could invalidate assumptions used to deter-mine the behavior 'of the radial"peaking factors.
bhe PYNGS CPC and COLSS systems are responsible for the safety and monitoring functions, respectively, of the reactor core. COLSS monitors the DNB Power Operating Limit (POL) and various operating parameters to help the operator main-tain plant operation within the limiting conditions for operation (LCO). Operat-ing within the LCO guarantees that in the event of an Anticipated Operational Occurrence (AOO), the CPCs will provide a reactor trip in time to prevent un-acceptable fuel damage.
The COLSS reser awk. l'GA M> Sop~h ~
the Required Overpower Margin (ROPM) to account for the Loss of Flow (LOF .transien4 When the COLSS is Out of Service (COOS), the monitoring function is performed via the CPC calculation of ONBR in conjunction with p'echnical Specification COOS Limit Lines(Figures3.2-2) which restricts the reactor power sufficiently to preserve the ROPM, The reduction of the CEA deviation penalties in accordance with the CEAC (Control Element Assembly Calculator) sensitivity reduction program has been performed. This task involved setting many of the inward single CEA deviation penalty factors to 1.0. An inward CEA deviation event in effect would not be accompanied by the application of the CEA deviation penalty in either the CPC DNB and LHR (Linear Heat Rate) calculations for those CEAs with the reduced penalty factors. The protection for an inward CEA deviation event is thus accounted for separately.
""""'CQNTRGLL&5YUSER
4 ATTACHMENT 12 A. DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE UEST The proposed amendment change expands the operating limits of Azimuthal Tilt with COLSS in service. The azimuthal tilt limits will be a step function of power with the upper limit of 0.20 at 20$ power and stepping down to 0.10 at 40% power, where it remains steady through to 100% power.
B. PURPOSE OF THE TECHNICAL SPECIFICATION The limitations on the Azimuthal Power Tilt are to ensure that design safety margins are maintained.
C. NEED FOR THE TECHNICAL SPECIFICATION AMENDMENT During a reactor power cutback event in Unit 1 the plant was unable to go above 20% power because the azimuthal tilt limit would have been exceeded.
They were required to remain below 20% power for approximately 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months until xenon burned out. This'elay could have been prevented and the azimuthal tilt corrected if the plant had been allowed to increase power. This would cause the xenon to burn out faster thus restoring the plant within the limits sooner. By imp lementin g the P ro P osed chan g e such dela s could be avoided.
D. BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION
1. The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
A discussion of these standards as they relate to the amendment request follows:
Standard 1--Involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because a reevaluation of the safety analysis pertaining to azimuthal tilt was conducted and the results of the reanalysis show that for the conditions of azimuthal tilt as defined in the new Figure 3.2-1A the safety analysis of the referenced cycle (Cycle 1) is bounding. Therefore there is no change to the probability or consequences of an accident previously evaluated in the FSAR.
0 Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated. The results of the reanalysis were found to be bounded by the reference cycle safety analysis. Relaxing the azimuthal power tilt limit at lower power levels will not create any new or different kinds of accidents.
Standard 3--Involve a significant reduction in a margin of safety.
The proposed change does not involve a significant reduction in the margin of safety.
limits and it A reanalysis was performed using the proposed was found that the results of the reanalysis were bounded tilt by the reference cycle safety analysis. Therefore the margin of safety is maintained.
2. The proposed amendment matches the guidance concerning the application of standards for determining whether or not a significant hazards consideration exists (51 FR 7751) by example:
(vi) A change which either may result in some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan: for example, a chan g e resultin g from the a PP lication of a small refinement of a previously used calculation model or design method.
E. SAFETY EVALUATION FOR THE AMENDMENT RE UEST The proposed Technical Specification amendment will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. The proposed change does not change or replace equipment or components important to safety.
The change is bounded by the existing safety analysis and will not increase the probability of an occurrence or consequences of an accident.
