ML18095A244
| ML18095A244 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 05/31/1990 |
| From: | Preston B Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML18095A245 | List: |
| References | |
| GL-83-43, GL-85-19, LCR-84-01, LCR-84-1, NLR-N90113, NUDOCS 9006060081 | |
| Download: ML18095A244 (20) | |
Text
PSEG *)
Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Nuclear Department MAY 3 1 1990 NLR-N90113 LCR 84-01, Rev. 3 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
REQUEST FOR AMENDMENT SALEM GENERATING STATION UNIT NOS. 1 AND 2 FACILITY OPERATING LICENSE NOS. DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 The attached License Change Request (LCR) is a resubmittal of LCR 84-01, Revision 3, which was sent via letter dated May 18, 1990 (NLR-N90069)., which includes a description, justification and significant hazards analysis for the proposed changes, is identical to that of the May 18, 1990 letter. contains the Technical Specification pages marked up with pen and ink changes, and has been revised to utilize the currently effective Technical Specification pages and to increase legibility.
The changes proposed in Attachment 2 to this letter contain no substantive differences from those proposed in the May 18, 1990 submittal.
PSE&G requests that the implementation date be established at 60 days following issuance of the approved 1 License Amendment.
Similar changes to the 10 CFR 50.59 review process, the Nuclear Safety Review organization and Radiological Environmental Monitoring Program for the Hope Creek Generating Station have been requested via a separate submittal.
In order to assure consistent implementation of the changes between the Salem and Hope Creek stations, PSE&G also requests that the License Amendments approving the changes common to both Stations be issued concurrently.
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Mr. J. c. Stone Licensing Project Manager Mr. T. Johnson 2
Senior Resident Inspector - Salem Mr. T. Martin, Administrator Region I Mr. K. Tosch, Chief Bureau of Nuclear Engineering Department of Environmental Protection CN 415 Trenton, New Jersey 08625 MAY 3 1 1990
NLR-N90113 ATTACHMENT 1 Docket#_ ~tJ -
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Date ~-/J>//90 Oba og,1 Regulatory DocketFiie
~of Ltr PROPOSED CHANGE TO THE TECHNICAL SPECIFICATIONS FACILITY OPERATING LICENSE NOS. DPR-70 AND DPR-75 SALEM GENERATING STATION UNIT NOS. 1 AND 2 DOCKET NOS! 50-272 AND 50-311 I.
Description of Change This proposed change supersedes License Change Request (LCR) 84-01, Revision 2, dated July 7, 1989 (NLR-N89050).
The specific proposed changes are as follows:
- 1.
Change definition 1.27 from "REPORTABLE OCCURRENCE" to "REPORTABLE EVENT", referencing 10CFR50.73.
- Also, editorial corrections are made to definition 1.32, "STAGGERED TEST BASIS".-
- 2.
(Unit 2 only).
Editorial revision of Specification 4.2.2.2(c)* to incorporate the change to the relationship between the limit for fractional thermal power operation and rated thermal power, as approved in Amendment 6 to the SGS Unit 2 Technical Specifications.
Revise Specification 4.2~2.2(e) and B3/4.2.3 to reference 6.9.1.9.
- 3.
Revision of Specifications 3.3.3.8, 3.3.3.9 and 3.11.3 to delete reference to Specification 6.9.1.9.b and to provide editorial corrections.
- 4.
Revise Table 4.4-2, Steam Generator Tube Inspection, Specifications 4.4.5.5 and B 3/4.4.5 (Unit 1) and Specifications 4.4.6.5 and B3/4.4.6 (Unit 2), to delete reference to Section 6.9.1 and add the reporting requirements applicable to steam generator tube degradation.
similarly revise Specification 4.6.1.6.2 for containment degradation reporting requirements.
Unit 2 Specification 4.8.1.1.4 is being revised to replace reference to Section 6.9.1 with 6.9.2 (Special Reports) for diesel generator failures.
- 5.
Revision of Section 3.4.8 (Unit i) and 3.4.9 (Unit 2) to eliminate those reactor coolant system specific activity limit requirements deemed unnecessary by NRC Generic
NLR-N90113 Letter 85-19.
The corresponding BASES sections are revised accordingly.
- 6.
Revision of Specifications 3.11.1.2, 3.11.1.3, 3.11.2.2, 3.11.2.3, 3.11.2.4, 3.11.4, 3.12.1, 3/4.12.2 and 3/4.12.3 to delete reference to Licensee Event Reports and/or to provide editorial corrections.
