ML20062E847

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Forwards Amend to 900821 Proposed Tech Spec Changes Re Fuel Parameters,Per NRC Request.Ref to Improved BWR Tech Specs Deleted & Request for Changes Re Reactor Water Level Safety Limit & Generic Ltr 90-02 Recommendations Withdrawn
ML20062E847
Person / Time
Site: Pilgrim
Issue date: 11/08/1990
From: Bird R
BOSTON EDISON CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20062E852 List:
References
BECO-90-135, GL-90-02, GL-90-2, NUDOCS 9011260086
Download: ML20062E847 (11)
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Text

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g: 10CFR50,90 nosnwanson Pilgrim Nuclear Fower station Rocky Hill Road Plymouth, Massachusetts 02360 Ralph G. Bird Novetaber 8. 1990 senior vice hesident - Nuclear g U.S. Nuclear Regulatory Commission Document Control Desk Hashington, DC 20555 License OPR-35 Docket 50-293 PROPOSED TECHNICAL SPECIFICATION CHANGE CONCERNING FUEL PARAMETERS AMENDED PROPOSAL _

He are amending our proposed Technical Specification change (ref. BECo Letter 90-101, dated August 21, 1990) concerning fuel parameters. To reflect comments made by the NRC, we are deleting references to the " Improved" BHR Technical Specifications and withdrawing our request 1) to make changes to the Reactor hater Level Safety Limit, 2) to incorporate the recommendations of Generic Letter 90-02, 3) to alter the actions regarding APRM gain adjustments.

Attachments A-E have been amended to reflect these changes to our initial request.

.G B RAH /njm/4606 Commonwealth of Massachusetts)

County of Plymouth )

Then personally appeared before me, Ralph G. Bird, who being duly sworn, did state that he is Senior Vice President - Nuclear of Boston Edison Company and that he is duly authorized to execute and file the submittal contained herein in the name and on behalf of Boston Edison Company and that the statements in said submittal are true to the best of his knowledge and belief.

My commission expires: Mc b sf/19f DATE d /M NOTARYP%LIC Attachments: A. Amended Description of Proposed Changes B. Amended List of Affected Pages '

C. Amended Marked-up Technical Specification Pages D. Amended Replacement Technical Specification Pages E. Core Operating Limits Report, Cycle 8. I 1 signed original and 37 copies 00

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BOSTON EDISON COMPANY U.S. Nuclear Regulatory Commission Page 2 cc: Mr. R. Eaton, Project Manager Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Mail Stop: 14D)

U. S. Nuclear Regulatory Commission 1 White Flint North 11555 Rockville Pike Rockville, MD 20852 U. S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Senior NRC Resident Inspector Pilgrim Nuclear Power Station Mr. Robert M. Hallisey, Director Radiation Control Program Massachusetts Department of Public Health 150 Tremont Street, 2nd floor Boston, MA 02111

Attachment A to BECo 90-135 DESCRIPTION OF PROPOSED CHANGES PROPOSED CHANGES:

These proposed changes to the Pilgrim Nuclear Power Station (PNPS) Techrical Specifications are concerned with fuel parameters and accomplish the frllowing:

1. Remove Cycle-soecific Parameter Limits As recommended by Generic Letter 88-16. " Removal of Cycle-specific Parameter Limits from Technical Specifications," dated October 4, 1988, this change proposes the relocation of cycle-specific parameter limits from Technical Specifications to the Core Operating Limits Report (COLR). The affected cycle-specific parameter limits include the following:

Flow-biased Average Power Range Monitor (APRH) Flux Scram Trip Setting flow-blased APRH Rod Block Trip Setting Rod Block Honitor Trip Setting Average Planar Linear Heat Generation Rate (APLHGR) Limits Linear Heat Generation Rate (LHGR) Limit Minimum Critical Power Ratio (MCPR) Operating Limits Power / Flow Operating Hap Fuel Design Features These cycle-specific parameter limits are proposed to be relocated to the COLR. The relocation of these parantters to the COLR is in accordance with the agreement made between the NRC and General Electric on the implementation of Generic Letter 88-16. This agreement is documented in the letter from J.S. Charnley, General Electric, to H.H. Hodges, NRC,

" Acceptance Implementation of Generic Letter 88-16," dated August 8, 1989.

