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Forwards Draft Tech Spec Page Changes Request Re Reactor Protection Sys.Proprietary XN-NF-86-91(P), Low Flow Trip Setpoint.... & Proprietary ANF-87-150(P),Vol 1, Palisades Modified Reactor... & Affidavits Also Encl.W/O Repts
	ML18054A208
Person / Time
Site:	Palisades
Issue date:	12/23/1987
From:	Kuemin J; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	; NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
	NUDOCS 8712290418
	Download: ML18054A208 (74)
	v • d • e

consumers Power POWERiNii MICHlliAN"S PROliRESS General Offices: 1945 West Parnell Road, Jackson, Ml 49201 * (517) 788-0550 December 23,, 1987 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET.50-255 - LICENSE DPR PALISADES PLANT-DRAFT TECHNICAL SPECIFICATIONS CHANGE REQUEST -

REACTOR PROTECTION SYSTEM During the next Palisades refueling outage, presently scheduled to begin in September 1988, modifications will be made t*o improve the capabilities of the Reactor Protection System (RPS). These modifications include the addition of a Variable High Power Trip (VHPT), the addition of an Axial Shape Index (ASI)

Alarm and modifications to the Thermal Margin/Low Pressure (TM/LP) Trip calculator. The modifications are summarized below. The modified RPS will provide comprehensive* protection of the. fuel for all reactivity insertion transients.

This draft Technical Specifications Change Request is being provided to initiate NRC Staff review. Consumers Power Company requests a meeting to discuss the modifications and proposed changes to the Technical Specifications described herein. We request the meeting be held in the final week of January 1988. We will work with the NRC Palisades Project Manager to set a final date. A final Technical Specifications Change Request will be submitted following resolution of NRC comments and completion of a Transient Analysis Report presently in preparation by Advanced Nuclear Fuels Corporation.

VHPT The purpose of this modification is to provide a capability for early detection and minimizing transients starting at reduced power levels.

Therefore, operator action to mitigate slow reactivity insertion transient (ie, boron dilution) will no longer be necessary~

Anticipated fuel assembly design changes and additional steam generator tube plugging will reduce thermal margin. The VHPT will enhance core protection and assure the required safety margins are maintained.

ASI - ALARM The ASI function is derived from the power range safety drawer excore indications for upper an4 lower neutron flux power. These are input into the ~

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.I OC1287-0051A-NL02-NL04

Palisades Plant Nuclear Regulatory Commission TSCR - Reactor Protection System December 23, 1987 2

Thermal Margin Calculators which calculate the Axial Shape Index (lower-upper/

lower+upper) and correct these results for geometry effects. The corrected value of ASI is then used in TM/LP calculations and trip determination. The positive and negative setpoint values for ASI will simultaneously be generated as a function of maximum power and compared to the measured ASI to generate an alarm.

The purpose of this modification is to monitor and protect the reactor from axial power distributions outside the limits of current and future safety evaluations. This modification is also part of the enhanced core protection required for anticipated modifications to the fuel assembly design.

TM/LP - TRIP The purpose of this modification is to enhance Thermal Margin/Low Pressure calculations by including actual monitoring for the Axial Shape Index, and reducing the dependency on RTD time responses by selecting the maximum of reactor power and thermal power. The current analog calculators are being replaced by programmable digital calculators. The additional thermal margin will allow maintaining licensed power levels with additional steam generator tube plugging or other modifications which would restrict coolant flow to the

.reactor.

T Inlet Max Alarm The purpose of this modification is to alert the operator of an impending limiting operating condition in order that appropriate action may be taken.

High Rate Trip Bypass The power range nuclear/instrument drawer will be modified to allow an adjusted nuclear power signal to be sent to the lower power (16.5%) bypass bistable. This will allow proper high rate trip bypass function above 15%

power without violating the technical specification minimum value.

Equipment to be removed or disabled (ie, function replaced by the Thermal Margin, Calculators)

A. Existing thermal power (~T power) calculators manufactured by Bell and Howell. These are categorized as "associated circuits" presently. They provide no input to the Reactor Protective Systems. They only provide input to the deviation meter (thermal power vs. nuclear power).

B. Existing TM/LP calculators mounted in the rear of the main control boards.

These calculators are comprised of GE/MAC function modules (summers, limit-ers, function generators). Their function will be replaced by the new Thermal Margin Calculators.

C. The existing high power trip signals to the RPS auxiliary trip units. The auxiliary trip units will be commanded by the new Thermal Margin Calculators (VHPT).

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Nuclear Regulatory Commission 3 Palisades Plant TSCR - Reactor Protection System December 23, 1987 D. The power ratio calculator mounted at the top of panel C-27. The Thermal Margin Calculators (ASI function) will store data to replace this function.

E. The Xl/XlO pushbuttons. These are no longer required as a result of the described modifications.

In 1982 Consumers Power Company initiated a Thermal Margin Improvement Program for the Palisades Plant. During this evaluation we determined that some nonconservative *assumptions were used in 1977 while performing the Control Rod Withdrawal Transient. Therefore, a new analysis was performed by Advanced Nuclear Fuels Corp. (then called Exxon Nuclear Company) and documented in XN-NF-83-57, Rod Withdrawal Transient Reanalysis for the Palisades Nuclear Reactor. This reanalysis consumed the additional thermal margin provided by the XNB DNB correlation and the calculated higher primary coolant flow at full power. Amendment No. 82 to the Palisades Technical Specifications incorporated the results of this reanalysis.

Several other licensing issues were identified in 1983 which required the additional thermal margin if a plant derate was to be avoided. These issues included Pressurized Thermal Shock, additional steam generator tube plugging, RTD time delay uncertainty, Asymmetric LOCA loads (Task A-2) and .

identification of reserve thermal margin to resolve future unknown licensing concerns. Resolution of Task A-2 is being concluded by performing a primary coolant piping Leak-Before-Break Analysis sponsored by the CE Owners Group.

RTD delay time is no longer an important issue since maintenance to the RTD thermowells has resulted in regaining fast RTD response times. Consumers Power measured the response time of several primary loop RTDs and obtained acceptable results. _

Since additional thermal margin was needed to institute a low leakage fuel management scheme and to provide allowance for additional steam generator tube plugging, hardware modification of the reactor protective system were investigated. Also, the Reactor Protective System currently installed in the Palisades Plant has some short comings. In order to ensure protection of the fuel during transient starting at 50% power level, excessively conservative assumptions must be made in establishing the trip setpoints. The trip setpoints are based upon part power conditions which restrict 100% full power operations. Also, the core power level is determined by measuring the hot leg and cold leg temperatures. Therefore, the indicated temperatures are affected by the time constant of the RTD which must be account~d for by making additional conservative assumption in establishing the trip setpoints.

Modifications are being made to the RPS to resolve the above issues. The high power trip is being replaced by a variable high power trip. An axial shape index alarm is being added to ensure plant operations are bounded by assumptions used to generate the inlet temperature LCO. Also, the current analog TM/LP trip is being replaced by a digital TM/LP trip which includes an axial shape index parameter. The VHPT and the new TM/LP trip were designed based upon descriptions in the C-E Setpoint Methodology document CENPD-199-NP Revision 1 - NP dated March 1982.

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Nuclear Regulatory Commission 4 Palisades Plant TSCR - Reactor Protection System December 23, 19,87 The proposed modifications were reviewed by ANF. After sev~ral discussions, the final modifications were finalized. ANF has also completed a reload analysis for the St. Lucie Plant and they are very familiar with the proposed trips. Essentially the same analytical procedures were used by ANF to develop the new TM/LP trip constants for Palisades that were used for the St. Lucie Plant.

The Advanced Nuclear Fuels Corp. (ANF) reports No. ANF-87-150(P), Volume 1, and XN-NF-86-91(P) attached hereto are considered proprietary and are requested to be withheld from public disclosure. In accordance with 10CFR2. 790 (b) affidavits, executed *by H. E. Williamson of ANF, attesting to the reports' proprietary nature are also attached. The first report provides a disposition of Standard Review Plan, Chapter 15, events with the modified reactor protection system. Those events which have been dispositioned as requiring reanalysis will be the subject of Volume 2 of the report which is still under development. Volume 2, a Transient Analysis Report, will be completed and submitted with the Technical Specifications Change Request following our January 1988 meeting. Chapter 15 events which are bounded by existing analyses and do not require additional analyses, are so noted in the report (Volume 1). The second report (No. XN-NF-86-91(P) is a Low Flow Trip Setpoint and Thermal Margin Analysis for Three Primary Coolant Pump Operation of the Palisades Reactor. It is referenced within the draft Technical Specifications. We have limited the distribution of these proprietary documents to the NRC Document Control, Palisades Project Manager, and Region III.

tJ~~

ames L Kuemin Staff Licensing Engineer CC Administrator, Region III, NRC NRC Resident Inspector - Palisades Attachment OC1287-0051A-NL02-NL04

ATTACHMENT Consumers Power Company Palisades Plant Docket 50-255 AFFIDAVITS ADVANCED NUCLEAR FUELS CORPORATION December 23, 1987 OC1287-0051A-NL02-NL04

A F F I DAV I T STATE OF WASHINGTON )

) SS.

COUNTY OF BENTON )

I, H. E. Williamson being duly sworn, hereby say and depose:

1. I am Manager, Licensing and Safety Engineering, for Advanced Nuclear Fuels Corporation, ("ANF"), and as such I am authorized to execute this Affidavit.

2. I am familiar with ANF's detailed document control system and policies which govern the protection and control of information.

3. I am familiar with the document x'N-NF-86-9l(P), entitled "Low Flow Trip Setpoint and Thermal...

Ma.rgin Analysis'.

for Three Primary Coolant Pump Operation of the Palisades Reactor," referred to as "Document." Information contained in this *Document has been classified* *by ANI' as proprietary in accordance with the control system and policies established by ANF for the control and protection of information.

4. The Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by ANF and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in the Document as proprietary and confidential.

5. The Document has been made available to the U.S. Nuclear Regulatory Commission in confidence, with the request that the information contained in the Document will not be disclosed or divulged.

2 -

6. The Document contains information which is vital to a competitive advantage of ANF and would be helpful to competitors of ANF when competing with ANF.

7. The information contained in the Document is considered to be proprietary by ANF because it reveals certain distinguishing aspects of the set point methodology and analysis which secure competitive advantage to ANF for fuel design optimization and marketability, and includes information utilized by ANF in its business which affords ANF an opportunity to obtain a competitive advantage over its competitors who do not or may not know or use the information contained, in the Document.

8. The disclosure of the proprietary information contained in the Docum.ent to a competitor would permit the competitor to reduce its expenditure of money and manpower and to imp rove its competitive position by giving it extremely valuable insights into the setpoint methodology and analysis and would result in substantial harm to the competitive position of ANF.

9. The Document contains proprietary information which is held in confidence by ANF and is not available in public sources.

10. In accordance with ANF's policies governing the protection and control of information, proprietary information contained in the Document has been made available, on a limited basis, to others outside ANF only as required and under suitable agreement providing for nondisclosure and limited use of the information.

11. ANF policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

12. This Document provides information which reveals the setpoint methodology and analysis developed by ANF over the past several years. ANF has invested inany thousands of dollars and many man-months of effort in developing the setpoint methodology and analysis revealed in the Document.

Assuming a competitor had available the same background data and incentives as ANF, the competitor might, at a minimum, develop the information for the same expenditure of manpower and money as ANF.

THAT the statements made hereinabove are, to the best of my knowledge, information, and belief, truthful and complete.

FURTHER AFFIANT SAYETH NOT .

