Text

Proposed TS Change, Various Administrative Changes
	ML023650621
Person / Time
Site:	Surry
Issue date:	12/19/2002
From:	Hartz L; Virginia Electric & Power Co (VEPCO)
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	02-721
	Download: ML023650621 (49)
	v • d • e

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 December 19, 2002 U.S. Nuclear Regulatory Commission Serial No.02-721 Attention: Document Control Desk SPS-LICJTJN RO Washington, D.C. 20555 Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS I AND 2 PROPOSED TECHNICAL SPECIFICATIONS CHANGE VARIOUS ADMINISTRATIVE CHANGES 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 Licenses Numbers DPR-32 and DPR-37 for Surry Power Station Units 1 and 2, respectively. The proposed change makes administrative changes to reflect revisions in regulations, correct typographical and editorial errors made in previous TS revisions, and correct TS references to corresponding Updated Final Safety Analysis Report (UFSAR) sections. A discussion of the proposed TS change is provided in . The marked-up and proposed TS pages reflecting the proposed change are provided in Attachments 2 and 3, respectively.

We have evaluated the proposed TS change and have determined that it does not involve a significant hazards consideration as defined in 10 CFR 50.92. The basis for this determination is provided in Attachment 1. We have also determined that 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 will occur. Therefore, the proposed amendment is eligible for categorical exclusion 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 Station Nuclear Safety and Operating Committee, and by the Management Safety Review Committee.

Ifyou have any questions or require additional information, please contact us.

Very truly yours, Lesli at Vice President- Nuclear Engineering A00

Attachments:

1. Discussion of Change

2. Mark-up of Technical Specifications Change

3. Proposed Technical Specifications Change Commitment made in this letter: None cc: U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center Suite 23T85 61 Forsyth Street, SW Atlanta, Georgia 30303 Mr. R. A. Musser NRC Senior Resident Inspector Surry Power Station Commissioner Bureau of Radiological Health 1500 East Main Street Suite 240 Richmond, VA 23218

SN: 02-721 Docket Nos.: 50-280/281

Subject:

Proposed TS Change - Various Admin. Changes 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 Engineering, 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 19th day of December, 2002.

My Commission Expires: March 31, 2004.

.)*Notary Public

Attachment I Discussion of Change Surry Power Station Units I and 2 Virginia Electric and Power Company (Dominion)

DISCUSSION OF CHANGE Introduction Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Dominion) requests a change to the Technical Specifications (TS) for Surry Power Station Units 1 and 2. The proposed change makes administrative changes to reflect revisions in regulations, correct typographical and editorial errors made in previous TS revisions, and correct TS references to corresponding Updated Final Safety Analysis Report (UFSAR) sections.

Conforming changes are also being made to the Bases as appropriate and are included for information.

The proposed 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 change.

Background

A revision to the Surry Power Station Unit 1 and 2 Technical Specifications is being proposed to address necessary administrative changes. The detailed descriptions of changes by TS item are listed in the next section, Description of Proposed Change. The proposed TS change has been divided into the below three groups of specific changes based on their commonality.

1. Administrative changes to the TS and Bases associated with revisions to the Code of Federal Regulations (Technical Specification sections 4.0.5.a, TS 4.0.3 Basis, 6.1 .C.1 .f.1 .b. 6.1 .C.1 .o.2. 6.1 .C.2.0.. 6.1 .C.2.g.2. 6.1 .C.2.g.3. 6.4.C)

Periodic changes to the Code of Federal Regulations often result in process and terminology changes that require revisions to the TS. This is because the specific wording of certain regulations is often reflected in the TS as means of implementing regulatory requirements at the station. Therefore, to maintain consistency with the wording contained in the revised regulations, certain TS that are associated with the revised regulations also require revision.

Regulations 10 CFR 50.55a, "Codes and standards," includes inservice inspection (IS) and inservice testing (IST) requirements; 10 CFR 50.59, "Changes, tests and experiments," addresses regulatory review requirements; and 10 CFR 50.73, "Licensee event report system," includes reportability requirements. These regulations have been revised thus necessitating changes to the TS to maintain consistency with the regulations. Specifically, the below changes are proposed.

a) The TS currently states that ISI and IST requirements shall be performed in Page 1 of 6

accordance with 10 CFR 50.55a(g). However, 10 CFR 50.55a was previously revised to separate IST and ISI requirements into 10 CFR 50.55a(f) and 10 CFR 50.55a(g), respectively. Consequently, the TS require revision to reflect the appropriate regulation citations.

b) Several administrative TS currently discuss reviews related to changes performed in accordance with 10 CFR 50.59. However, 10 CFR 50.59 has been revised to include different terminology and requirements than previously used in the regulation and the associated TS for performing changes, tests and experiments at the station without obtaining NRC review (e.g., the term "unreviewed safety question" is no longer used.) The TS terminology requires revision to reflect the revised 10 CFR 50.59 wording.

c) A TS Basis currently discusses event reportability in accordance with 10 CFR 50.73 requirements. However, 10 CFR 50.73 has been revised to include reportability exceptions. The TS Basis requires revision to clarify that 10 CFR 50.73 reportability requirements include exceptions.

2. Correction of editorial errors included in previous TS amendments and Bases changes (Technical Specification sections 3.1.A.1.d.1(b), 3.1.A.1.d.2, 3.10 Basis, 3.17.4.b.1. 3.17.4.c.3, 3.17.5.a.2. 3.17.5.c.2, 3.17.5.d.1, Table 4.1-1 Item 32.a. Table 4.1-2A Item 9. 6.4.L, 6.6.A.2. 6.8.A.1 and 6.8.B.1)

During the preparation and implementation of previous TS revisions, editorial errors occasionally occur that require subsequent correction. Examples include TS numbering errors, obsolete references to deleted TS sections, etc. Specifically, the below changes are proposed.

a) Inaccurate references to previously renumbered sections, sections relocated to a different TS page, and sections relocated to the VEPCO Operational Quality Assurance Program Topical Report require correction.

b) A test frequency reference included in the "Remark" section of a TS table requires revision from "monthly" to "quarterly." The test frequency change was approved in a previous TS amendment but was not revised in the 'Remark" section of the associated TS table as required.

c) A fire protection TS surveillance relocated, in a previous TS amendment, to the Updated Final Safety Analysis Report in accordance with Generic Letters 86-10 and 88-12 was erroneously not deleted from the TS. This previously relocated TS surveillance requires deletion.

d) An item numbering mistake due to a typographical error introduced in a previous TS amendment requires correction.

Page 2 of 6

3. Correction of UFSAR references included in the TS (Technical Specification Bases 3.8 and 4.4)

Revision of the UFSAR can impact the TS in that the TS Bases include references to associated UFSAR sections. As the UFSAR is revised over time, section numbering can change and render an associated TS reference obsolete. Therefore, the TS Bases require revision to correct obsolete references to UFSAR sections.

