ML080940287

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Proposed License Amendment Request Revision of Technical Specifications Design Features
ML080940287
Person / Time
Site: Surry  Dominion icon.png
Issue date: 04/02/2008
From: Hartz L
Virginia Electric & Power Co (VEPCO)
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
08-0168
Download: ML080940287 (52)
v • d • e


Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 April 2, 2008 10CFR50.90 U. S. Nuclear Regulatory Commission Serial No. 08-0168 ATIN: Document Control Desk NLOS/GDM R2 Washington, D. C. 20555 Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 PROPOSED LICENSE AMENDMENT REQUEST REVISION OF TECHNICAL SPECIFICATIONS DESIGN FEATURES Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Dominion) requests amendments, in the form of changes to the Technical Specifications (TS) to Facility Operating License Numbers DPR-32 and DPR-37 for Surry Power Station Units 1 and 2, respectively. The proposed change revises TS Section 5.0, "Design Features,"

to delete certain design details and descriptions included in TS 5.0 that are already contained in the Updated Final Safety Analysis Report (UFSAR), or are redundant to existing TS requirements, and are not required to be included in the TS by 10 CFR 50.36(c)(4). The proposed change also revises the format of, and incorporates design descriptions into, TS 5.0 consistent with NRC policy and NUREG-1431, Standard Technical Specifications, Westinghouse Plants, Revision 3.0, to the extent practical. An editorial change is also proposed to address a minor TS discrepancy introduced by a previous license amendment. A discussion of the proposed change is provided in Attachment 1. The marked-up and typed proposed TS pages are provided in Attachments 2 and 3, respectively.

We have evaluated the proposed amendment and have determined that it does not involve a significant hazards consideration as defined in 10 CFR 50.92. The basis for our determination is included in Attachment 1. We have also determined that operation with the proposed change will not result in any significant increase in the amount of effluents that may be released offsite and no significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed amendment is eligible for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed change. The proposed TS change has been reviewed and approved by the Facility Safety Review Committee.

As this is an administrative change, NRC approval of the proposed TS change is requested by September 30,2008 with a 30 day implementation period.

Serial No. 08-0168 Docket Nos. 50-280/281 Page 2 of 3 If you have any questions or require additional information, please contact Mr. Gary D.

Miller at (804) 273-2771.

Sincerely, Vice President - Nuclear Support Services Attachments

1. Discussion of Change
2. Proposed Technical Specifications Pages (Mark-Up)
3. Proposed Technical Specifications Pages (Typed)

Commitments made in this letter: None COMMONWEALTH OF VIRGINIA )

)

COUNTY OF HENRICO )

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Leslie N. Hartz, who is Vice President - Nuclear Support Services, of Virginia Electric and Power Company. She has affirmed before me that she is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of her knowledge and belief.

Acknowledged before me this 0< AID day of {2p~ ,2008.

My Commission Expires: ~b ~< Joo8 .

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Notary Public (SEAL) ...,.....,

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Serial No. 08-0168 Docket Nos. 50-280/281 Page 3 of 3 cc: U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 NRC Senior Resident Inspector Surry Power Station State Health Commissioner Virginia Department of Health James Madison Building - i h Floor 109 Governor Street Room 730 Richmond, Virginia 23219 Mr. S. P. Lingam NRC Project Manager - Surry U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 8G9A Rockville, Maryland 20852 Mr. R. A. Jervey NRC Project Manager - North Anna U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 8G9A Rockville, Maryland 20852

Serial No. 08-0168 Docket Nos. 50-280/281 ATTACHMENT 1 DISCUSSION OF CHANGE Virginia Electric and Power Company (Dominion)

Surry Power Station Units 1 and 2

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1 DISCUSSION OF CHANGE

1.

SUMMARY

DESCRIPTION Virginia Electric and Power Company (Dominion) proposes a change to the Surry Power Station Units 1 and 2 Technical Specifications (TS) pursuant to 10 CFR 50.90.

The proposed change revises TS Section 5.0, "Design Features," to delete certain design details and descriptions included in TS 5.0 that are already contained in the Updated Final Safety Analysis Report (UFSAR), or are adequately addressed by existing TS requirements, and are not required to be included in the TS by 10 CFR 50.36(c)(4). The proposed change also revises the format of, and incorporates design descriptions into, TS 5.0 consistent with NRC policy and NUREG-1431, Standard Technical Specifications, Westinghouse Plants, Revision 3.0 (Reference 1), to the extent practical. An editorial change is also proposed to address a minor TS discrepancy introduced by a previous license amendment.

The proposed TS change has been reviewed, and it has been determined that no significant hazards consideration exists as defined in 10 CFR 50.92. In addition, it has been determined that the change qualifies for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9); therefore, no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed TS change.

2. DETAILED DESCRIPTION 2.1 Proposed Change The following specific changes to the Surry Units 1 and 2 TS are proposed:
  • TS TABLE OF CONTENTS

- Revise TS 5.1 title from "SITE" to "SITE LOCATION."

- Delete TS 5.2 CONTAINMENT.

- Revise TS 5.3 title from "REACTOR" to "REACTOR CORE."

- Renumber TS 5.3 and 5.4 as TS 5.2 and 5.3, respectively.

- Renumber TS page numbers as TS 5.0-1, TS 5.0-2, etc.

- Revise TS 5.1 , SITE, to:

o Change the section title from, "SITE," to "SITE LOCATION."

o Delete the Applicability, Objective and References sections, as well as the Specification header.

Page 1 of 20

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1 o Delete the discussion associated with the site exclusion area boundary.

o Delete TS Figure 5.1-1, Map Defining Unrestricted Areas for Radioactive Gaseous and Liquid Effluents.

- Delete TS 5.2, CONTAINMENT.

- Revise TS 5.3, REACTOR, to:

o Change the section title from "REACTOR" to "REACTOR CORE," and renumber as TS 5.2.

o Delete the Applicability, and Objective sections, as well as the Specification header.

o Delete TS 5.3.A.1 through 5.3.A.6 and replace with new TS 5.2.1, Fuel Assemblies, and 5.2.2, Control Rod Assemblies.

o Delete TS 5.3.8, Reactor Coolant System.

- Revise TS 5.4, FUEL STORAGE, to:

o Renumber as TS 5.3.

o Delete the Applicability, Objective and References sections, as well as the Specification header.

o Delete TS 5.4.A and first sentence of TS 5.4.8.

o Reformat TS 5.4.8 as new TS 5.3.1.1 and 5.3.1.2 to conform to NUREG-1431 content and format.

o Add new TS 5.3.1.1.a and 5.3.1.2.a for the spent fuel and new fuel storage racks, respectively, to establish an enrichment limit of 4.3 weight percent for stored fuel.

o Add new TS 5.3.1.2.b to establish a Keff limit ~O.95 for new fuel stored in the new fuel storage racks if fully flooded with unborated water.

o Retain the TS 5.4.8 discussion regarding spent fuel pool regional storage requirements as new TS 5.3.1.3.

o Renumber TS 5.4.C, which specifies the spent fuel pool boron concentration limit, as new TS 5.3.2. The word "pit" is also changed to "pool" for consistency in TS terminology. The associated footnote is deleted since it is obsolete.

o Replace TS 5.4.0 with new TS 5.3.3 to establish a limit for spent fuel pool level.

o Add new TS 5.3.4 to establish a limit for spent fuel pool storage capacity.

o Change figure number from TS Figure 5.4-1 to TS Figure 5.3-1.

Page 2 of20

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1

  • TS 6.2.C CORE OPERATING LIMITS REPORT Delete the reference to TS 5.3.A.6.b associated with the Moderator Temperature Coefficient contained in TS 6.2.C.1 since TS 5.3.A.6 is being deleted .
  • TS 6.4 UNIT OPERATING PROCEDURES AND PROGRAMS Revise TS 6.4 to add item C as "Deleted" to correct a TS numbering inconsistency.