SWORN TO AND SUBSCRIBED before me this '1ft day of

£).~

~~~~---'-~~~~

, 1987.

( ' .

,~*~;/1~

I . - .

- , NOT_ARY'oPUBLI C

' .. \

A F F I DAV I T STATE OF WASHINGTON SS.

COUNTY OF BENTON I, H. E. Williamson being duly sworn, hereby say and depose:

1. I am Manager, Licensing and Safety Engineering, for Advanced Nuclear Fuels Corporation, ("ANF"), and as such I am authorized to execute this Affidavit.

2. I am familiar with ANF's detailed document control system and policies which govern the protection and control of information.

3. I am familiar with the document ANF-87-150(P), Volume l, entitled "Palisade.s Modified Reactor Prote.ction. System.Report-Disposition of Standard Review Pl an Chapter 15 Events," referred to as "Document."

1, * *'t lnforma~ ion contained in this Document .*has been' classified by ANF as proprietary in accordance with the control system and policies established by ANF for the control and protection of information.

4. The Document contains information of a proprietary and

confidential nature and is of the type customarily held in confidence by ANF and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in the Document as proprietary and confidential.

5. The Document has been made available to the U.S. Nuclear Regulatory Commission in confidence, with the request that the information contained in the Document will not be disclosed or divulged.

6. The Document contains information which is vital to a competitive advantage of ANF and would be helpful to competitors of ANF when competing with ANF.

7. The information contained in the Document is considered to be proprietary by ANF because it reveals certain distinguishing aspects of the SRP Chapter 15 event disposition which secure competitive advantage to ANF for fuel design optimization and marketability, and includes information utilized by ANF in its business which affords ANF an opportunity to obtain a

.competitive advantage over its competitors who do not or may not know or use the information contained in the Document.

8. The disclosure of the proprietari information contained in the Document to a compet.itor *wo.uld permit the competitor.to .reduce its expenditure of money and manpower and to imp rove its competitive position by giving it extremely va.luable :'insights into the* SRP Chapter,

is event disposition and would restilt in substantial harm to the competitive position of ANF.

9. The Document contains proprietary information which is held in confidence by ANF and is not available in public sources.

10. In accordance with ANF's policies governing the protection and control of information, proprietary information contained in the Document has been made available, on a limited basis, to others outside ANF only as required and under suitable agreement providing for nondisclosure and limited use of the information.

11. ANF policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

3

12. This Document provides information which reveals the SRP Chapter 15 event disposition developed. by ANF over the past several years.

ANF has invested many thousands of dollars and many man-months of effort in developing the SRP Chapter 15 event disposition revealed in the Document.

Assuming a competitor had available the same background data and incentives as ANF, the competitor might, at a minimu~, develop the information for the same expenditure of manpower and money as ANF.

THAT the statements made hereinabove are, to the best of my knowledge, information, and belief, truthful and complete.

FURTHER AFFIANT SAYETH NOT.

SWORN TO AND SUBSCRIBED before me this C/_;o..., day of

£)~ '1987.

\I' DRAFT CONSUMERS POWER COMPANY DOCKET 50-255 - LICENSE DPR PALISADES PLANT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS For the reasons hereinafter set forth, it is requested that the Technical Specifications contained in the Provisional Operating License DPR-20, Docket 50-255, issued to Consumers Power Company on October 16, 1972, for the Palisades Plant be changed as described in Section I below:

I. Changes and Discussion of Changes A. Chapter 1 Specifications In 1.1, Change "Axial Offset" to "Axial Offset or Axial Shape Index" The thermal margin analysis performed by Advanced Nuclear Fuel, Inc (ANF) to support the revised TM/LP trip and the revised inlet temperature LCO incorporates the Axial Shape Index (AS!)

terminology. This change represents an Administrative Change that is more descriptive.

B. Chapter 2 Specifications

1. In 2.1, Applicability, Insert "4-pump" between "during" and*

"operation."

This specification *will be applicable to the normal mode of plant operations with four primary coolant pumps operating.

Two pump operations have been deleted while three pump operating limits are given in Section 2.3.

2. In 2 .1, Specification, Delete existing Figure 2-1 "Reactor Core Safety Limits - 2 Pump Operation" and Figure 2-2 "Reactor Core Safety Limits - 3 Pump Operation." Renumber Figure 2.:.3 "Reactor Core Safety Limits - 4 Pump Operation" as Figure 2-1 and revise safety limit lines. Remove reference to two pump and three pump operations from the text. Add the adjective "maximum" to "cold leg temperatures" in the first sentence.

A decision was made to delete the 2 pump operating mode,

therefore, the revised RPS hardware does not include a mode selector position for 2 pump operations. Three pump operating mode is discussed in Specification 2.3 and the safety limit lines shown in the new Figure 2-1 are based upon the thermal margin analysis performed by ANF to support this modification as described in report number ANF-87-150(P), Volume 2. The safety limit lines shown in Figure 2-1 are appropriate for the specified ASI. The adjective "maximum" was added to provide additional clarification.

TSOP1287-0254-NL04

-* 2

3. In 2.lf Basis, Replace all ~eferences t~ th~ W-3 DNB correlation and its safety limit of 1.3 with the XNB DNB correlation with a limit of 1.17. Delete discussions pertaining to two-pump and three-pump operations. Change the wording "loci of points of thermal power, primary coolant system pressure and average temperature of variou*s pump combinations" to "loci of points of thermal power, primary coolant system pressure, maximum cold leg temperature and a specified ASI." Reference to Figure 2-2 and 2-3 ~ere deleted.

Also, a new sentence was added to the last paragraph and references 1 and 2 were changed.

Reference to the W-3 correlation has been deleted because the majori~y of the transients in Chapter 14 of the Palisades FSAR were analyzed or are lrnunded by transients analyzed using the XNB DNB correlation. The primary exception is the steam line break event. Reference to two-pump and three-pump operations

.and reference to Figure 2-2 and Figure 2-3 were deleted to be consistent with Section 2.1, Applicability. The wording "average temperature" was replaced with "maximum cold leg temperature" to correct an error and the phrase "at a specified ASI" was added because the new TM/LP trip includes a term that is a function of the measured Axial Shape Index.

The sentence added to the end of the Basis justifies using the XNB DNB correlation for Palisades thermal margin analysis.

This statement was transferred from the fourth paragraph from the 2.3 Basis. Reference 1 was changed to the appropriate document containing the development of the XNB DNB correlation and Reference 2 was changed to XNB DNB application document.

4. In 2.2, Basis, Delete the phrase "the pressurizer power-operated relief valves at 2400 psia and".

The PORVs are inoperable during power operations.

5. in Table 2.3.1, Reactor Protection System Trip Setting Limits, delete column pertaining to 2-pump operations, replace Item 1, "High Power" with "Variable High Power" and provide appropriate set points for 4-pump and 3-pump operations.

Revise set point for Item 2, "Primary Coolant Flow" for 3-pump operations, revise set points for Item 4, "Thermal Margin/Low Pressure" for 3-pump and 4-pump operations. Revise Notes 1, 3, and 4.

The column giving set points for 2-pump operations was deleted since this mode of operation is being deleted from the Technical Specifications. The new RPS hardware incorporates a Variable High Power trip which limits power operations less than or equal to 10% above the indicated power level to provide protection for such events such as the boron dilution transient. The appropriate set points for both 4-pump and TSOP1287-0254-NL04

3 3-pump operations have been given. The 3-pump thermal margin analysis also justified lowering the low flow trip from 71% to 60% of the 4-pump flow. The referenced figure for Item 4, "Th~rmal Margin/Low Pressure" was revised from Figure 2-3 to Figure 2-1 to conform with previous changes made above. The phrase "for a specified ASI" was added since the new TM/LP trip is a function of the measured ASI. For 3-pump operations, "High Power Level Trip" was replaced with "Variable High Power Trip" due to the hardware changes.

The former note 1 was deleted since the times ten selector switch is to be removed during this plant modification. The new note 1 clarifies the trip setpoint for power operation at less than 20% power. The first sentence of Note 3 was deleted since the hot leg temperature is not used in the TM/LP trip equation (see Page 2-7 of the Specifications) and the units for the cold leg temperature is given in the new Figure 2-1.

Everything following the first "1750 psia" in Note 3 was deleted since the safety analysis is based upon this assumption. Reference to 2-pump operation was deleted from note 4 since power operations with less than 3 PCP's is being deleted from the Technical Specifications.

6. In 2.3, Basis, Item 1, change the heading from "High Power" to "V~riable High Power," and replace the text in the first paragraph with a new paragraph describing how the VHPT functions and deleted the third and fourth paragraphs.

Reference 1 was deleted and Reference 2 was updated (see page changes).

The RPS modification includes a VHPT which replaces the existing High Power trip. The discussion pertaining to different trip set points for 3-pump operation was deleted because they are adequately described in the basis se_ction for the Low Primary Coolant Flow Trip. The fourth paragraph was deleted since the XlO switch will be removed by the plant modification. Reference 1 was deleted due to changes made to the FSAR and Reference 2 was changed to the new basis document.

7. In 2.3, Basis, Item 2, delete all references to 2-pump operation and segregate discussions pertaining to 3-pump operations into one paragJ;aph at the end of the subsection.

The notation ref erring to Reference 3 was changed to Reference 4 and Reference 3 was moved to the first sentence of the section. The notation referring to Reference 5 was moved to the second paragraph. Also, References 4 and 5 were updated (see page changes).

The references to 2-pump operations were deleted for the previously stated reasons. Segregation of the discussion on 3-pump operations was done for clarification. The historical TSOP1287-0254-NL04

-~

4 flow rate data for different pump combinations was deleted from the FSAR, therefore, the notation referring to Reference 5 was moved. Reference 4 and Reference 5 were changed to the new basis documents. The notation referring to Reference 3 was moved to a more appropriate sentence.

8. In 2.3, Basis, Item 3, delete two sentences beginning with "The power-operated" . through "operations of the safety valves." Reference 11 was updated.

The LTOP system is disarmed when the PCS average temperature is greater than 430°F. Reference 11 was changed to the new basis document.

9. In 2.3, Basis, Item 4, essentially the entire basis description for the TM/LP trip was rewritten and reference to 2-pump operation was deleted. Reference 7, 13, 14 and 15 were deleted and Reference 12 was updated (see page changes).

This section was rewritten because the existing analog .

hardware has been replaced by a microprocessor and the form of the trip function was altered significantly. Reference to 2-pump operation was deleted since this mode of power operation is no longer allowed by the Technical Specifications.

FSAR Section 3.3.6 has been deleted, therefore Reference 7 was deleted. Reference 12 was changed to_ the new basis document.

Reference 13 was replaced by Reference 12 and Reference 14 is no longer needed since the event was reanalyzed in ANF-87-150(P), Volume 2, Section 15.4. Reference 15 was moved to Section 2 .1. (Note: The TM/LP constants will be finalized with the completion of the Transient Analysis Report).

10. In 2.3, Basis, Item 5, change "provide a 15 minute margin before auxiliary feedwater is required" to "allow.a safe and orderly plant shutdown arid to prevent steam generator dryout assuming minimum auxiliary feedwater capacity." Also, Reference 9 was updated.

Steam generator water level is not an explicit acceptance criterion. However, the analysis shows that sufficient steam generator water level is maintained to ensure an adequate heat sink until the primary coolant system temperature and pressure are reduced below the initiation threshold of the shutdown cooling system.

Reference 9 was changed to the new basis document.

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')

5

11. In 2.3, References Reference 10 was changed to "FSAR, Section 7.2.3.9."

The FSAR was revised to include previous amendments, therefore, the new reference is more appropriate.

C. Chapter 3

1. In 3.1.1, Operable Components, Item a, insert the words "with a minimum flow rate of 1500 gpm" following "shutdown cooling pump."

The phrase was added to ensure that the plant operations are bounded by assumptions used in the safety analysis pertaining to boron dilution events.

2. In 3.1.1, O~erable Components, Item b, change "above 5% of rated power to 11 above hot shutdown" and delete the parenthetical exception.

This change is made to prevent continual reactor operation except when all the reactor coolant pumps are operating. The present plant safety analysis supports plant operation when all four coolant pumps are operating. The present specification allows continual operation up to 5% power with no requirements with regard to primary coolant pump operation.

Thus as a result of this change, the reactor cannot be made critical unless four pumps are operating. The parenthetical was deleted since the requirements are adequately specified in Section 2.3.

3. In 3 .1.1, Operable Components, .Item b, move the third.

paragraph of Item c to Item b after making the following change: 1) delete the first sentence, 2) replace "Following loss of a pump" with "Before removing a pump from service,"

3) replace "or more pumps" with "pump," 4) replace "the pumps to" with "the pump to," S) in the last two sentences change "hot standby" to "hot shutdown" (two places) and 6) add "and power operations with less than three pumps is not permitted" the the last sentence.

This paragraph was moved from Item c to Item b to place all of the pump operability requirements in one specification. The first sentence was deleted because it would have been redundant. The second change was made because the reactor power must be reduced below the 3 pump allowed power level to select the 3 pump trip set points before a pump can be removed from service. Reference to 1 or 2 pump operation was deleted because required safety analyses have not been performed.