Description of Proposed Change All changes proposed are applicable to the Technical Specifications and Bases of Unit 1 and Unit 2 Operating Licenses (OLs) DPR-32 and DPR-37, respectively. The specific changes proposed are as follows:

" TS 3.1.A.1.d.1(b), 3.1.A.1.d.2, TS 3.10 Basis page 3.10-6a, and TS 3.17.4.b.1, 3.17.4.c.3, 3.17.5.a.2, 3.17.5.c.2, and 3.17.5.d.1 are being revised to correct references to TS 3.10.A items that were renumbered in TS Amendments 230/230 dated March 8, 2002.

"* On TS page 3.8-5, the TS 3.8 Basis reference to UFSAR section 4.3.2, Reactor Coolant Pump, is being revised to reference UFSAR section 4.2.2.4.

"* On TS page 3.8-5, the TS 3.8 Basis reference to UFSAR section 5.5.2, Isolation Design, is being revised to reference UFSAR section 5.2.2.

"* On TS page 3.8-5, the TS 3.8 Basis reference to UFSAR section 6.3.2, Containment Vacuum System, is being revised to reference UFSAR section 5.3.4.

"* TS 4.0.5.a is being revised from a reference to 10 CFR 50.55a(g) for ISI and IST, to reference 10 CFR 50.55a(f) for IST and 10 CFR 50.55a(g) for ISI.

"* On page TS 4.0-5, the TS 4.0.3 Basis is being revised to clarify that an event may not be reportable in accordance with 10 CFR 50.73 exceptions to reportability.

"* TS Table 4.1-1, item 32.a uRemark7 is being revised to reflect the accurate surveillance frequency of "quarterly" rather than "monthly". This error was introduced when the related surveillance frequency was previously changed in TS Amendments 228/228 dated August 31, 2001.

" The fire protection (FP) pump surveillance included in Table 4.1-2A, item 9 should have been deleted as part of TS Amendments 217/217 dated December 16, 1998.

FP requirements were previously relocated to the UFSAR and subsequently to the Technical Requirements Manual.

"* On TS page 4.4-2, the TS 4.4 Basis reference to UFSAR section 5.4 is being revised to UFSAR section 5.5, Containment Tests and Inspections.

"* Due to a previous revision of 10 CFR 50.59, the regulatory terminology included in Page 3 of 6

TS 6.1.C.1.f.1.b, 6.1.C.1.g.2, 6.1.C.2.g.1, 6.1.C.2.g.2, 6.1.C.2.g.3 and 6.4 C is being revised for consistency with the regulation.

" TS 6.4.L is being revised to correct an item numbering error. This error (item numbers 1, 2 and 3 became item numbers 1, 2 and 2 as the result of typographical error) was introduced in TS Amendments 229/229 dated December 18, 2001.

" The Note on TS page 6.6-2 states that footnotes are located on TS page 6.6-12.

However, the footnotes were previously relocated from TS page 6.6-12 to TS page 6.6-11 by TS Amendments 208/208 dated April 18, 1996. The Note on TS page 6.6 2 is being revised to reference the correct page.

" TS 6.8.A.1 and 6.8.B.1 reference TS 6.5.B.12, which has been previously relocated to the VEPCO Operational Quality Assurance Program, UFSAR Chapter 17 by TS Amendments 211/211 dated July 15, 1997. TS 6.8.A.1 and 6.8.B.1 are being revised to correct this reference error.

Safety Implications of the Proposed Change The proposed change to the Surry Technical Specifications is administrative in nature and makes editorial changes to reflect changes in regulations, correct errors from previous TS revisions, and revise TS references to UFSAR sections. The proposed change does not alter the operation of the station in any way, nor are any plant modifications being proposed. Furthermore, the current Surry licensing and design bases are not being changed, nor is the margin of safety assumed in the plant accident analyses being affected. Consequently, there is no safety significance associated with the proposed administrative change.

Evaluation of Significant Hazards Consideration Dominion has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed administrative change to the Surry Power Station Units 1 and 2 Technical Specifications (TS) and Bases. The proposed change to the Surry TS makes administrative revisions to reflect changes in regulations, corrects editorial and typographical errors from previous TS revisions, and revises TS cross-references to Updated Final Safety Analysis Report (UFSAR) sections. Due to the strictly administrative nature of the proposed TS change, we have determined that a significant hazards consideration does not exist. The basis for this determination is provided as follows:

1. Does the proposed license amendment involve a siqnificant increase in the probability or consequences of an accident Previously evaluated?

The proposed change is administrative in nature and as such does not impact the condition or performance of any plant structure, system or component. The proposed administrative change does not affect the initiators of any previously analyzed event nor Page 4 of 6

the assumed mitigation of accident or transient events. As a result, the proposed change to the Surry Technical Specifications does not involve any increase in the probability or the consequences of any accident or malfunction of equipment important to safety previously evaluated since neither accident probabilities or consequences are being affected by this proposed administrative change.

2. Does the proposed license amendment create the Possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change is administrative in nature, and therefore does not involve any changes in station operation or physical modifications to the plant. In addition, no changes are being made in the methods used to respond to plant transients that have been previously analyzed. No changes are being made to plant parameters within which the plant is normally operated or in the setpoints, which initiate protective or mitigative actions, and no new failure modes are being introduced. Therefore, the proposed administrative change to the Surry Technical Specifications does not create the possibility of a new or different kind of accident or malfunction of equipment important to safety from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

The proposed change is administrative in nature, and does not impact station operation or any plant structure, system or component that is relied upon for accident mitigation.

Furthermore, the margin of safety assumed in the plant safety analysis is not affected in any way by the administrative "cleanup" of the Surry Technical Specifications.

Therefore, the proposed administrative change to the Surry Technical Specifications does not involve a reduction in a margin of safety.

Environmental Assessment This amendment request meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) as follows:

(i) The amendment involves no significant hazards consideration.

As described above, the proposed administrative TS change does not involve a 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 administrative TS change 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. Plant operation is not affected in any manner by this proposed administrative change. Therefore, there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

Page 5 of 6

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

The proposed administrative TS change does not involve plant physical changes or changes in the method of plant operation. Therefore, there is no significant increase in individual or cumulative occupational radiation exposure.

Based on the above assessment, Dominion concludes that the proposed change meets the criteria specified in 10 CFR 51.22 for a categorical exclusion from the requirements of 10 CFR 51.22 relative to requiring a specific environmental assessment or impact statement by the Commission.

Conclusion The proposed TS change is administrative in nature, and merely "cleans up" previous editorial and typographical errors and updates the TS to reflect current regulation terminology. Neither station design nor operation is being affected. The Station Nuclear Safety and Operating Committee (SNSOC) and the Management Safety Review Committee (MSRC) have reviewed the proposed change and have concluded that this change does not involve a significant hazards consideration and will not endanger the health and safety of the public.