The proposed Technical Specification amendment will not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR. By determining that the results of the reanalysis were bounded by the reference cycle safety analysis the field of accidents or malfunctions have not changed. Therefore there is no increase in the probability for an accident or malfunction of a different type than any previously evaluated in the FSAR.
The proposed Technical Specification amendment will not reduce the margin of safety as defined in the basis for the Technical Specifications. To determine the impact of the change to the azimuthal tilt limits, a reanalysis was performed. The results of the reanalysis were bounded by the reference cycle safety analysis and therefore the margin of safety has been maintained.
f ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental question because operation of PVNGS Unit 2, in accordance with this change, would not:
1. Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the staff's testimony to the Atomic Safety and Licensing Board; or
2. Result in a significant change in effluents or power levels; or
3. Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.
MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES Limiting Conditions For Operation And Surveillance Requirements:
3/4 2-3 B 3/4 2-2 3/4 2-4
INDEX LIST OF FIGURES PAGE
3. 1" 1A SHUTDOWN MARGIN VERSUS COLO LEG TEMPERATURE............ 3/4 1-2a
3. 1-1 ALLOWABLE MTC MODES 1 AND 2 3/4 1"5
3. 1-2 MINIMUM BORATED WATER VOLUMES. 3/4 1-12
3. 1-2A PART LENGTH CEA INSERTION LIMIT VS THERMAL POWER....... 3/4 1-23 3.1-28 CORE POWER LIMIT AFTER CEA DEVIATION..... 3/4 1-24
3. 1-3 CEA INSERTION LIMITS VS THERMAL POWER (COLSS IN SERVICE).................... 3/4 1-31
3. 1" 4 CEA INSERTION LIMITS VS THERMAL POWER i" la (COLSS OUT OF SERVICE).... 3/4 1"32 3, ALCMuVERL VOWtl. flan T 'llHTT V5 THRAHALPO~SK 'LC.OL.bb 14 %4LM I44) >/e i 3~ 2 1 DNBR MARGIN OPERATING LIMIT BASED ON COLSS (COLSS IN SERVICE}............ 3/4 2-6
3. 2-2 DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATOR (COLSS OUT OF SERVICE}......... 3/4 2-7.
3.2"3 REACTOR COOLANT COLD LEG TEMPERATURE VS CORE POWER LEVELYN ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 2-10 3 3~ 1 DNBR MARGIN OPERATING LIMIT BASED ON COLSS FOR BOTH CEAC'S INOPERABLE... 3/4 3"10
3. 4-1 DOSE E(UIVALENT I"131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY
> 1.0 pCi/GRAN DOSE EQUIVALENT I-131................... 3/4 4-27
3. 4-2 REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITATIONS FOR 0 TO 10 YEARS OF FULL POWER OPERATION..............,.... 3/4 4-29
4. 7" 1 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST.............. 3/4 7-26 B 3/4.4-1 NIL-DUCTILITYTRANSITION TEMPERATURE INCREASE AS A FUNCTION OF fAST (E > 1 MeV) NEUTRON FLUENCE (550 F IRRADIATION). B 3/4 4-10
5. 1-1 SITE AND EXCLUSION BOUNDARIES 5"2
5. 1-2 LOW POPULATION ZONE 5-3 5.1-3 GASEOUS RELEASE POINTS 5-4 6 2-1 OFFSITE ORGANIZATION . 6" 3
6. 2-2 ONSITE ORGANIZATION 6-4 PALO VERDE - UNIT 2 AMENDMENT NO. 13
0 POWER DISTRIBUTION LIMITS
- 3/4.2.3 AZEHUTHAL POWER LIHITING CONDITION TILT " T FOR OPERATION 3.2.3 The AZIHUTHAL POWER TILT (T ) shall be less than or equal to the AZIHUTHAL POMER TILT Allowance used in the Core Protection Calculators (CPCs).
APPLICABILITY: MODE 1 above 20io of RATED THERHAL POWER~.