Revise 3.11.2.4 to state that Specification 3.0.3 is not applicable.
- 7.
Revision to tables 3.12-2 and 4.12-1 to change the reporting levels and lower limits of detection of radioactivity concentrations of tritium and I-131 for the radiological environmental monitoring program.
The changes will allow higher activity levels provided that drinking water pathways are not involved.
Also, an editorial correction to table 4.12-1 is made to change "Cs-136 11 to "Cs-134 11 *
- 8.
Change "General Manager - Nuclear Safety Review" to "General Manager - Quality Assurance and Nuclear Safety" throughout Chapter 6.
- 9.
Revise Section 6.2.2 to delete Specification 6.2.2(c),
which is redundant to Table 6.2-1; delete references to "Health Physics technician" and provide editorial corrections to Unit 1.
Revise table 6.2-1 to include the minimum shift staffing requirements for the radiation protection technicians.
- 10. Revision of Specifications 6.2.3.1, 6.3.1 and 6.4.2 to reflect shift Technical Advisor accountability to the "Senior Nuclear" Shift Supervisor, Radiation Protection Manager qualifications, and organizational responsibility for Fire Brigade training, respectively.
Revise Section 6.4.1 to replace reference to 10 CFR Part 55, Appendix A and supplemental requirements, with a general reference to 10 CFR Part
- 55.
- 11. Revision to Specifications 6.5.1.2, 6.5.1.3 and 6.5.1.5 to change the SORC composition and rules concerning alternate members and quorum requirements, respectively.
- 12. Revise Specification 6.5.1.6 to simplify the SORC responsibilities with respect to internal distribution of reports.
Revise 6.5.1.6(a) to require SORC review of procedure changes only if a 10CFR50.59 safety evaluation is required.
Revise 6.5.1.6(f) and 6.5.1.6(i) to delete reference to the General Manager - Nuclear Safety Review.
Revise 6.5.1.6(j), 6.5.1.6(k) and 6.5.1.6(1) to require SORC review of changes to the Security Plan,
NLR-N90113 Emergency Plan or Fire Protection Plan and their implementing procedures only if a 10CFR50.59 safety evaluation is required (or an evaluation under 10CFR50.54(p) for Security Plan or 10CFR50.54(q) for Emergency Plan changes).
Revise Specification 6.5.3.2(d) to replace the reference to safety significant issues with a reference to 10CFR50.59.
Provide editorial clarification to Specification 6.5.1.8(a), with regard to SORC recommendations of approval or disapproval of items to the General Manager -
Salem Operations.
- 13. Revise Section 6.5.2 to reflect the proposed Nuclear Safety department configuration.
The Manager - Offsite Safety Review and the Manager - Onsite Safety Review will be replaced by the Manager - Nuclear Safety, who will have management responsibility for both the Offsite Safety Review (OSR) staff and the Onsite Safety Review Group ( SRG).
- 14. Revise Specification 6.5.2.3 to clarify the use of consultants by the Nuclear Safety Department.
Provide editorial revisions to Specification 6.5.2.4.
Delete the miscellaneous OSR activities at the end of Specification 6.5.2.4.1.
Revise 6.5.2.4.2(g) to add the phrase "that could affect nuclear safety."
- 15. Revision to Specification 6.5.2.4.3(i) and 6.5.2.4.3(j) to specify that either offsite fire protection engineers or independent fire protection consultants will be utilized for the fire protection and loss prevention program implementation audits, with an outside consultant being used at least once per 36 months.
- 16. Delete Specification 6.5.2.7 and insert new Specification 6.5.2.4.4.
This will clarify that the record requirements described therein pertain specifically to the OSR staff. The proposed change will also increase the allowable time periods for forwarding reports of reviews and audits.
- 17. Revise Section 6.5.3 to replace "NSR" with "the OSR staff".
- 18. Revision of the title of Section 6.9 to delete reference to REPORTABLE OCCURRENCES.
Revision to Section 6.9.1 and 6.9.2 to include the correct NRC mailing addresses and refer specifically to USNRC Region I.
Revision to Specification 6.9.1.4 to delete requirements for submittal of the initial Annual Report.
NLR-N90113 19. Revise Specification 6.9.1.5a to replace "film badge measurements" with "self reading dosimeter measurements 11
- Addition of Specification 6.9.1.5c to include primary coolant specific activity analyses results among the annual reports.
