LR-N05-0036, Request for Changes to Technical Specifications Deletion of Reactor Coolant System Volume
| ML050560048 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 02/15/2005 |
| From: | Gallagher M Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| +kBR1SISP20051121, LCR S04-08, LR-N05-0036 | |
| Download: ML050560048 (17) | |
Text
TREA AM SENSIA IV
.PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 FEB 1 5 2005 LCR S04-08 Nuclear LLC U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS DELETION OF REACTOR COOLANT SYSTEM VOLUME SALEM NUCLEAR GENERATING STATION UNITS 1 and 2 FACILITY OPERATING LICENSES DPR-70 and DPR-75 DOCKET NOs. 50-272 and 50-311 Pursuant to 10 CFR 50.90, PSEG Nuclear, LLC (PSEG) hereby requests a revision to the Technical Specifications (T/S) for the Salem Nuclear-Generating Station, Units 1 and 2. In accordance with 10 CFR 50.91 (b)(1), a copy of this submittal has been sent to the State of New Jersey.
PSEG Nuclear proposes to revise the Salem Unit 1 and 2 Technical Specifications to reflect the deletion of Reactor Coolant System (RCS) Volume from Design Features Section 5.4.2. Information concerning the RCS Volume is included in the Salem Updated Final Safety Analysis Report (UFSAR), and any changes to the information are controlled in accordance with 10 CFR 50.59. By letter dated March 1, 2000 (TAC Nos.
MA7756 and MA7757), Donald C. Cook Nuclear Plant has received NRC approval of a similar request.
PSEG has evaluated the proposed changes in accordance with 10 CFR 50.91 (a)(1),
using the criteria in 10 CFR 50.92 (c), and has determined this request involves no sig-nificant hazards considerations. This amendment to the Salem T/S meets the criteria of 10 CFR 51.22 (c)(9) for categorical exclusion from an environmental impact statement.
The requested changes are provided in Attachment 1 to this letter. The proposed marked up Technical Specification pages are provided in Attachment 2.
Should you have any questions regarding this request, please contact Mr. Steve Man-non at 856-339-1129.
95-2168 REV. 7/99
I Document Control Desk LR-N05-0036 2
fEB 1 5 2005 I declare under penalty of perjury that the foregoing is true and correct.
Sincerely, Executed on Michael P. Gallagher Vice President-EngfTech Support Attachments (2) cc Mr. S. J. Collins, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U. S. Nuclear Regulatory Commission ATTN: Mr. Daniel Collins, Licensing Project Manager - Salem Mail Stop 08C2 Washington, DC 20555-0001 USNRC Senior Resident Inspector - Salem (X24)
Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering PO Box 415 Trenton, NJ 08625
I EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS DELETION OF RCS VOLUME SALEM NUCLEAR GENERATING STATION UNITS I AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOs. 50-272 AND 50-311 ATTACHMENT 1 EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS DELETION OF RCS VOLUME
LR-NO5-0036 ATTACHMENT I EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS DELETION OF RCS VOLUME Table of Contents
- 1.
DESCRIPTION.............................................................................................2
- 2.
PROPOSED CHANGES.......................................... 2
- 3.
BACKGROUND..........................................
2
- 4.
EVALUATION...............................................................................................2
- 5.
REGULATORY SAFETY ANALYSIS
.......................................... 4 5.1 No Significant Hazards Consideration........................................... 4 5.2 Applicable Regulatory Requirements/Criteria.................................... 6
- 6.
ENVIRONMENTAL ASSESMENT/IMPACT STATEMENT
.......................... 7
- 7.
REFERENCES.............................................................................................7 LR-NO5-0036 ATTACHMENT I EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS DELETION OF RCS VOLUME
1.0 DESCRIPTION
The amendment request deletes Technical Specification (T/S) 5.4.2, "Reactor Coolant System Volume," regarding the Reactor Coolant System (RCS) vol-ume information. This information'is not required to be in the T/S for compli-ance with 10 CFR 50.36(c)(4). Information concerning the RCS volume is in-cluded in the Salem Updated Final Safety Analyses Report (UFSAR), and any changes to the information are controlled in accordance with 10 CFR 50.59.
2.0 PROPOSED CHANGE
S Specifically the proposed changes would revise the following:
2.1 Delete Section 5.4.2, Volume, located in Page 5-5 for Salem Unit 1 and 2.2 Delete Section 5.4.2, Volume, located in Page 5-4 for Salem Unit 2.
