ML18101B022

Application for Amends to Licenses DPR-70 & DPR-75, Representing Changes to RCS - Chemistry Spec 3/4.4.7 & Spec 3/4.4.8

ML18101B022
Person / Time
Site:	Salem
Issue date:	09/28/1995
From:	Storz L Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML18101B023	List: ML18101B022 ML18101B024
References
LCR-95-12, LR-N95054, NUDOCS 9510050020
Download: ML18101B022 (12)

Text

, ,~-I Public Service Electric and Gas Company Louis F. Storz Senior Vice President - Nuclear Operations Public Service Electric and Gas Comp~E p 28* ;99;36, Hancocks Bridge, NJ 08038 609-339-5700 LR-N95054 LCR 95-05 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:

LICENSE AMENDMENT APPLICATION REACTOR COOLANT SYSTEM - CHEMISTRY SALEM GENERATING STATION UNIT NOS. 1 and 2 FACILITY OPERATING LICENSE NOS. DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 This letter submits an application for amendment to Appendix A of Facility Operating Licenses DPR-70 and DPR-75 for Salem Generating Station Unit Nos. 1 and 2, respectively, and is being filed in accordance with 10CFR50. 90. Pursuant to the requirements of 10CFR50.91(b) (1), a copy of this request for amendment has been sent to the State of New Jersey.

The proposed Technical Specification revisions c::ontained herein represent changes

• to "Reactor Coolant System Chemistry" Specification 3/4.4.7 and Specification 3/4.4.8, (Salem Units 1 and Unit 2, respectively) . The proposed changes will remove these sections of the Technical Specifications and relocates the Bases to the Salem Updated Final Safety Analysis Report (UFSAR) and the Surveillance Requirements and Limiting Conditions for Operation (LCOs) to applicable plant procedures controlled by the 10CFR50.59 process. The Applicability will be changed from "At all times" to "Modes 1,2,3,4,5 and 6".

A description of the requested amendment, supporting information and analyses for the changes, and the basis for a no significant hazards consideration determination are provided in Attachment 1.

The Technical Specification pages affected and that would be removed by the proposed change are provided in Attachments 2 and 3, with pen and ink changes, for Salem Units 1 and 2, respectively.

The proposed changes have been evaluated in accordance with 10CFR50. 91 (a) (1), using the criteria in 10CFR50. 92 (c), and it has been determ1ned that this request involves no significant hazards considerations.

Printed on Recycled Paper 9510050020 950928 PDR ADOCK 05000272 f\(JD \ I I p PDR

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Document Control Desk SEP 2 8 1995 LR-N95054 In accordance with the guidance presented in the "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" as published in the Federal Register (58 FR 39132),

the proposed changes have received a multidisciplinary review by responsible technical supervisory personnel, including onsite operations personnel.

Upon NRC approval of this proposed changes, PSE&G requests that the amendment be made effective on the date of issuance, but implemented within sixty days to provide sufficient time for associated administrative activities.

Should you have any questions regarding this request, please contact us.

Sincerely,

~;,(~

Affidavit Attachments (3)

Document Control Desk SEP 2 8 1995

. LR-N95054 C Mr. T. T. Martin, Administrator - Region I U. s. Nuclear Regulatory Commission 475 Allendale Road King of Prvssia, PA 19406 Mr. L. N. Olshan, Licensing Project Manager - Salem

u. s. Nuclear Regulatory Commission One White Flint North Mail Stop 14E21 11555 Rockville Pike Rockville, MD 20852 Mr. c. s. Marschall (S09)

USNRC Senior Resident Inspector Mr. K. Tosch, Manager IV NJ Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625

REF: LR-N95054 LCR 95-05 STATE OF NEW JERSEY )

)SS.

COUNTY OF SALEM )

L. F. Storz, being duly sworn according to law, deposes and says:

I am Senior Vice President - Nuclear Operations of Public Service Electric and Gas Company, and as such, I find the matters set forth in the above referenced letter, concerning the Salem Generating Station, Unit Nos. 1 and 2, are true to the best of my knowledge, information and belief.