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6 The present specif1cation requires hot standby which allows unlimited operation of the reactor up to 2% power. This change requires the plant to be in hot shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of turning off a primary coolant pump. Thus the time is limited to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the reactor can be operated after a pump has been shut off.

The last phrase was added to enforce previous statements pertaining to 1-pump and 2-pump operations.

4. In 3.1.1, Operable Components, Item c, delete the first sentence, delete the parenthetical in the second sentence, change "126.9" to "124.3" and delete the third sentence beginning with "In the event."

The allowed power and flow rates for 4-pump and 3-pump operations are adequately specified in Section 2.3. The parenthetical describing the primary coolant flow rate measurement technique was deleted because the heat balance method has been adopted. The new safety analysis calculations were performed using a full power core flow rate based upon a hot zero power flow rate of 124.3 M-lbm/hr. The procedure for changing the TM/LP trip was deleted since-adequate justification does not exist at the present time.

5. In 3.1.1, Operable Comgonents, Item e(l), change "operating

transient differential 1 to "operating differential" and change "1530 psi" to "1380 psi."

The maximum transient differential pressure cannot be controlled by operator actions. Therefore, it is being deleted and. replaced by the operating differential pressure limit of 1380 psi. The 1530 psi limit was added to the Palisades technical specification by Amendment No. 20 (April 26, 1976) before RG 1.121 was issued.

The structural integrity of the steam generator tubes are assured by appropriately selecting the tube plugging criteria.

The NRC has approved our plugging criteria in the SER dated June 11, 1984. Since the requirements pf RG 1.121 are satisfied, the current transient differential limit of 1530 psi is not required. Also, the Combustion Engineering Standard Technical Specifications (NUREG-0212, Rev 2)

reference RG 1.121 in Section 4.4.6.4 as the basis for determining the plugging limit.

6. In 3.1.1, Operable Components, Item g, revise the inlet temperature LCO equation and replace the note by an ASI limiting statement with reference to a new Figure 3.0, ASI LCO for T Inlet Function. Delete the previous Figure 3.0, Reactor Inlet Temperature vs Operating Pressure. Also, add the following action statement.

TSOP1287-0254-NL04

7 "When the ASI exceeds the limits specified-in Figure 3.0, within 15 minutes, initiate corrective actions to restore the ASI to th~ acceptable region. Restore the ASI to acceptable values within one hour or be at less than 70% of rated power within the following two hours."

The ASI restraints were added to allow higher core inlet temperatures at full power operating conditions. The revised inlet temperature LCO was developed in the transient analysis report ANF-87-150(P), Vol 2.

The inlet temperature LCO provided protection against penetrating DNB during the most limiting transient from full power operation. The most limiting transient for Palisades is the inadvertent drop of a full length control rod without a reactor trip. The transient analysis report shows that adequate thermal margin is available for power level < 70% to allow ASI to exceed values expected to bound plant operations.

7. In 3 .1.1, Basis, following the sentence ending with "operated at rated capacity.", insert the following sentence "By imposing a minimum shutdown cooling pump flow rate of 1500 gpm, sufficient time is provided for the operator t~ 5erminate the boron dilution under asymmetric flow conditions 6 ."

Added Reference 6 to the list of references for Section 3.1.1.

This statement is supported by the analysis documented in ANF-87-150(P), Volume 2, Section 15.4.6.3.2.

8. In 3.1.1, Basis, add a second paragraph as follows:

"The FSAR safety analysis was performed assuming four primary coolant pumps were operating for accidents that occur during reactor operation. Therefore, reactor startup above hot shutdown is not permitted unless all four primary coolant pumps are operating. Operation with less than four primary coolant pumps is permitted for a limited time to allow the restart of a stopped pump or for reactor internals vibration monitoring and testing."

This change to the basis explains that the FSAR safety analysis is based on four.primary coolant*pumps operating* and that a limited time is allowed for resta~t of a pump or for testing.

11

9. In 3.1.1, Basis, add reference number (3) 11 to "1380 psi" in the third paragraph.

The stated reference demonstrates steam generator tube integrity by meeting Regulatory Guide 1.121, August 1976, requirements.

TSOP1287-0254-NL04

8

10. In 3.1.1, Basis, delete the last paragraph beginning with "The maximum transient."

The maximum steam generator transient differential pressure limit is being deleted. The basis for the deletion is described above.

11. In 3.1.1, Basis, in the paragraph beginning with "The transient analysis" change "126.9" to, "124.3."

This is the vessel flow rate used in the new transient analysis performed to establish the proposed RPS set points.

12. In 3.1.1, Basis, replace the sentence beginning with "A DNB analysis" with the following sentence:.

"A DNB analysis was performed in a parametric fashion to determine the core inlet temperature as a function of pressure and flow for which the minimum DNBR is equal to 1.17. *This analysis includes the following uncertainties and allowance:

2% of rated power for power measurement*; +/-0 .06 for ASI measurement; +/-50 psi for pressurizer pressure; +/-7°F for inlet temperature; and 3% measurement and 3% bypass for core flow.

In addition, transient biases were included in the derivation of the folloy!~g equation for limiting reactor inlet

temperature: * . .

These are the assumptions documented in the transient analysis report ANF-87-150(P), Volume 2, Section 15.0.7.1, for the development of the inlet temperature LCO.

13. In 3.1.1, Basis, Replace the existing inlet temperature LCO equation with the following equation:

T I l t S 543.3 + 0.575(P-2060) + 0.00005(P-2060)**2 +

n e l.173(W-120) - 0.0102(W-120)**2 The new RPS hardware along with using the XNB DNB correlation provides additional thermal margin which allows higher inlet temperatures. The supporting analysis is documented in ANF-87-150(P), Volume 2, Section 15.0.7.1.

14. In 3 .1.1, Basis, Delete the sentence beginning with "A temperatu*r~ measurement uncertainty" and delete the sentence beginning with "The nominal full power temperature."

The inlet temperature LCO equation incorporates all of the appropriate uncertainty and bias factors.

15. In 3.1.1, Basis, Change "1850" "2250" and "130" to, "1800,"

"2200" and"135," respectively and add the following statement: "ASI as shown in Figure 3.0."

TSOP1287-0254-NL04

9 These are the limits of validity for the revised inlet temperature LCO equation.

16. In 3.1.1, Basis, Add the following paragraph describing the ASi alarm following the above stated limits for the inlet temperature LCO equation.

"The Axial Shape Index alarm channel is being used to monitor the ASI to ensure that the assumed axial power profiles used in the development of the inlet temperature LCO bound measured axial power profiles. The signal representing core power (Q) is the auctioneered higher of the neutron flux power and the Delta-T power.* The measured ASI calculated from the excore detector signals and adjusted for shape annealing and the core power constitute an ordered pair (Q,YI). An alarm signal is activated before the ordered pair exceed the boundaries specified in Figure 3 .0."

The modified RPS will provide a new alarm when the measured ASI is not within the acceptable range defined in the new Figure 3.0.

17. In 3.1, References, change Reference 3 from XN-NF-77-18, to Palisades 1983/1984 Steam Generator Evaluation and Repair Program Report, Section 4, April 19, 1984.

The transient differential limit is being replaced by the operating differential limit of 1380 psi.

18. In 3.1, References, change Reference Number 4 from "XN-NF-77-22" to "ANF-87-lSO(P), Volume 2, Section 15.0.7.1.

Reference 4 was changed to the new design basis document.

19. In 3.1.7, Basis, change Reference 2 to "ANF-87-lSO(P), Volume 2, Section--rs:2'.l.

The loss of load transient reanalysis is documented in the new reference.

20. In Figure 3-6, delete "TWO OR THREE PUMP OPERATION" curve and change title of bottom curve to "THREE OR FOUR PUMP OPERATION."

Two pump operating mode is being deleted from the Palisades Technical Specifications and the new transient reanalysis used only the bottom.curve.

21. In 3.10, Basis, delete Reference 5 (two places).

Reference 5 did not provide a basis for the statement in the last paragraph. The statement does not require additional TSOP1287-0254-NL04

10 support. Since Reference 5 was not referred to in the basis discussion, it was deleted from the list of references.

22. In 3.11.2, Basis, change "The APL considers both LOCA and DNB based LHR limits," to, "The APL considers LOCA based LHR limits."

With the current modification of the RPS; AS! monitoring has become an integral parameter in the TM/LP trip. Therefore, the APL function does not have to provide DNB protection.

23. In 3.11, References, change Reference 2 to "ANF-87-150(P),

Volume 2, Table 15.0.7.2-1."

Reference 2 was changed to the new basis document.

24. In 3.17, Basis, 1) in sentence beginning with "If the bypass is not affected" change "high power" to, "variable high power" and 2) in the sentence beginning with "At rated power" change "high power" to "variable high power."

The high power trip has been replaced with the variable high power trip.

25. In 3.17, Basis, delete "in the event that a turbine runback signal is required from the power range channels" from the first paragraph on page 3-77.

The deleted sentence is obsolete as the turbine runback feature was disabled prior to being licensed at rated power of 2530 MWt.

26. In 3.17, References, change Reference 1 to "FSAR, Section 7.2.2" and delete Reference 2.

FSAR Section 7.2.2 is the appropriate document for Reference 1 and Reference 2 is not used in the text.

27. In Table 3.17.1, Item 2, change "High...,Power Level" to, "

Variable High Power Level."

The high power trip will be replaced with a variable high*

power trip.

28. In Table 3.17.4, Item 16, change "Excore Detector" to "Excore Detector Deviation Alarms," change "None" to "Not Required Below 50% of Rated Power," change footnote "g" to state "Calculate the Quadrant Power Tilt using the excore readings at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the excore detectors deviation alarms are inoperable, or at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months using symmetric incore detectors when the difference between the excore and incore measured Quadrant Power Tilt exceeds 2%."

TSOP1287-0254-NL04

11 This change will clarify Item 16. Requiring calculations every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and not requiring the alarms to be operable below 50% of rated power is consistent with Technical Spe~ification 3.23.3. Addition of calculating Quadrant Power Tilt using symmetric incore detectors when the difference between the excore and incore measured Quadrant Power Tilt exceeds 2% is consistent with current practice.

29. In Table 3.17.4, add new Item 17, Axial Shape Index (see page changes).

ASI is a new feature of the modified RPS.

30. In 3.23.1, Action 3, add to "to monitor LHR" to the second line of the paragraph.

The words were added to clarify when the action is applicable *.

31. In 3. 23. 1, Basis, delete "In addition, the limitation enveloped by the design power distribution."

Prior to the inclusion of ASI monitoring in the TM/LP trip, it was necessary to limit linear heat rates (LHR) in order to ensure that thermal margin was not compromised when operating with LHR's that differed from those used in the safety analysis. However, with the current modification to the RPS, ASI monitoring has become an integral parameter in the TM/LP trip. Therefore, with the modified RPS, if thermal margin is threatened by LHR, the reactor will trip provided that the radial peaking factor limits (Table 3. 23-2) are ~bserv.ed.

32. In 3.23.1, References, delete Reference 2, Reference 3 and Reference 5.

Statements made based upon these references were deleted.

33. In Figure 3.23-1, delete the line labeled "DNB" and remove the excess shading.

See the preceding justification for Section 3.23.1, Basis.

34. In 3.23.2, Radial Peaking Factors, change "the following quantity (1.0 + 0.5(1-P)) where P is the" to, "the following quantity. The quantity is (1.0 + 0.3(1-P)) for P ~ .5 and the quantity is 1.15 for P < .5. P is the... "

The allowed radial peaking factor at partial powers was reduced to provide additional thermal margin for transients initiated at power levels less than full power.

TSOP1287-0254-NL04

.e 12 35~ In 3.23.2, Action, F

Charige "[l - 2(.....E - 1)) x Rated Power" FL to . "[l - 3 33 (F

.....!: - 1 )] x Rated Power" FL This change is required to implement the preceding radial peaking factor LCO change.

D. Chapter 4

1. Change the following in Table 4.1.1:

a. In Item l(d), add footnote 6. (Delete obsolete footnote 6

in three places in Table).

One of the changes as a result of the RPS Modification is the addition of a Variable High Power Trip (VHPT). The VHPT setpoint is maintained by the Variable High Power Function in the Thermal Margin Calculator. The addition of footnote 6 assures that this function is tested to verify proper performance at least once every 18 months.

The test will be performed by applying known power inputs (flux power and/or ~T power) to the Variable High Power Function. The Variable High Power Function will be tested for both 3 and 4 primary coolant pump operation.

Performing the above testing every.18 months is deemed adequate, since the Variable High Power Function is also being tested less extensively on a monthly basis during the performance of Item l(c).

The obsolete footnote (6) referred to a 1981 deferred surveillance.

b. In Item l(c), delete the reference to footnote 4.

The VHPT setting will be tested for the operating pump combination and power level only. Since the High Power Trip wili be changed to a VHP'i', verification of trip settings involves more extensive testing which will now be covered by Item l(d). The monthly testing required by Item l(c) will provide assurance that the settings and trip circuitry are performing per design. Values for minimum and maximum setpoints (see Table 2.3.1) for both 3 and 4 primary coolant pump operation will be checked TSOP1287-0254-NL04

13 during performance of Item 14(a)~ The testing and checks performed by Items l(d) and 14(a) provide adequate assurance that the 3 primary coolant pump VHPT settings will perform per design.

c. In Item 4(b)(l), change "known resistance substituted for RTD coincident with known pressure input."

to, "known resistance substituted for RTD coincident with known pressure and power input."

Thermal Margin/Low Pressure (TM/LP) trip settings will be a function of core inlet temperature, power level, and excore measured axial offset (ASI). Previous equipment used to calculate TM/LP settings did no_t use an ASI input, and core power was indirectly input based on core inlet and exit temperatures.

The TM/LP temperature input calibration will require a known power input (lower core power, upper core power and total core -power) in addition to a known pressure input.

Surveillance requirements for power and ASI inputs into TM/LP are adequately covered by Items 1, 12 and 13 and need not be reiterated by Item 4 for TM/LP.