References

"* UFSAR Section 4.2.2.4

"* UFSAR Section 5.2.2

"* UFSAR Section 5.3.4

"* UFSAR Section 5.4

"* UFSAR Section 5.5

"* UFSAR Section 17.2

"* TS Amendments 208/208

"* TS Amendments 211/211

"* TS Amendments 217/217

"* TS Amendments 228/228

"* TS Amendments 229/229

"* TS Amendments 230/230 Page 6 of 6

Attachment 2 Mark-up of Technical Specifications Change Surry Power Station Units I and 2 Virginia Electric and Power Company (Dominion)

TS 3.1-2 01-9~-5

b. If an unscheduled loss of one or more reactor coolant pumps occurs while operating below 10% RATED POWER (P-7) and results in less than two pumps in service, the affected plant shall be shutdown and the reactor made subcritical by inserting all control banks into the core. The shutdown rods may remain withdrawn.

c. When the average reactor coolant loop temperature is greater than 350*F, the following conditions shall be met:

1. At least two reactor coolant loops shall be OPERABLE.

2. At least one reactor coolant loop shall be in operation.

d. When the average reactor coolant loop temperature is less than or equal to 350 0 F, the following conditions shall be met:

1. A minimum of two non-isolated loops, consisting of any combination of reactor coolant loops or residual heat removal loops, shall be OPERABLE, except as specified below:

(a) One RHR loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other RHR loop is OPERABLE and in operation.

(b) During REFUELING OPERATIONS the residual heat removal loop may be removed from operation as scified in TS A .'.

I J [I* 2. At least one reactor coolant loop or one residual heat removal loop shall be i=3-TF Iin operation, except as specified in Specification DIP 4il a6 cc C.,S AmnAndfrnt Nor-,_. 201 a.-d 204

"TS 3.8-5 If the containment air partial pressure rises to a point above the allowable value the reactor shall be brought to the HOT SHUTDOWN condition. If a LOCA occurs at the time the containment air partial pressure is at the maximum allowable value, the maximum containment pressure will be less than design pressure (45 psig), the containment will depressurize to 0.5 psig within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and less than 0.0 psig within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The radiological consequences analysis demonstrates acceptable results provided the containment pressure does not exceed 0.5 psig for the interval from 1 to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following the Design Basis Accident.

If the containment air partial pressure cannot be maintained greater than or equal to 9.0 psia, the reactor shall be brought to the HOT SHUTDOWN condition. The shell and dome plate liner of the containment are capable of withstanding an internal pressure as low as 3 psia, and the bottom mat liner is capable of withstanding an internal pressure as low as 8 psia.

References UFSAR Section Reactor Coolant Pump UFSAR Section 5.2 Containment Isolation 1UFSAR Section 5.2.1 Bases UFSAR Section iation Design UFSAR Section . Containment Vacuum System Ame~ndment Nee. 230 and 230

TS 3.10-6a

,,e-00&-02 During refueling, the reactor refueling water cavity is filled with approximately 220,000 gal of water b ted to at east 2,300 ppm boron. The boron concentration of this water, established by Specin-catn-3O6..., is sufficient to maintain the reactor subcritical by at least 5% Ak/k in the COLD SHUTDOWN condition with all control rod assemblies inserted. This includes a 1% Ak/k and a 50 ppm boron concentration allowance for uncertainty. This concentration is also sufficient to maintain the core subcritical with no control rod assemblies inserted into the reactor. Checks are performed during the reload design and safety analysis process to ensure the K-effective is equal to or less than 0.95 for each core. Periodi .,cheks nf refueling water boron concentration assure the proper shutdown margin. Mon. allows the Control Room Operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.

In addition to the above safeguards, interlocks are used during refueling to assure safe handling of the fuel assemblies. An excess weight interlock is provided on the lifting hoist to prevent movement of more than one fuel assembly at a time. The spent fuel transfer mechanism can accommodate only one fuel assembly at a time.

IRA!E SIi oil~

Amendmen Nes. 230 and 230

"TS 3.17-2

b. Before opening the hot leg loop stop valve.

1) The boron concentration of the isolated loop shall be greater than or equal to the boron concentration corresponding to the sh argin requirements of Specification 1.0.C.2 or a sappicaer the J active volume of the Reactor Coolant System. Verification of this condition shall be completed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to opening the hot leg 4 stop valve in the isolated loop.

c. Before opening the cold leg loop stop valve.

1) The hot leg loop stop valve shall be open with relief line flow established for at least 90 minutes at greater than or equal to 125 gpm.

2) The cold leg temperature of the isolated loop shall be at least 70°F and within 20°F of the highest cold leg temperature of the active loops.

Verification of this condition shall be completed within 30 minutes prior to opening the cold leg stop valve in the isolated loop.

3) The boron concentration of the isolated loop shall be greater than or equal to the boron concenrti onding to the shutdown margin ill . requirements of Specification 1.0.C.2 or . , as applicable for the N Is5 Spciiato .21 S*-active volume of the Reactor Coolant System. Verification of this E.§ condition shall be completed after relief line flow for at least 90 minutes at greater than or equal to 125 gpm and within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to opening the cold leg stop valve in the isolated loop.

LO

5. Whenever an isolated and drained reactor coolant loop is filled from the active

[rr-!n~i*_ volume of the RCS, the following conditions shall apply:

I .*a. Seal injection may be initiated to the reactor coolant pump in the isolated loop a'. provided that:

1) The isolated loop is drained. Verification of this condition shall be completed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to initiating seal injection.

Am-enzdment Nos. 22-6 and 226

TS 3.17-3

2) The boron concentration of the source for reactor coolant pump seal injection shall be greater than or equal to the boron concentration corresponding to the shut n requirements of Specification 1.O.C.2 or *-(1k9-, asapplicame-for the active volume of the Reactor Coolant System. If using the Volume Control Tank (VCT) as the source for reactor coolant pump seal injection, verification of the boron concentration shall be completed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to initiating seal injection and every hour thereafter during the loop backfill evolution.

b. The cold leg loop stop valve may be energized and/or opened to backfill the loop from the active volume of the Reactor Coolant System provided that:

1) The isolated loop is drained or reactor coolant pump seal injection has been initiated in accordance with Specification 3.17.5.a above.

Verification of the loop being drained shall be completed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to partially opening the cold leg stop valve in the isolated loop.

2) The Reactor Coolant System level is at least 18 ft.

3) A source range nuclear instrumentation channel is OPERABLE with audible indication in the control room.

c. Backfilling of the isolated loop may continue provided that:

1) The Reactor Coolant System level is maintained at or above 18 ft. If Reactor Coolant System level is not maintained at or above 18 ft. the loop stop valve shall be closed.

2) The boron concentration of the reactor coolant pump seal injection source is greater than or equal to the boron concert sponding to the shutdown margin requirements of Specification 1.T.C. or.... as applicable for the active volume of the Reactor Coolant System. If the boron concentration is not maintained greater than or equal to the required boron concentration noted above, the loop stop valve on the loop being backfilled shall be closed and either drain the loop or apply Specification 3.17.4.