Subscribed and Sworn to before me day o~pfu~ 1995 I

0 4~ 60];UIY'._ KIMBERLY JO BROWN NOTARY PUBLIC OF NEW JERSEY My Commission expires on My Commission Expires April 21, 19119

'.'

ATTACHMENT 1 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS LICENSE AMENDMENT APPLICATION REACTOR COOLANT SYSTEM CHEMISTRY SALEM GENERATING STATION UNIT NOS. 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 LR-N95054 DOCKET NOS. 50-272 and 50-311 LCR 95-05 I. DESCRIPTION OF THE PROPOSED CHANGES The proposed changes relocate "Reactor Coolant System Chemistry" Technical Specification 3/4.4.7 (Salem Unit 1) and 3/4.4.8 (Salem Unit 2)and their associated Bases. It would remove these sections of the Technical Specifications and relocate the requirements to licensee documents controlled by the 10CFR50. 59 process. Additionally, the Applicability would be changed from "At all times" to "Modes 1,2,3,4,5 and 6 11

• II. REASONS FOR THE CHANGES The proposed changes to the Technical Specifications conform with the NRC's "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors", dated July 22, 1993 (58 FR 39132). During the development of NUREG-1431 "Standard Technical Specifications, Westinghouse Plants" the NRC issued a letter to the B&W Owners Group, T.E. Murley to w.s. Wilgus dated May 9, 1988, which identified in W-Table 2 "Westinghouse Standard Technical Specifications, LCOs Which May Be Relocated", the Reactor Coolant System Chemistry as a Limiting Condition for Operation (LCO) which may be relocated to other licensee-controlled documentation contingent upon NRC staff approval of the location of and controls over the relocated requirement.

Through this submittal, PSE&G is requesting implementation of the NRC' s recommendation regarding the relocation of the Reactor Coolant System Chemistry specification. The submittal will relocate the Bases to the Salem Generating station Unit Nos. 1 and 2 Updated Final Safety Analysis Report (UFSAR) and the Surveillance Requirements and LCOs to the applicable plant procedures. The Applicability will be changed from "At all times" to "Modes 1,2,3,4,5 and 6 11

• Removal of these specifications from the Technical Specifications is expected to produce an improvement in the safety of Salem Generating Station through a reduction in unnecessary plant transients and more efficient use of NRC and PSE&G resources. It would also reduce regulatory burden, Page 1 of 6

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ATTACHMENT 1

• LR-N95054

. REVISION OF SURVEILLANCE REQUIREMENTS FOR LCR 95-05 AREAS INSIDE CONTAINMENT since future changes to the Reactor Coolant System Chemistry functions would be conducted using the 10CFR50.59 process.

III. JUSTIFICATION FOR THE CHANGES The limitations on Reactor Coolant Chemistry ensure that corrosion of the Reactor Coolant System is minimized and that the potential for Reactor Coolant System leakage or failure due to stress corrosion is reduced. Maintaining the chloride, fluoride and dissolved oxygen chemistry within the limits specified in the Salem Generating Station Unit Nos. 1 and 2 UFSAR, Table 5. 2-28, "Reactor Coolant Water Chemistry Specifications", provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the chloride, fluoride and dissolved oxygen limits are time and temperature dependent. A review of the Reactor Coolant system Chemistry for chlorides, fluorides and dissolved oxygen for Salem Generating station Unit Nos. 1 and 2 for the past 5 years shows that the Action statement has not been entered while at normal power operation. During a Unit 2 shutdown in November 1991, while in Mode 3 and cooling down, the Reactor Coolant System chloride level was above Technical Specification limits for approximately 62 hours7.175926e-4 days <br />0.0172 hours <br />1.025132e-4 weeks <br />2.3591e-5 months <br />.

Moving all fuel to the Fuel Pool is a common outage activity.

This condition is commonly referred to as the "Mode Undefined." This Mode allows relief from certain equipment operability requirements in the Technical Specifications and permits draining the reactor vessel to facilitate outage work.