- 20. Deletion of Specifications 6.9.1.7, 6.9.1.8 and 6.9.1.9 to reflect reporting requirements of 10CFR50.73, consistent with NRC Generic Letter 83-43.
Renumbering of Specifications 6.9.1.10 and 6.9.1.11 and other editorial revisions.
- 21. Revision of the requirements of Specification 6.9.1.7 (formerly 6.9.1.10) regarding radiological sampling location maps submitted with the Annual Radiological Environmental Operating Report.
- 22. Inclusion of Specification 6.9.1.9 to revise the scheduler requirements for submittal of the Radial Peaking Factor Limit Report.
For Unit 2, revise B3/4.2.3 to reference 6.9.1.9.
- 23. Revision of Section 6.10.l to change "REPORTABLE OCCURRENCES" to "REPORTABLE EVENTS", to delete the five year storage requirement for reactor te:;;ts and experiments and reflect the record storage requirements of 10CFR50.59.
- 24. Revision of Section 6.10.2 to reflect the succession of the Nuclear Review Board by the Offsite Safety Review staff, to reflect the record storage requirements of 10CFR50.59 and to provide editorial corrections.
- 25. Addition of a footnote to Sections 6.12.1 and 6.12.2 to define radiation intensity as it is used to determine whether an area is a High Radiation Area.
6.12.1 is also revised to delete reference to "Health Physicist" and to provide editorial corrections.
6.12.2 is being revised to reflect the "Senior Nuclear Shift Supervisor" title.
Revision 2 of this change request proposed adding the provisions of 10CFR20.203(c) (4), to Section 6.12.1.
These provisions allow a high radiation area established for thirty days or less to be controlled via direct surveillance.
A telephone conference call was held between PSE&G and NRC (NRR and Region I) personnel, on March 13, 1990, to discuss this proposed change.
The consensus reached was that the provisions of 10CFR20.203(c) (4) could be applied to high radiation areas without changing the current Technical
NLR-N90113 Specifications.
Therefore, the change has been removed from this revision of the change request.
PSE&G requests that the agreement that 10CFR20.203(c) (4) does not conflict with the present Technical Specifications be documented in the NRC Safety Evaluation Report for this License Change Request.
- 27. (Unit 1 only).
Revision of Section 6.16 to delete references to past deadlines for Environmental Qualification compliance.
II.
Reason for the Proposed Change License Change Request (LCR) 84-01 was initially submitted to the NRC on March 27, 1984, in response to the revised reporting requirements of Generic Letter 83-43.
Revision 1 to LCR 84-01 was subsequently submitted on October 6, 1986 and added changes to reflect organizational changes, reflect revised primary coolant specific activity reporting requirements of Generic Letter 85-19 and to provide editorial corrections to the Technical Specifications.
Revision 2 was submitted on July 7, 1989 to reflect a recent management reorganization, to propose additional changes to administrative controls and to facilitate NRC review by providing greater description and justification for the changes proposed in the previous revisions.
Revision 3 is being submitted in response to NRC comments and includes additional changes generated by internally PSE&G.
Reasons for the specific changes listed above, excluding editorial changes, are as follows:
- 1.
"REPORTABLE EVENT" is phraseology consistent with Generic Letter 83-43 and 10CFR50.73.
- 2.
The correct multiplier for the peaking factor relationship is 0.3, which was approved per Amendment 6 to the Salem Unit 2 Technical Specifications.
11 0.3 11 was handwritten into the Technical Specifications; this LCR is typing 11 0.3 11 into the appropriate place. Proposed Specification 6.9.1.9 is the appropriate reference regarding the Radial Peaking Factor Limit Report.
- 3.
Due to the implementation of the LER rule (Generic Letter 83-43), Specification 6.9.1.9.p no longer pertains to thirty-day written reports.
Therefore, the references to this Specification are being deleted.
NLR-N90113 4.
The Technical Specifications for steam generators, containment integrity and diesel generators currently reference the prompt notification requirements of Section 6.9.1.
Since Section 6.9.1 is being revised to reflect 10CFR50 reporting requirements (see #20), the reporting requirements of the individual specifications are being revised for consistency.
- 5.
Due to improved fuel design, fuel management and the reporting requirements of 10CFR50.72(b) (1) (ii), the NRC has modified the primary coolant specific activity Technical Specification requirements.
Generic Letter 85-19 provides guidance in this regard and has been used as a model for the applicable proposed revisions in this LCR.