Only the actual cycle-specific parameters are relocated to the COLR, with a COLR reference in the Technical Specifications. The related surveillance requirements and action statements for exceeding the parameters remain in the Technical Specifications. Bases are provided in Technical Specifications for each parameter.

Ttis change is intended to reduce the burden on licensee and NRC resources in the processing of license amendments for each new fuel cycle to update cycle-specific parameter limits in the Technical Specifications. Such license amendments are developed using NRC-approved methodologies and are consistent with all applicable limits in the Final Safety Analysis Report (FSAR). Therefore, additional NRC review of the updates to the cycle-specific parameters for each new fuel cycle is not necessary. A new reporting requirement for the submittal of COLR revisions to the NRC allows continued trending of these cycle-specific limits without the necessity of prior NRC review and approval.

2. Uparaded MCPR Safety Limit By letter from A.C. Thadani, NRC, to J.S. Charnley., Ge.%rhi Electric, dated December 27, 1987, the NRC approved Amer.dment 14 to GESTAR-II (NEDE-24011-P-A, " General Electric Standard Application for Reactor fuel") that Page 1 of 8 J

upgrades the MCPR safety liQit for D-lattice BWR's using high enrichment fuel. The current MCPR safety limit of 1.07 was established by NEDE-24131,

Basis for 8x8 Retrofit Fuel Thermal Analysis Application," dated September 1978 and was based on fuel design characteristics typical of those used at the time. However, core reloads using high bundle R-factor General  !

Electric fuel has resulted in increased conservatisms and has permitted the  ;

upgrade of the HCPR safety limit from 1.07 to 1.04. Consequently, this  !

change proposes to upgrade the HCPR safety limit in the PNPS Technical  ;

Specifications from 1.07 to 1.04. Additionally, the MCPR operating limits  ;

in new COLR Table 3.3-1 would also be upgraded by the same amount (.03) to  !

take advantage of this increased margin of safety. l

3. Administrative Chanaes The proposed changes include many editorial changes to update the Table of Contents, correct grammatical and spelling errors, correct references to the Final Safety Analysis Report (FSAR), reformat portions of the Technical Specification, and add text inadvertently deleted in a previous amendment.

These changes add to the clarity and readability of Technical Specifications and do not impact any margins of safety.  ;

DETAILED DESCRIPTION OF CHANGES:

The following detailed description of chan es is provided for each of the pro)osed changes discussed above. In addi ion, a list of the affected Tec1nical Specification pages is provided in Attachment B. The marked-up pages of the current Technical Specifications are provided in Attachment C.

The proposed replacement Technical Specification pages are provided in Attachment D.

1. Remove Cvele-soecific Parameter Limits
  • The maximum APRM scram trip setting of 1201 of rated thermal power in the last paragraph of the current Technical Specification 2.1. A.1.a is relocated to new Footnote 15 in Technical Specification Table 3.1.1.
  • The APRM rod block trip setting in refuel and startup modes in current Technical Specification 2.1.B.2 is relocated to Footnote 2 of Technical Specification Table 3.2.C-2.
  • Technical Specification Figure 2.1.1 is deletted because it contains a graphical representation of the APRM scram and rod block trip settings defined in the COLR. The figure contains no new information or requirements. The reference to this figure in current Bases 2.1 was also deleted.

Page 2 of 8 /

1

e The reference in Teehnical Specification 3.1.B.1 to current Technical Specifications 2.1.A.1.a and 2.1.B.1 is revised to indicate the relocation of the APRM trip setpoints to the COLR.

  • The APRM scram trip setpoint definition provided in Technical Specification Table 3.1.1 is replaced with a reference in the new Footnote 15 to the COLR. Footnote 14 is deleted because it contains formula definitions that have been relocated to the COLR.
  • The rod block monitor trip setpoint and the reference to current Technical Specification 2.1.B in Technical Specification Table 3.2.C-2 are replaced with a reference to the COLR in new Footnote 1 of Table 3.2.C-2.
  • The reference to Technical Specification Figures 3.11-1 through 3.11-7 for the applicable limiting values of APLHR in Technical Specification 3.11.A is revised to indicate their relocation to the COLR. Also, related text describing the figures and Technical Specification Figures 3.11-1 through 3.11-7 are deleted. The portion of this deleted information that is applicable to the current operating cycle is relocated to the COLR.
  • A reference to the COLR is added to Technical Specification 3 ll.C for the MCPR operating limit values. Related Technical Specification Table 3.11-1 and Figure 3.11-8 are relocated to the COLR.
  • The power / flow operating map in Technical Specification Figure 3.11-9 is relocated to the COLR. The reference to Figure 3.11-9 in Technical Specification 3.11.D is revised to indicate its relocation to the COLR.