The TM/LP utilizes the power and ASI as inputs into the Power Peaking Function and the Axial Function respectively. During performance of Item 4(b)(l), the Power Peaking Function and the Axial Function will be tested to verify proper performance.

d. In Item 4(c), delete reference to footnote 4.

Footnote 4 is no longer applicable for TM/LP, since the same TM/LP settings are applied to both 3 and 4 primary coolant pump operations. For 3 pump operation, TM/LP is protected by the VHPT and the 1750 psia minimum low pressure setting.

e. Add a new Item 12 to the table.

One of the changes as a result of the. RPS Modification is.

the addition of the Axial Shape Index (ASI) alarm. ASI is monitored against setpoints to assure that the core inlet temperature equation remains valid, and ASI is used as an input into TM/LP.

TSOP1287-0254-NL04

14 The ASI alarm setpoints are calculated by the LOCA Peaking Function and the Local Power Density Function, using a total core power input. The surveillance requirements for the power input are adequately covered by Items 1 and 13 and need not be reiterated by Item 12 for ASI.

The ASI value is calculated by the ASI Function based on lower core and upper core power inputs from the power range safety channels. The ASI value is checked against the total core axial offset measured by the incores at least every 31 days of power operation. This surveillance requirement is covered by 4.18.2.l(b) and need not be reiterated by Item 12, ASI.

The surveillance requirement of Item 12(a) will test the Axial Shape Index Function, LOCA Peaking Function, and the Local Power Density Function to verify proper performance.

The ASI alarming function will also be verified. The surveillance requirements given above, along with the constant check required by Item 14, provide adequate assurance that the ASI alarming function is performing per design and the ASI value is accurate.

f. Add a new Item 13 to the table.

/

./ I This item has been added since 6T Power is now a part of the RPS. Items 13(a) and (b) provide assurance that the 6T Power value is accurate. The temperature input used to calculate 6T Power is calibrated by Item 4(b)(l) and need not be reiterated by Item 13 for 6T Power. The surveil-lance requirement of Item 13(c) will provide adequate assurance that the circuitry used for the 6T Power calculation is performing per design.

g. Add a new Item 14 to the table.

At least once every 92 days the constants used by the Thermal Margin Calculator will be verified. The constants are used in calculating 6T Power, TM/LP temperature input, ASI Function, Axial Function, Power Peaking Function, LOCA Peaking Function, Local Power Density Function, VHPT minimum and maximum trip settings, TM/LP minimum pressure setting, TM/LP calculated pressure settings, and core inlet temperature Setpoirtt.

The Surveillance requirements of Items l(c), l(d),

4(a)(l), 4(b)(l), 4(c), 12(a), 13(a), 13(b), 13(c) as .well as 4.18.2.l(b) indirectly provide assurance that the correct constants are being.used. Additionally, changes to constants are administratively controlled, and a keylock must be* used to modify any constants. The surveillance requirement of Item 14(a) provides adequate TSOP1287-0254-NL04

15 assurance that correct constants are being used by the Thermal Margin Calculator.

h. Iri Footnote 3, change "Adjust the nuclear gain pot on the tiT cabinet until readout agrees with heat balance calculations".

to, "Adjust the nuclear power or tiT power until readout agrees with heat balance calculations when above 15% of rated power".

This footnote clarifies that the adjustment is applicable to both nuclear power and tiT power. Nuclear adjustment will be performed using the gain adjustment on Panel C-27.

The tiT power will be adjusted by changing a constant in the Thermal Margin Calculator. Normally the "BIAS" constant in the tiT power equation will be used to adjust tiT power. If necessary, the "Ka" constant may also be changed to adjust tiT power. Changes to constants will be controlled through Operating Procedures.

Footnote. 3 was also expanded to only require the daily check and adjustment, if necessary, to be performed when above 15% of rated power, which is consistent with the Standard CE Technical Specifications. Above 15% power, the heat balance is considered accurate enough to make the daily check meaningful.

2. Change the following in Table 4.1.3:

a. In Item 9, change "Flux - T Power Comparator" to, "F:I.ux - tiT Power Comparator".

In addition, the surveillance method for Item 9(b) is changed from "Internal Test Signal" to, "Use simulated signals".

The Flux - tiT Power Comparator will no longer have an internal test signal capability. A simulated signal will be used to verify the performance of the Flux - tiT Power alarm.*

b. In Item B(a), change " *.. all rod drive control system interlocks .*. "

to, " .*. all manual rod drive control system interlocks *.. "

The Reactor Regulating System is not used to provide automatic control of the reactor, therefore there is no TSOP1287-0254-NL04

16 need to test the automatic rod drive control system interlocks.

c. Delete footnotes (1) & (2) a.nd reference to them in Items 2c, 3c and lOa. The obsolete footnotes refer to past refueling outages.

3. Change the following in Section 4.15:

a. Change "primary steam flow" to, "primary system flow".

b. Change " **. shall be made with four primary coolant pumps in operation before the reactor is made critical".

to, " ..* shall be made with four primary coolant pumps in operation".

By omitting "before the reactor is critical", more than one method may be used to verify primary system flowrate.

The surveillance requirement of 4.15 is presently met by performing a flow measurement at hot zero power using the primary coolant pump A pressure method. A more accurate flow measurement method is the calorimetric method, which is used by most Combustion Engineering plants. The words "before the reactor is made critical" must be deleted to allow use of the

4. Change the following in 4.18.2~1:

a. Change Item b from, "The excore measured AO shall be compared to the incore measured AO. If the difference is greater than 0.02, the excore monitoring system shall be recalibrated" to "Individual excore channel measured AO shall be compared to the total core AO measured by the incores. If the difference is greater than 0.02, the excore monitoring system shall be recalibrated."

Excore measured AO was previously determined using an average value from all power range safety channels. The average value was referred to as the Power Ratio and was recorded on NR-0100. The Power Ratio was multiplied by the Shape Annealing Factor (SAF) to obtain an excore measured AO. The new equipment being installed as a result of the RPS Modification allows the excore measured AO to be determined from each individual power range TSOP1287-0254-NL04

17 safety channel. Section 4.18.2.l(b) was rewritten to cover this enhanced monitoring capability.

5. Charige the following in 4. 19 .1. 2:

a. Change Item a from "Prior to use, verify that the measured AO has not deviated from the target AO by more than 0.05 in the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ."

to "Prior to use, verify that the measured AO has not deviated from the target AO by more than 0.05 in the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for each operable channel using the previous 24 hourly recorded values".

b. Change Item d from "Once per hour, verify that the measured AO is within 0.05 of the established target AO."

to "Continuously verify that the measured AO is within 0.05 of the established target AO for at least 3 of the 4, 2 of the 3 or 2 of the 2 operable channels, whichever is the applicable case."

Due. to the equipment changes as stated in the discussion.

regarding Specification 4.18.2;1 changes, Specification 4.19.1.2 also had to be rewritten since each individual power range safety channel is now capable of monitoring Linear Heat Rate.

Specification 4.1~.l.2(a) was expanded to require the last 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months history be checked for each operable channel.

Specification 4.19.l.2(a) was also expanded to clarify that the last 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months history is verified by reviewing the last 24 hourly recorded values. Previously, the last 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months history c6uld be verified to be within 0.05 continuously using the Power Ratio Recorder, NR-0100.

This continuous 0.05 verification can no longer be performed, since the new equipment only records the last 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months history in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months intervals. Reviewing the previous 24 hourly recorded values* is sufficient to determine that the axial power distribution is near equilibrium, and significant xenon oscillations are not present.

Specification 4.19.l.2(d) was changed to require continuous verification of the +/-0.05 limit while using the excore monitoring system to monitor Linear Heat Rate. The monitoring capabilities of the new equipment will now TSOP1287-0254-NL04

18 allow this continuous verification. The ASI alarm setpoints in the Thermal Margin Calculator will be reset to conservative values to meet the +/-0.05 requirement and Figure 3.0 requirements, whichever are more restrictive.

Changes that have been made to Specification 4.19.1.2 which allow use of each operable channel for monitoring LHR and allow coincident signals to be the basis for determining when the measured AO has deviated from the target AO by more than 0.05 are supported by XN-NF-80-47.

XN-NF-80-47, "Palisades Power Distribution Control Procedures, was the original basis of 4.19.1.2.

Since the last 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months history can not be verified contin-uously, Specification 4.19.l.2(a) conservatively requires that each operable channel be within the 0.05 limit over the last 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Two coincident signals indicating greater than the 0.05 limit will not be required. Any one value outside the 0.05 limit will mean that the excores can not be used for monitoring LHR.

II DISCUSSION This Technical Specifications Change is the result of several major changes to the Reactor Protective System (RPS) and Nuclear Instrumentation System which are summarized as follows:

A. Combining the Existing Thermal Margin/Low Pressure (TM/LP), Thermal Power (dT) and Power Ratio Calculators into a Digitalized Thermal Margin Calculator Including:

1. Modifying the High Power Trip into a Variable High Power Trip (VHPT)

2. Adding Axial Shape Index Alarm (ASIA)

3. Modifying the TM/LP calculator logic B. Modifying the High Rate Trip Bypass Nuclear Power input to use Adjusted Nuclear Power Signal.

Each of these will be discussed individually as follows:

A. Installation of the Thermal Margin Calculator The Thermal Margin/Low Pressure Trip is designed to protect the fuel against failure due to the loss of heat transfer capability between TSOP1287-0254-NL04

19 the fuel and the cooling water. Poor heat tran~fer conditions can be caused by increased heat flux (power level) or decreased .inlet subcooling (increased.inlet temperature). If the power level rises rapidly above the high flux trip set point, the reactor will be tripped and the fuel will be protected. If the power level increases slowly with increasing core inlet temperature, thermal margin can be quickly lost and the fuel could experience a boiling crisis if a TM/LP trip was not provided.

The existing TM/LP trip is overly conservative in order to ensure protection of the fuel during transients starting at 50% power level. Thermal power level is determined by measuring the hot and cold leg temperatures which are affected by the time lag of the RTDs. This time lag is accounted. for by making more conservative assumptions when estab.lishing the.trip set points. With the new calculator installed, these assumptions are lessened by the addition of a penalty factor (Qd b) in the TM/LP logic and by utilizing the larger of either thermaY power or flux power. The reason for replacing the original TM/LP trip system is to: 1) obtain current technology equipment, 2) procure a system capable of measuring both thermal and nuclear power and selecting the higher value of the two, and 3) to provide a means of protecting the fuel at all power levels without placing excessive restrictions on full power operations.

The VHPT and the ASIA are designed to reduce the conservative assumptions which establish the TM/LP trip set points. Because of the ability to monitor actual plant operating conditions, fewer "worst case" assumptions must be made during the transient analysis.

The VHPT system will limit power increases to approximately 10%

above the power level at the start of the transient by continuously calculating*a VHPT set point, which is a fixed level above the existing reactor power level. During power reductions the set point follows power level down to a pre~established minimum. During power increases the set point remains fixed until manually reset by the operator. Resetting VHPT establishes a new set point at the fixed level above the existing power level. This is repeated until reaching a maximum allowable set point. The ASIA will aid the operator in limiting local power increases caused by axial power shifts while maintaining a constant core power level.

This modification provides reactor monitoring inputs resulting in a more accurate determination of plant operating conditions. The.VHPT and ASIA allow a reduction in conservative assumptions which establish the TM/LP trip setpoints. Less conservative TM/LP set-points can be used while still maintaining the same thermal limits.

Thus, there is still a 95% probability at a 95% confidence level that a Departure from Nucleate Boiling will not occur for all conditions.

This modification improves several existing functions by replacing analog calculators with digital ones while retaining four independent channels. The calculator's operation, during power and TSOP1287-0254-NL04

20 system failures,* to a safe condition ensures that a single calculator failure will not cause or prevent a safety action.

Software failure due to progrannning errors is ensured against by the thorough verification and validation effort performed under ANSI/IEEE 7.4.3.2.

Hardware Description The Thermal Margin Calculator (TMC) is a self-contained, microprocessor-based instrument which monitors and displays the nuclear reactor's coolant temperature as well as reactor power. The TMC provides trips which are based on functions of reactor power and temperature. These trips act to protect the reactor by providing safety control signals to the RPS. The TMC alarm system provides "positive operator control" for enhanced safety and plant efficiency.

The signal inputs to the TMC are analog voltages representing:

1) the temperature of the reactor coolant entering the core from each loop, 2) the average temperature of the reactor coolant leaving the core, 3) the neutro~ flux over the upper half of the core, and*

the neutron flux over the lower half of the core for each of four quadrants, and 4) the total flux over the core (Phi). The difference of the two temperatures, cited above (dT), and the calibrated sunnnation of the two neutron fluxes, cited above are independent measurements of reactor power. Setting limits combining core power and core inlet temperature ensures that the reactor fuel will not experience boiling heat transfer conditions which could lead to fuel failure.