A.-Amndrent Nes. 22m id 226

"TS 3.17-4 05-22-01

3) A source range nuclear instrumentation channel is OPERABLE and continuously monitored with audible indication in the control room during the backfill evolution. Should the count rate increase by more than a factor of two over the initial count rate, the cold leg loop stop valve shall be closed and no attempt made to open the cold leg stop valve until the reason for the count rate increase has been determined.

d. When the isolated loop is full, the cold leg loop stop valve can be fully opened and the hot leg loop stop valve opened provided that:

1) The boron concentration of the isolated loop is greater than or equal to the boron concentration correspon to the shutd margin requirements of Specification 1.0.C.2 or oer the active volume of the Reactor Coolant System. If the VCT was used as the source for reactor coolant pump seal injection, this condition shall be verified within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to fully opening the loop stop valves. If the boron concentration in the isolated loop does not meet the condition above, close the loop stop Svalve and either drain the loop or apply Specification 3.17.4.

i 2) The hot and cold leg loop stop valves are opened within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the isolated loop is filled. If the loop stop valves are not fully open within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , close the loop stop valves and either drain the loop or apply Specification 3.17.4.

Basis The Reactor Coolant System may be operated with isolated loops in COLD SHUTDOWN or REFUELING SHUTDOWN in order to perform maintenance. A loop stop valve in any loop can be closed for up to two hours without restriction for testing or maintenance in these operating conditions. While operating with a loop isolated, AC power is removed from the loop stop valves and their breakers locked opened to prevent inadvertent opening. When the isolated loop is returned to service, the coolant in the isolated loop Amen -AndmentNse. 226 ind 22

TS 4.0-2 e-+12-93 4.0.5 Surveillance requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:

a. Inservice inspection of ASME Code Class 1, 2, and 3 components adisepie tesfing ef ASME Code CGlas 1, 2, and 3 pumps a 'av"cshall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code I

and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).

b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:

T 0

Lr / 13 f f.ir2 Ay/I *.1 Pg %v5 IY-w a!ils M U1 I zz-ýi 6r'!

% 1 I!ij-. YLS -441) to V a

r-r -

/~~~r*.

L'1]) )

t VJ-VVI A1A.Y I

L1 VV 'I IVVj-A I Ce gi] d K

HE*

vi i4-:- i)-¶ -it P3e~

V CfvL 9 O, ruIlo Stendment Ncs. 175 and 174N

TS 4.0-5 is identified that a surveillance has not been performed and not at the time that the allowed surveillance interval was exceeded. Completion of the surveillance requirement within the allowable outage time limits of the Action Statement requirements restores compliance with the requirements of Specification 4.0.3.

However, this does not negate the fact that the failure to have performed the surveillance within the allowed surveillance interval, defined by the provisions of Specification 4.0.2, was a violation of the operability requirements of a Limiting Condition for Operation. Further, the failure to perform a surveillance within the provisions of Specification 4.0.2 is a violation of a Technical Specification requirement and is, therefore, a reportable event under the requirements of 10 CFR 50.73(a)(2)(i)(B )because it is a condition prohibited by the plant's I !s Technical Secifications. l If the allowable outage time limits of the Action Statement requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or a shutdown is required to comply with Action Statement requirements, e.g., Specification 3.0.1, a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance is provided to permit a delay in implementing the Action Statement requirements. This provides an adequate time limit to complete surveillance requirements that have not been performed. The purpose of this allowance is to permit the completion of a surveillance before a shutdown is required to comply with Action Statement requirements or before other remedial measures would be required that may Wii preclude completion of a surveillance. The basis for this allowance includes consideration for plant conditions, adequate planning, availability of personnel, the time required to perform the surveillance, and the safety significance of the delay in completing the required surveillance. This BEii Affcndi.nn Noc. 175 --- 1"1

TABLE 4.1-1 (Continued)

MINIMUM FREQUENCIES FOR CHECK, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS Channel Description Check Calibrate Test Remarks

32. Auxiliary Feedwater

a. Steam Generator Water Level Low-Low

b. RCP Undervoltage S

S R

R Q(1)

R(1)(2)

1) The auto start of the turbine driven pump is not included in the wt each startump.

test but is tested within 31 days prior to

Ž. - tf

1) The actuation logic and relays are tested within 31 days

,, $1 prior to each startup.

2) Setpoint verification not required.

c. S.I. (All Safety Injection surveillance requirements)

d. Station Blackout N.A. R N.A.

e. Main Feedwater Pump Trip N.A. N.A. R

33. Loss of Power

a. 4.16 KV Emergency Bus Undervoltage N.A. R Q(1) 1)Setpoint verification not required.

(Loss of Voltage)

b. 4.16 KV Emergency Bus Undervoltage N.A. R Q(1) 1)Setpoint verification not required.

(Degraded Voltage)

34. Deleted

35. Manual Reactor Trip N.A. N.A. R The test shall independently verify the operability of the undervoltage and shunt trip attachments for the manual reactor trip function. The test shall also verify the operability of the bypass breaker trip circuit.

36. Reactor Trip Bypass Breaker N.A. N.A. M(1), 1) Remote manual undervoltage trip immediately after R(2) placing the bypass breaker into service, but prior to commencing reactor trip system testing or required maintenance.

2) Automatic undervoltage trip.

S37. Safety Injection Input to RPS N.A. N.A. R S38. Reactor Coolant Pump Breaker Position Trip N.A. N.A. R 60

)0 IThis page was published electronically for useon the MTNO systm. Differences between this page and a page from the hardco. version w of the Technica it -1 e differeces ns i Iappaance onthly.

Iversion of the Suchdifferences Technical areintentional Specifcations andame has beenconfirmedthe byresult of managing Configuraton anelectronic.Imaster of the station's lechnicalt Secifications. Te accracy ofte content o the M Nanagement. IN I

TABLE 4.1-2A MINIMUM FREQUENCY FOR EQUIPMENT TESTS FSAR SECTION DESCRIPTION TEST FREQUENCY REFERENCE

1. Control Rod Assemblies Rod drop times of all full Prior to reactor criticality: 7 length rods at hot conditions a. For all rods following each removal of the reactor vessel head

b. For specially affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and

c. Once per 18 months

2. Control Rod Assemblies Partial movement of all rods Quarterly 7

3. Refueling Water Chemical Addition Functional Once per 18 months 6 Tank

4. Pressurizer Safety Valves Setpoint Per TS 4.0.5 4

5. Main Steam Safety Valves Setpoint Per TS 4.0.5 10

6. Containment Isolation Trip

Functional Once per 18 months 5

7. Refueling System Interlock,,s

Functional Prior to refueling 9.12

8. Service Water Systems (::i. ,)

Functional Once per 18 months 9.9

9. !Firn Prcno:tw&n Purnp =ndy,'z:"Suppy S"n-,zrnIM4nnt th y 4-44_

10. Primary System Leakage

Evaluate Daily 4

11. Diesel Fuel Supply

Fuel Inventory 5 days/week 8.5

12. Deleted

13. Main Steam Line Trip Valves Functional Before each startup (TS 4.7) 10 (Full Closure) The provisions of Specification 4.0.4.

are not applicable

)

I hspage was ublished electronically for ueon theMINDsystem. Dfeecsbtenti page an.pgfrmteh ~ esnofheecialpcfctom reifrnc II of thestation's Technncal3pecf°cat"ons. Theaccuracy the content ofthtMNnD I pag ony. Such differencesarer itentionat an are, the result of managing aparancv an electronc master aron Ofthe Technicat Specifications has beenconfirmed by Configuration Management. I I

TS 4.4-2

-696 The containment is designed for a maximum pressure of 45 psig. The containment is maintained at a subatmospheric air partial pressure consistent with TS Figure 3.8-1 depending upon the cooldown capability of the Engineered Safeguards and will not rise above 45 psig for any postulated loss-of-coolant accident.