- 6.
The phrase "in lieu of a Licensee Event Report" is superfluous and its removal is consistent with Generic Letter 83-43.
Since Specification 3.11.2.4 applies at all times, Specification 3.0.3 is not applicable to it.
- 7.
The changes are consistent with NUREG 0472, "Standard Radiological Effluent Technical Specifications for Pressurized Water Reactors," Revision 3, draft.
The current values of allowable activity levels for tritium and I-131 are based on 40CFR141 limits for drinking water pathways.
Where such pathways are not potentially affected, it is proposed that the lower limits of detection and reporting levels be increased.
As discussed in the Justification section below, these changes do not affect compliance of PSE&G's Radiological Environmental Monitoring Program with 10CFR20.
Cesium-136 was inadvertently included in table 4.12-1, instead of Cs-134, which is the correct isotope per Regulatory Guide 4.8 and NUREG 0472.
An editorial change is being made accordingly.
- 8.
In the current organization, the General Manager -
Quality Assurance and Nuclear Safety Review has replaced the General Manager - Nuclear Safety Review with regard to responsibility for the Onsite and Offsite Safety Review Groups.
This License Change Request proposes changing the title to "General Manager - Quality Assurance and Nuclear Safety" to more accurately reflect the responsibilities of the position (i.e., safety related review and audit functions).
- 9.
Radiation protection technician" is the title consistent with the current organizational configuration, not "Health physics technician".
"Radiation protection technician" is being added to
NLR-N90113 table 6.2-1 to be consistent with the current requirement (Specification 6.2.2.c) to have a technician present whenever fuel is in the reactor.
Specification 6.2.2.c is being deleted to eliminate redundancy to revised table 6.2-1.
- 10. The Shift Technical Advisors' reporting responsibility is to the Senior Nuclear Shift Supervisor.
The person designated as the Radiation Protection Manager (RPM) is responsible to the General Manager-Salem Operations, for implementation of the Radiation Protection Program.
Therefore, the RPM is qualified per Regulatory Guide 1.8, September 1975.
The Radiation Protection Engineer usually serves as the RPM.
The more generic designated Radiation Protection Manager" is being proposed to allow another qualified person (e.g.,
Radiation Protection/Chemistry Manager) to become the RPM in the absence of the Radiation Protection Engineer.
Responsibility for Fire Brigade training has been transferred from Manager - Nuclear ~raining to Manager -
Site Protection to consolidate administration of the Fire Protection Program.
Appendix A to 10 CFR Part 55 has been deleted.
The requirements for the operator requalif ication program formerly contained in 10CFR55 Appendix A are now contained in 10CFR55.59.
The proposed change references the applicable requirements (i.e., 10 CFR Part 55) and reduces the likelihood of requiring an additional Technical Specification change by not being too specific.
Operator requalification is still performed in accordance with the NRC-approved requalification program.
- 11. The title "Assistant General Manager -
Salem Operations" position has been deleted in the recent reorganization.
"I&C Engineer" is being deleted becauss Instrumentation and Controls is no longer a separate department.
Maintenance and Technical department personnel are responsible for I&C.
The Maintenance Engineer -
Controls and the Technical department I&C Group head are both Salem SORC members.
Radiation Protection/Chemistry Manager is a member of Salem Station management and is therefore being formally added to the SORC membership list.
Since there is now more than one Operating Engineer and more than one Maintenance Engineer, specification 6.5.1.2 is being revised accordingly.
It is also being revised to clarify that Technical department Group Heads (not all engineers in the Technical department) are SORC members.
NLR-N90113 Specification 6.5.1.3 ALTERNATES is being revised to delete the statement that Vice Chairmen shall be members of Station management, which is redundant since Vice Chairmen are listed in 6.5.1.2.
The Senior Nuclear Shift Supervisor position is being included as an alternate to the Operations Manager and Operating Engineers, based on expertise in plant operations.
Also, Specification 6.5.1.3d is being deleted.
This will allow a member and an alternate from the same department to make up part of the voting quorum.
Per proposed Specification 6.5.1.5, the number of allowable alternates per meeting is being increased from two to three.
This increase is intended to make it easier to attain a SORC quorum and to distribute the amount of time spent at SORC meetings among a greater number of Station management personnel.
By placing restrictions on the use of alternate members and quorum composition, the proposed change will require that at least three departments are represented in the voting quorum for any given SORC meeting.