Digital outputs are provided when any of the following conditions are encountered:

1. The reactor coolant temperature is outside of acceptable limits (alarm).

2. The difference between upper and lower neutron flux is greater.

than acceptable limits (AS! RPS alarm).

3. The reactor power is greater than acceptable limits (VHP RPS trip).

Analog output signals are provided which represent the following:

1. The calculated VHPT set point signal for indication.

2. The difference between the measurement of reactor power by neutron flux level (Phi) and the measurement of reactor power by temperature difference (dT) for indication (dT-flux deviation).

TSOP1287-0254-NL04

21

3. The minimum allowable reactor primary pressure as calculated from reactor coolant temperature, reactor power and the axial

offset of the reactor power over the height of the core. This pressure is compared with the low pressure trip set point. The larger of .these signals is used as a reference for pretrip and trip set points for Thermal Margin Low Pressure, DNB trip in the reactor protection system (TM/LP pressure signal).

The TMC inst,rument consists of an embedded computer system, a video monitor, a data entry keypad, two security keyswitches, and four function selection keys. The video monitor and the function keys

allow the operator to observe the status of TMC functions, to input the values of the adjustable parameters and trips, as well as to regulate the functions monitored by the TMC.

This modification will install one TMC in each of the four safety channels, each powered from its respective safety related instrument bus. The inputs and outputs from each TMC are either to IEEE Class lE devices or are provided with Class lE isolation devices.

The failure of any input to a TMC will not affect the other independent channel TMC's, nor will the failure of one of the TMCs.

Thus, a single failure will not initiate nor prevent a safety action*

from occurring .since both the VHPT and TMLP signals are utilized in the RPS in a two out of four logic to provide protective actions.

The TMC operates in three modes: 1) normal, 2) test, and 3) data modify. The mode of operation is selected at one of the two security keyswitches. In "normal" mode, the TMC performs its standard calculations and safety monitoring functions, while in the "test" mode, the TMC hardware automatically 'tests to insure the proper functioning of the analog and digital ports, computer memory, and data entry *keypad. The test results are displayed on the video monitor. An additional type of system testing is a continually performed on-line test operating in background during normal operation to check the Random Access (RAM) and the Read Only Memories (ROM) for memory errors.

In "data modify" mode, the numeric keypad is used to modify the values of constants and to change the internal time clock. Also in this mode, the second keylock is accessed and the numerical constants for calculations may be changed to represent three or' four primary coolant pump operations. The use of these keys and the

- changing of TMC constants will*be administrative-ly controlled.

The TMC has been designed, constructed and tested to meet IEEE standards for Class lE electrical equipment, seismic qualification and for safety systems. The digital software has been.tested per a joint ANSI/IEEE standard to verify the proper execution of its logic and the operation of its outputs to safe or trip conditions on loss of power and on failure of the operating system.

TSOP1287-0254-NL04

22 B. Modify the High Rate Trip Bypass Input Signal According to the Palisades Technical Specifications, the RPS High Rate Trip is to.be bypassed at 15% through bistable action of the power range safety channels. The nuclear signal that is fed into.

this bypass bistable is uncompensated (not calorimetrically adjusted). Therefore, as calorimetrics are performed on a daily basis, the adjusted nuclear power signal is greater in magnitude than the uncompensated signal. This allows the high rate trip bypass bistable to actuate at a nuclear power level at less than 15%, thus violating Technical Specifications. For this reason the bypass set point was set at 16.5% to assure the 15% limit would not not be exceeded. However, due to exceeding that value the bypass was set at 18% until this modification can be implemented.

Up until the installation of the VHPT modification, the adjusted nuclear power signal could not be fed to the 15% bistable since the adjusted signal could be gained by a factor of 10 at low power levels, thus allowing the bypass to occur at less than 15%. The VHPT ,modification allows removal of the power range safety channel Xl-XlO switches which, in turn, will avoid early high rate trip bypassing.

This modification only changes the input to this circuit from unadjusted to adjusted nuclear power. All adjustments are made under controlled conditions using approved plant procedures.

This modification will ensure that the margin of safety as defined in the basis for the Tech Specs is not reduced by ensuring that the set point for the bypass is adjusted each time the nuclear instrumentation is adjusted.

TSOP1287-0254-NL04

ATTACHMENT Consumers Power Company Palisades Plant Docket 50-255 DRAFT TECHNICAL SPECIFICATIONS PAGE CHANGES REACTOR PROTECTION SYSTEM December 23, 1987 37 Pages TSP1287-0253-NL04

1.1 REACTOR OPERATING CONDITIONS (Cont'd)

Axial Offset or Axial Shape Index I The difference between the power in the lower half of the core and the upper half of the core divided by the sum of the powers in the lower half and upper half of the core.

Narrow Water Gap Fuel Rod A fuel rod adjacent to the narrow interfuel assembly water gap (a gap not containing a control rod).

N Narrow Water Gap Fuel Rod Peaking Factor - F r

The maximum product of the ratio of individual fuel assembly power to core average fuel assembly power times the highest narrow water gap fuel rod local peaking factor integrated over the total core height including tilt.

l.-2a Draft TSP1287-0253-NL04

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2. t SAFETY. LIMITS - REACJ'_OR_ ~Q'lrn Applicability This specification applies to the limiting combinations of reactor power, prima~y coolant system flow, temperature and pressure during 4-pump operation. I Objective To maintain the integrity of the fuel cladding and prevent the release of significant amounts of fission products to the primary coolant.

Specifications The reactor power level shall not exceed the allowable limft for the pressurizer pressure and the maximum cold leg temperatures (shown in /

Figures 2-1 for a specified ASI) for 4-pump operation. I The safety limit is exceeded if the point defined by the combination of primary coolant cold leg temperature and power level is at any time above the appropriate pressurizer pressure line.

Basis To maintain the integrity of the fuel cladding and prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions. This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat, transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant temperature, The upper boundary of the nucleate boiling regime is termed "departu-re from nucleate boiling" (DNB). At this point, there is a sharp reduction of the heat transfer coefficient, which would result in high-cladding temperatures and the possibility of cladding failure. Although DNB is not an observable parameter during reactor operation, the observable parameters of thermal power, primary coolant flow, temperature and pressure, can be related to DNB through the use of the XNB DNB

I Correlatio~~.,(l) The XNB DNB Correlation has been developed to predict /

DNB and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of the heat flux t~at would cause DNB at a particular core location. to the actual heat flux, is indicative of the margin to DNB. The minimum

_y?lueof the DNBR, during steady-state operation, normal operational transients, and anticipated -transients is limited to. L 17. A DNBif of /

1.17 corresponds to a 95% probability at* 95% confidence level that I 2-1 Draft TSP1287-0253-NL04

2.1 SAFETY LIMITS - REACTOR CORE (Contd)

- -nNB will not occur which- is* considered an appropriate margin to- DNB for

I all operating conditions. (l) The curves of Figure 2-1 represent the I loci of points of thermal power, primary coolant system pressure, I maximum cold leg temperature at a specified ASI for which the DNBR is /

> 1.17. The area of safe operation is below these lines. I I

I I

I The reactor protective system is designed to prevent any anticipated combination of transient conditions for primary coolant system temperature, pressure and thermal power level that would result in a DNBR of less than 1.17( 3 ) The XNB DNB correlation has been shown to be I

_applicable to the Palisades Plant in Reference 2. I I

References. I I

(1) XN-NF-621(P). I (2) XN-NF-709 I (3) Updated FSAR, Section 14.1. /

2-2 Draft TSP1287-0253-NL04

2.2 SAFETY LIMITS - PRIMARY COOLANT SYSTEM PRESSURE Applicability Applies to the limit on primary coolant system pressure.

Objective To maintain the integrity of the primary coolant system and to prevent the release of significant amounts of fission product activity to the primary coolant.

Specification The primary coolant system pressure shall not exceed 2750 psia when.

there are fuel assemblies in the reactor vessel.

Basis The primary coolant system(l) serves as a barrier to prevent radionuclides in the primary coolant from reaching the atmosphere. In the event of a fuel cladding failure, the primary coolant system is the foremost barrier against the release of fission products. Establishing a system pressure limit helps to assure the continued integrity of both the primary coolant system and the fuel cladding. The maximum transient pressure allowable in the primary coolant system pressure vessel under the ASME Code,Section III, is 110% of design pressure.

The maximum transient pressure allowable in the primary coolant system piping, valves and fittings under ASA Section B31.l is 120% of design pressure. Thus, the safety limit of 2750 psia (110% of the 2500 psia 2

design pressure) has been established.( ) The settings and capacity of the secondary coolant system safety valves (985-1025 psig)( 3 ), the reactor high-pressure trip (~2400 psia) and the primary safety valves (2500-2580 psia)( 4 ) have been established to assure never reaching the primary coolant system pressure safety limit. The initial hydrostatic test was conducted at 3125 psia (125% of design pressure) to verify the integrity of the primary coolant system. Additional assurance that the nuclear steam supply system (NSSS) pressure does not exceed the safety limit is provided by setting the secondary coolant system steam dump and I bypass valves at 900 psia.

References (1) Updated FSAR, Section 4. '/

(2) _ Update_d FSAR, Section 4.3. I (3) Updated FSAR, Section 4.3.4. I (4) Updated FSAR, Section 4.3.9. I 2-3

Draft TSP1287-0253-NL04

TABLE 2.3.1

I I

Reactor Prot_g_ctj._y~_~yst_em Trip Setting Limits I I

I Four Primary Coolant Three Primary Coot~yt I Pumps Operating Pumps Operating I I

1. Varia~H High ~10% above core power, ~10% above core power I Power with a minimum setpoint with a minimum setpoint I of ~30% of rated power of ~15% rated power I and a maximum of ~106.5% and a maximum of ~49% I of rated power of rated power I _J I

2. Primary ~95% of Primary Coolant ~60% of Primary Cool- I Coolant Flow( 2 ) Flow With Four Pumps ant Flow With Four I Operating Pumps Operating I I

3. High Pressure ~2255 Psia ~2255 Psia I Pressurizer I I

4. Thermal ¥2r§f n/Low PT ~ Applicable Limits Replaced by Variable I Pressure ' specified in Figure 2-1 High Power Trip and I for a specified ASI 1750 Psia Minimum Low- I Pressure Setting I I

5. Steam Generator Not Lower Than the Cen- Not Lower Than the Cen- I Low Water Level ter Line of Feed-Water ter Line of Feed-Water I Ring Which ls Located Ring Which Is Located I 6'-0" Below Normal 6'-0" Below Normal I Water Level Water Level I I

6. Steam Generaf 2} ~500 Psia ~500 Psia I Low Pressure I I

7. Containment High ~3.70 Psig ~3.70 Psig I Pressure I

/

(1) The VHPT can be > 10% above core power for power levels ~ 20% of rated I power.

4 I (2) May be bypassed below 10- % of rated power provided auto bypass removal circuitry is operable. . For low power physics te,sts, thermal margin/low pressure and low steam generator pressure trips may be bypassed until their react points are reached (approximately 1750 psia and 500 Efia_,

respectively), provided automatic bypass removal circuitry at 10 % rated power is operable.

(3) -Minimum trip setting shall be 1750 psia. I (4) Operation with three pumps is permitted to provide a limited time for I repair/pump restart, to provide for an orderly shutdown or to provide for I the conduct of reactor internals noise monitoring test measurements. I I

2-5 Draft TSP1287-0253-NL04

2.3 LIMITING SAFETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM (Contd)

__ BasiS The reactor protective system consists of four instrument channels to monitor selected plant conditions which will cause a reactor trip if any of these conditions deviate from a preselected operating range to the degree that a safety limit may be reached.

1. Variable High Power - The variable high power trip (VHPT) is I incorporated in the reactor protection system to provide a reactor I trip for transients exhibiting a core power increase starting from I any initial power level (such as the boron dilution transient). I The VHPT system provides a trip setpoint no more than a I predetermined amount above the indicated core power. Operator I action is required t6 increase the setpoint as core powe~ is I increased; the setpoint is automatically decreased as core power I decreases. Provisions have been made to select different set points I for three pump and four pump operations. I During normal plant operation with all primary coolant pumps operating, reactor trip is initiated when the reactor power level reaches 106.5% of indicated rated power. Adding to this the possible variation in trip point due to calibration *and instrument errors, the maximum actual steady state power at which a trip would be actuated is 112%, which was used for the purpose of safety analysis. (1)

I I

I

2. Primary Coolant System Low Flow - A reactor trip is provided to I protect the core agains~ 9NB should the coolant flow suddenly decrease significantly.

3 Flow in each of the four coolant loops I is determined from a measurement of pressure drop from inlet to outlet of the steam generators. The total flow through the reactor core is measured by summing the loop pressure drops across the steam generators and correlating this pressure sum with the pump calibration flow curves. The percent of normal core flow is shown in the following table: I 4 Pumps 100.0%

3 Pumps 74.7%

I During four-pump operation, the low-flow trip setting of 95%

insures that the reactor cannot operate when the flow rate is less thtH)93% of the nominal value considering instrument errors. I 2-6 Draft TSP1287-0253-NL04

2.3 LIMITING SAFETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM (Contd).

Provisions are made in the reactor protective system to permit I operation of the reactor at reduced power if one coolant pump is I taken out of service. These low-flow and high-flux settings have I been derived in consideration of instrument errors and response I times of equipment involved to assure that thermal margin and flow I stability wf transients.

st be maintained during normal operation and anticipated For reactor operation with one coolant pump I