ACTION:
ar With the measured AZIHUTHAL POMER TILT determined to exceed the The/ ~;9;~
to ~,
AZIHUTHAL POMER TILT Allowance used in within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either correct AZIHUTHAL POWER TILT A11owance used in the CPCs but less than or equal the power tilt or adjust the the CPCs to greater than or equal to the measured value.
F>grcrg 3,2 /P wig
(.'OLS5 In St ~Vi'Ce O3- Mith the measured AZIHUTHAL POWER TILT determined to exceed t5,jo wi t'4 CQL.SS 1. Due to misalignment of either a part-length or full-'length CEA, out o9 Sev viCt'. within 30 minutes verify that the Core Operating Limit Supervisory System (COLSS) (when COLSS is being used to monitor the core power distribution per Specifications 4.2. 1 and 4.2.4) is detecting the CEA misalignment.
2. Verify that the AZIMUTHAL POMER TILT is within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or reduce THERHAL POMER to less than 50io of RATED THERHAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and verify that the Variable Overpower Trip Setpoint has been reduced as appropriate within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
3. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above SOX of RATED THERHAL POMER may proceed provided that the AZIHUTHAL POWER TILT is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95/o Qr greater RATED THERMAL POMER.
~See Special Test Exception 3. 10.2.
3"
- PALO VERDE - UNIT 2 3/4 2-3 t=am~CLLED BY U<<~ r
0, POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.
4.2.3.2 .-The AZIMUTHAL POWER TILT shall be determined to be within the limit above 20K of RATED THERMAL POWER by:
gn Seruiee
a. Continuously monitoring the tilt with COLSS when the COLSS is 6PBRA&EE.
b.
i~~. o~t tilt Calculating the at least aR.service.
once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the COLSS is
c. Verifying at'east once per 31 days, that the COLSS Azimuthal Tilt Alarm is actuated at an AZIMUTHAL POWER TILT less than or equal to the AZIMUTHAL POWER EILT Allowance used in the CPCs.
d. Using the incore detectors at least once per 31 EFPD to independently confirm the validity of the COLSS calculated AZIMUTHAL POWER TILT.
PALO VERDE - UNIT 2 3/4 2-4
FIGURE 3.2 1A AZIMUTHALPOWER TILT LIMIT vs THERMAL POWER (coLss IN sERvIcE) 100 90 A
7 80 I
M U
T 70 H
A RELQN L
UNAGGEPT E OPERATI 0 SO VO 30 20 10 20 30 VO 50 60 70 80 90 100 PERCENT QF RATED THERMAL PQWER PALQ YERDE - UNIT 2
i CONTROLLED BY USER POWER DISTRIBUTION LIMITS BASES 3/4.2.2 PLANAR RADIAL PEAKING FACTORS Limiting tne values of the PLANAR RADIAL PEAKING FACTORS (F xy ) used :n the COLSS and CPCs to values equal to or gr eater than the measur ed PLANAR RAO'AL PEAKING FACTORS (F ) provides assurance that the limits calculated by COLSS xy and the CPCs remain valid. Data from the incore detectors are used -or determining the measured PLANAR RADIAL PEAKING FACTORS. A minimum core oower at 20% of RATED THERMAL POWER is assumed in determining the PLANAR RADIAL PEAKING FACTORS. The 20% RATED THERMAL POWER threshold is due to the neutron flux detector system being inaccurate below 20% core power. Core noise level at low power is too large to obtain usable detector readings. The periodic surveillance requirements for determining the measured PLANAR RADIAL PEAKING FACTORS provides assurance that the PLANAR RADIAL PEAKING FACTORS usea in COLSS and the CPCs remain valid throughout the fuel cycle.. Determining tne measured PLANAR RADIAL PEAKING FACTORS afte'r'ach fuel loading orior :o exceeding 70% of RATED THERMAL POWER, provides'dditional assurance tnat tne core was properly loaded.