2.2 Background TS 5.0, Design Features, contains design information regarding the plant site (TS 5.1),

the containment structure, penetrations and systems (TS 5.2), the reactor core and Reactor Coolant System (TS 5.3), and fuel storage (TS 5.4). A significant amount of this information is already contained in the UFSAR, or is adequately addressed by other TS requirements, and is not required to be included in the TS Design Features section by 10 CFR 50.36(c)(4), Design features. Likewise, much of this information does not conform to the TS guidance provided in NUREG-1431, Standard Technical Specifications, Westinghouse Plants, Revision 3.0. Consequently, it is proposed that TS 5.0 be revised to remove the information that is already, and more appropriately, included in the UFSAR, or is redundant to or adequately addressed by other TS requirements, and to revise and reformat the remaining TS information to conform to NUREG-1431 improved STS. The deleted TS information that is retained in the UFSAR will continue to be subject to and controlled by the requirements of 10 CFR 50.59, Changes, Tests and Experiments, should any future changes to the information be necessary. TS requirements are also being added to conform to NUREG-1431 guidance.

Separately, the implementation of a previous license amendment into the Surry Units 1 and 2 TS resulted in a minor editorial discrepancy that requires correction.

Consequently, the information within the specifications discussed below is proposed to be: 1) deleted since the information is already included in the Surry UFSAR or is adequately addressed by other existing TS requirements, and is not required by 10 CFR 50.36(c)(4), 2) revised to reflect the content and/or format of NUREG-1431, or

3) revised to make an editorial correction within the TS.
3. TECHNICAL EVALUATION The following discussion provides justification for the proposed changes discussed in Section 2.1 above. A cross-reference is provided in Table 1 indicating where information being deleted from the TS is included in the UFSAR, or is adequately addressed by other TS, and the associated industry precedent(s) for TS deletion/revision. Applicable precedent references are also provided at the end of each TS section discussion where appropriate.

Page 3 of 20

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1 TS 5.0 DESIGN FEATURES TS 5.1 through 5.4 contain general statements regarding the applicable requirements entitled "Applicability" and "Objective," as well as the section title "Specification." These TS also include a list of UFSAR references at the end of each section. The purpose of the general statements and headings is to provide an introduction to the Design Features, and the purpose of the reference section is to provide background information for the requirements. This format appears to be a carryover from the original Surry TS format used for the TS contained in Sections 3 (LCOs) and 4 (Surveillance Requirements). However, this information is not required to understand and apply the Design Features, and NUREG-1431 does not include these general statements and headings or the list of references. The TS descriptions are adequate to understand and apply the Design Features; therefore, these items are being deleted.

TS 5.1 SITE Consistent with the NUREG-1431 improved STS format, TS 5.1, "SITE," is renamed "SITE LOCATION." TS 5.1 is also revised to delete the existing discussion of the site exclusion area boundary and TS Figure 5.1-1, Map Defining Unrestricted Areas for Radioactive Gaseous and Liquid Effluents. The site exclusion area boundary text is duplicative of information already contained in the Surry UFSAR (Section 2.1.2, Exclusion Area Authority and Control), and TS Figure 5.1-1 is identical to UFSAR Figure 2.1-4, Site Boundary and Unrestricted Areas. Since changes to the UFSAR are subject to 10 CFR 50.59 requirements, any future changes to the information would be properly evaluated and adequately controlled. Furthermore, the deleted text and figure are not required to be included in the TS by 10 CFR 50.36(c)(4), and NUREG-1431 does not include this information in the Design Features section. Therefore, it is proposed that the site boundary discussion and Figure 5.1-1 be deleted from the TS. The existing site location description is retained and unchanged. Industry precedents associated with this change are provided in Table 1 and References 2 and 3.

TS 5.2 CONTAINMENT TS 5.2, CONTAINMENT, is deleted in its entirety, including subsections 5.2.A, Structure; 5.2.B, Containment Penetrations; and 5.2.C, Containment Systems. The design features, i.e., system design information referred to within these specifications, are duplicated within the UFSAR or, in certain cases, appropriately controlled by other applicable TS LCOs. The removal of these details from the TS is acceptable because this type of information is not necessary to be included in the TS to provide adequate protection of public health and safety. The removed information will be maintained in the UFSAR, which is controlled by 10 CFR 50.59 and therefore ensures that any changes will be properly evaluated. In addition, Surry TS contain requirements regarding containment OPERABILITY in TS 3.8 and containment leakage in TS 4.4 to ensure the containment is capable of performing its design function.

Page 4 of 20

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1 TS 5.2.C, Containment Systems, contains the minimum spray flows for the Containment Spray (CS) and Recirculation Spray (RS) Subsystems. While these values are significant, they are input parameters to the plant safety analyses for Surry, and it is these analyses and their results that are controlling. Therefore, maintaining the CS and RS Subsystems minimum flow values in the design features section has little additional safety benefit. The NRC came to the same conclusion regarding a safety analysis type value [Reactor Coolant System (RCS) volume] that was being removed from another licensee's TS Design Features section (Reference 10). Furthermore, CS and RS system operability and performance are adequately assured by existing TS 3.4, Spray Systems, and TS 4.5, Spray Systems Tests, which ensure the CS and RS systems will adequately perform their design functions.

Finally, the information contained in TS 5.2 does not meet the criteria of 10 CFR 50.36(c)(4) for items to be included within the TS, and the elimination of this information from the Design Features section has been previously approved by the NRC Staff for other licensees on that basis. Also, the proposed change is consistent with the guidance provided in NUREG-1431, which does not include containment specifications in the Design Features section of the improved STS. Applicable industry precedents associated with this change are provided in Table 1 and References 2,4,5, 6 and 10.

TS 5.3 REACTOR Consistent with the NUREG-1431 improved STS format, TS 5.3, "REACTOR," is renumbered as TS 5.2 and renamed "REACTOR CORE."

Existing TS 5.3.A.1, 2 and 3 contain details of fuel assembly design, such as number of fuel rods per fuel assembly, that the fuel rods are pressurized with helium, the approximate weight of the uranium dioxide fuel, the initial core loading average and maximum enrichment and the number of enrichments in the initial core, and the maximum enrichment of reload fuel. The existing TS wording regarding the number of fuel assemblies in each reactor core (Le., 157) is retained. However, the other information is replaced with the more general statement of, "Each assembly shall consist of a matrix of Zircaloy or ZIRLO fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material." This proposed change to the TS to eliminate the detailed description of fuel assemblies is permissible since this information is duplicative of information already contained in the UFSAR, and because this type of information is not required to be included in the TS by 10 CFR 50.36(c)(4).

The proposed change is also consistent with the guidance provided in NUREG-1431, which does not include this information in the Design Features section of the improved STS. Since the deleted TS information is contained in the UFSAR, it will be adequately controlled in accordance with 10 CFR 50.59, which ensures changes are properly evaluated. The fuel enrichment limit currently specified in TS 5.3.A.3 is retained in TS 5.3.1.1.a and 5.3.1.2.a for spent and new fuel, respectively.

Page 5 of 20

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1 New TS 5.2.A.1 (Le., revised TS 5.3.A.1) includes the statement, "Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with 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 non-limiting core locations." The purpose of this statement is to reiterate that all aspects of reactor core design must be performed in accordance with NRC staff approved methods and with approved computer codes.

Surry TS Amendment 102/102, dated August 25, 1985, (Reference 7) previously revised TS 5.3.A.1 to allow the use of fuel assemblies which have been reconstituted to replace leaking fuel rods with non-fuel rods (e.g., Zircaloy or stainless steel). However, the current TS 5.3.A.1 does not specifically require the use of NRC-approved codes and methods. On July 31, 1992, the NRC issued Generic Letter (GL) 90-02 Supplement 1, "Alternative Requirements for Fuel Assemblies in the Design Features Section of Technical Specifications," (Reference 8) which included a model TS and stated that licensees may propose the model TS wording as noted above. In a letter dated March 14, 1995 (Reference 9), Dominion committed to comply with the intent of the GL 90-02 Supplement 1 TS wording for Surry Power Station but did not revise the TS at that time to include the proposed wording. The statement that fuel designs must be limited to those analyzed with NRC staff approved codes and methods is consistent with 10 CFR 50.59 and previously stated NRC policy. One slight difference from the GL wording is being made. The phrase "nonlimiting core locations" is used instead of the phrase "nonlimiting core regions." This wording is consistent with that approved for the North Anna Power Station Technical Specifications in Amendment 204/185 dated May 9, 1997 (Reference 16). Therefore, the proposed change does not result in any change in operation of the plant, and the change is designated as administrative because it does not result in a technical change to the specifications.