Performing water chemistry in accordance with the Technical Specifications in the Mode Undefined is an unnecessary task that does not provide commensurate safety benefit. When the Reactor Coolant System recirculation is removed from service and the reactor vessel is drained, samples are stagnant for lack of recirculation and represent coolant when the system was last in operation. The Bases for the Technical Specification is to minimize and reduce potential for Reactor Coolant System leakage or failure due to stress corrosion.

High stress and high temperature conditions which lead to stress corrosion, are absent in the Mode Undefined. After the reactor vessel is refilled and prior to fuel movement back to the reactor vessel, chlorides and fluorides are sampled to verify that Reactor Coolant System Chemistry is within limits.

Page 2 of 6

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ATTACHMENT 1 LR-N95054

. REVISION OF SURVEILLANCE REQUIREMENTS FOR LCR 95-05 AREAS INSIDE CONTAINMENT Additionally, the plant procedure requirements will continue to provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

With the conclusion of the Technical Specification Improvement Program, the NRC issued it's Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors (58 FR 39132). The NRC concluded that the purpose of the Technical Specifications is to impose those conditions or limitations upon reactor operation necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public heal th and safety. This was achieved by identifying those issues that are of controlling importance to safety and establishing on them certain limiting conditions for operation which cannot be changed without prior NRC approval.

This policy statement included a set of criteria which delineated those constraints on design and operation derived from the UFSAR and that belonged in the Technical Specifications in accordance with 10CFRS0.36. Specifically, these criteria are as follows:

Criterion 1 Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Licensee Determination Reactor Coolant System Chemistry is a chemical analysis of Reactor Coolant System water for chlorides, fluorides and dissolved oxygen. It is not installed chemistry analysis instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or' presents a challenge to the integrity of a fission product barrier.

Page 3 of 6

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ATTACHMENT 1 e

• LR-N95054

. REVJ:SJ:ON OF SURVEJ:LLANCE REQUJ:REMENTS FOR LCR 95-05 AREAS INSIDE CONTAINMENT Licensee Determination A review of the appropriate Salem design bases documentation (i.e. , UFSAR) determined that Reactor Coolant System Chemistry is not a process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis as discussed in the Accident Analysis Section of the UFSAR that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3 A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Licensee Determination A review of the appropriate Salem design bases documentation (i.e. UFSAR) determin.ed that Reactor Coolant System Chemistry is not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient discussed in the Accident Analysis Section of the UFSAR that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4 A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

Licensee Determination A review of the Reactor Coolant System Chemistry for chlorides, fluorides and dissolved oxygen for Salem Generating Station Unit Nos. 1 and 2 for the past 5 years showed that the Action statement has never been entered while at normal power operation. During a Unit 2 shutdown in November 1991, while '

in Mode 3 and cooling down, the Reactor Coolant System chloride level was above Technical Specification limits for Page 4 of 6

ATTACHMENT 1 LR-N95054

. REVISION OF SURVEILLANCE REQUIREMENTS FOR LCR 95-05

.AREAS INSIDE CONTAINMENT approximately 62 hours7.175926e-4 days <br />0.0172 hours <br />1.025132e-4 weeks <br />2.3591e-5 months <br />. A review of the Probability Risk Assessment and the Individual Plant Examination for Salem Generating £tation showed that Reactor Coolant System Chemistry is not considered as a potential initiating event.

In summary, water chemistry, other than reactivity control, is not considered in the accident sequence evaluations.

Therefore, Reactor Coolant System Chemistry is not a structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

In accordance with the NRC's criteria for determining those specifications which may be relocated to licensee-controlled documentation, PSE&G believes that this submittal for removing the Reactor Coolant system Chemistry specification from the Technical Specifications is justified. These changes will reduce regulatory burden while maintaining existing levels of plant safety.