The initial test pressure for the Type A test is 47.0 psig to allow for containment expansion and equalization. A review was performed to determine the effects of pressurizing containment above its design pressure of 45.0 psig. This review was based on the original containment test at 52 psig. During that test, the calculated stresses were found to be well within the allowable yield strength of the structural reinforcing bars, therefore performance of the Type A test at 47 psig will have no detrimental effect on the containment structure.

All loss-of-coolant accident evaluations have been based on an integrated containment leakage rate not to exceed 0.1% of containment volume per 24 hr.

The above specification satisfies the conditions of 10 CFR 50.54(o) which stated that primary reactor containments shall meet the containment leakage test requirements set forth in Appendix J.

The limitations on closure and leak rate for the containment airlocks are required to meet the restrictions on containment integrity and containment leak rate. Surveillance testing of the airlock seals provides assurance that the overall airlock leakage will not become excessive due to seal damage during the intervals between airlock leakage tests.

ad~ References Section . - 1 E*.,l,,ti,. cf Containment efUFSAR Tests and Inspections -ef t;~

X UFSAR Sectiop, 7.5.1 Design Bases of Engineered Safeguards Instrumentation UFSAR Section 14.5 Loss of Coolant Accident 10 CFR 50 Appendix J "Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors" A nd

.... nt" o. 2-U a4ýJ .0

TS 6.1-7

--0649-9&

f. Responsibilities The SNSOC shall be responsible for:

1. Review of a) all new normal, abnormal, and emergency operating procedures and all new maintenance procedures, b) all procedure changes L* that require a evaluation, and c) any other procedures or changes thereto as determined by the Site Vice President which affect nuclear safety.

2. Review of all new test and experiment procedures that affect nuclear safety.

3. Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.

4. Review of proposed changes to Technical Specifications and shall submit recommended changes to the Site Vice President. ':h

5. Investigation of all violations of the Technical Specifications, including LeWthe preparation and forwarding of reports covering evaluation and Ft-X recommendations to prevent recurrence to the Vice President - Nuclear 11

iOperations and to the Management Safety Review Committee.

_ ,6. Review of all Reportable Events and special reports submitted to the NRC.

7. Review of facility operations to detect potential nuclear safety hazards.

8. Performance of special reviews, investigations or analyses and report Hc UMif thereon as requested by the Chairman of the SNSOC or Site Vice President.

Zw*iI A--A dmentNor.

.. 2t5 a-fnd 215

TS 6.1-8 9-.Dlet6-9d

9. Deleted.

10. Deleted.

11. Review of every unplanned onsite release of radioactive material to the environs exceeding the limits of Specification 3.11, including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vice President - Nuclear Operations and to the Management Safety Review Committee.

12. Review of changes to the Process Control Program and the Offsite Dose Calculation Manual.

13. Review of the Fire Protection Program and implementing procedures and shall submit recommended Program changes to the designated offsite management responsible for reviewing changes that pertain to Fire Protection.

g. Authority The SNSOC shall:

1. Provide written approval or disapproval of items considered under (1) through (3) above. SNSOC approval shall be certified in writing by either the Manager - Station Operations and Maintenance or the Manager Station Safety and Licensing.

2. Render determinations in writing with regard to whether or not each item considered under (1) through (5) abovefconst-,t-te -A-;;.

3. Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Vice President I -_CA*=C- Nuclear Operations and to the Management Safety. Review Committee of disagreement between SNSOC and the Site Vice President; however, the Site Vice President shall have responsibility for resolution of such

,g *,j-c disagreements pursuant to 6. LA above.

EI-* h. Records RICE02 The SNSOC shall maintain written minutes of each meeting and copies shall be p

,e* provided to the Vice President - Nuclear Operations and to the Management

-*_.Et Safety Review Committee.

Amoned-mznt Nezz. 21P -nd 21:7

TS 6.1-10

-*4-24-9J

e. Meeting Frequency The MSRC shall meet at least once per calendar quarter.

f. Quorum The minimum quorum of the MSRC necessary for the performance of the MSRC review and audit functions of these Technical Specifications shall consist of the Chairman or his designated alternate and at least 50% of the MSRC members including alternates. No more than a minority of the quorum shall have line responsibility for operation of the unit.

g. Review The MSRC shall be responsible for the review of:

as programmatically discussed in the Updated Final Safety I Analysis Report for 1) changes to procedures, equipment or systems and 2) tests or f f ' ti v4itsescompleted

experime p roof f t the provision sOf t h e s under m a nd 50.59, gr aSection 10 CFR, to assess the to ve rify t a ~ e r , e ; s <

i Ssdcagn *il*

or systems which

as eedin Section 50.59, 10 CFR.

" ""*'*3. Proposed tests or experiments . ........ .. un

....... ,ei,,*d*'f oy . ..... o as defined in Section 50.59, 10 CFR.

- ,4. Proposed changes to Technical Specifications or the Operating Licenses.

B.E.

-j AmnAmnt Nc 197 and 97

TS 6.4-3

-2+--95

2. The requirements of 6.4.B.1 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 Supervisor 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. All procedures described in 6.4.A and 6.4.B shall be reviewed and approved by the Station Nuclear Safety and Operating Committee (SNSOC) prior to implementation.

': r/ Subsequent procedure changes that require a za evaluation shall also be reviewed 1' 1 and approved by SNSOC prior to implementation. All other changes shall be M independently reviewed and approved as discussed in the Updated Final Safety S..Analysis Report.

F111 Ah11 ..11 Jment Nor. 197 and 1!7

TS 6.4-7 L. Iodine Monitoring The licensee shall implement a program which will ensure the capability to accurately determine the airborne iodine concentration in vital area under accident conditions.

This program shall include the following:

1. Training of personnel,

2. Procedures for monitoring, and

.- Provisions for maintenance of sampling and analysis equipment.

M. Deleted r11E r5V -..

M1111

.99 11111 Affien4awRt Nr.22 ýiJ 2

TS 6.6-2

-Gi-25-94 resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operations),

supplementary reports shall be submitted at least every 3 months until all three events have been completed.

2. Annual Reports'I T

a. A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man-rem exposure according to work and job 2

functions , e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements.

Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

Ij~j F Note: Footnotes 1 and 2 are located on page T'SctiI A-n--r*at N a. s"*185 a-d 15

TS 6.8-1

-96419-98 6.8 PROCESS CONTROL PROGRAM AND OFFSITE DOSE CALCULATION MANUAL A. Process Control Program (PCP)

Changes to the PCP:

1. Shall be documented and records of reviews performed shall be retained as dThis doclmentation shall contain

~a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and

b. A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

2. Shall require review and acceptance by the SNSOC and the approval of the Site Vice President prior to implementation.

B. Offsite Dose Calculation Manual (ODCM)

Changes to the ODCM:

1. Shall be documented and records of reviews performed shall be retained as

£jIIj required byft-+/-fti. .5. 1. This documentation shall contai

.r.

11 f41'qvifph. VS),4 Q` 'Cs C- of wn uu

a. Sufficient information to support the change together with the appropriate 7* Ianalyses or evaluations justifying the change(s) and Y:11 TOu~r aNJ ~

Am... -er.: Nos. 215 ftnd-2115-

Attachment 3 Proposed Technical Specifications Change Surry Power Station Units I and 2 Virginia Electric and Power Company (Dominion)

TABULATION OF CHANGES License No. DPR-32 / Docket No. 50-280 License No. DPR-37 / Docket No. 50-281 Summary of Changes:

The proposed change to the Surry Power Station Technical Specifications is being made to reflect revisions in regulations, correct typographical and editorial errors made in previous TS revisions, and correct TS references to corresponding Updated Final Safety Analysis Report (UFSAR) sections.

DELETE DATED SUBSTITUTE TS 3.1-2 09-01-95 TS 3.1-2 TS 3.8-5 03-08-02 TS 3.8-5 TS 3.10-6a 03-08-02 TS 3.10-6a TS 3.17-2 05-22-01 TS 3.17-2 TS 3.17-3 05-22-01 TS 3.17-3 TS 3.17-4 05-22-01 TS 3.17-4 TS 4.0-2 03-12-93 TS 4.0-2 TS 4.0-5 03-12-93 TS 4.0-5 TS 4.1-8a 08-31-01 TS 4.1-8a TS 4.1-9b 06-11-98 TS 4.1-9b TS 4.4-2 04-18-96 TS 4.4-2 TS 6.1-7 06-19-98 TS 6.1-7 TS 6.1-8 12-16-98 TS 6.1-8

-TS 6.1-10 04-21-95 TS 6.1-10 TS 6.4-3 04-21-95 TS 6.4-3 TS 6.4-7 12-18-01 TS 6.4-7 TS 6.6-2 01-25-94 TS 6.6-2 TS 6.8-1 06-19-98 TS 6.8-1

TS 3.1-2

b. If an unscheduled loss of one or more reactor coolant pumps occurs while operating below 10% RATED POWER (P-7) and results in less than two pumps in service, the affected plant shall be shutdown and the reactor made subcritical by inserting all control banks into the core. The shutdown rods may remain withdrawn.

c. When the average reactor coolant loop temperature is greater than 350 0 F, the following conditions shall be met:

1. At least two reactor coolant loops shall be OPERABLE.

2. At least one reactor coolant loop shall be in operation.

d. When the average reactor coolant loop temperature is less than or equal to 350°F, the following conditions shall be met:

1. A minimum of two non-isolated loops, consisting of any combination of reactor coolant loops or residual heat removal loops, shall be OPERABLE, except as specified below:

(a) One RHR loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other RHR loop is OPERABLE and in operation.

(b) During REFUELING OPERATIONS the residual heat removal loop may be removed from operation as specified in TS 3.10.A.4.

2. At least one reactor coolant loop or one residual heat removal loop shall be in operation, except as specified in Specification 3. 10.A.4.

Amendment Nos.

TS 3.8-5 If the containment air partial pressure rises to a point above the allowable value the reactor shall be brought to the HOT SHUTDOWN condition. If a LOCA occurs at the time the containment air partial pressure is at the maximum allowable value, the maximum containment pressure will be less than design pressure (45 psig), the containment will depressurize to 0.5 psig within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and less than 0.0 psig within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The radiological consequences analysis demonstrates acceptable results provided the containment pressure does not exceed 0.5 psig for the interval from I to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following the Design Basis Accident.

If the containment air partial pressure cannot be maintained greater than or equal to 9.0 psia, the reactor shall be brought to the HOT SHUTDOWN condition. The shell and dome plate liner of the containment are capable of withstanding an internal pressure as low as 3 psia, and the bottom mat liner is capable of withstanding an internal pressure as low as 8 psia.

References UFSAR Section 4.2.2.4 Reactor Coolant Pump UFSAR Section 5.2 Containment Isolation UFSAR Section 5.2.1 Design Bases UFSAR Section 5.2.2 Isolation Design UFSAR Section 5.3.4 Containment Vacuum System Amendment Nos.

TS 3.10-6a During refueling, the reactor refueling water cavity is filled with approximately 220,000 gal of water borated to at least 2,300 ppm boron. The boron concentration of this water, established by Specification 3. 10.A.7, is sufficient to maintain the reactor subcritical by at least 5% Ak/k in the COLD SHUTDOWN condition with all control rod assemblies inserted. This includes a 1% Ak/k and a 50 ppm boron concentration allowance for uncertainty. This concentration is also sufficient to maintain the core subcritical with no control rod assemblies inserted into the reactor. Checks are performed during the reload design and safety analysis process to ensure the K-effective is equal to or less than 0.95 for each core. Periodic checks of refueling water boron concentration assure the proper shutdown margin. Specification 3.10.A.8 allows the Control Room Operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.

In addition to the above safeguards, interlocks are used during refueling to assure safe handling of the fuel assemblies. An excess weight interlock is provided on the lifting hoist to prevent movement of more than one fuel assembly at a time. The spent fuel transfer mechanism can accommodate only one fuel assembly at a time.

Amendment Nos.

TS 3.17-2

b. Before opening the hot leg loop stop valve.

1) The boron concentration of the isolated loop shall be greater than or equal to the boron concentration corresponding to the shutdown margin requirements of Specification 1.O.C.2 or 3.10.A.7, as applicable for the active volume of the Reactor Coolant System. Verification of this condition shall be completed within I hour prior to opening the hot leg stop valve in the isolated loop.

c. Before opening the cold leg loop stop valve.

1) The hot leg loop stop valve shall be open with relief line flow established for at least 90 minutes at greater than or equal to 125 gpm.

2) The cold leg temperature of the isolated loop shall be at least 70°F and within 20°F of the highest cold leg temperature of the active loops.

Verification of this condition shall be completed within 30 minutes prior to opening the cold leg stop valve in the isolated loop.

3) The boron concentration of the isolated loop shall be greater than or equal to the boron concentration corresponding to the shutdown margin requirements of Specification 1.O.C.2 or 3.10.A.7, as applicable for the active volume of the Reactor Coolant System. Verification of this condition shall be completed after relief line flow for at least 90 minutes at greater than or equal to 125 gpm and within I hour prior to opening the cold leg stop valve in the isolated loop.

5. Whenever an isolated and drained reactor coolant loop is filled from the active volume of the RCS, the following conditions shall apply:

a. Seal injection may be initiated to the reactor coolant pump in the isolated loop provided that:

1) The isolated loop is drained. Verification of this condition shall be completed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to initiating seal injection.