I inoperative, the low-flow trip points and the overpower trip points I must be manually changed to the specified values for the selected I pump condition by means of set point selector switches. The trip I points are shown in Table 2.3.1. I

3. High Pressurizer Pressure - A reactor trip for high pressurizer I pressure is provided in conjunction with the primary and secondary safety valves to prevent primary system overpressure (Specification 3.1.7). In the event of loss of load without reactor trip, the temperature and pressure of the primary coolant system would increase due to the reduction in the heat removed from the coolant via the steamgenerators. This setting is consistent with the I trip point assumed in the accident analysis. (U)

4. Thermal Margin/Low-Pressure Trip - The TM/LP trip system monitors I core power, reactor coolant maximum inlet temperature, core coolant I system pressure and axial shape index. The low pressure trip limit I (P ) is calculated using the following equation for comparison I var . (12) with the measured reactor coolant pressure. /

I Pvar = ****** (QA)(QR ) + ****** T 1 in - ****

I I

Where: I I

QRl *** + *** Q ~ 1.0 I Q Q ~ 1.0 and I I

QA .****(ASI) + **** *** :S ASI ~ **** I

= -.****(ASI) + ** **** - *** ;S ASI ;S *** I

-.****(ASI) + ***** -.*** ~ ASI ;S -.*** I I

The calculated limit is then compared to a fixed low pressure trip I limit (Pi). The auctioneered highest of these signals becomes the I

__ ~rip lim1fcn(Ptri). P_t i __ is _compar:ed t<:> the ~eas~red _reactor I coolant pressurep(P) ana R trip signal is generated when p is less I - --

than or equal to P i

A pre-trip alarm is also generated when P /

is less than or eqS~rPto the pre-trip setting P i + ~P. I tr p 2-7 Draft TSP1287-0253-NL04

2.3 LIMITING SAFETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM (Continued)

Basis (Continued)

The TM/LP trip set points are derived from the 4-pump operation l core thermal limits (Figure 2-1) through application of appropriate I allowances for measurement uncertainties and processing errors. i A pressure allowance of 165 psi is assumed to account for: I instrument drift in both power and inlet temperatures; I calorimetric power measurement; inlet temperature measurement; and I primary system pressure measurement. Uncertainties accounted for. I that are not a part of the 165 psi term include allowances for: I assembly power tilt; fuel pellet manufacturing tolerances; core I flow measurement uncertainty and core bypass; inlet temperature I measurement time delays; and ASI measurement. Each of these I allowances and uncertainties are included in the development of I the TM/LP trip set point used in the accident analysis. I For three-pump operation, power is limited to 39% of rated power I for a maximum of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . During this mode of operation, the I high power level trip in conjunction with the TM/LP trip (minimum I set point = 1750 psia) and the secondary system safety valves ( )

I 5

(set at 1000 psia) assure that adequate DNB margin is maintained. I 1

I I

I

5. Low Steam Generator Water Level - The low steam generator water level reactor trip protects against the loss of feed-water flow accidents and assures that the design pressure of the primary coolant system will not be exceeded. The specified set point assures that there will be sufficient water inventory in the steam generator at the time of trip to allow a safe and orderly I plant shutdown and to prevent steam generator dryout assuming I minimum auxiliary feedwater capacity. <9 ) I The setting listed in Table 2.3.1 assures that the heat transfer surface (tubes) is covered with water when the reactor is critical.

2-8 Draft TSP1287-0253-NL04

2.3 LIMITING SAFETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM (Contd)

Basis (Contd)

6. Low Steam Generator Pressure - A reactor trip on low steam generator secondary pressure is provided to protect against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the primary coolant.* The setting of 500 psia is sufficiently below the rated load operating point of 739 psia so as not to interfere with normal operation, but still high enough to provide the required protection in the event of excessively high steam flow. This setting was used in the accident analysis. ( 8 ) .

7. Containment High Pressure - A reactor trip on containment high pressure is provided to assure that the reactor is shut down upon the initiation of the safety injection system. The setting of this trip is identical to that of the containment high-pressure safety injection signal. (lO)

. 8. Low Power Physics Testing - For low power physics tests, certain tests will require the reactor to be critical at low temperature

(> 260°F) and low pressure(> 415 psia). For these certain tests only, the thermal margin/low-pressure, and low steam generator pressure trips may be_ bypassed in order that reactor power can be increased for improved data acquisition. Special operating precautions will be in effect during these tests in acc.ordance with approved written testing procedures. At reactor power levels below

-1 .

10 % of rated power, the thermal margin/low-pressure trip is not required to prevent fuel rod thermal limits from being exceeded.

The low steam generator pressure trip is not required because the low steam generator pressure will not allow a severe reactor cooldown, should a steam line break occur during these tests.

References (1) deleted I (2) ANF-87-105(P), Volume 2, Table 15.0.7-1 I (3) FSAR, Section 7.2.3.3. I (4) ANF-87-150(P), Volume 2, Section 15.3 I (5) XN-NF-86-.91 (P) I (6) deleted I (7) deleted I (8) XN-NF-77-18, Section 3.8 I (9) ANF-87-150(P), Volume 2, Section 15.2.7 I (10) FSAR, Section 7. 2. 3. 9: I (11) ANF-87-150(P), Volume 2, Section 15.2.1 I (12) ANF-87-150(P), Volume 2, Section 15.0.7.2 I I

I I

(next page is 2-13) 2-9 Draft TSP1287-0253-NL04

FIGURE 2-1 Reactor Core Safety Limits 4 Pump Operation GRAPH TO BE PROVIDED AFTER COMPLETION OF SAFETY ANALYSIS 2-13 Draft TSP1287-0253-NL04

3.1 PRIMARY COOLANT SYSTEM Applicability Applies to the operable status of the primary coolant system.

Objective

To specify certain conditions of the primary coolant system which must be met to assure safe reactor operation.

Specifications 3 .1.1 Operable Components

a. At least one* primary coolant pump or one shutdown cooling pump with a minimum flow rate of 1500 gpm shall be in operation I whenever a change is being made in the boron concentration of the primary coolant.

b. Four primary coolant pumps shall be in* operation whenever the I reactor is operated continually above hot shutdown. I*

I Before removing a pump from service, thermal power shall be I reduced as specified in Table 2.3.1 and appropriate corrective I action implemented. With one pump out of service, return the I pump to service (return to four-pump operation) or be in hot I shutdown (or below) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Start-up (above hot I shutdown) with less than four pumps is not permitted and power I operation with less than three pumps is not permitted. I I

c. The measured four primarg coolant pumps operating reactor vessel I flow shall be 124.3 x 10 lb/hr or greater, when corrected to I 532°F. I I

I I

d. Both steam generators shall be capable of performing their heat transfer function whenever the average temperature of the primary coolant is above 325°F.

e. Maximum primary system pressure differentials shall not exceed the following:

(1) Maximum steam generator operating differential of 1380. psi. I 3-lb Draft TSP1287-0253-NL04

3 .1. PRIMARY COOLANT SYSTEM (Continued) 3 .1.1 Operable Components (C~ntinued)

(2) Hydrostatic tests shall be conducted in accordance wtth applicable paragraphs of Section XI ASME Boiler &

Pressure Vessel Code (1974). Such tests shall be conducted with sufficient pressure on the secondary side of the steam generators to restrict primary to secondary

pressure differential to a maximum of 1380 psi. Maximum hydrostatic test pressure shall not exceed 1.1 Po plus SO psi where Po is nominal operating pressure:

(3) Primary side leak tests shall be conducted at normal operating pressure. The temperature shall be consistent with applicable fracture toughness criteria for ferritic materials and shall be selected such that the differential pressure across the steam generator tubes is not greater than 1380 psi.

(4) Maximum secondary hydrostatic test pressure shall not exceed 1250 psia. A minimum temperature of 100°F is required. Only ten cycles are permitted.

(5) Maximum secondary leak test pressure shall not exceed 1000 psia. A minimum temperature of 100°F is required.

(6) In performing the tests identified in 3.l.l.e(4) and 3.1.1.e(S), above, the secondary pressure shall not exceed the primary pressure by more than 350 psi.

f. Nominal primary system operation pressure shall not exceed 2100 psia.

g. The reactor inlet temperature (indicated) shall not exceed the value given by the following equation at steady state 100%

power operation:

Ti 1 ~ 543.3 + .057S(P-2060) + O.OOOOS(P-2060)**2 + l.173(W-120) - I n et .0102(W-120)**2 I Where: Tinlet = reactor inlet temperature in F 0 P = nominal operating pressure in psia .

6 W = total recirculating mass flow in 10 lb/h corrected to the operating temperature conditions.

Note: This equation is valid when ASI is maintained within I the bqundaries shown in Figure 3.0 I When the ASI exceeds the limits specified in Figure 3.0, within I 15 minutes, initiate corrective actions to restore the ASI to I the acceptable region. Restore the ASI to acceptable values I within one hour or be at less than 70% of rated power within I two hours. I 3-lc Draft TSP1287-0253-NL04

3.1 PRIMARY COOLANT SYSTEM (Cont'd) 3.. 1. l Operable Components (Cont'd)

h. A reactor coolant pump shall not be started with one or more of the PCS cold leg temperatures ~ 250°F unless 1) the pressurizer water volume is less than 700 cubic feet or 2) the secondary water temperature of each steam generator is less than 70°F .above each of the PCS cold leg temperatures.

i. The PCS shall not be heated or maintained above 325°F unless a minimum of 375 kW of pressurizer heater capacity is available from both buses lD and lE. Should heater capacity from either bus lD and lE fall below 375 kW, either restore the inoperable heaters to provide at least 375 kW of heater capacity from both buses lD and lE within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be. in hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Basis When primary coolant boron concentration is being changed, the process must be uniform throughout the primary coolant system volume to prevent stratification of primary coolant at lower boron concentration which could result in a reactivity insertion.

Sufficient mixing of the primary coolant is assured if one shutdown cooling or one primary coolant pump is in operation. (l) The shutdown coolant pump will circulate the primary system volume in less than 60 minutes when operated at rated capacity. By imposing a I minimum shutdown cooling pump flow rate of 1500 gpm, sufficient time I is provided for the operatof 6jo terminate the boron dilution under I asymmetric flow conditions. The pressurizer volume is relatively I inactive, therefore will tend to have a boron concentration higher than rest of the primary coolant system during a dilution operation.

Administrative procedures will provide for use of pressurizer sprays to maintain a nominal spread between the boron concentration in the pressurizer and the primary system during the addition of boro~. ( 2 )

The FSAR safety analysis was performed assuming four primary coolant I pumps were operating for accidents that occur during reactor I operation. Therefore, reactor startup above hot shutdown is not I permitted unless all four primary coolant pumps are operating. I Operation with less than four primary coolant pumps is permitted for I a limited time to allow the restart of a stopped pump or for reactor. I internals vibration monitoring and testing. I Both steam .generators are required to be operable whenever the temperature of the primary coolant is greater than the design temperature of the shutdown cooling system to assure a redundant heat removal system for the reactor.

3-ld Draft TSP1287-0253-NL04

3.1 PRIMARY COOLANT SYSTEM (Contd)

Calculations have been performed to demonstrate that a pressure differential of 1380 psi( 3 ) can be withstood by a tube uniformily /

thinned to 36% of its original nominal wall thickness (64% degradation), while maintaining:

(1) A factor of safety of three between the actual pressure differential and the pressure differential required to cause bursting.

(2) Stresses within the yield stress for Inconel 600 at operating temperature.

I (3) Acceptable stresses during accident conditions.

I I

I Secondary side hydrostatic and leak testing requirements are consistent with ASME BPV Section XI (1971). The differential maintains stresses in the steam generator tube walls within code allowable stresses.

The minimum temperature of 100°F for pressurizing the steam generator ..

secondary side is set by the NDTT of the mayway cover of + 40°F.

The transient analyses were performed assuming a vessel flow at hot zero power (532°F) of 124.3 x 10 6 lb/hr minus 6% to account for flow /

measurement uncertainty and core flow bypass. A DNB analysis was I performed in a para-metric fashion to determine the core inlet /

temperature as a function of pressure and flow for which the /

minimum DNBR is equal to 1 .17. This analysis includes. the following uncertainties and allowances: 2% of rated power for power /

measurement; +/-0.06 for ASI measurement; +/-50 psi for pressurizer I pressure; +/-7°F for inlet temperature; and 3% measurement and 3% I bypass for core flow. In addition, transient biases were included in I the derivatio~ yf the following equation for limiting reactor inlet 4

I temperature: . I I

Tinlet 2 543.3 + .0575(P-2060) + 0.00005(P-2060)**2 + l.173(W-120) - /

.0102(W-120)**2 I I

The limits of validity of this equation are: I 1800 < Pressure < 2200 Psia I 100.0-x 10 6 < Vessel Flow < 135 x 106 Lb/h /

ASI as shown in Figure 3.0 I 3-2 Draft TSP1287-0253-NL04

3.1 e

PRIMARY COOLANT SYSTEM (Contd)

The Axial Shape Index alarm channel is being used to monitor the AS! I to ensure that the* assumed axial power profiles used in the I development of the inlet temperature LCO bound measured axial power I profiles. The signal represent'ing core power (Q) is the auctioneered I higher of the neutron flux power and the Delta-T power. The measured I AS! calculated from the excore detector signals and adjusted for I shape annealing and the core power constitute an ordered pair (Q,Y ). I An alarm signal is activated before the ordered pair exceed the 1 I boundaries specified in Figure 3.0. I The restrictions on starting a Reactor Coolant Pump with one or more PCS cbld legs S 250°F are provided to prevent PCS pressure transients, caused by energy additions from the secondary system, which would exceed the limits of Appendix G to 10 CFR Part 50. The PCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 70°F above each of the PCS cold leg temperatures. <5 ).

References (1) FSAR, Sections 6.1.2.3 and 14.3.2 (2) FSAR, Section 4~3.7 (3) Palisades 1983/1984 Steam Generator Evaluation and Repair Program I Report, Section 4, April 19, 1984 I (4) ANF-87-150(P), Volume 2, Section 15.0.7.1 /

(5) "Palisades Plant Overpressurization Analysis," June, 1977, and "Palisades Plant Primary Coolant System Overpressurization Subsystem Description," October, 1977 (6) ANF-87-150(P), Volume 2, Section 15.4.6.3.2 /