i e Details of the reactor core design are relocated from Technical Specification 5.2 to the COLR.

  • As recommended by Generic Letter 88-16, new Technical Spe;ification 6.9.A.4 is added to list the NRC-approved analytical % thods to be used to determine the core operating limits for each rcioad. The requirement that core operating limits meet all applicable limits of the safety analysis is also added. A new reporting requirement to submit COLR revisions to the NRC upon issuance is added.
2. Uoaraded MCPR Safety limit

!

  • As part of the effort to remove cycle-specific parameter limits, the l MCPR operating limits in current Technical Specification Table 3.11-1

( are relocated to the COLR. Note that these MCPR operating limits in new COLR Table 3.3-1 have been ievised by the same amount (.03) as the MCPR safety limit.  ;

l_ Page 3 of 8 L

3. Administrative Chanaes l 1

4

- The applicability and objective sections of current Technical  !

Specifications 1.1/2.1 and 1.2/2.2 are deleted because they contain no requirements or information required for operation of PNPS.

- The MCPR safety limit in current Technical Specification 1.1.A is relocated to new Technical Specification 2.1.2. The applicability conditions of reactor steam dome pressure and core flow are revised to be consistent with existing Technical Specification bases and stated in psig to be easily compared to plant instrumentation.

- The core thermal power safety limit in current Technical Specification 1.1.B is relocated to new Technical Specification 2.1.1. The applictbility conditions of reactor steam dome pressure  !

and core flow are restated to be consistent with existing Technical Specification bases and stated in psig to be easily compared to plant instrumentation.

- Current Technical Specification 1.1.C Power Transient, is deleted because it is redundant to the requirements of 10CFR50.36(c)(1)(ii)(A) and 10CFR50.73(b)(3). In the case that reactor scram is accomplished by indirect means,10CFR50 :equires an analysis be performed to determine whether safety liraits were exceeded when the direct scram signal failed to perform as expected. Thus, current Technical Specificati,,n 1.1.C makes no new requirements and may be deleted.

- The following scram trip settings are deleted from the current Technical Specifications listed below because they are provided in current Technical Specification Table 3.1.1.

Scram Trio Settina Deleted Technical SoecificatinD l APRM (refuel or startup modes) 2.1.A.I.b Intermediate Range Monitor (IRM) 2.1.A.I.c Reactor Low Hater level 2.1.C Turbine Stop Valve Cloiure 2.1.D Turbine Control Valve Fast Closure 2.1.E Condenser low Vacuum 2.1.F Hain Steam Isolation Valve Closure 2.1.G

- The main steam isolatioF setpoint on main steam line low pressure in current Technical Sptcification 2.1.H is deleted because it is provided in current lechnical Specification Table 3.2.A. i

- The core standby cooling system (CSCS) initiation setpoint for reactor low-low water level in current Technical Specification 2.1.I is deleted because it is provided in current Technical Specification Table 3.2.B.

Page 4 of 8 i

- The bcses are rewritten for consistency with the changes discussed above for current Technical Specifications 1.1/2.1 and 1.2/2.2 Some bases are relocated to 6.he bases for current Technical Specifications 3.1, 3.2, and 3.6 to accompany the location of the respective trip settings. Ir some cases, bases paragraphs are deleted because appropriate tases are provided elsewhere in the document.

- The reactor steam dome pres',ure safety limit in current Technical Specification 1.2 is reloct.ted to new Technical Specification 2.1.4.

- The reactor vessel high pressure scram trip setting in current Technical Specification 2.2.A is deleted because it is provided in current Technical Specification Table 3.1.1.