TS 5.3.A.4 contains details of burnable poison rods. Burnable poison rods are included in a core design to control core power distribution. The TS already include requirements on core power distribution, such as Fo{Z) and FNDoH, in TS 3.12. Also, NUREG-1431 does not contain a description of burnable poison rods nor is this information required by 10 CFR 50.36{c){4). The proposed change eliminates the description of burnable poison rods since it is duplicative of information contained in the UFSAR. The UFSAR is controlled in accordance with 10 CFR 50.59, which ensures changes are properly evaluated.

TS 5.3.A.5 contains details of control rod design, such as the nominal length of absorber material and control rod cladding material. The improved TS do not contain these details and, instead, provides a general statement of, "The control material shall be silver indium cadmium as approved by the NRC." Furthermore, the TS include requirements on control rod OPERABILITY in TS 3.12. This change is acceptable because the removed information is duplicative of information contained in the UFSAR and will be controlled in accordance with 10 CFR 50.59, which ensures changes are properly evaluated. The revised specification is renumbered as TS 5.2.2.

Page 6 of 20

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1 TS 5.3.A.6 addresses hot channel factor and moderator temperature coefficient limits and the requirement for the reactor core to be made subcritical. This specification is being deleted. This proposed change is acceptable because the technical requirements have not changed. Specifically, as indicated in TS 5.3.A.6.a, the hot channel factors requirements are provided in TS 3.12.B.1 and must be met. The moderator temperature coefficient requirements in TS 5.3.A.6.b are repeated verbatim in TS 3.1.E.

Finally, the TS 5.3.A.6.c subcriticality requirement is contained in TS 3.12.A.3.C.

Consequently, TS 5.3.A.6 is redundant to other TS requirements and therefore unnecessary. An associated reference to TS 5.3.A.6 contained in TS 6.2.C.1 is also being deleted. Thus, this change is administrative in nature because it does not result in a technical change to the specifications, and merely deletes redundant TS requirements, which are not required by 10 CFR 50.36(c)(4) and are not contained in NUREG-1431.

TS 5.3.B describes certain RCS design criteria associated with code requirements, seismic qualifications and RCS volume. The proposed change deletes TS 5.3.B from the TS in its entirety because: 1) this information is duplicative of information contained in the UFSAR, 2) NUREG-1431 does not contain this information, and 3) this information does not satisfy any of the inclusion criteria specified in 10 CFR 50.36(c)(4).

Also, while the RCS volume value is significant, it is an input parameter to the plant safety analyses for Surry, and it is these analyses and their results that are controlling.

Therefore, maintaining the RCS volume in the Design Features section has little additional safety benefit. The NRC came to the same conclusion regarding an RCS volume value that was being removed from another licensee's TS Design Features section (Reference 10). Furthermore, TS requirements on RCS OPERABILITY are contained in TS 3.1, and surveillance requirements associated with RCS Operational LEAKAGE are included in TS 4.13. These TS requirements adequately control RCS parameters, such as pressure, temperature, and pressure boundary degradation, which could have a significant impact on safety. Also, this change is acceptable because the removed information will be adequately controlled in the UFSAR. The UFSAR is controlled in accordance with 10 CFR 50.59, which ensures changes are properly evaluated Applicable industry precedents associated with this change are provided in Table 1 and References 2, 5, 6, 10, 11, 12 and 16.

TS 5.4 FUEL STORAGE Consistent with the NUREG-1431 improved STS format, TS 5.4, "FUEL STORAGE," is renumbered as TS 5.3.

Current TS 5.4.A and 5.4.B contain descriptions of the Fuel Building, the spent fuel storage racks, and the new fuel storage racks, such as materials of construction, seismic design, fuel orientation, and operational controls. The proposed change eliminates details of the seismic and structural design of the Fuel Building and spent Page 7 of 20

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1 fuel storage racks and reformats the section to conform to NUREG-1431 gUidance.

However, the proposed TS change retains requirements on the nuclear design characteristics of the spent fuel and new fuel storage racks, such as maximum center-to-center fuel cell spacing and maximum enrichment. Also, the deleted TS information is contained in the UFSAR and therefore controlled by 10 CFR 50.59, which ensures changes are properly evaluated. The current TS requirements associated with spent fuel regional storage contained in TS 5.4.8 and Figure 5.4-1, and the spent fuel pool boron concentration limit included in TS 5.4.C, are being retained as TS 5.3.1.1.d and TS 5.3.2, respectively. The footnote associated with the spent fuel pool boron concentration limit is being deleted since it is associated with an earlier operating cycle, and thus no longer provides pertinent or meaningful information.

Current TS 5.4.D addresses how draining of the spent fuel pool is prevented. TS 5.4.0 is being replaced by proposed TS 5.3.3 which states, "The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below the elevation 41 feet, 2 inches mean sea level, USGS datum." The proposed change revises the TS by eliminating details of how the spent fuel storage pool is operated to prevent draining of the spent fuel pool below the required level. The removal of these details, which are related to system operation, from the TS is acceptable because this type of information is not necessary to be included in the TS to provide adequate protection of public health and safety. The TS retain the requirement that the spent fuel storage pool is designed and maintained to prevent inadvertent draining below the required level. It is not necessary or appropriate for the TS to state how this requirement is implemented.

Proposed TS 5.3.4, Capacity, states that, "The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1044 fuel assemblies." The current TS do not contain this information. This change is acceptable because it provides appropriate limits for the spent fuel pool storage racks. The Atomic Energy Act of 1954 (Public Law 83-703), Section 2232, states that technical specifications shall include information of the "amount, kind, and source" of the special nuclear material required. Therefore, it is necessary to state the limit on the amount of spent fuel which may be stored in the spent fuel pool. This proposed change is more restrictive because it adds a limit to the Surry TS that does not currently exist.

An applicable industry precedent associated with this change is provided in Table 1 and Reference 6.

TS 6.4 UNIT OPERATING PROCEDURES AND PROGRAMS License Amendment 244/243 for Surry Units 1 and 2, respectively, (Reference 13) revised the Administrative Controls section of the TS to support implementation of Topical Report DOM-QA-1, "Dominion Nuclear Facility Quality Assurance Program Description." As part of the TS revision, the organization and responsibilities of the onsite and offsite safety review committees were relocated from the TS to the Topical Report. This required two items in TS 6.4, items TS 6.4.C and 6.4.G, to be relocated to Page 8 of 20

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1 the Topical Report. The intent was to retain the TS 6.4 "C" and "G" item designators with the word "Deleted" inserted in place of the relocated TS requirements to preclude renumbering the subsequent TS items and their associated references within other TS.

While this was done for TS 6.4.G, TS 6.4.C was inadvertently deleted in its entirety, thus introducing a TS numbering inconsistency into TS 6.4. Consequently, the proposed change reinserts TS item designator C into TS 6.4 with the word "Deleted" to restore the correct TS numbering sequence.

Page 9 of 20

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1 TABLE 1 TS 5.0 DESIGN FEATURES INFORMATION CONTAINED IN THE UFSAR OR OTHER TS AND ASSOCIATED PRECEDENTS FOR REMOVAL SURRY UNITS 1 AND 2 TS Section Deleted/Revised TS Location of Deleted TS Precedents Information Information (where applicable) 5.1 Site Location Maintenance of Site Location description and deletion of Exclusion Area TS and Map Defining Unrestricted Areas for EAB Discussion UFSAR 2.1.2 Radioactive Gaseous and Liquid Effluents

  • NUREG-1431, Revision 3.0 Fig. 5.1-1 Map Defining Unrestricted UFSAR Fig. 2.1-4 Areas for Radioactive
  • Letter from the USNRC to K. W. Singer of Tennessee Valley Gaseous and Liquid Authority, dated August 2, 2006, "Sequoyah Nuclear Plant, Effluents Units 1 and 2 - Issuance of Amendments Regarding Technical Specification Changes to Cyclic and Transient Limits with Design Features Revision (TAC Nos. MC8532 and MC8533) (TS05-02)."
  • Letter from the USNRC to J. A. Stall of Florida Power and Light Company, dated February 12, 2002, Turkey Point Units 3 and 4 - Issuance of Amendments Regarding Removal of Site Area and Plant Area Maps from Technical Specifications (TAC Nos. MB1968 and MB1969)."