IV. DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION PSE&G has, pursuant to 10CFR50. 92, reviewed the proposed amendment to determine whether our request involves a significant hazards consideration. We have determined that operation of Salem Generating Station Unit 1 and Unit 2, in accordance with the proposed changes:

1. Will not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes involve no hardware changes, no changes to the operation of any systems or components, and no changes to existing structures. Specifically, changing the Applicability from "At all times" to "Modes 1,2,3,4,5 and 6 11 by this submittal will not alter established chemistry for chlorides, fluorides and dissolved oxygen of the Reactor Coolant System. The relocation of this Surveillance Requirement/LCOs and Bases to plant procedures and the UFSAR respectively, will continue to ensure that the chemistry analysis of the Reactor Coolant System water is monitored and controlled. Changing the Applicability from "At all times" to "Modes 1, 2, 3, 4, 5 and 6" represent changes that do not affect plant safety and do not alter existing accident analyses.

Page 5 of 6

ATTACHMENT 1

• REVISION OF SURVEILLANCE REQUIREMENTS FOR LR-N95054 LCR 95-05 AREAS INSIDE CONTAINMENT

2. Will not create the possibility of a new or different kind of accident from any previously evaluated.

The proposed changes are procedural in nature concerning the location of the descriptive information and surveillance requirements for Reactor Coolant System Chemistry. Removing these specifications from the Technical Specifications and placing them in the UFSAR and plant procedures will not alter the maintenance of the Reactor Coolant System Chemistry or the ability to monitor its intended functions. Therefore, these changes will not create a new or unevaluated accident or operating condition.

3. Will not involve a significant reduction in a margin of safety.

The proposed changes relocate the Reactor Coolant System Chemistry Requirements/Leos from the Technical Specifications to the UFSAR and plant procedures in accordance with guidance provided by the NRC Final Policy Statement (58 FR 39132) regarding the improvement of Technical Specifications. The requirements that will reside in the UFSAR and plant procedures for the Reactor Coolant System Chemistry will ensure that the ability to determine chloride, fluoride and dissolved oxygen concentrations in the Reactor Coolant System is properly maintained and that the maintenance of the Reactor Coolant System Chemistry will be commensurate with its safety significance. Therefore, the proposed changes will not involve a significant reduction in any margins of safety.

V. CONCLUSIONS Based on the above, PSE&G has determined that the proposed change does not involve a significant hazards consideration.

Page 6 of 6

• ATTACHMENT 2 e

TECHNICAL SPECIFICATION PAGE WITH PEN AND INK CHANGES LICENSE AMENDMENT APPLICATION REACTOR COOLANT SYSTEM - CHEMISTRY SALEM GENERATING STATION UNIT 1 FACILITY OPERATING LICENSE DPR-70 LR-N95054 DOCKET NO. 50-272 LCR 95-05 The following Technical Specification for Facility Operating License No. DPR-70 is affected by this License Amendment Request:

Technical Specification Pages INDEX V, XII 3/4.4.7 3/4 4-17, 3/4 4-18, 3/4 4-19, B 3/4 4-4 INSERTS INSERT A 3/4.4.7 DELETED INSERT B 3/4.4.7 DELETED INSERT C PAGES 3/4 4-17 THROUGH 3/4 4-19 DELETED INSERT D 3/4 4-17 through 3/4 4-19 INSERT E 3/4.4.7 THIS SECTION DELETED

• ATTACHMENT 3

• TECHNICAL SPECIFICATION PAGE WITH PEN AND INK CHANGES LICENSE AMENDMENT APPLICATION REACTOR COOLANT SYSTEM - CHEMISTRY SALEM GENERATING STATION UNIT 2 FACILITY OPERATING LICENSE DPR-75 LR-N95054 DOCKET NO. 50-311 LCR 95-05 The following Technical Specification for Facility Operating License No. DPR-75 is affected by this License Amendment Request:

Technical Specification INDEX V, XII 3/4.4.8 3/4 4-20, 3/4 4-21, 3/4 4-22, B 3/4 4-5 INSERTS INSERT F 3/4.4.8 DELETED INSERT G 3/4.4.8 DELETED INSERT H PAGES 3/4 4-20 THROUGH 3/4 4-22 DELETED INSERT I 3/4 4-20 through 3/4 4-22 INSERT J 3/4.4.8 THIS SECTION DELETED