3.0 BACKGROUND
Unit 1 and Unit 2 T/S 5.4.2 lists the approximate, total combined RCS vol-ume at a nominal Tavg of 5730F. T/S 5.4.2 also includes an adjustment to ac-count for previously evaluated steam generator tube plugging limits. These nominal values do not reflect the actual RCS volumes that will exist when the unit is restarted following the Unit 2 steam generator replacement currently scheduled for 2008. Therefore, the T/S 5.4.2 values for RCS volume would need to be revised at a future date.
The UFSAR includes values for total RCS volume and RCS component and piping volumes that are more detailed and complete than the approximate RCS volumes listed in Unit I and Unit 2 T/S Section 5.4.2. These more de-tailed values are normally used as design inputs to the actual UFSAR Chapter 15 accident analyses, and include values for RCS volume at previously evalu-ated steam generator tube plugging limits. Therefore, T/S 5.4.2 is redundant to the UFSAR.
4.0 EVALUATION 10 CFR 50.36(c)(4) governs the contents of Technical Specification (T/S) Sec-tion 5.0, 'Design Features." 10 CFR 50.36(c)(4) states, "Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)
(1), (2), and (3) of this section."
LR-NO5-0036 ATTACHMENT I EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS DELETION OF RCS VOLUME As stated in 10 CFR 50.36(c)(2)(ii)(B), the T/S limiting conditions for operation must be established for 'process variables, design feature, or operating restric-tion that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." T/S Section 3/4.4, Reactor Coolant System," includes the limiting conditions for operation related to the RCS, and includes informa-tion either limiting changes to, or derived from, RCS volume.
Changes to the actual RCS volume can result from physical modifications to RCS components, changes to procedures affecting pressurizer pressures and levels, or by plugging of steam generator tubes. Changes to the facility and procedures are required to be evaluated in accordance with 10 CFR 50.59, which ensures that changes to RCS volume as a result of physical modifica-tions and procedure changes are evaluated for impact on the plant accident analyses.
Since detailed RCS information already exists in the UFSAR (Ex. Table 5.1-1),
and any method by which the RCS volume could be changed is required to be evaluated in accordance with 10 CFR 50.59, then including this information in the T/S is not necessary. RCS nominal Tavg is only included in T/S Section 5.4.2 as a reference value associated with RCS Volume and it's deletion does not represent a change to the RCS Temperature limitations which are included in other Sections of the Salem Technical Specifications.
The original Salem Technical Specifications were developed prior to the most recent guidance provided in NUREG-1431, "Standard Technical Specifications
-Westinghouse Plants." NUREG-1431 does not include RCS volume infor-mation in T/S Section 4, "Design Features," as this information does not meet the criteria for inclusion in the T/S, and is not considered necessary for com-pliance with 10 CFR 50.36(c)(4).
The proposed change to remove this information from T/S does not affect any accident initiators or precursors. Elimination of the RCS volume information from the T/S does not change the methods for plant operation or actions to be taken in the event of an accident. The deletion of the RCS volume information from the T/S does not change the methods of plant operation or modify plant systems, structures, or components. No new methods of plant operation are created. As such, the proposed change does not affect any accident initiators or precursors or create new accident initiators or precursors. The deletion of the RCS volume, including the reference to RCS Tavg nominal value of 5730 F, from the T/S does not affect safety limits or limiting safety system settings. The proposed change does not modify the quantity of radioactive material available for release in the event of an accident. As such, the proposed change will not LR-NO5-0036 ATTACHMENT I EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS DELETION OF RCS VOLUME affect any previous safety margin assumptions or conditions. The actual vol-ume of the RCS is not affected by the change, only the location of the text de-scribing the volume. More detailed and complete RCS component and piping volume information is included in the UFSAR, and any changes to that infor-mation would be evaluated prior to implementation in accordance with 10 CFR 50.59.
5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration As required by 10 CFR 50.91(a), PSEG provides its analysis of the no sig-nificant hazards consideration. According to 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards consid-eration if operation of the facility in accordance with the proposed amend-ment would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Create the possibility of a new or different kind of accident from any ac-cident previously evaluated.
- 3. Involve a significant reduction in a margin of safety.
The determinations that the criteria set forth in 10 CFR 50.92 are met for this amendment request are indicated below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response
No The proposed change to remove this information from T/S does not af-fect any accident initiators or precursors. Elimination of the RCS vol-ume information from the T/S does not change the methods for plant operation or actions to be taken in the event of an accident.