Page 10 of 20

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1 TABLE 1 TS 5.0 DESIGN FEATURES INFORMATION CONTAINED IN THE UFSAR OR OTHER TS AND ASSOCIATED PRECEDENTS FOR REMOVAL SURRY UNITS 1 AND 2 TS Section I Deleted/Revised TS Location of Deleted TS Precedents Information Information (where applicable) 5.2 I Containment Deletion of TS 5.2 Containment

  • Letter from the USNRC to K. W. Singer of Tennessee Valley UFSAR 1.4.10 Authority, dated August 2, 2006, "Sequoyah Nuclear Plant, UFSAR 1.4.49 Units 1 and 2 - Issuance of Amendments Regarding UFSAR 5.1 Technical Specification Changes to Cyclic and Transient UFSAR 5.4 Limits with Design Features Revision (TAC Nos. MC8532 and UFSAR 15.2 MC8533)(TS05-02)."

UFSAR 15.5

  • Letter from USNRC to J. A. Spina, Nine Mile Point Nuclear Station, LLC, dated June 6, 2005, "Nine Mile Point Nuclear Station, Unit NO.1 - Issuance of Amendment Re: Relocation of Design Features from the Technical Specifications to the 5.2.B I Containment Penetrations UFSAR 5.2 Updated Final Safety Analysis Report (TAC No. MC4928)."
  • Letter from USNRC to C. Randy Hutchinson, Arkansas Nuclear One, Unit No.2, dated May 19, 1999, "Issuance of Amendment Re: Design Features and Administrative Controls," (TAC No. MA2403).
  • Letter from the USNRC to R. E. Denton, Baltimore Gas and Electric, dated March 14, 1995, "

Subject:

Issuance of 5.2.C I Containment Systems UFSAR 5.3 Amendments for Calvert Cliffs Nuclear Power Plant, Unit NO.1 UFSAR 6.2 UFSAR 6.3 (TAC No. M88429) and Unit NO.2 (TAC No. M88230)"

  • Letter from USNRC to C. R. Hutchinson, Entergy Operations Incorporated, dated April 16, 1997, "Issuance of Amendment No. 181 to Facility Operating License No. NPF Arkansas Nuclear One, Unit No.2 (TAC No. M97534)." (Amendment No. 181)

Page 11 of 20

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1 TABLE 1 TS 5.0 DESIGN FEATURES INFORMATION CONTAINED IN THE UFSAR OR OTHER TS AND ASSOCIATED PRECEDENTS FOR REMOVAL SURRY UNITS 1 AND 2 TS Section Deleted/Revised TS Location of Deleted TS Precedents Information Information (where applicable) 5.3 Reactor Revised Reactor Core (Fuel Assemblies/Control Rod Assemblies) TS Wording 5.3.A Reactor Core UFSAR Table 1.3-1 UFSAR 3.3

  • GL 90-02, Supplement 1, dated July 31, 1992, "Alternative Requirements for Fuel Assemblies in the Design Features TS3.1.E Section of Technical Specifications."

TS 3.12 COLR

  • Letter from USNRC to C. Randy Hutchinson, Arkansas Nuclear One, Unit No.2, dated May 19, 1999, "Issuance of Amendment Re: Design Features and Administrative Controls," (TAC No. MA2403).
  • Letter from the USNRC to R. E. Denton, Baltimore Gas and Electric, dated March 14, 1995, "

Subject:

Issuance of Amendments for Calvert Cliffs Nuclear Power Plant, Unit No.

1 (TAC No. M88429) and Unit NO.2 (TAC No. M88230)."

  • Letter from USNRC to T. C. McMeekin, Duke Power Company, dated July 19, 1993, "Issuance of Amendments -

McGuire Nuclear Station, Units 1 and 2 (TAC Nos. M86015 and M86016)."

  • Letter from USNRC to Virginia Electric and Power Company dated May 9,1997 (Serial No.97-310), "

Subject:

North Anna Units 1 and 2 - Issuance of Amendments Re: Demonstration Fuel Assemblies (TAC Nos. M96530 and M96531)."

Page 12 of 20

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1 TABLE 1 TS 5.0 DESIGN FEATURES INFORMATION CONTAINED IN THE UFSAR OR OTHER TS AND ASSOCIATED PRECEDENTS FOR REMOVAL SURRY UNITS 1 AND 2 TS Section Deleted/Revised TS Location of Deleted TS Precedents Information Information (where applicable)

Deletion of TS 5.3.8 Reactor Coolant System 5.3.B Reactor Coolant System UFSAR Table 1.3-1

  • Letter from the USNRC to K. W. Singer of Tennessee Valley UFSAR Appendix 15A Authority, dated August 2, 2006, "Sequoyah Nuclear Plant, Units 1 and 2 - Issuance of Amendments Regarding Technical Specification Changes to Cyclic and Transient Limits with Design Features Revision (TAC Nos. MC8532 and MC8533)(TS05-02)."
  • Letter from USNRC to C. R Hutchinson, Arkansas Nuclear One, Unit No.2, dated May 19,1999, "Issuance of Amendment Re: Design Features and Administrative Controls," (TAC No. MA2403).
  • Letter from USNRC to W. T. Cottle, STP Nuclear Operating Company, dated November 18, 1998, "South Texas Project, Units 1 and 2 - Amendment Nos. 98 and 85 to Facility Operating License Nos. NPF-76 and NPF-80 (TAC Nos.

MA2502 and MA2503)."

  • Letter from the USNRC to R. E. Denton, Baltimore Gas and Electric, dated March 14, 1995, "

Subject:

Issuance of Amendments for Calvert Cliffs Nuclear Power Plant, Unit No.

1 (TAC No. M88429) and Unit NO.2 (TAC No. M88230)."

Deletion of RCS Volume

  • Letter from USNRC to C. R. Hutchinson, Entergy Operations Incorporated, dated April 16, 1997, "Issuance of Amendment No. 181 to Facility Operating License No. NPF Arkansas Nuclear One, Unit NO.2 (TAC No. M97534)." (Amendment No. 181)

Page 13 of 20

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1 TABLE 1 TS 5.0 DESIGN FEATURES INFORMATION CONTAINED IN THE UFSAR OR OTHER TS AND ASSOCIATED PRECEDENTS FOR REMOVAL SURRY UNITS 1 AND 2 TS Section Deleted/Revised TS Location of Deleted TS Precedents Information Information (where applicable) 5.4 Fuel Storage UFSAR 1.4.66

  • Letter from the USNRC to R. E. Denton, Baltimore Gas and Electric, dated March 14, 1995, "

Subject:

Issuance of UFSAR Appendix 9A Amendments for Calvert Cliffs Nuclear Power Plant, Unit No.

1 (TAC No. M88429) and Unit No.2 (TAC No. M88230)."

Page 14 of 20

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to develop TS, which are included as a part of the operating license (OL). 10 CFR 50.36, Technical specifications, sets forth the content of the TS. This regulation requires the TS to include items in specific categories including, (1) safety limits, limiting safety system settings, and limiting control settings, (2) limiting conditions for operation (LCOs), (3) surveillance requirements, (4) design features, and administrative controls. 10 CFR 50.36 does not specify the particular specifications to be included as part of a plant's OL. By letter dated May 9, 1988, (Reference 14) the NRC described results of an NRC staff review to determine which LCOs should be included in the TS. This ultimately resulted in four criteria being developed, as described in the "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" (Reference 15), which were later codified in 10 CFR 50.36(c)(2)(ii). The Final Policy Statement also included the position that specifications that do not meet the 10 CFR 50.36 applicability criteria may be proposed for removal and relocation to licensee-controlled documents such as the Final Safety Analysis Report (FSAR).