Amendment Nos.

TS 3.17-3

2) The boron concentration of the source for reactor coolant pump seal injection shall be greater than or equal to the boron concentration corresponding to the shutdown margin requirements of Specification 1.O.C.2 or 3.1O.A.7, as applicable for the active volume of the Reactor Coolant System. If using the Volume Control Tank (VCT) as the source for reactor coolant pump seal injection, verification of the boron concentration shall be completed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to initiating seal injection and every hour thereafter during the loop backfill evolution.

b. The cold leg loop stop valve may be energized and/or opened to backfill the loop from the active volume of the Reactor Coolant System provided that:

1) The isolated loop is drained or reactor coolant pump seal injection has been initiated in accordance with Specification 3.17.5.a above.

Verification of the loop being drained shall be completed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to partially opening the cold leg stop valve in the isolated loop.

2) The Reactor Coolant System level is at least 18 ft.

3) A source range nuclear instrumentation channel is OPERABLE with audible indication in the control room.

c. Backfilling of the isolated loop may continue provided that:

1) The Reactor Coolant System level is maintained at or above 18 ft. If Reactor Coolant System level is not maintained at or above 18 ft. the loop stop valve shall be closed.

2) The boron concentration of the reactor coolant pump seal injection source is greater than or equal to the boron concentration corresponding to the shutdown margin requirements of Specification 1.O.C.2 or 3.10.A.7, as applicable for the active volume of the Reactor Coolant System. If the boron concentration is not maintained greater than or equal to the required boron concentration noted above, the loop stop valve on the loop being backfilled shall be closed and either drain the loop or apply Specification 3.17.4.

Amendment Nos.

TS 3.17-4

3) A source range nuclear instrumentation channel is OPERABLE and continuously monitored with audible indication in the control room during the backfill evolution. Should the count rate increase by more than a factor of two over the initial count rate, the cold leg loop stop valve shall be closed and no attempt made to open the cold leg stop valve until the reason for the count rate increase has been determined.

d. When the isolated loop is full, the cold leg loop stop valve can be fully opened and the hot leg loop stop valve opened provided that:

1) The boron concentration of the isolated loop is greater than or equal to the boron concentration corresponding to the shutdown margin requirements of Specification 1.O.C.2 or 3.1O.A.7, as applicable for the active volume of the Reactor Coolant System. If the VCT was used as the source for reactor coolant pump seal injection, this condition shall be verified within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to fully opening the loop stop valves. If the boron concentration in the isolated loop does not meet the condition above, close the loop stop valve and either drain the loop or apply Specification 3.17.4.

2) The hot and cold leg loop stop valves are opened within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the isolated loop is filled. If the loop stop valves are not fully open within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , close the loop stop valves and either drain the loop or apply Specification 3.17.4.

Basis The Reactor Coolant System may be operated with isolated loops in COLD SHUTDOWN or REFUELING SHUTDOWN in order to perform maintenance. A loop stop valve in any loop can be closed for up to two hours without restriction for testing or maintenance in these operating conditions. While operating with a loop isolated, AC power is removed from the loop stop valves and their breakers locked opened to prevent inadvertent opening. When the isolated loop is returned to service, the coolant in the isolated loop Amendment Nos.

TS 4.0-2 4.0.5 Surveillance requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:

a. Inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(f), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(f)(6)(i). Inservice inspection of ASME Code Class 1, 2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).

b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:

Amendment Nos.

TS 4.0-5 is identified that a surveillance has not been performed and not at the time that the allowed surveillance interval was exceeded. Completion of the surveillance requirement within the allowable outage time limits of the Action Statement requirements restores compliance with the requirements of Specification 4.0.3.

However, this does not negate the fact that the failure to have performed the surveillance within the allowed surveillance interval, defined by the provisions of Specification 4.0.2, was a violation of the operability requirements of a Limiting Condition for Operation. Further, the failure to perform a surveillance within the provisions of Specification 4.0.2 is a violation of a Technical Specification requirement and is, therefore, a reportable event under the requirements of 10 CFR 50.73(a)(2)(i)(B), unless it meets an exception listed therein, because it is a condition prohibited by the plant's Technical Specifications.

If the allowable outage time limits of the Action Statement requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or a shutdown is required to comply with Action Statement requirements, e.g., Specification 3.0.1, a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance is provided to permit a delay in implementing the Action Statement requirements. This provides an adequate time limit to complete surveillance requirements that have not been performed. The purpose of this allowance is to permit the completion of a surveillance before a shutdown is required to comply with Action Statement requirements or before other remedial measures would be required that may preclude completion of a surveillance. The basis for this allowance includes consideration for plant conditions, adequate planning, availability of personnel, the time required to perform the surveillance, and the safety significance of the delay in completing the required surveillance. This Amendment Nos.

TABLE 4.1-1(Continued)

MINIMUM FREQUENCIES FOR CHECK, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS Channel Description Check Calibrate Test Remarks

32. Auxiliary Feedwater

a. Steam Generator Water Level Low-Low S R Q (1) I) The auto start of the turbine driven pump is not included in the quarterly test, but is tested within 31 days prior to I each startup.

b. RCP Undervoltage S R R(1)(2) 1)The actuation logic and relays are tested within 31 days prior to each startup.

2) Setpoint verification not required.

c. S.I. (All Safety Injection surveillance requirements)

d. Station Blackout N.A. R N.A.

e. Main Feedwater Pump Trip N.A. N.A. R

33. Loss of Power

a. 4.16 KV Emergency Bus Undervoltage N.A. R Q(1) 1) Setpoint verification not required.

(Loss of Voltage)

b. 4.16 KV Emergency Bus Undervoltage N.A. R Q(I) 1) Setpoint verification not required.

(Degraded Voltage)

34. Deleted

35. Manual Reactor Trip N.A. N.A. R The test shall independently verify the operability of the undervoltage and shunt trip attachments for the manual reactor trip function. The test shall also verify the operability of the bypass breaker trip circuit.

36. Reactor Trip Bypass Breaker N.A. N.A. M(l), 1) Remote manual undervoltage trip immediately after R(2) placing the bypass breaker into service, but prior to z0 commencing reactor trip system testing or required maintenance.

F1

2) Automatic undervoltage trip.

37. Safety Injection Input to RPS N.A. N.A. R

38. Reactor Coolant Pump Breaker Position Trip N.A. N.A. R 00

TABLE 4.1-2A MINIMUM FREQUENCY FOR EQUIPMENT TESTS FSAR SECTION DESCRIPTION TEST FREQUENCY REFERENCE

1. Control Rod Assemblies Rod drop times of all full Prior to reactor criticality: 7 length rods at hot conditions a. For all rods following each removal of the reactor vessel head

b. For specially affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and

c. Once per 18 months

2. Control Rod Assemblies Partial movement of all rods Quarterly 7

3. Refueling Water Chemical Addition Functional Once per 18 months 6 Tank

4. Pressurizer Safety Valves Setpoint Per TS 4.0.5 4

5. Main Steam Safety Valves Setpoint Per TS 4.0.5 10

6. Containment Isolation Trip "*Functional Once per 18 months 5

7. Refueling System Interlocks "*Functional Prior to refueling 9.12

8. Service Water System "*Functional Once per 18 months 9.9

9. Deleted 4

I

10. Primary System Leakage "*Evaluate Daily

11. Diesel Fuel Supply "*Fuel Inventory 5 days/week 8.5 z0 12. Deleted

13. Main Steam Line Trip Valves Functional Before each startup (TS 4.7) 10 (Full Closure) The provisions of Specification 4.0.4.