- The relief / safety valve and safety valve settings in current Technical Specifications 2.2.B and 2.2.C are relocated to Technical Specification 3.6,0.1. The reference to current Technical Specification 2.2 in Technical Snecification 4.6.D.1 is deleted because the safety valve setpoints are reic:ated to Technical Specification 3.6 D.1.

- The actions to be taken in the event a safety limit is violated are relocated from current Technical Specification 6.7 to new Technical Specification 2.2.

  • The Table of Contents is updated.
  • Grammatical and editorial changes are included in the affected Technical Specification bases to improve readability. An FSAR reference in the bases for the relief / safety valve settings (new Technical Specification Bases 3.6.D) is corrected, e The following editorial changes are made to Technical Specifications to correct grammar, spelling, punctuation, and format.

- Heading for Technical Specification 3/4.1 revised.

- Names of trip functions in Technical Specification Table 3.1.1 are made consistent with bases.

- Heading for Technical Specification 3.6.C corrected on Page 126.

- Reference to Specification 3.6,0.1 corrected in Specification 3.6,0.2.

- Abbreviation spelled out in Technical Specification 3.6,0.3.

- Format of heading corrected on Page 127.

- Punctuation corrected in Technical Specification 3.6.G.I.

- Punctuation added to Technical Specification 4.11 on Page 20Sa.

Hord added to applicability paragraph of Technical Specification 4.11 to clarify meaning.

- Unnecessary specification numbers deleted from Technical Specification 3.11.C and 4.11.C.

- Punctuation revised in Technical Specification 3.11.D.

- Format of Technical Specification 6.9.A revised to increase clarity.

- Heading for Technical Specification 6.9.B added to indicate the section was previously deleted.

Page 5 of 8

o The word " delta" was added to Technical Specification 4.6.E.3 because the delta sy.nbol was inadvertently deleted in Amendment 42 to the Technical Specifications.

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS:

The Code of Federal Regulations (10CFR50.91) requires licensees requesting an amendment to provide an analysis, using the standards in 10CFR50.92, that determines whethe a significant hazards consideration exists. The following analyses are provided in accordance with 10CFR50.91 and 10CFR50.92 for these proposed Technical Specification changes.

1. Remove Cycle-soecific Parameter limits A. The proposed changes do not involve a significant increase in the

,)tobability or consequences of an accident previously evaluated because the cycle-specific limits will still be determined by analyzing the same postulated events previously analyzed. The removal of the cycle-specific limits from the Technical Specifications has no influence or impact on a Design Basis Accident occurrence. Each Design Basis Transient and accident analysis previously addressed will be examined wito respect to changes in the cycle dependent parameters using the NRC-approved reload design methodologies to ensure that the transient evaluation of new reloads are bounded by previously accepted analyses.

This examination, which will be performed per the requirements of 10CFR50.59, will ensure future reloads will not involve a significant increase in the probability or consequences of an accident previously evaluated. The plant will continue to operate within the limits specified in the Core Operating Limits Report (COLR) and to take the same actions when, or if, the limits are exceeded as required by the current Technical Specifications.

B. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated because no physical alterations of plant configuration, changes to setpoints, or safety limits are proposed. As stated above, the removal of the cycle-specific limits does not influence, impact, nor contribute in any way to the probability or consequences of any accident. The cycle-specific limits will be calculated using the NRC-approved methods. The Technical Specifications will continue to require operation within the required core operating limits and appropriate actions will be taken when, or if, limits are exceeded.

C. The proposed changes do not involve a significant reduction in a safety margin because they do not affect any operating practices, limits, or safety-related equipment. The margin of safety presently provided by the current Technical Specifications remains unchanged. The proposed amendment still requires operation within the core limits as obtained from the NRC-approved reload design methodologies and appropriate actions to be taken if Ibits are violated. The development of the limits for future reloads will continue to conform to those methods described in the NRC-approved documentation. In addition, each future reload will involve a safety review to assure that operation of the plant within the cycle-spccific limits will not involve a significant reduction in a margin of safety.