Prior to development of the improved STS, there was little guidance or standardization as to what design features should be included in the TS. Evaluation during improved STS development identified the features required to be retained in the TS, and those that could be relocated to other licensing basis documents, e.g., the UFSAR. With respect to design features, 10 CFR 50.36(c)(4) provides guidance as to which design features should be included in TS. It 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."

System design information, e.g., specifications and figures, not included within the improved STS for the respective unit design, does not meet the criteria of 10 CFR 50.36(c)(4) for inclusion as design features and may be relocated from the TS to other licensee-controlled documents. The design features proposed for removal in the proposed change are included in the TS but do not meet the applicability requirements in 10 CFR 50.36. Several licensees have made similar changes during conversion to vendor-specific Standard TSs, such that information is relocated to the FSAR, which includes 10 CFR 50.59 review requirements. The FSAR meets the NRC expectations for a licensee-controlled document for the purpose of removal and relocated specifications that do not meet 10 CFR 50.36 criteria. Therefore, the removal of the design features identified in the proposed TS change is acceptable because they do not meet the criteria of 10 CFR 50.36(c)(4).

NUREG-1431 provides generic recommendations for requirements associated with the operation of Westinghouse Electric Company designed nuclear steam supply systems such as Surry. The design features proposed for removal from the Surry TS are not Page 15 of 20

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1 included in NUREG-1431. Consequently, the removal of certain design features from the Surry TS and reliance on other licensee-controlled documents is consistent with NUREG-1431.

Therefore, Dominion proposes to revise the specifications in TS Section 5.0 to reflect the NUREG-1431 improved STS, to the extent practical, and 10 CFR 50.36(c)(4) inclusion criteria. The TS items proposed for deletion are already included within the UFSAR or are addressed by other TS requirements, which will ensure that any future changes are evaluated under 10 CFR 50.59 or 10 CFR 50.90, as applicable.

Consequently, the proposed change meets the regulations and is consistent with NUREG-1431 guidance, industry precedents, and NRC Policy. Therefore, the proposed change is acceptable.

4.2 Precedents Industry precedents in support of the proposed change are referenced in the Section 3 Technical Evaluation discussion above, included in Table 1 with the associated TS, and listed in Section 6, References, below.

4.3 Determination of No Significant Hazards Consideration Virginia Electric and Power Company (Dominion) proposes a change to the Surry Units 1 and 2 Technical Specifications (TS) to revise Section 5.0, "Design Features" to delete information that is not required to be included in TS 5.0 by 10 CFR 50.36(c)(4).

The TS information being deleted is either included in other licensing basis documents, e.g., the Updated Final Safety Analysis Report, or is adequately addressed by other TS requirements. Also, new TS limits associated with fuel storage are being incorporated into TS 5.0 consistent with the Improved Standard Technical Specifications for Westinghouse Plants contained in NUREG-1431, Revision 3.0. Other administrative changes are also proposed involving reformatting, renumbering, and rewording of the Design Features TS for consistency with NUREG-1431 and existing NRC Policy. An editorial discrepancy introduced during the implementation of an earlier license amendment is also being corrected.

In accordance with the criteria set forth in 10 CFR 50.92, Dominion has evaluated the proposed TS change and determined that the change does not represent a significant hazards consideration. The following is provided in support of this conclusion:

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change to Section 5.0, "Design Features," deletes certain details from the TS that are not required to be maintained in the TS by 10 CFR 50.36(c)(4), adds new TS limits that meet the 10 CFR 50.36(c)(4)

Page 16 of 20

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1 inclusion criteria and revises the TS for consistency with NUREG-1431, Revision 3.0. The remaining change addresses a minor editorial discrepancy.

The proposed change does not add or modify any plant system, structures or component and has no impact on plant equipment operation. Thus, the proposed change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accident or transient events.

Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

Since the proposed change is administrative in nature, it does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility. The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

Response: No.

The proposed TS change is administrative in nature and as such does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, and the dose analysis acceptance criteria are not affected. The proposed change does not result in plant operation in a configuration outside the analyses or design basis and does not adversely affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition. Therefore, the proposed TS change does not involve a significant reduction in a margin of safety.

Based on the above, Dominion concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

Page 17 of 20

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1 4.4 Conclusion Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by implementation of the proposed TS change, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5. ENVIRONMENTAL CONSIDERATION The proposed amendment is confined to administrative changes only. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(1 0) as follows:

(i) The proposed change involves no significant hazards consideration.

As described in Section 4.3 above, the proposed change involves no significant hazards consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

The proposed change is administrative in nature and therefore does not involve the installation of any new equipment or the modification of any equipment that may affect the types or amounts of effluents that may be released offsite.

Therefore, there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

(iii) There is no significant increase in individual or cumulative occupation radiation exposure.

The proposed change is administrative in nature and therefore does not involve physical plant changes or introduce any new modes of plant operation.

Therefore, there is no significant increase in individual or cumulative occupational radiation exposure.

Based on the above, Dominion concludes that, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Page 18 of 20

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1

6. REFERENCES
1. NUREG-1431, Standard Technical Specifications, Westinghouse Plants, Volume 1, Rev. 3.0, dated June 2004.
2. Letter from the USNRC to K. W. Singer of Tennessee Valley Authority, dated August 2, 2006, "Sequoyah Nuclear Plant, Units 1 and 2 - Issuance of Amendments Regarding Technical Specification Changes to Cyclic and Transient Limits with Design Features Revision (TAC Nos. MC8532 and MC8533) (TS05-02)."
3. Letter from the USNRC to J. A. Stall of Florida Power and Light Company, dated February 12, 2002, "Turkey Point Units 3 and 4 - Issuance of Amendments Regarding Removal of Site Area and Plant Area Maps from Technical Specifications (TAC Nos. MB1968 and MB1969)."
4. Letter from USNRC to J. A. Spina, Nine Mile Point Nuclear Station, LLC, dated June 6, 2005, "Nine Mile Point Nuclear Station, Unit No. 1 - Issuance of Amendment Re: Relocation of Design Features from the Technical Specifications to the Updated Final Safety Analysis Report (TAC No. MC4928)."
5. Letter from USNRC to C. Randy Hutchinson, Arkansas Nuclear One, Unit No.2, dated May 19, 1999, "Issuance of Amendment Re: Design Features and Administrative Controls, (TAC No. MA2403)."
6. Letter from the USNRC to R. E. Denton, Baltimore Gas and Electric, dated March 14, 1995, "

Subject:

Issuance of Amendments for Calvert Cliffs Nuclear Power Plant, Unit NO.1 (TAC No. M88429) and Unit No.2 (TAC No. M88230)."

7. Letter from USNRC to W. L. Stewart, Virginia Electric and Power Company, dated August 26, 1885 (Serial No.85-649), Amendment 102/102 regarding the Use of Reconstituted Fuel Assemblies.
8. Supplement 1 to Generic Letter 90-02, "Alternative Requirements for Fuel Assemblies in the Design Features Section of Technical Specifications," dated July 31, 1992.
9. Letter from J. P. O'Hanlon, Virginia Electric and Power Company, to the USNRC dated March 14, 1995 (Serial No.95-107), "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Use of Reconstituted Fuel Assemblies and Application of NRC-Approved Methodology."
10. Letter from USNRC to C. Randy Hutchinson, Arkansas Nuclear One, Unit No.2, dated April 16, 1997, "Issuance of Amendment 181 to Facility Operating License No. NPF-6, Arkansas Nuclear One, Unit No.2, (TAC No. M97534)."

Page 19 of 20

Serial No. 08-0168 Docket Nos. 50-280/281 Attachment 1

11. Letter from USNRC to W. T. Cottle, STP Nuclear Operating Company, dated November 18, 1998, "South Texas Project, Units 1 and 2 - Amendment Nos. 98 and 85 to Facility Operating License Nos. NPF-76 and NPF-80 (TAC Nos.

MA2502 and MA2503)."

12. Letter from USNRC to T. C. McMeekin, Duke Power Company, dated July 19, 1993, "Issuance of Amendments - McGuire Nuclear Station, Units 1 and 2 (TAC Nos. M86015 and M86016)."
13. Letter from USNRC to D. A. Christian, Virginia Electric and Power Company, dated September 15, 2005 (Serial No.05-659), "

Subject:

Surry Power Station, Units 1 and 2 - Issuance of Amendments to Revise the Administrative Controls Section (TAC Nos. MC4412 AND MC4413)."