3-3 Draft TSP1287-0253-NL04

FIGURE 3-0 H._i~I LCO FOR Tinlet FUNCTION

1. 25 F

R H

('*.

T UNACCEPTABLE I

0 OPERATIONS N 1 / ..... r, ii t. .j w 0 /"

I w

o' F ,/"

I R... "l*

H T '//

..t BREAK POINTS E 0.75 ./ ACCEPTABLE * .

./

D .I 1 OPERATIONS 1. -. 342, 0.7 p ..

,.., _:I 080, 1. 0 '

0 IJJ 3. +.653, 1. 0 t:J E

'1 Ill R HJ ct ' u., ..C'J

..r*, 5* -0.25 0 ().,_.I *;1c:

w.J 0.5 0.75

'*' I AXIAL *SHAPE INDEX

3.1 PRIMARY COOLANT SYSTEM (Contd) 3 .1. 7 Primary and Secondary Safety Valves Specifications.

a. The reactor shall not be made critical unless all three pressurizer ~afety valves are operable with their lift settings maintained between 2500 psia and 2580 psia (+/- 1%).

b. A minimum of one operable safety valve shall be installed on the pressurizer whenever the reactor head is on the vessel.

c. Whenever the reactor is in power operation, a minimum of 23 secondary system safety valves shall be operable with their lift settings between 985 psig (+/- 10 psig) and 1025 (+/- 1%) psig.

Basis The primary and secondary safety valves pass sufficient steam to limit the primary system pressure to 110 percent of design (2750 psia) following a complete loss of turbine generator (l) load without simultaneous reactor trip while operating at 2650 MWt.

The reactor.is assumed to trip on a "High Primary Coolant System Pressure" signal. To determine the maximum steam flow, the only other pressure relieving system assumed operational is the secondary system safety valves. Conservative values for all system parameters, delay times and core moderator coefficient are assumed.

Overpressure protection is provided to the portions of the primary coolant system which are at the highest pressure considering pump head, flow pressure drops and elevation heads.

If no residual heat were removed by any of the means available, the amount of steam which could be generated at safety valve lift pressure would be less than half of one valve's capacity. One valve, therefore, provides adequate defens~ against overpres-surization when the reactor is subcritical.

The total relief capacity of the 24 secondary system safety valves is 6

11.7 x 10 lb/h. This is based on a steam flow equivalent to an NSSS power level of 2650 MWt at the nominal 1000 psia* valve lift pressure.

At the power rating of 2530 MW , a relief capacity of less than 6 . t 11.2 x 10 lb/h is required to prevent overpressurizatlon of the secondary system of loss of load conditions, and 23 yalves provide

. 6 (1 2) relieving capability of 11.2 x 10 lb/h. '

The ASME Boiler and Pressure Vessel Code,Section III, 1971 edition, Paragraph NC-7614.2(a) allows the specified tolerances in the lift pressures of safety valves.

References (1) FSAR, Sections 4.3.4 and 4.3.7.

(2) ANF-87-150(P), Volume 2, Section 15.2.1 I 3-25 Draft TSP1287-0253-NL04

THREE OR FOUR PUMP OPERATION 100 90

~

80 70 0

~

N

~ 60 Q

.,. so cIU 40 I

!u c

Ill 30 20

~

10

-~ ......

.40..,.. (!)

0 20 .IO IO 100: O 20 I:\ 40 1

..,.. \!I 0~*------**------*1-20 40.--------11

... GI°....----....

.801----........1001 CONTROL ROD INSERTION LIMITS

' u

. 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS (Contd)

Basis (Contd)

For a control rod. misaligned up to 8 inches from the remainder of the banks, hot channel factors will be well within design limits:

If a control rod is misaligned by more than 8 inches, the maximum reactor power will be reduced so that hot channel factors, shutdown margin and ejected rod worth limits are met. If in-core detectors are not available to measure power distribution and rod misalignments >8 inches exist, then reactor power must not exceed 75% of rated power to insure that hot channel conditions are met.

Continued operation with that rod fully inserted will only be permitted if the hot channel factors, shutdown margin and ejected rod worth limits are satisfied.

In the event a withdrawn control rod cannot be tripped, shutdown margin requirements will be maintained by increasing the boron concentration by an amount equivalent in reactivity to that control rod. The deviations permitted by Specification 3.10.7 are required in order that the control rod worth values used in the reactor physics calculations, the plant safety analysis, and the Technical Specifications can be verified. These deviations will only be in effect for the time period required for the test being performed.

The testing_ interval during which these deviations will be in effect will be kept to a minimum and special operating precautions will-be in effect during these deviations in accordance with approved written testing procedures.

Violation of the power dependent insertion Jimits, when it is necessary to rapidly reduce power to avoid or.minimize a situation harmful to plant personnel or equipment, is acceptable due to the brief period of time that such a violation would be expected to exist, and due to the fact that it is unlikely that core operating limits such as thermal margin and shutdown margin would be violated as a result of the rapid rod insertion. Core thermal margin will actually inc.rease as a result of the rapid rod insertion. In addition, the required shutdown margin will most likely not be violated as a result of the rapid rod insertion because present power dependent insertion limits result in shutdown margin in excess of that required by the safety analysis. I References (1) FSAR, Section 14.

(2) FSAR, Section 3.3.3.

(3) FSAR, Sectfon 7 :4. 2. 2.

(4) FSAR, Section 7.3.3.6.

I 3-63 Draft TSP1287-0253-NL04

.* POWER DISTRIBUTION INSTRUMENTATION 3.11.2 EXCORE POWER DISTRIBUTION MONITORING SYSTEM LIMITING CONDITION FOR OPERATION Basis (Contd)

Surveillance requirements ensure that the instruments are calibrated to agree with the incore measurements and that the target AO is based on the current operating conditions. Updating the Excore Monitoring APL ensures that the core LHR limits are protected within the +/- 0.05 band on AO. The APL considers LOCA based LHR limits, and factors I are included to account for changes in radial power shape and LHR limits over the calibration interval.

The APL is determined from the following:

LHR(Z)TS APL = [ ] x Rated Power LHR(Z)Max x V(Z) x Ep(Z) x 1.02 Min Where:

(1) LHR(Z)TS is the limiting LHR vs Core Height (from Section 3.23.1),

(2) LHR(Z)M is the measured peak LHR including uncertainties vs ax Cor'e Height, (3) V(Z) is the function (shown in Figure 3.11-1),

(4) E (Z) is a factor to account for the reduction of allowed LHR p

in the peak rod with increased exposure (Figure 3.23.2) such that:

For fuel rod burnups less-than 27.0 GWd/MT - Ep 1.0 For fuel rod burnups greater than 27.0 GWd/MT but less than.

33.0 GWd/MT - E 1.0 + 0.0064 x LHR p

For fuel rod burnups greater than 33.0 GWd/MT - - E p

= 1.0 +

0.0012 x LHR Where LHR is the measured fuel rod average LHR in kW/ft, 3-66b Draft TSP1287-0253-NL04

POWER DISTRIBUTION INSTRUMENTATION 3.11.2 EXCORE POWER DISTRIBUTION MONITORING SYSTEM LIMITING CONDITION FOR OPERATION Basis (Contd)

(S) The factor of 1.02 is an allowance for the effects of upburn, (6) The quantity in brackets is the minimum value for the entire core at any elevation (excluding the top and bottom 10% of core) considering limits for peak rods, interior fuel rods and narrow water gap fuel rods. E (Z) is only applied if the p

minimum value is based on limits for the peak rod. If the quantity in brackets is greater than one, the APL shall be the rated power level.

Reference (1) XN-NF-80-47 (2) ANF-87-lSO(P), Volume 2, Table 15.0.7.2-1 I 3-66c Draft TSPl287-0253-NL04

3.17 INSTRUMENTATION AND CONTROL SYSTEMS (Contd)

If the bypass is not effected, the out-of-service channel (Power Removed) assumes a tripped condition (except high rate-of-change power, variable high power and high pressurizer pressure), (l) which /

results in a one-out-of-three channel logic. If, in the 2 of 4 logic system of either the reactor protective system or the engineered safeguards system, one channel is bypassed and a second channel manually placed in a tripped condition, the resulting logic is 1 of 2. At rated power, the minimum operable variable high power I level channels is 3 in order to provide adequate flux tilt detection.

If only 2 channels are operable, the reactor power level is reduced t'o 70% rated power which protects the reactor from possibly exceeding design peaking factors due to undetected flux tilts and from exceeding dropped rod peaking factors. I The engineered safeguards system provides a 2 out of 4 logic on the signal used to actuate the equipment connected to each of the 2 emergency diesel generator units.

Two start-up channels are available any time reactivity changes are deliberately being introduced into the reactor and the neutron power is not visible on the log~range nucle~r instrumentation or above

-4 .

10 % of rated power. This ensures that redundant start-up instrumentation is available to operators to monitor effects of reactivity changes when neutron power levels are only visible on the start-up channels. In the event only one start-up range channel is available and the neutron power. level is sufficiently high that it is being monitored by both channels of log-range instrumentation, a startup can be performed in accordance with footnote (d) of Table 3.17.4.

References (1) FSAR, Section 7.2.2. I I

3-77 Draft TSP1287-0253-NL04

Table 3.17.1 Instru~entation Operating Requirements for Reactor Protective System Minimum Minimum . Permissible Operable Degree of Bypass No Functional Unit Channels Redundancy Conditions 1 Manual (Trip 2 None None Buttons) 2 Variable High None I Power Level I 3 Log Range 2 1 Below 10- 4 7.(e) or Channels AbovecHi. Rated Power Except as Noted in (c) 4 Thermal Margin/ 1 Below 10- 4 7.(e) of Low~Pressurizer Rated Power(a)

Pressure 5 High-Pressurizer 1 None Pressure 6 Low Flow Loop 1 Below 10- 4 7.(e) of Rated Power(a) 7 Loss of Load 1 None None 8 Low Steam Gen- l/Steam None erator Water Generator Level 9 Low Steam Gen- l/Steam Below 10- 4 7.(e) of erator Pressure Generator Rated Power(a) 10 High Containment 1 None Pressure (a) Bypass automatically removed.

(b) One of the inoperable channels must be in the tripped condition.

(c) Two channels required if TM/LP, low steam generator or low-flow channels are bypassed.

(d) If only two channels are operable, load shall be reduced to 70% or less of rated power. _4 -1 (e) For low power physics testing, 10 i. may be increased to 10 i..

3-78 Draft TSP1287-0253-NL04

Table 3.17.4 (Cont'd)

Minimum Minimum Permissible Operable Degree of Bypass No Functional Unit Channels Redundancy Conditions

8. Pressurizer Water 2 1 Not required in Level (LI-0102) Cold or Refuel-ing Shutdown

9. Pressurizer Code 1 per None Not Required Safety Relief Valves Valve below 325°F Position Indication (Acoustic Monitor or Temperature Indication)

10. Power Operated Relief 1 per None Not required when Valves (Acoustic Valve PORV isolation valve Monitor or Temperature is closed and its Indication) indication system is operable

11. PORV Isolation Valves 1 per None Not required when Position Indication Valve reactor is depressurized and vented through a vent ~1.3 sq.in .

12. Subcooling Margin 1 None . Not required Monitor . below 515°F

13. Auxiliary Feed Flow 1 per flow (h) None Not required Rate Indication Control below 325°F Valve

14. Auxiliary Feedwater 2 per sterm 1 Not required Actuation System generator e) below.325°F Sensor Channels

15. Auxiliary Feedwater 1 Not required Actuation System below 325°F Actuation Channels

16. Excore Detector None Not Required Below 50% I Deviation Alarms of Rated Power I

17. Axial Shape Index 2 1 Not Required Below 50% I of Rated Power I (e) Auxiliary Feedwater System Actuation System Sensor Channels contain pump auto initiation circuitry . . If two sensor channels for one steam generator are inoperable, one of the steam generator low level bistable modules in one of the inoperable channels must be in the tripped condition.

3-8la Draft TSP1287-0253-NL04

Table 3.17.4 (Cont'd)

(f) With one Auxiliary Feedwater Actuation System Actuation Channel inoperable, in lieu of the requirement of 3.17.2, provide a second licensed operator in the control room within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . With both inoperable, in lieu of following the requirements of 3.17.2, start and maintain in operation the turbine driven auxiliary feed pump.

(g) Calculate the Quadrant Power Tilt using the excore readings at I least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the excore detectors deviation alarms I are inoperable, or at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months using symmetric incore I detectors when the difference between the excore and the incore I measured Quadrant Power Tilt exceeds 2%. I (h) With two flow rate indicators inoperable for a given control valve, the control valve shall be considered inoperable and the requirements of 3.S.2(e) apply.

3-8lb Draft TSP1287-0253-NL04

t POWER DISTRIBUTION LIMITS 3.23.1 LINEAR HEAT RATE (LHR)

LIMITING CONDITION FOR OPERATION ACTION 3:

If the incore alarm system is inoperable and the excore monitoring system is not being used to monitor LHR, operation at less than or I equal to 85% of rated power may continue provided that incore readings are recorded manually. Readings shall be taken on a minimum of 10 individual detectors per quadrant (to include 50% of the total number of detectors in a 10-hour period) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and at least every 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thereafter. If readings indicate a local power level equal to or greater than the alarm setpoints, the action specified in ACTION 1 above shall be taken.

Basis The limitation of LHR ensures that, in the event of a LOCA, the peak temperature ot the cladding will not exceed 2200°F. (l) I I

I Either of the two core power distribution monitoring systems (the incore alarm system or the excore monitoring system) provides adequate monitoring of the core power distribution and is capable of verifying that the LHR does not exceed its limits. The incore alarm system performs this function by continuously monitoring the local power at many points throughout the core and comparing the measurements to predetermined setpoints above which- the limit on LHR could be exceeded. The excore monitoring system performs this function by providing comparison of the measured core AO with predetermi~ed AO limits based on incore measurements. An Excore Monitoring Allowable Power Level (APL), which may be less than rated power, is applied when using the excore monitoring system to ensure

that the AO limits adequately restrict the LHR to less than the 4

limiting values. < )

If the incore alarm system and the excore monitoring system are both inoperable, power will be reduced to provide margin betwe~~ the actual peak LHR and the LHR limits and the incore readings will be manualiy col.lected at the terminal blocks in the control room utilizing a suitable signal detector. If this is not feasible with the manpower available, the reactor power will be reduced to a point below which it is improbable that the LHR limits could be exceeded.

3-104 Draft TSP 1287-0253-NL04

POWER DISTRIBUTION LIMITS 3.23.1 LINEAR HEAT RATE (LHR)

LIMITING CONDITION FOR OPERATION Basis (Contd)

The time interval of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and the minimum of 10 detectors per quadrant are sufficient to maintain adequate surveillance of the core power distribution to detect significant changes until the monitoring systems are returned to service.