H3 are not applicable En' oC.

TS 4.4-2 The containment is designed for a maximum pressure of 45 psig. The containment is maintained at a subatmospheric air partial pressure consistent with TS Figure 3.8-1 depending upon the cooldown capability of the Engineered Safeguards and will not rise above 45 psig for any postulated loss-of-coolant accident.

The initial test pressure for the Type A test is 47.0 psig to allow for containment expansion and equalization. A review was performed to determine the effects of pressurizing containment above its design pressure of 45.0 psig. This review was based on the original containment test at 52 psig. During that test, the calculated stresses were found to be well within the allowable yield strength of the structural reinforcing bars, therefore performance of the Type A test at 47 psig will have no detrimental effect on the containment structure.

All loss-of-coolant accident evaluations have been based on an integrated containment leakage rate not to exceed 0.1% of containment volume per 24 hr.

The above specification satisfies the conditions of 10 CFR 50.54(o) which stated that primary reactor containments shall meet the containment leakage test requirements set forth in Appendix J.

The limitations on closure and leak rate for the containment airlocks are required to meet the restrictions on containment integrity and containment leak rate. Surveillance testing of the airlock seals provides assurance that the overall airlock leakage will not become excessive due to seal damage during the intervals between airlock leakage tests.

References UFSAR Section 5.5 Containment Tests and Inspections UFSAR Section 7.5.1 Design Bases of Engineered Safeguards Instrumentation UFSAR Section 14.5 Loss of Coolant Accident 10 CFR 50 Appendix J "Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors" Amendment Nos.

TS 6.1-7

f. Responsibilities The SNSOC shall be responsible for:

1. Review of a) all new normal, abnormal, and emergency operating procedures and all new maintenance procedures, b) all procedure changes that require a regulatory evaluation, and c) any other procedures or changes thereto as determined by the Site Vice President which affect nuclear safety.

2. Review of all new test and experiment procedures that affect nuclear safety.

3. Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.

4. Review of proposed changes to Technical Specifications and shall submit recommended changes to the Site Vice President.

5. Investigation of all violations of the Technical Specifications, including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Vice President - Nuclear Operations and to the Management Safety Review Committee.

6. Review of all Reportable Events and special reports submitted to the NRC.

7. Review of facility operations to detect potential nuclear safety hazards.

8. Performance of special reviews, investigations or analyses and report thereon as requested by the Chairman of the SNSOC or Site Vice President.

Amendment Nos.

TS 6.1-8

9. Deleted.

10. Deleted.

11. Review of every unplanned onsite release of radioactive material to the environs exceeding the limits of Specification 3.11, including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vice President - Nuclear Operations and to the Management Safety Review Committee.

12. Review of changes to the Process Control Program and the Offsite Dose Calculation Manual.

13. Review of the Fire Protection Program and implementing procedures and shall submit recommended Program changes to the designated offsite management responsible for reviewing changes that pertain to Fire Protection.

g. Authority The SNSOC shall:

1. Provide written approval or disapproval of items considered under (1) through (3) above. SNSOC approval shall be certified in writing by either the Manager - Station Operations and Maintenance or the Manager Station Safety and Licensing.

2. Render determinations in writing with regard to whether or not each item considered under (1) through (5) above requires a license amendment request.

3. Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Vice President Nuclear Operations and to the Management Safety Review Committee of disagreement between SNSOC and the Site Vice President; however, the Site Vice President shall have responsibility for resolution of such disagreements pursuant to 6.1.A above.

h. Records The SNSOC shall maintain written minutes of each meeting and copies shall be provided to the Vice President - Nuclear Operations and to the Management Safety Review Committee.

Amendment Nos.

TS 6.1-10

e. Meeting Frequency The MSRC shall meet at least once per calendar quarter.

f. Quorum The minimum quorum of the MSRC necessary for the performance of the MSRC review and audit functions of these Technical Specifications shall consist of the Chairman or his designated alternate and at least 50% of the MSRC members including alternates. No more than a minority of the quorum shall have line responsibility for operation of the unit.

g. Review The MSRC shall be responsible for the review of:

1. Regulatory reviews as programmatically discussed in the Updated Final Safety Analysis Report for 1) changes to procedures, equipment or systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to assess the effectiveness of the safety and regulatory review program and to verify it is effective in identifying changes that require a license amendment pursuant to Section 50.59, 10 CFR.

2. Proposed changes to procedures, equipment or systems which require a license amendment as defined in Section 50.59, 10 CFR.

3. Proposed tests or experiments which require a license amendment as defined in Section 50.59, 10 CFR.

4. Proposed changes to Technical Specifications or the Operating Licenses.

Amendment Nos.

TS 6.4-3

2. The requirements of 6.4.B.1 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 Supervisor 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. All procedures described in 6.4.A and 6.4.B shall be reviewed and approved by the Station Nuclear Safety and Operating Committee (SNSOC) prior to implementation.

Subsequent procedure changes that require a regulatory evaluation shall also be reviewed and approved by SNSOC prior to implementation. All other changes shall be independently reviewed and approved as discussed in the Updated Final Safety Analysis Report.

Amendment Nos.

TS 6.4-7 L. Iodine Monitoring The licensee shall implement a program which will ensure the capability to accurately determine the airborne iodine concentration in vital area under accident conditions.

This program shall include the following:

1. Training of personnel,

2. Procedures for monitoring, and

3. Provisions for maintenance of sampling and analysis equipment.

M. Deleted Amendment Nos.

TS 6.6-2 resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operations),

supplementary reports shall be submitted at least every 3 months until all three events have been completed.

2. Annual ReportsI

a. A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man-rem exposure according to work and job functions2 , e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements.

Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

Note: Footnotes I and 2 are located on page TS 6.6-11.

Amendment Nos.

TS 6.8-1 6.8 PROCESS CONTROL PROGRAM AND OFFSITE DOSE CALCULATION MANUAL A. Process Control Program (PCP)

Changes to the PCP:

1. Shall be documented and records of reviews performed shall be retained as required by the Operational Quality Assurance Program Topical Report. This documentation shall contain:

a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and

b. A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

2. Shall require review and acceptance by the SNSOC and the approval of the Site Vice President prior to implementation.

B. Offsite Dose Calculation Manual (ODCM)

Changes to the ODCM:

1. Shall be documented and records of reviews performed shall be retained as required by the Operational Quality Assurance Program Topical Report. This documentation shall contain:

a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and Amendment Nos.

ML023650621

Text

Subject:

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