14. Thomas E. Murley, Director, Office of Nuclear Reactor Regulation, US. Nuclear Regulatory Commission, letter to Walter S. Wilgus, Chairman, The B&W Owners Group, "NRC Staff Review of Nuclear Steam Supply System Vendor Owners Groups' Application of the Commission's Interim Policy Statement Criteria to Standard Technical Specifications," dated May 9, 1988.
15. US. Nuclear Regulatory Commission, "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" (Final Policy Statement), 58 FR 39132, dated July 22, 1993.
16. Letter from USNRC to Virginia Electric and Power Company dated May 9, 1997 (Serial No.97-310), "

Subject:

North Anna Units 1 and 2 - Issuance of Amendments Re: Demonstration Fuel Assemblies (TAC Nos. M96530 and M96531 )."

Page 20 of 20

Serial No. 08-0168 Docket Nos. 50-280 and 281 ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATIONS PAGES (MARK-UP)

Virginia Electric and Power Company (Dominion)

Surry Power Station Units 1 and 2

TS iii B3-2987 TECHNICAL SPECIFICATION TABLE OF CONTENTS SECTION TITLE PAGE 4.15 AUGMENTED INSERVICE INSPECTION PROGRAM FOR HIGH TS 4.15-1 ENERGY LINES OUTSIDE OF CONTAINMENT 4.16 LEAKAGE TESTING OF MISCELLANEOUS RADIOACTIVE TS 4.16-1 MATERIALS SOURCES 4.17 SHOCK SUPPRESSORS (SNUBBERS) TS 4.17-1 4.18 DELETED 4.19 STEAM GENERATOR (SG) TUBE INTEGRITY TS 4.19-1 e 4.20 CONTROL ROOM AIR FILTRATION SYSTEM TS 4.20-1

_..J-~

5.0 DESIGN FEATURES -./~ TS 5:9-1

--~

,/

..r --.r-5.1 SITE Lex.: A', 10 JJ TS 5!J-l

~.2 CON'fAINM8NT 'TS 5.2-1 REACTOR CoRE: TS 5j-l FUEL STORAGE TS

/1-1..

M' 6.0 ADMINISTRATIVE CONTROLS TS 6.1-1 6.1 ORGANIZATION, SAFETY AND OPERATION REVIEW TS 6.1-1 6.2 GENERAL NOTIFICATION AND REPORTING REQUIREMENTS TS 6.2-1 6.3 ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED TS 6.3-1 6.4 UNIT OPERATING PROCEDURES AND PROGRAMS TS 6.4-1 6.5 STAnON OPERATING RECORDS TS 6.5-1 6.6 STAnON REPORTING REQUIREMENTS TS 6.6-1 6.7 ENVIRONMENTAL QUALIFICAnONS TS 6.7-1 6.8 PROCESS CONTROL PROGRAM AND OFFSITE DOSE TS 6.8-1 CALCULAnON MANUAL Amendment Nos. ~ 1 and 250

TS 5:++

98-10-06 5.0 DESIGN FEATURES 5.1 SITE Loe A\ lOtJ I

( or the Surry Power Station.

\

')

Objective

. e which will affect the ov

/

S ecification _._---"~~

The Surry Power Station is located in Surry County, Virginia, on property owned by Virginia Electric and Power Company on a point of land called Gravel Neck which juts into the James River. It is approximately 46 miles SE of Richmond, Virginia, 17 miles NW of Newport News, Virginia, and 25 miles NW of Norfolk, Virginia. 'Hte-stfe.-e*:ekl~*HEa'@tHletiitQ~l-Hl-4E~~~r--""'*r

-the~'Sfte*beundary line shown en-the site map in T8 Figure 5.1 1. The the di~ttm:ee ter Vliit 1, v/hieh i~ eontfolling.

-Referenc~

....---~.......

FSAR s IOn 2.0 Site "",

AR sectio:~~Pti:Y---

,-/

Amendment Nos. 249 and 248-

11 Inch =Approximately 1,000 Feetl STATION A. Gaseous Release

1. Process Vent - 131 Ft. - Mixed Mode JAMES 2. Vent-Vent Slacks RIVER Ground level B. liquid leaves Site RADWASTE FACILITY C. Gaseous Release Ground level

- - - - Site Bou-.... ~Area Unre~f~~Gaseous AI or Beyond is Effluents I Site Boundary I 1_ - - ........

I I I I Maximum Individual OCcupancy Within Site 1 I ndary:

1. Canal Bank Fishing = 160 HrlYr Liquid Maximum Individual Occupancy Within Site Boundary:
1. Boat Fishing Discharge Canal = 800 Hr/Yr

... .... 1 Site Boundary .........

" ... ------------------------------------- >-3 CI.l "Ij

~

rIO'

~"

~

~

C\.I 0

__ ="'1 C>

MAP DE ING UNREST ':IGURE5.1=

TED ARE FOR RADIO CTIVE

'. I

'NUl (1l

~ ~;.....

o SEOUS AND QUID EFFL NTS (J)

I

e.~

~

TS 5 -1 3 7-72 5.2 CONTAI Applicability atures of the reactor containment structures and ontainment systems relating to operational and Objective To define ths significant design f tures of the reactor contai ent structures and containment systems.

Specifications A. Structure

1. A containment encloses each reactor and Reactor Coolant per limit for leakage of radioactive materials occurs. Each structure R ovides biological ielding for both normal operation and accident situations cture is designed for an internal
2. ainment structure is designed for a reactor operating at the ultimate ted
3. structure is designed to withstand an internal des* n pressure of 45 psig acting simu aneously with: (1) loads resulting from an Earthquake having a ho *zontal ground acceleration of 0.07 g t zero period with an assumed structural dampin factor of 5 percent, or (2) loa resulting from a Design Basis Earthquake having a h izontal ground acceler ion of 0.15 g at zero period B. Containment Penetrations
1. All penetrations through the containme structure for pipe, electrical conductors, 2.

The actuation system is desig d such that no sin e component failure will prevent in the Technical Specifications

a. A safety injec *on signal closes all trip valves whic are located in normally open lines connecting the reactor coolant loops nd penetrating the

TS 5.2-3/'

08-03-95

,/'

?

b. tainment pressure isolation signal closes the automatic tripytllves in pen lines penetrating the containment which are not r~ired to be open to contro containment pressure to perform an orderly r~ic;tor shut down without actuatio of the consequence limiting safeguard,slin case of a small Reactor Coolant S

/

c. A further rise in pressure, indica¥g a major loss-of-coolant accident, produces a co tainment high-high essure isolation signal which closes all normally open li s which penetr te the containment which have not been closed by 2-b above.
d. Isolation can be accomplished ally from the control in the Main Control Room if any of the automatic si fail to actuate the above valves.

C. Containment Systems

1. Following a loss-of-coolant ccident, the Conta ment Spray Subsystems distribute at least 2,600 gpm borated ater spray containing s ium hydroxide for iodine removal within the containme atmosphere. The Recirculat n Spray Subsystems recirculate at least 3,000 gpm f water from the containment sum .

Amendment Nos. 293 aed 203 _

TS 5.2-4 y17-72

2. e Containment Ventilation System is designed for cont" ed operation during a tota loss-of-coolant accident. It may, however, cont" e to operate with References FSAR Section 5.2 FSAR Section 5.3 FSAR Section 5.4 FSAR Section 7.2.2 FSAR Section 15.2.4 FSAR Section 15.5.1 Technical Specification

/

TS-573-l-

  • 7 27 9~

'011- REACTOR

/ Applicability

\ t System, and Safety Inject" n System.

I ential in providin

, or safe system "J operations. ,-----f

/

\ . ~

"'. A. Reactor Core ./

'---= ------.r--..

The reactor core con aIn;-approxim~6,200 lbs ~~.

fonn of slightly enriched uranium dioxide pellets. The pellets are encapsulated~

Zircaloy-4 or ZIRLO tubing to form rods. All fuel rod re pressurized with I \

helium during fabrication. T pt for fuel assemblies which may eplace leaking fuel r s with non-fueled rods (e.g. . aloyor Itial core is 2.51 weight p cent ofU-235. Three

_ _-,uel e .chments are used in t e initial core. The highes percent of U-235.