The quantity of radioactive material available for release in the event of an accident is not increased.
Barriers to release of radioactive material are not eliminated or oth-erwise changed. More detailed RCS component and piping volume information is included in the Salem UFSAR, and changes to that in-LR-NO5-0036 ATTACHMENT I EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS DELETION OF RCS VOLUME formation would be evaluated prior to implementation in accor-dance with 10 CFR 50.59.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of accidents previously evaluated.
- 2. Does the proposed amendment create the possibility of a new or differ-ent kind of accident from any accident previously evaluated?
Response
No The deletion of the RCS volume information from the T/S does not change the methods of plant operation or modify plant systems, structures, or components. No new methods of plant operation are created. As such, the proposed change does not affect any accident initiators or precursors or create new accident initiators or precur-sors. More detailed and complete RCS component and piping vol-ume information is included in the Salem UFSAR, and any changes to that information would be evaluated prior to implementation in ac-cordance with 10 CFR 50.59.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response
No The deletion of the RCS volume information from the T/S does not affect safety limits or limiting safety system settings. Plant opera-tional parameters are not affected. The proposed change does not modify the quantity of radioactive material available for release in the event of an accident. As such, the change will not affect any previ-ous safety margin assumptions or conditions. The actual volume of the RCS is not affected by the change, only the location of the text describing the volume. More detailed and complete RCS component and piping volume information is included in the Salem UFSAR, and any changes to that information would be evaluated prior to implementation in accordance with 10 CFR 50.59.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
LR-NO5-0036 ATTACHMENT I EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS DELETION OF RCS VOLUME Based on the above, PSEG concludes that the proposed amendment pre-sents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and accordingly, a finding of "no significant hazards con-sideration" is justified.
5.2 Applicable Regulatorv Requirements/Criteria 10 CFR 50.36 Technical Specifications (c) Technical specifications will include items in the following categories:
(4) Design features. Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c) (1), (2), and (3) of this section.
Response
Since detailed RCS information already exists in the UFSAR, and any method by which the RCS volume could be changed is required to be evaluated in accordance with 10 CFR 50.59, then including this information in the T/S is not necessary to ensure that a significant effect on safety does not occur. In addition, since T/S Section 3/4.4 already includes the limiting conditions for operation related to the RCS, and includes information either limiting changes to, or derived from, RCS volume, then including RCS vol-ume in T/S Section 5.0 is not required as allowed by 10 CFR 50.36(c)(4).
Summary Based on the evaluation, PSEG believes that the proposed T/S changes do not reduce the level of safety currently maintained by the T/S, is consistent with NUREG-1431, and is in accordance with 10 CFR 50.36.
CONCLUSION PSEG has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issu-ance of the amendments will not be inimical to the common defense and secu-rity or to the health and safety of the public.
LR-NO5-0036 ATTACHMENT I EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS DELETION OF RCS VOLUME
6.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.22 (b), PSEG has evaluated this license amendment request to determine whether or not it meets the criteria for categorical exclu-sion set forth in 10 CFR 51.22 (c)(9) of the regulations.
PSEG has concluded that implementation of this amendment will have no ad-verse impact upon Salem Units I and 2; neither will it contribute to any signifi-cant additional quantity nor the type of effluent being available for adverse en-vironmental impact or personnel exposure. The change does not introduce any new effluents or significantly increase the quantities of existing effluents.
As such, the change cannot significantly affect the types or amounts of any ef-fluents that may be released offsite.
Therefore, it has been determined that there is:
- 1. No significant hazards consideration,
- 2. No significant change in the types, or significant increase in the amounts, of any effluents that may be released offsite, and
- 3. No significant increase in individual or cumulative occupational radiation exposures involved.
Therefore, this amendment request to the Salem Technical Specifications meets the criteria of 10 CFR 51.22 (c)(9) for categorical exclusion from an en-vironmental impact statement.
7.0 REFERENCES
7.1 Code of Federal Regulations, 10CFR 50.36.
7.2 Improved Standard Technical Specifications for Westinghouse Plants, NUREG 1431.
7.3 PSEG Salem Units 1 and 2, Updated Final Safety Analysis Report.
7.4 PSEG Salem Units 1 and 2, Technical Specifications.
7.5 Issuance of Amendments-Donald C. Cook Nuclear Plant, Units 1 and 2 (TAC Nos. MA7756 and MA 7757, dated March 1, 2000) Deletion of T/S 5.4.2, RCS Volume from the Technical Specifications.