3/4.2.3 AZIMUTHAL POWER TILT - T Q The limitations on the AZIMUTHAL POWER TILT are proviaed t ens.:.e tnat e
g
-
II-~,4 ore S,2-/1 design safety margins are maintained. An AZIMUTHAL POWER'TILT creater tnan R=& is not expected ahd if it should occur, operation is restr'ctaa -o only those conditions required to identify the cause of the tilt. The ti-- is
~
w C'OL g5 normally calculated by COLSS. A minimum core power of 20% of RATED '-:-"RMAL lN SCMVICQ POWER is assumed by the CPCs in its input to COLSS for calculation o-E,W O,IO ~%4 AZIMUTHAL POWER TILT. The 20% RATED THERMAL POWER threshold is due :o the COtSS OutLPV neutron flux detector system being inaccurate below 20,O core power. ore Strv <c8 noise level at low power is too large to obtain usable detector read'.ngs. The surveillance requirements specified when COLSS is out of service provide an acceptable means of detecting the presence of a steady-sta e tilt. :-t is necessary to explicitly account for power asymmetries because the racial peaking factors used in the core'power distribution calculations are oasea on an untilted power distribution.
The AZIMUTHAL POWER TILT is equal to Pt lt t'lt '0 ~here AZIMUTHAL POWER TILT. is measured by assuming that the ratio of the power at any core location in the presence of a tilt to the unti lted power at the location is of the form:
Pt'lt 1 + T g cos (e - eo)
P t lt where:
T q
is the peak fractional tilt amplitude at the core periphery g is the radial normalizing factor 8 is the azimuthal core location eo is the azimuthal core location of maximum tilt PALQ vERDE - UgiQQTRQLLEPDA2QY
0 ATTACHMENT 13 A. DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE VEST The proposed amendment ensures the Refueling Actuation Signal (RAS) trip value of the Refueling Water Storage Tank for recirculation is maintained at the midpoint of the allowable operational values by removing the "greater than" sign from the trip value as set forth in Technical Specification (T.S.) 3.3.2 Table 3.3-4.
B. PURPOSE OF THE TECHNICAL SPECIFICATION The purpose Safety of T.S. 3.3.2 Features Actuation
's action to ensure that (1) the associated Engineered and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.
Ih f C. NEED FOR THE TECHNICAL SPECIFICATION AMENDMENT The proposed change to T.S. 3.3.2 Table 3.3-4 will eliminate an abiquity concerning the level setpoint in relation to the allowable range.
D. BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION
1. The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
A discussion of these standards, as they relate to the amendment request, follows:
Standard 1--Involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change does not involve a significant increase in the probability or consequences of an'ccident previously evaluated because, by maintaining the RAS trip value at the midpoint of the allowable band, the proposed change is. more restrictive. This, in turn, limits the
0 l '
h operation of,, the 'Refueling Water Storage Tank such that a maximum assurance of protecting the pumps 'from cavitating is provided. Since the change is still within the limits of the allowable values, the possibility,, of consequences of an accident previously evaluated will not be increased. .1 i Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because, by maintaining the trip value at the midpoint of the allowable band, the proposed change is more restrictive. Since the change reduces the allowable values of the trip to a single value, which was part of the original safety analysis, the possibility of a new or different kind of accident from any accident previously evaluated will not be created.
Standard 3--Involve a significant reduction in a margin of safety.
The proposed change does not involve a significant reduction in a margin of safety because, by maintaining the trip value at the midpoint of the allowable band, the proposed change is more restrictive. By restricting the allowed operation of the Tank even further within the allowable trip values, the Unit does not experience as many possible accidents as before.
Therefore, the change will not reduce the margin of safety.
2. The proposed amendment matches the guidance concerning the application of standards for determining whether or not a significant hazards consideration exists (51 FR 7751) by example:
ii) A change that constitutes an additional limitation, restriction or control not presently included in the Technical Specifications: for example, a more stringent surveillance requirement.
SAFETY EVALUATION FOR THE AMENDMENT RE UEST The proposed Technical Specification amendment will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. The proposed change does not change or replace equipment or components important to safety.