-/ .. '-...", ~,.~

_.---=-----'v ----:::h:::O=----=::::~ ---- ~.. ~~/~~

r?~~::-Iu~ ~r:~\ ::~~4"."_, 1.5'7 J:."/cs>e~b.l,e,. k,,1)

".;,,~h'j 5 h.. \ \ t,:,,,,,.,, -t c+ ~ ,",,,+,-,'X c+ ;t, ve.. /Oj C- ']

)2jRLO LeI vod~ w',th ~ Ir"C+*~1 Ccj'<'\(Cls.,'+.~n ot Vl~}u,,'~i /

C < S \ ') ~ +Ij e " r, C he ,I LA ... " " . u,...... cI '6 ><.' ,Ie (<..J 0 2.) '" 5 .( ",I

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f

~~_.,., /~ ---~._--/

Amendment No:Z02 Rfl:8 2G2-

INSERT 1 Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with 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 non-limiting core locations.

3. Reload fuel will be similar in design to the initial core. The enrichment of reload",\_

"\,

fuel will not exceed 4.3 weight percent ofU-235. 1\\

4. Burnable poison rods are inc orated in the initial core. There are 816 poison rods \

\

\

in the form of 12 rod sters, which are located in vacant control rod assembly \

i guide thimbles. e burnable poison rods consist yrex clad with stainless steel.

5. blies in the reactor core. The full-length co rol rod assemblies contain 44-inch length of silver-indium-3admium alloy J

clad with stainless steel. /

6. Surry uni~ 4, Surry Unit 2, Cycle 3, a /J following criteria at all times during th~ eration lifetime. _//
a. Hot channel factor limits as specified in Section 3.12 shall be met. ~

~-~

\i

~~----~

~e I e c.c-\:c.. ~ Cov I;Z ~h~ \\ <fi", -\0 {I/\ -4' 8 Co." -ho1 ('0 d csseV""6I,e;,). -fJJe CA:J .... -I>-ol rt"o-- +e Y~"d I S' he.11 be.

Amendment Nos. ~J14 aIltJ 214-..

-'FS 5.3-"3 03"=02=94

/

b. The mo erator temperature coefficient in the power operating range is less than or e ual to the limits specified in the CORE OPERATING LIMITS e maximum upper limit for the moderator temperature coeffici t shall be:
1) t less than 50% of RATED POWER, or
2) 50% of RATED
c. Capable of made subcritical accordance with Specification 3.l2.A.3.C.

B. Reactor Coolant System

1. The design of the Reactor Coola mplies with the code requirements specified in Section 4 of the UFSAR
2. structures of the Reactor Coolant System are designed to Class 1 seismi requ' rements, and have been designed to withstand:
a. Primary operating stres es combined wit the Operational seismic stresses resulting from a hori ontal ground accelera 'on of 0.07g and a simultaneous vertical ground ac leration of 2/3 the horizon 1, with the stresses maintained
b. Primary ope ting stresses when combined with th Design Basis Earthquake seismic str sses resulting from a horizontal ground a eleration of 0.15g and a ous vertical ground

~~

' j'-V Amendment Nes. 189 ~

.:rS 5.3 4 ,

3-17-72 the stresses such that the function of the component or sys m shall not e impaired as to prevent a safe and orderly shutdown of the unit.

3. System, at rated operating

TS~

2 2§ 8+

5/3 FUEL STORAGE (f~icability~~-----~~" - __ ",,--_,~ __--~

Applies to the design of the new and spent fuel storage areas.

I f

Objective

/'

" spects of fuel storage relating t revention of criticality in fuel storage areas; to the reactor; and to prevention of inadvertent draining

/

. A. The reinforced co Basis Earthqua t.,;-

he spent fuel pit has a stainless eel liner to ensure against loss of water.

B. The new and spent fuel storage ~ are designed so that it is impossible to insert

...._ assemblies in other than the prescribed locations. ew uel is store "6rtieaUy in an array with a distance of 21 inches between assemblies to assure keff ~ 0.98 with fuel of the

.[ highest anticipated enrichment in place assuming optimum moderation.

  • Spent fuel is stored v6ftietlll~in an array with a distance of 14 inches between

\

\ '-.

  • E.G., an aqueous foam envelopment as the result of fire fighting.

.~~,,"-....-------,,/ ----- j' ,~~

Amendment Nos. 66 & 65

.__ /~~L\ '.~ ----v~.~/ TS ~

)'0 ~ ( J'\"t-£.

<<2 \..~,.. K' / {)6-19-98

"~~(v~~blies

, ) I.-

to ensure ""If ~ 0,95, even if unborated water were used to fill the s;ent fuel

\ \~~r-//; / / storage pit. The spent fuel pool is divided into a two-region storage pool. Region 1 comprises the first three rows of fuel racks (324 storage locations) adjacent to the Fuel Building Trolley Load Block. Region 2 comprises the remainder of the fuel racks in the fuel pool. During spent fuel cask handling, Region I is limited to storage of spent fuel assemblies which have decayed at least 150 days after discharge and shall be restricted to those assemblies in the "acceptable" domain of Figure 5 A-I. Administrative controls with written procedures will be employed in the selection and placement of these assemblies.

The enrichment of the fuel stored in the spent fuel racks shall not exceed 4.3 weight percent of U-235.

1. Whenever there is spent fuel in the spent fuel pit, the pit shall be filled with borated water at a boron concentration not less than 2300tp'pm to match that used in the reactor cavity ./

. and refueling canal during refueling operations. ._ _______'

.'--..-_ft

~ e- ~ --..:::::::=>- ....../~

/ D. The only drain which can be connected to the spent fuel storage area is that in the reactor y-step procedures us~nng refueling ensure that the -valve er tube which connec e spent fuel storage area with t reacto~ity

(\ is closed efore draining of the avity commences. In addition, e procedures require

\ placi g the bolted blank flange on the fuel transfer tube as soon as the reactor cavity is

~--------"-~='

0r>eIMifl:g CYble 1..0.

FSAR SeetioR 9.5 FaelPit Cooling S¥~tem FSAIt Section 9.12 Fueillandling ~ystem Amendment Nos. 214 !tREl214.

INSERT 2 5.3.1 Criticality 5.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 4.3 weight percent;
b. keff ~ 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Appendix 9A of the UFSAR; and
c. A nominal 14 inch center to center distance between fuel assemblies placed in the storage racks.

5.3.1.2 The new fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 4.3 weight percent;
b. keff ~ 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties calculated in accordance with the methodology described in Virginia Electric and Power Company letter dated November 5, 1997 (Serial No.97-614);
c. keff ~ 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties calculated in accordance with the methodology described in Virginia Electric and Power Company letter dated November 5, 1997 (Serial No.97-614); and
d. A nominal 21 inch center to center distance between fuel assemblies placed in the storage racks.

5.3.1.3 The spent fuel pool is divided into a two-region storage pool.

Region 1 comprises the first three rows of fuel racks (324 storage locations) adjacent to the Fuel Building Trolley Load Block. Region 2 comprises the remainder of the fuel racks in the fuel pool. During spent fuel cask handling, Region 1 is limited to storage of spent fuel assemblies which have decayed at least 150 days after discharge and shall be restricted to those assemblies in the "acceptable" domain of Figure 5.3-1. Administrative controls with written procedures will be employed in the selection and placement of these assemblies.

5.3.2 Boron Concentration Whenever there is spent fuel in the spent fuel storage pool, the pool shall be filled with borated water at a boron concentration not less than 2300 ppm to match that used in the reactor cavity and refueling canal during refueling operations.

INSERT 3 5.3.3 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 41 feet, 2 inches mean sea level, USGS datum.

INSERT 4 5.3.4 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1044 fuel assemblies.