l Page 6 of 8

2. Unaraded Minimum Critical Power Ratio (MCPR) Safety Limit I A. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The NRC-approved methodology used to derive the upgraded MCPR safety limit of 1.04 applied the same criteria as that used to derive the current MCPR safety limit of 1.07. The upgraded MCPR safety limit value of 1.04 ensures fuel cladding protection equivalent to that provided with the 1.07 safety limit is maintained. In the safety evaluation for Amendment 14 to NEDE-240ll-P-A (GESTAR-II), dated December 27, 1987, the NRC approved the use of the 1.04 MCPR safety limit for D-lattice BHRs subject to the following constraints: 1) the fuel has a beginning of life R-factor of greater than or equal to 1.04 and consists of fuel types P8 x BR, BP8 x BR, GE8 x 8E or GE8 x BEB, 2) the fuel is at least 2.80 weight percent U-235 bundle average enrichment, and 3) the lower enrichment bundles residing in the core have operated for at least 2 cycles. Because the Pilgrim Nuclear Power Station currently meets these constraints and will meet them in future reloads, the 1.04 MCPR safety limit provides the same degree of assurance for fuel cladding integrity as the 1.07 HCPR safety limit did for previous reload cores. Thus, the consequences of accidents previously evaluated are not significantly increased. The MCPR safety limit does not affect any physical system or equipment that could change the probability of an accident. Therefore, the proposed change does not involve a significant increase in the probability of any accident previously evaluated.

B. Adoption of the proposed MCPR safety limit value does not affect the function of any component or system. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

C. The use of the 1.04 HCPR safety limit reflects the utilization of current General Electric fuel designs and does provide the same margin of safety as 1.07 does with older Wneral Electric fuel types as discussed in the previously referenced NRC safety evaluation. Because equivalent fuel cladding protection is provided with the 1.04 HCPR safety limit, the design criterion that 99.9 percent of all fuel rods do not experience boiling transition following any des'gn basis transient is met. Therefore, the proposed change does not involve a significant reduction in the margin of safety.

3. Adminstrative Chanaes A. Proposea changes are made to improve the format and readability. The proposed changes include relocating sections of the existing technical specifications and editorial changes to update the Table of Contents, correct grammatical and spelling errors, correct a reference to the Final Safety Analysis Report (FSAR), make the Technical Specification format consistent, and add text inadvertently deleted in a previous amendment. These changes are considered to be entirely administrative in nature. In addition, the following minor modificatioas were included:
1. The conditions for applicability for the MCPR and thermal power safety limits are revised to be consistent with Technical Specification bases and restated in psig to be easily compa.ed to plant instrumentatior.. The result of this change is to increase Page 7 of 8

the range of applicability of the MCPR safety linit (and correspondingly decrease the range of applicability for the thermal power safety limit) by rea*. tor steam dome pressure of 0.3 psid.

Specifically, the reactor steam dome pressure of 800 psia converts to 785.3 psig, which is tourded off to 785 psig and results in a difference of 0.3 psid. This change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Current Technical Specification 1.1.C is deleted because it is .

redundant to the requirements of 10CFR50.36(c)(1)(li)(A) and j 10CFR50.73(b)(3). In the case that reactor scram is accomplished by -

indirect means, 10CFR50 requires an analysis be performed to determine whether safety limits were exceeded when the direct scram signal failed to perform as expected. Thus, current Technical Specification 1.1.C contains no unique requirements and may be deleted.

As discussed above, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

B. The proposed changes do not create the possiblity of a new or different kind of accident from any accident previously evaluated because no plant design or configuration changes are involved.

C. The change in the range of applicability of the MCPR and thermal power safety limits of 0.3 psi does not involve a significant reduction in a safety margin. Accordingly, the proposed changes do not involve a significant reduction in a safety margin because they do not affect any operating practices, limits, or safety-related equipment.

These changes have been reviewed and approved by the Operations Review Committee and reviewed by the Nuclear Safety Review and Audit Committee.

Schedule for Chance This change will be implemented within 30 days following Boston Edison's receipt of its approval by the Commission.

1 Page 8 of 8

ATTACHMENT B to BEco 90 135  !

List of revised Technical Specifications Pages:

1.11.111,1, 6, 7, 8, 9,10, 26, 27. 29, 36, 37, 38, 39 40, 404, 40b,  !

40c, 554, 71, 72, 126, 127, 127a, 145, 146, 205a, 205b, 20$c, 205d, 205e,  !

'205f, 206m, 217, 218, and 219.  ;

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