The change only limits the allowable values of the trip to a single value and is more restrictive by maintaining the trip value at the midpoint of the allowable band. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR will not be increased.
The proposed Technical Specification amendment will not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR. The proposed change is more restrictive by maintaining the trip value at the midpoint of the allowable band. Since the change reduces the allowable values of the trip to a single value which was part of the original safety analysis, the possibility of a different accident or malfunction will not be created.
The proposed Technical Specification amendment will not reduce the margin of safety as defined in the basis 'for the Technical Specifications. The proposed change is more restrictive by maintaining the trip value at the midpoint of the allowable band. By restricting the allowed operation of the Tank even further within the allowable trip values, the Unit does not experience as many possible accidents as before. Therefore, the change will not reduce the margin of safety.
F. ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental question because operation of PVNGS Unit 2, in accordance with this change, would not:
1. Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as by the staff's testimony to the Atomic Safety and Licensing 'odified Board; or
2. Result in a significant change in effluents or power levels; or
3. Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.
MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES Limiting Conditions For Operation And Surveillance Requirements:
0 Ihlili 3. 3-4 ((:oni,iree(li CHGINCCREO snFETY FEnTURES nciunTION SYSIEH INSTRUHEH1ATIOH TRIP VALUCS ESFA SYSTEH FUHCTIOHAL UNIT TRIP VALUES ALLOMAOLE VALUES RL'C I RCULAT ION (RAS)
h. Seiisor/Irip Unils Refiieliiig Maler Storage Tank - Low . 7.4X of Span 7.9 > X of Span > 6.9
0. ESFA System Logic Not Applicable Hot Applicable C. Actuation System Not Applicable Not Applicable Vl. AUXILIARY FEEOMATER (SG-l)(AFAS-1)
h. Sensor/Tr i p Uni ls
l. Steam Generator ffl Level = Low > 25.OX MR(') > 25.3X MR
2. Steam Generator d Pressure- < 105. psid < 192 psid SG2 > SGl
0. ESFA System Logic Hot Applicable Hot Applicable C. Ac lua lion Sys lems Hot Applicable Hot Applicable VII. nuxILInRY FCEOMATFR (SG-2)(AFAS-2)
h. Sensor/Trip Units
l. Sleam Generator tl2 Level - Low (') X (')
2. Sleam Geneialor h Pressure- <=185 psid < 192 psid SGl > SG2
0. ESFA System Logic Hot Applicable Hol Applicable C. Actuation Syslems Not Applicable Hot Applicable VIII. LOSS OF POWER
h. I.I6 kV Emergency 0iis Undervoltage (l.oss of Voltage) > 3250 > 3250 vol ls volts'930
0. 4.16 kV Imergeiicy 0us Un(lervol tage to 3740 volts 2930 to 3744 volts (I)egrade<l Vol loge) . wilh a 35-second willi a 35- second maximiim lime de)ay maximum lime delay Ix. coHTRoI. RooH CssCNT InL FII.TRnTIoH < 2 x 10- Iici/cc < 2 x 10-s Iici/cc
ATTACHMENT 14 A. DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE VEST The proposed amendment is a number of administrative changes for the following Technical Specifications (T.S.):
Bases 3/4.3.1 and 3/4.3.2
1) page 3-2 remove Cycle 1 specific information no longer needed for Cycle 2 Bases 2.2.1
1) page 2-2 remove reference to CESSAR for description of the method of calculation for the trip variables for DNBR-Low and Local Power Density High trips and replace with the correct CE Topicals
2) page 2-3 update the latest revision used for calculating the PVNGS trip setpoint values B. PURPOSE OF THE TECHNICAL SPECIFICATION The purpose of T.S. 3.3 ' is to ensure that (1) the associated Engineered Safety Features Actuation action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.
C. NEED FOR THE TECHNICAL SPECIFICATION AMENDMENT The administrative changes are required to ensure clarity and conciseness.