3 TS Figure 5 t-l O~ 1998 35000 ~---,..---------r---'-----;-----r-----r----r--r-,

225~' ppm ~.(oron in P 01 Water 30000 I

I I

25000 I

.K<O*r

- ~ccept ble j Regi n:

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20000 I

I RegiOn

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C

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en

c>- 15000 E

Q) l/I ,  ;

l/I unaccrPtabl~

Re,ion I 10000 J 5000 o

(")

a 1.9 2.1 2.3 2.5 2.7 2.9 3.1 3.3 3.5 3.7 3.9 4.1 4.3 (f)

I-Initial Fuel En . ent (w/o U-235)

~I MINIMUM FUEL EXPOSUR Fi ure 5J-l ITIAL ENRICHMENT TO PREVENT CRITICALITY IN DAMAGED RACKS Amendment Nos. 214 and 214

TS 6.2-1 89-1S-t):;

6.2 GENERAL NOTIFICATION AND REPORTING REQUIREMENTS Specification A. The following action shall be taken for Reportable Events:

A report shall be submitted pursuant to the requirements of Section 50.73 to 10 CFR.

B. Immediate notifications shall be made in accordance with Section 50.72 to 10 CFR.

C. CORE OPERATING LIMITS REPORT Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle.

Parameter limits for the following Technical Specifications are defined in the CORE OPERATING LIMITS REPORT:

1. TS 3.1.~MOderator Temperature Coefficient
2. TS 3.12.A.2 and TS 3.12.A.3 - Control Bank Insertion Limits
3. TS 3.12.B.1 and TS 3.12.B.2 - Power Distribution Limits Amendment Nos. 144 and 243

TS 6.4-3 09 15-05 '

2. The requirements of 6.4.B.l above, shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem/hr, but less than 500 rads/hr at one meter from a radiation source or any surface through which radiation penetrates. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Manager on duty and/or the senior station p individual assigned the responsibility for health physics and radiation protection.
3. Written procedures shall be established, implemented, and maintained covering the activities referenced below:
a. Process Control Program implementation.
b. Offsite Dose Calculation Manual implementation.

Amendment Nos. ~44 and 243

Serial No. 08-0168 Docket Nos. 50-280 and 281 ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATIONS PAGES (TYPED)

Virginia Electric and Power Company (Dominion)

Surry Power Station Units 1 and 2

TS iii TECHNICAL SPECIFICAnON TABLE OF CONTENTS SECTION TITLE PAGE 4.15 AUGMENTED INSERVICE INSPECTION PROGRAM FOR HIGH TS 4.15-1 ENERGY LINES OUTSIDE OF CONTAINMENT 4.16 LEAKAGE TESTING OF MISCELLANEOUS RADIOACTIVE TS 4.16-1 MATERIALS SOURCES 4.17 SHOCK SUPPRESSORS (SNUBBERS) TS 4.17-1 4.18 DELETED 4.19 STEAM GENERATOR (SG) TUBE INTEGRITY TS 4.19-1 4.20 CONTROL ROOM AIR FILTRATION SYSTEM TS 4.20-1 5.0 DESIGN FEATURES TS 5.0-1 5.1 SITE LOCATION TS 5.0-1 5.2 REACTOR CORE TS 5.0-1 5.3 FUEL STORAGE TS 5.0-2 6.0 ADMINISTRATIVE CONTROLS TS 6.1-1 6.1 ORGANIZATION, SAFETY AND OPERATION REVIEW TS 6.1-1 6.2 GENERAL NOTIFICATION AND REPORTING REQUIREMENTS TS 6.2-1 6.3 ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED TS 6.3-1 6.4 UNIT OPERATING PROCEDURES AND PROGRAMS TS 6.4-1 6.5 STATION OPERATING RECORDS TS 6.5-1 6.6 STATION REPORTING REQUIREMENTS TS 6.6-1 6.7 ENVIRONMENTAL QUALIFICATIONS TS 6.7-1 6.8 PROCESS CONTROL PROGRAM AND OFFSITE DOSE TS 6.8-1 CALCULATION MANUAL Amendment Nos.

TS 5.0-1 5.0 DESIGN FEATURES 5.1 SITE LOCATION The Surry Power Station is located in Surry County, Virginia, on property owned by Virginia Electric and Power Company on a point of land called Gravel Neck which juts into the James River. It is approximately 46 miles SE of Richmond, Virginia, 17 miles NW of Newport News, Virginia, and 25 miles NW of Norfolk, Virginia.

5.2 REACTOR CORE 5.2.1 Fuel Assemblies The reactor shall contain 157 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy or ZIRLO fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U0 2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with 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 non-limiting core locations.

5.2.2 Control Rod Assemblies The reactor core shall contain 48 control rod assemblies. The control material shall be silver indium cadmium, as approved by the NRC.

Amendment Nos.

TS 5.0-2 5.3 FUEL STORAGE 5.3.1 Criticality 5.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 4.3 weight percent;
b. keff ~ 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Appendix 9A of the UFSAR; and
c. A nominal 14 inch center to center distance between fuel assemblies placed in the storage racks.

5.3.1.2 The new fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 4.3 weight percent;
b. keff ~ 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties calculated in accordance with the methodology described in Virginia Electric and Power Company letter dated November 5, 1997 (Serial No.97-614);
c. keff $. 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties calculated in accordance with the methodology described in Virginia Electric and Power Company letter dated November 5, 1997 (Serial No.97-614); and
d. A nominal 21 inch center to center distance between fuel assemblies placed in the storage racks.

5.3.1.3 The spent fuel pool is divided into a two-region storage pool. Region 1 comprises the first three rows of fuel racks (324 storage locations) adjacent to the Fuel Building Trolley Load Block. Region 2 comprises the remainder of the fuel racks in the fuel pool. During spent fuel cask handling, Region 1 is limited to storage of spent fuel assemblies which have decayed at least 150 days after discharge and shall be restricted to those assemblies in the "acceptable" domain of Figure 5.3-1. Administrative controls with written procedures will be employed in the selection and placement of these assemblies.

TS 5.0-3 5.3 FUEL STORAGE (CONTINUED) 5.3.2 Boron Concentration Whenever there is spent fuel in the spent fuel storage pool, the pool shall be filled with borated water at a boron concentration not less than 2300 ppm to match that used in the reactor cavity and refueling canal during refueling operations.

5.3.3 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 41 feet, 2 inches mean sea level, USGS datum.

5.3.4 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1044 fuel assemblies.

TS Figure 5.3-1 35000 22501 ppm ~.(oron in PT'Wath 30000 25000

. K<O.95

. Regi~n: I 5' Accept~ble I-

!: I Region C 20000

~Co

s

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al

c 15000 E

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Re$ion I 10000 +__ l 5000 OIL-------~--'------'----'-----'-----'-~____'____'"~ _ _'____ _____'

1.9 2.1 2.3 2.5 2.7 2.9 3.1 3.3 3.5 3.7 3.9 4.1 4.3 Initial Fuel Enrichment (w/o U-235)

Figure 5.3-1 MINIMUM FUEL EXPOSURE VERSUS INITIAL ENRICHMENT TO PREVENT CRITICALITY IN DAMAGED RACKS Amendment Nos.

TS 6.2-1 6.2 GENERAL NOTIFICATION AND REPORTING REQUIREMENTS Specification A. The following action shall be taken for Reportable Events:

A report shall be submitted pursuant to the requirements of Section 50.73 to 10 CFR.

B. Immediate notifications shall be made in accordance with Section 50.72 to 10 CFR.

C. CORE OPERATING LIMITS REPORT Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle.

Parameter limits for the following Technical Specifications are defined in the CORE OPERATING LIMITS REPORT:

1. TS 3.I.E - Moderator Temperature Coefficient
2. TS 3.I2.A.2 and TS 3.I2.A.3 - Control Bank Insertion Limits
3. TS 3.l2.B.l and TS 3.12.B.2 - Power Distribution Limits Amendment Nos.

TS 6.4-3

2. The requirements of 6.4.B.I above, shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem/hr, but less than 500 rads/hr at one meter from a radiation source or any surface through which radiation penetrates. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Manager on duty and/or the senior station individual assigned the responsibility for health physics and radiation protection.
3. Written procedures shall be established, implemented, and maintained covering the activities referenced below:
a. Process Control Program implementation.
b. Offsite Dose Calculation Manual implementation.

C. Deleted Amendment Nos.

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