To ensure that the design margin of safety is maintained, the determination of both the incore alarm setpoints and the APL takes into account a measurement uncertainty factor of 1.10, an engineering uncertainty factor of 1.03, a thermal power measurement uncertainty factor of 1.02 and allowance for quadrant tilt.

References

. (1) XN-NF-77-24 (2) deleted I (3) deleted I (4) XN-NF-80-47 I

(next page is 3-107) 3-105 Draft TSP1287-0253-NL04

~

1.1 I I I I I

' I r-*r I I .,....--y--1-*-1-"~~-r~~**-***r-*

~

....*1 - -..... ..

UMACCEPTABLE -.

lo 1 ......... 0.PERATION lo .....

lo ......

- ' J ; *~

(l ....

lo-l ie 3 ACCEPTABLE ..........

I- .*\ R OPERATION .

JC *._. I 1_ !-' .... .,_ -

w *i "' *. ...., .'

....I 0 ... .:

O> 0 B . .., .

E'REAY\ *., POINTS ....

- (). i lo- " -

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-5l ie (\

'*'I b- -

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-i ..

~

~ lo lo

i 0.5 lo

. ' I I . I I ' .

I ' .

c 0 0.2 0.4 0.6 0.8 1 LOCATIOI Of AXIAL POWEi PEAi (FRACTION OF ACTIVE FUEL HEIGHT)

ALloWABLE lHI AS A FUNCTION ..

Palisades FIGllllE 3. 23-1 OF PEAi POWER lOCATIOH Technlcal Spec;:lflcQtlone

e POWER DISTRIBUTION LIMITS 3.23.2 RADIAL PEAKING FACTORS LIMITING CONDITION FOR OPERATION The radial peaking factors~. FT, ~and Ft.H shall be less than or r r r r equal to.the value in Table 3.23-2 times the following quantity. I The quantity is [1.0 + 0.3 (1 - P)] for P ~ .5 and the quantity is I 1.15 for P < .5. P is the core thermal power in fraction of rated I power.

APPLICABILITY: Power operation above 50% of rated power.

ACTION:

With any radial peaking factor exceeding its limit within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , reduce thermal power to less than the lowest value of:

F I

[l - 3

33 (_!. - 1 )] x Rated Power I FL I Where F r

is the measured value of either yAr' FTr' ~r or Frt.H arid FL is the corresponding limit from Table 3.23-2.

Basis

_A T t.H N The limitations on r--, F , F and F are provided to ensure r r r r that assumptions used in the analysis for establishing DNB margin, LHR and the thermal margin/low-pressure and high-power trip set points remain valid during operation. Data from the incore detectors are used for determining the measured radial peaking factors. The periodic surveillance requirements for determining the measured radial peaking factors provide assurance that they remain within prescribed limits. Determining the measured radial peaking factors after each fuel loading prior to exceeding 50% of rated power provides additional assurance that the core is properly loaded.

3-111 Draft TSP1287-0253-NL04

TABLE 4.1.1 Minimum Frequencies for Checks, C.alibrations and Testing of Reactor Protective System(S)

Surveillance Channel Description Function Frequency Surveillance Method

1. Power Range Safety Channels a. Check s a. Comparison of four*power channel readings.

b. Check(3) D b. Channel adjustment to agree with heat balance calculation. Repeat whenever flux-AT power comparators alarms.

c. Test M(2) c. Internal test signal.

d. Calibrate (6) R d. Channel alignment through measurement/adjustment of internal test points.

2. Wide-Range Logarithmic a. Check s a. Comparison of both wide-range readings.

Neutron*Monitors b. Test p b. Internal test signal.

3. Reactor Coolant Flow a. Check s a. Comparison of four separate total flow indications.

b. Calibrate R b. Known*differential pressure applied to sensors. I

c. Test M(2) c. Bistable trip tester.(1)(4)

4. Thermal Margin/Low a. Check: s a. Check:

Pressurizer Pressure (1) Temperature (1) Comparison of four separate calculated Input trip pressure set point indications.

(2) Pressure (2) Comparison of four pressurizer pressure Input indications. Same as S(a) below.)

b. Calibrate R b. Calibrate:

(1) Temperature (1) Known resistance substituted for RTD coinci-Input dent with known pressure and power input. I (2) Pressure (2) Part of S(b) below.

Input

c. Test M(2) c. Bistable trip tester.(l) I

s. High-Pressurizer Pressure a. Check s a. Comparison of four separate pressure indicati~ns.

b. Calibrate R b. Known pressure applied to sensors. I

c. Test M(2) c. Bistable trip tester.(1) 4-3 Draft TSP1287-0253-NL04

- ~

TABLE 4.1.1

~

M~nimum Frequencies for Checks, Calibrations and Testing of Reactor Protective System(S) (Contd)

Surveillance Channel Description Function Frequency Surveillance Method

6. Steam Generator Level a. Check s a. Comparison of four level indications per generator.

b. Calibrate R b. Known differential pressure applied to sensors.

c. Test M(2) c. Bistable trip tester.(1)

7. Steam Generator Pressure a. Check s a. Comparisons of four pressure indications per generator. e

b. Calibrate R b. Known pressure applied to sensors. I

c. Test M(2) c. Bistable trip tester.(1)

8. Containment Pressure a *. Calibrate R a. Known' pressure applied to sensors.

b. Test M(2) b. Simulate pressure switch action.

9. Loss of Load a. Test p a. Manually trip turbine auto stop oil relays.

10. Manual Trips a. Test p a. Manually test both circuits.

11. Reactor Protection System Logic Units

a. Test M(2) a. Internal test circuits.

e

12. Axial Shape Index (ASI) a. Test R a. Known power inputs applied to Thermal I Margin Calculator. I I

13. liT Power a. Check s a. Same as l(a). I

b. Check ( 3) D b. Same as l(b). I

c. Test R c. Known temperature imputs applied _to I Thermal Margin Calculator. I 4-4 Draft TSP l 287-0253-NL0'4

TABLE 4.1.1 Minimtim Frequencies for Checks, Calibrations and Testing of Reactor Protective System(5) (Contd)

Surveillance Charinel Description Function Frequency Surveillance Method

14. Thermal Margin*Calculator a. Check Q a. Verify constants. I NOTES: (l)The bistable trip tester injects a signal into the bistable and provides a precision readout of the trip set point.

(2)All monthly tests will be done on only one of four channels at a time to prevent reactor trip.

(3)Adjust the nuclear power or ~T power until readout agrees with heat balance calculations when above 15% of rated I power'. I (4)Trip setting for operating pump combination only. Settings for other than operating pump combinations must be tested during routine monthly testing performed when shut down and within four hours after resuming operation with a different pump combination if the setting for that combination has not been tested within the previous month.

(5)It is not necessary to perform the specified testing during prolonged periods in the refueling shutdown condition If this occurs, omitted testing will be performed prior to returning the plant to service.

(6)Also includes testing variable high power function in the Thermal Margin Calculator. I FREQUENCY NOTIATION Notation Frequency s At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

w At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 6 months.

R At least once per 18 months.

p Prior to each start-up i f not done previous week.

NA Not applicable.

4-5 Draft TSP1287-0253-NL04

TABLE 4.1.3 Minimum Frequencies for Checks, Calibrations and Testing of Miscellaneous Instrumentation and Controls Surveillance Channel Description ~~~-F_u_n_c_t_i_o_n~~~~ Frequency Surveillance Method

1. Start-Up Range Neutron . a. Check s a. Comparison of both channel count rate indications when Monitors in service.

b. Test p b. Internal test signals.

2. Primary Rod Position a. Check s a. Comparison of output data with secondary RPIS Indication System b. Check M b. Check of power dependent >> sertion limits monitoring system.

c. Calibrate R c. Physically measured rod drive position used to verify I system accuracy. Check rod position interlocks.

3. Secondary Rod Position a. Check s a. Comparison of output data with primary RPIS.

Indication System . b. Check M b. Same as 2(b) above.

c. Calibrate R c. Same as 2(c) above, including out-of-sequence alarm I function.

4. Area Monitors a. Check D a. Normal readings observed and internal test signals used Note: Process Monitor to verify instrument operation.

Surveillance Requirements b. Calibrate R b. Exposure to known external radiation source.

are located in Tables c. Test M c. Detector exposed to remote operated radiation check 4.24-1 and 4.24-2, source.

5. Emergency Plan Radiation a. Calibrate A a. Exposure to known radiation source.

ii struments b. Test M b. Battery check.

6. Environmental Monitors a. Check M a. Operational check.

b. Calibrate A b. Verify airflow indicator.

1.* Pressurizer Level a. Check s a *. Comparison of six independent level readings.

Instruments b. Calibrate R b. Known differential pressure applied to sensor.

c. Test M c. Signal to meter relay adjusted with test device.

4-10 Draft TSP1287-0253-NL04

TABLE 4 .1.3 Minimum Frequencies for Checks, Calibrations and Testing of Miscellaneous Instrumentation and Controls (Continued)

Surveillance Channel Description Function Frequency Surveillance Method

8. Control Rod Drive System a. Test R a. Verify proper operation of all manual I

interlocks rod drive control system interlocks, using simulated signals where necessary.

b. Test p b. Same as 8(a) above, if not done within three months.

9. Flux-AT Power Comparator a. Calibrate R a. Use simulated signals.

b. Test M b. Use simulated signals. I

10. Calorimetric Instrumentation a. Calibrate R a. Known differential pressure applied to I feedwater flow sensors.

11. Containment Building a. Test R a. Expose sensor to high humidity Humidity Detectors atmosphere.

12. Interlocks - Isolation Valves a. Calibrate R a. Known pressure applied to sensor.

on shutdown Cooling Line

13. Service Water Break Detector a. Test 'R a. Known differential pressure applied to in Containment Sensors.

I I

4-11 Draft TSP1287-0253-NL0r

4 .15. Primary System Flow Measurement Applicability Applies to the measurement of primary system flow rate with four primary coolant pumps in operation.

Objective To provide *assurance that the primary system flow rate is equal to.

or above the flow rate requ~red in 3.1.l(c).

Specification After each refueling outage, or after plugging 10 or more steam generator tubes, a primary system flow measurement shall be made with four primary coolant pumps in operation. I Basis*

This surveillance program assures that the reactor coolant flow is consistent with that assumed as the basis for Specification 3.Ll.(c).

4-70 Draft TSP1287-0253-NL04

POWER DISTRIBUTION INSTRUMENTATION 4.18.2 EXCORE MONITORING SYSTEM SURVEILLANCE REQUIREMENTS 4.18.2.1 At least every 31 days of power operation:

a. A target AO and excore monitoring allowable power level shall be determined using excore and incore detector readings at steady state near equilibrium conditions.

b. Individual excore channel measured AO shall be compared to the I total core AO measured by the incores. If the difference is I greater than 0.02, the excore monitoring system shall be recalibrated.

c. The excore measured Quadrant Power Tilt shall be compared to the incore measured Quadrant Power Tilt. If the difference is greater than 2%, the excore monitoring system shall be recalibrated.

4-82 Draft TSP1287-0253-NL04

4.19 4.19.1 POWER DISTRIBUTION LIMITS LINEAR HEAT RATES SURVEILLANCE REQUIREMENTS 4.19.1.l When using the incore alarm system to monitor LHR, prior to operation above 50% of rated power and every 7 days of power operation thereafter, incore alarms shall be set based on a measured power distribution.

4.19.1.2 When using the excore monitoring system to monitor LHR:

a. Prior to use, verify that the measured AO has not deviated from the target AO by more than 0.05 in the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for I each operable channel using the previous 24 hourly recorded I values. I

b. Once per day, verify that the measured Quadrant Power Tilt is less than or equal to 3%.

c. Once per hour, verify that the power is less than or equal to the APL and not more than 10% of rated power greater than the power level used in determining the APL.

d. Continuously verify that the measured AO is within 0.05 of the I established target AO for at least 3 of the 4, 2 of the 3 or I 2 of the 2 operable channels, whichever is the applicable case. I 4-83 Draft TSP1287-0253-NL04

MEMO ROUTE SLIP See me about thla. For concurrence*

. Fo~ AEC-93 (Rev. May 14, 1947) AECM 0240 Note and return.

TO (Name and unit) INITIALS TO (Name and unit) INITIALS

~ .

DATE*

TO (Name and unit) INITIALS DATE FROM (Name and unit) REMARKS P~oj ~"f f'let ~ 4.~ ~ ~r---=C:.-t?_n'1;....,.-+-:....:._~::::J....jL.....-!...!:::-:~V'-\:..:.!.....::::e:___!~::.!:!::=--L.!~==-~~..._--J Pa ks <<deJ f )t:l .;.f r--w~r_,*+--=e~,;--><L.----!...::::......1-..L-.-.:__--L.1'---~..::::.._.L..!..!:::;~~-:.....L----I lJoe-ke+- so -:l. ~-.!>;_. ::i:. h~ v.e --

USE OTHER SIDE FOR ADDITIOHAL REMARKS GPO: IHI O-lM-111

.. *- ~..-....-.._............... :..........-....-...... -*--~-----*

/~.

a/*

consumers Power Gene111I Offlc*: 19415 w- Pern1111 Roecl, Jedtson, Ml 49201 * (5171 788-0550 June 17, 1988 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT -

REACTOR PROTECTION SYSTEM MODIFICATION AND TECHNICAL SPECIFICATIONS CHANGES (TAC NO 66901).

Attachment 1 provides is a partial response to the Nuclear Regulatory Commission request for additional information (RAI) concerning the Palisades Reactor Protection System modification and Technical Specifications Change Request (TSCR). In addition to the resp.onse included in Attachment 2 are 4 Technical Specifications page changes containing minor revisions. Consumers Power Company previously submitted a draft TSCR and a TSCR on December 23, 1987, and March 25, 1988 respectively. These attached page changes replace those in the March 25, 1988 submittal. We expect to submit our response for the remaining_

items in the RAI in about* . - - : ~

one*--- *-week

~- .. - -

. --- .-.c_ - . . . -* --

Attachment 3 contains a description of th~ Reactor ~rotection -System Modification. Attachment 4 provides requested wiring diagrams.

Attachments 5 and 6 contain proprietary GAMMA-METRICS (G-M) reports.

Affidavits executed by Clinton L Lingren of G-M are included, in accordance with 10CFR2.790. The original affidavit supporting the proprietary nature of the G-M Instruction Manual (Attachment 5) was submitted with our March 25, 1988 Technical Specification Change Request. However, the manuals were withheld at that time by Consumers Power Company. As noted in the June 10, 1988 GAMMA-METRICS letter enclosed in Attachment 6 and Consumers.

Power Company letter dated April 7, 1988 the NRC is authorized to copy these proprietary doc\Jments for internal NR~ distribution and review.

Also, enclosed with this submittal... are non-proprietary Advanced Nuclear Fuels (ANF) reports numbers ANF-87-150(NP) Volume 1 and 2 and XN-NF-86-9l(NP).

These replace previous proprietary and non-proprietary reports transmit_ted to the NRC on December 23, 1987, March 25, 1988 and May 4, 1988. With submittal OC0688-0037-NL02

l;; . *i of these reports we request that the previously transmitted proprietarv ancL.

-n~n-proprietary reports either be returned or destroyed. The enclosed reports are identical to the proprietary versions currently in your possession. ANF has consented to release the proprietary information in these reports in order to support review of the proposed Palisades Technical Specifica~ions Change and completion of the Reactor Protection System modification.

James L Kuemin (Signed)

James L Kuemin Staff Licensing Engineer CC Administrator, Region 111, NRC NRC Resident Inspector - Palisades Attachments OC0688-0037-NL02

ML18054A208

Text

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