The change to Bases 3/4.3.1 removes information which pertained to Cycle 1 and is no longer valid for Cycle 2. The change to Bases 2.2.1 changes the source of the description of the method of calculation for the trip variables for DNBR-Low and Local Power Density High trips from the CESSAR to the correct CE Topicals and updates the T.S. to the latest revision of CEN - 286 (V), Rev 2.
1 t
f D. 'BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION
1. The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment. to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
A discussion of these standards, as they relate to the amendment request, follows:
Standard 1--Involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the proposed changes are administrative in nature. They eliminate incorrect and superfluous information, thus ensuring that the Technical Specifications are concise and understandable. Therefore, the changes ensure that the possibility of an accident previously evaluated will not be increased.
Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes will not create'he possibility of a new or different kind of accident previously evaluated because the proposed changes are administrative in nature. They eliminate incorrect and superfluous information thus ensuring that the Technical Specifications are concise and understandable. Therefore, the changes ensure that the possibility of a new or different kind of accident from any accident previously evaluated will not be created.
Standard 3--Involve a significant reduction in a margin of safety.
The proposed changes do not involve a significant reduction in a margin of safety because the proposed changes are administrative in nature.
They eliminate incorrect and superfluous information thus ensuring that the Technical Specifications are concise and understandable. Therefore, the changes ensure that the margin of safety is maintained.
2 ~ The proposed amendment matches the guidance concerning the application of standards for determining whether or not a significant hazards consideration exists (51 FR 7751) by example:
(i) A purely administrative change to Technical Specifications: for example, a change to achieve consistency throughout the Technical Specifications, correction of an error, or a change in nomenclature.
0 0
E. SAFETY EVALUATION FOR THE AMENDMENT RE VEST The proposed Technical Specification amendment will not: increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. The proposed change does not change any equipment or components important to safety. The proposed changes are administ:rative in nature. They eliminate incorrect and superfluous information thus ensuring that the Technical Specificat:ions are concise and understandable. Therefore, the changes ensure that: the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in t:he FSAR will not be increased.
The proposed Technical Specification amendment will not create the possibility for an accident'r'malfunction .of a", different type than any previously evaluated in the FSAR. The'roposed changes are administrat:ive in nature.
They eliminate incorrect and superfluous information, thus ensuring that the Technical Specifications are concise and, ,understandable. Therefore, the changes ensure that the pos'sibility 'of a differ'ent accident or malfunction will not be created.
The proposed Technical Specification amendment ,will not,reduce the margin of safety as defined in the basis for the Technical Specifications. The proposed changes are administrative in nature. They eliminate incorrect and superfluous information thus ensuring that the Technical Specifications are concise and understandable. Therefore, the changes ensure that the margin of safety is maintained.
F. ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental question because operation of PVNGS Unit 2, in accordance with this change, would not:
1. Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the staff's testimony to the Atomic Safety and Licensing Board; or
2. Result in a significant'hange in effluents or power levels; or
3. Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.
MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES Limiting Conditions For Operation And Surveillance Requirements:
B 3/4 3-2 B 2-2 B 3/4 3-1 B 2-3
0 4 r, I
l 'I
CGNTRGLLEB BY USER .
BASES Limiting Safety System Settings for the Low DNBR, High Local Power Density, High Logarithmic Power Level, Low Pressurizer Pressure and High Linear Power Level trips, and Limiting Conditions for Operation on DNBR and kM/ft margin are specified such that there is a high degree of confidence that the specified acceptable fuel design limits, are not exceeded during normal operation and design basis anticipated operational occurrences.
2. 1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the, release of radionuclides contained in the reactor coolant from reaching the
'containment atmosphere.
The Reactor Coolant System components are designed to Section III, 1974 Edition, Summer 1975 Addendum, of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110K,(2750 psia) of design pressure. The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code 'requirements.
The entire Reactor Coolant System 'is hydrotested at 3125 psia to demonstrate integrity. prior to initial operation.