SALEM NUCLEAR GENERATING STATION UNITS I AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOs. 50-272 AND 50-311 ATTACHMENT 2 MARKED-UP TECHNICAL SPECIFICATIONS CHANGES DELETION OF RCS VOLUME
a.
SALEM NUCLEAR GENERATING STATION, UNITS 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 REVISION TO TECHNICAL SPECIFICATIONS DELETION OF RCS VOLUME TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License No. DPR-70 are affected by this change request:
Technical Specification Page Index, Section 5.4 XVII Section 5.4 5-5 The following Technical Specifications for Facility Operating License No. DPR-75 are affected by this change request:
Technical Specification Page Index, Section 5.4 XVII Section 5.4 5-4
INDEX DESIGN FEATURES SECTION PAGE 5.1 SITE Exclusion Area................... 5-1 Low Population Zone...............
. 5-1 Unrestricted Areas for Radioactive Gasious and Liquid Effluents.................
5-1 5.2 CONTAINMENT Configuration 5-1 Design Pressure and Temperature
.......... 5-4 5.3 REACTOR CORE Fuel Assemblies
.................. 5-4 Control Rod Assemblies............... 5-4 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature
.......... 5-4 5.5 METEOROLOGICAL TOWER LOCATION
........... 5-5 5.6 FUEL STORAGE Criticality.
5-5 Drainage...................... 5-6a Capacity.
5-6a 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT
........ 5-6a SALEM -
UNIT 1 XVII Amendment No.
DESIGN FEATURES
- a.
In accordance with the code requirements specified in Section 4.1 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,
- b.
For a pressure of 2485 psig, and
- c.
For a temperature of 650EF, except for the pressurizer which is 680EF.
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.t e.m Hea Tavjg of 573 0F.
5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
5.6 FUEL STORAGE CRITICALITY 5.6.1.1 The new fuel storage racks are designed and shall be maintained with:
- a.
A maximum Keff equivalent of 0.95 with the storage racks flooded with unborated water.
- b.
A nominal 21.0 inch center-to-center distance between fuel assemblies.
- c.
Unirradiated fuel assemblies with enrichments less than or equal to 4.25 weight percent (w/o) U-235 with no requirements for Integral Fuel Burnable Absorber (IFBA) pins.
Unirradiated fuel assemblies with enrichments (E) greater than 4.25 w/o U-235 and less than or equal to 5.0 w/o U-235 which contain a minimum number of Integral Fuel Burnable Absorber (IFBA) pins.
This minimum number of IFBA pins shall have an equivalent reactivity hold-down which is greater than or equal to the reactivity hold down associated with N IFBA pins, at a nominal 2.35 mg B-10/linear inch'loading (1.5X), determined by the equation below:
N = 42.67 ( E B 4.25 SALEM -
UNIT 1 5-5 Amendment No. l2.?'
I.
INDEX DESIGN FEATURES SECTION PAGE 5.1 SITE Exclusion Area............................................. 5-1 Low Population Zone........................................ 5-1 Unrestricted Areas for Radioactive Gaseous and Liquid Effluents............................................... 5-1 II 5.2 CONTAINMENT Configuration.............................................. 5-1 Design Pressure and Temperature............................ 5-4 5.3 REACTOR CORE Fuel Assemblies............................................ 5-4 Control Rod Assemblies..................................... 5-4 5.4 REACTOR COOLANT SYSTEM DesiqaPressur and Temperature 5-4 5.5 METEOROLOGICAL TOWER LOCATION.............................. 5-5 5.6 FUEL STORAGE Criticality Drainage.....................................
Capacity.....................................
5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT...........
5-5 5-5a 5-5a 5-5b II I
SALEM - UNIT 2 XVI I Amendment No I
DESIGN FEATURES
== =--
==
= = =
= = = =
= = =
DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment is designed and shall be maintained for a maximum internal pressure of 47 psig.
Containment air temperatures up to 351.30 F are acceptable providing the containment pressure is in accordance with that described in the UFSAR.
5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 193 fuel assemblies.
Each assembly shall consist of a matrix of zircaloy or ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used.
Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases.
A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 53 full length and no part length control rod assemblies. The full length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing.
5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE l
5.4.1 The reactor coolant system is designed and shall be maintained:
- a.
In accordance with the code requirement specified in Section 4.1 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,
- b.
For a pressure of 2485 psig, and
- c.
For a temperature of 650'F, except for the pressurizer which is 680mF.
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