2.2.1 REACTOR TRIP SETPOENTS.
Reactor.:Trip Setpgints .specified in Table-2.'2-1 are the 'valiies"z't "",
f'r F'".The which the Reactor Trips are set each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and .Reactor Coolant System are prevented from exce'eding their Safety Limits during normal operation'nd
'esign basis anticipated operational occurrences,and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.
Operation with a trip set less conservative than i'ts Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed 'for each trip in the safety analyses.
The DNBR - Low and Local Power Density - High 'are digital~y generated
~
.trip setpoints based on Safety Limits of 1.231 and 21 kM/ft, respectively.
Since these trips are digitally gener ated by the Core Protection Calculators,
.,the.Anp values are not subject to drifts common to trips generated by analog type equipment. The Allowable Values for these trips are therefore the same as the Trip Setpoints.
To maintain the margins of safety assumed in the safety analyses, the calculations of the trip variables for the DNBR " Low and Local Power Density-High trips include the measuremer't, calculational and processor uncertainties and dynamic allowances as defined i~'66SSQMy 4
/the latest applicaole revision of CEN-3OS-P, "Functional Design Requirements for a Core Protection Calculator" and r
CEN-304-P, "Functional Design Requirements for a Control Element Assembly Calculator."
PALO VERDE " UNIT 2 B 2-2
,:g.gy]TPQLLEQ BY USER
CGiNTRGLLEB 8'( USER BASES REACTOR TRIP SETPOINTS (Continued)
The methodology for the calculation of the PVNGS trip setpoint values, plant protection system, is discussed in the CE Document No. CEN-286(V)<dated
)t4v. 7 Manual Reactor Tri The Manual reactor trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.
Variable Over ower Tri A reactor trip on Variable Overpower is provided to protect the reactor core during rapid positive reactivity addition excursions. This trip function will trip the reactor when the indicated neutron flux power exceeds either a rate limited setpoint at a great enough rate or reaches a preset ceiling. The flux signal used is the average of three linear subchannel flux signals originating in each nuclear, instrument safety channel. These trip setpoints are provided in Table 2.2-1.
Lo arithmic Power Level - Hi h
~
Logarithmic Power .Level.- High trip is provided to protect .the
'he integrity of fuel"cladding and the Reactor Coolant System pressure boundary in
'the event of an unplanned criticality from 'a shutdown condition. A reactor trip is initiated by the Logarithmic'ower Level - High trip unless this trip is manually bypassed by the operator. The operator may manually bypass this trip when the THERMAL'POMER level- is above 10-~X of RATED THERMAL POMER; this bypass is automatically removed when the THERMAL POMER level decreases to 10-~X of RATED THERMAL POMER.
Pressurizer Pressure - Hi h The Pressurizer Pressure - High trip, in conjunction with the pressurizer safety valves and main steam safety valves, provides Reactor Coolant'System protection against overpressurization in the event of loss of load without:.:.; '" "-
reactor trip. This trip s setpoint is below the nominal lift setting of the pressurizer safety valves and its operation minimizes the undesirable opera-tion of the pressurizer safety valves: \
Pressurizer Pressure - Low Pressurizer Pressure - Low trip is provided to trip the reactor and
'he to assist the Engineered Safety Features System in the event of a decrease in Reactor Coolant System inventory and in the event of an increase in heat PALO VERDE - UNIT 2 B 2-3 TPQ LLEW Q'f USER
0'I I
0
CONTROLLED BY USER 3/4. 3 IHSTRUt1EHTATION BASES 3/4.3. 1 and 3/4.3.2 REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURES ACTUAT CsiV SYS I =H IHSTRUiiEHTATiOH The OPERABII ITY of the reactor protective and Engineered Safety FeazJres Actuation Svstems instrumentation and bypasses ensures that (1) the associated Engineered Safety Features Actuation action and/or reactor trip will be initiated when- the parameter monitored by each channel or combination thereof reaiches its setpoint, (2) the spe'ci fied coincidence logic is maintained, (3) su-. .icient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.
The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent witn the assumptions used in the safety analyses.
Response time tes .i ng of resistance temperature devices, wnicn are a part of the reactor protective system, shall be performea by using in-siiu loop current test tecnniques or anorher NRC approved method.
The'ore Protection Calculator (CPC), addressable constants are provioed to allow calibration of the CPC system to more accurate indications oi power level, RCS flow rate, axial flux shape, radial peaking factoJ s and CEA deviation penalties. Administrative controls on changes and periodic checking oi addressable constant values (see also Technical Spec',fica-.ions 3.3. 1 and 6.8. 1) ensure that inadvertent misloading of addressable cons chants into ice CPCs is unlikely.
The design of the Control Element. Assembly Calculators (CEAC) provides
'r'eactor, protection. in the event one or both CEACs become inoperable. If one CEAC is in test or inoperable, verification of CEA position is performed at least every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . If the second CEAC fails, the CPCs in conjunc-.ion wl7h plant Technic=-1 Specifications will use DNBR and LPD penalty fac-ors and increased .DNBR and LPD margi n to restrict reactor operation to a power level that will ensure safe operation of the plant. If the margins are not maintained, a 'reactor trip will occur.
The value of the DNBR in Specification 2. 1 is conservatively compensated for measurement uncertainties. Therefore,. the actual RCS total flow rate determined by the reactor coolant pump differential pressure instrumentation or by calorimetric calculations does not have to be conservatively compensated for measurement uncertainties.
An ana ys> den~~soeciiy a minimum w w sch an addi-tional power reduction i en-s-f-thee s a CEA misalignment with Cc 1ce PALO VERDE - U i
0 CONTROLLED BY USER INSTRUflEN ATION s
BASES REACTOR PROTECTi IVE AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUflENTATION (,Continued)
The analysis determined a Power Operating Limit (POL) power and assumed A misalignment occurred from this power level. The power penalty factor~ at wo 'ccommodate changes in radial peaks and one hour xenon redistribuvfon that would cur if there were a CEA misalignment with CEACs out of serv e. The quotient the POL power and the CEA misalignment Power Penalt actor is the maximum power Oym power) at which DNBR SAFDL violation wi 11 ccur even if there is a CEA mi alignment from POL conditions. Below s power, extra thermal margin will be available to the plant. Thus or CEA misalignment, power reduction below tlat limiting power is unnecessary.
The lowest core power for POL was c~ culated to be 70<'f rated power.
This was based on the following wo OL'SS fluid conditions.
High Temoerature Low Pressure 1785 p ia
-.3 Unaer f1 ow~iacti on: 0. 865 Low F I ow<< 95 of full flo Hig <<Radial Peak '.70 (Bank 5+4+PLR; 4IL ='-'0.". Power)
Tge surveillance requirements specified for these sys emsWe'ns ne t h- the ovej-ail sysiem functional capability is maintained comparable to the Woicinal sign standards. The periodic surveillance tests'erformed a he m nimum
'requencies are sufficient to demonstrate this capability.
The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated wi-h eacn channel is completed within the time limit assumed in the safety analyses.
~
No credit was taken in the analyses for those channels with response times
'indicated as not applicable. The response times in Table 3.3-2 are made up of the time to generate the trip signal at the detector (sensor response time) and the time for the signal to interrupt power to the CEA drive mechanism (signal or trip delay time).
Response time mav be demonstrated by any series of sequential, overlapping, or total channel test, measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either (1) in place, onsite, or offsite test measurements or (2) utilizing replacement sensors with certified response times.
3/4. 3. 3 t<ONITORIHG IHSTRUitENTATIOH
'/4.
v (1)
3. 3. 1 RADIATION I!OHITORING INSTRUt1ENTATION The OPERABII ITY of the radi ati on moni tor ing channels ensures that:
the radiation levels are continually measured in the areas served by the "CONTROLL&SY USER

ML17300B115

Text

Navigation menu

Search