This compensates for the increased number of members from the individual departments resulting from the proposed changes to SORC composition.
- 12. The SORC responsibilities of Specification 6.5.1.6 are being revised to delete references to internal distribution of reports, which is addressed more appropriately in other sections.
For example, Specification 6.5.1.9 addresses distribution of SORC meeting minutes to the General Manager ~ QA and Nuclear Safety Review and to the Vice President and Chief Nuclear Officer; 6.5.2 addresses the Nuclear Safety Department's review responsibilities; 6.5.3 addresses review and control of activities in general.
Therefore, references to SORC interfaces with other Nuclear Department entities are being removed from Specification 6.5.1.6.
Specifications 6.5.1.6(a) and 6.5.3.2(d) are being revised to reflect PSE&G's intention to discontinue use of the significant safety issue determination in favor of consolidating the procedure revision screening and review process under the requirements of 10CFR50.59.
PSE&G is presently finalizing a procedure for performing 10CFR50.59 reviews and safety evaluations, which will provide criteria and examples for determining whether 10CFR50.59 applies to a specific procedure change.
If a 50.59 evaluation is required, the change is presented to SORC and forwarded to Offsite Safety Review (OSR)
- If the originator of the change determines that 10CFR50.59 is not applicable, peer review and appropriate manager approvals are still required prior to implementation,
NLR-N90113 thereby assuring that the correct determination regarding applicability is made.
Although not specifically required to do so by Technical Specifications, the Offsite Safety Review staff periodically reviews procedure changes which did not require 10CFR50.59 safety evaluations.
The proposed changes to Specifications 6.5.1.6(f) and 6.5.1.6(i) more accurately describes SORC's responsibility with regard to special reviews and investigations of Technical Specification violations.
While the General Manager - Quality Assurance and Nuclear Safety may request that SORC perform such a review, SORC is ultimately responsible to the General Manager -
Salem Operations.
current Specifications 6.5.1.6(j), 6.5.1.6(k) and 6.5.1.6(1) require SORC review of all changes to the Security Plan, Emergency Plan and Fire Protection Program Plan, respectively, and their implementing procedures.
The proposed change is intended to waive SORC review if the revisions are not safety significant.
This will eliminate unnecessary SORC reviews of revisions that are editorial or are otherwise insignificant with respect to safety.
- 13. Section 6.5.2 currently describes the Onsite Safety Review Group (SRG) as consisting of the Manager - Onsite Safety Review and four dedicated, full time engineers.
Since the SRG actually consists of the Onsite Safety Review Engineer and three dedicated, full time engineers, PSE&G recently received an NRC violation for changing the SRG without prior Commission approval (Inspection Report 50-272/89-03, 50-311/89-03).
The Onsite Safety Review Engineer has review as well as management responsibilities within the Onsite Safety Review Group (SRG).
PSE&G is proposing to replace the position of Manager - Offsite Safety Review with that of Manager - Nuclear Safety, who will have management oversight for both the Onsite Safety Review Group (SRG) and the Offsite Safety Review staff (OSR).
In this configuration the SRG and OSR staff will each maintain their present staffing level.
- 14. The revision to Specification 6.5.2.3 provides more flexibility to the Nuclear Safety Department with regard to the use of outside consultants/experts.
Per Specification 6.5.2.2, reviewers from organizations external to the Nuclear Safety Department are required to meet the same qualification requirements as the OSR staff.
The miscellaneous activities at the end of Specification 6.5.2.4.1 are actually the responsibility of the Onsite
NLR-N90113 Safety Review Group.
Since 6.5.2.4.1 pertains to the OSR staff, those activities are being deleted.
Specification 6.5.2.4.2(g) is being revised to indicate that OSR review of anticipated deficiencies includes aspects of design or operation that may not necessarily be classified as safety related but may still affect nuclear safety.
- 15. The revision to Specification 6.5.2.4.3(i) provides more specific information regarding the personnel who perform the fire protection audits.
- 16. The record requirements currently described for Nuclear Safety Review are relevant to Offsite Safety Review activities in particular. Relaxing the time constraints for internal distribution of the results of reviews performed per Specification 6.5.2.4 and audit reports performed by consultants will provide schedular flexibility in report preparation without compromising management oversight of the audit and review process.
- 17. "NSR" is being replaced by "the OSR staff" in Section 6.5.3 because the 10CFR50.59 review activities described therein are performed specifically by the Offsite Safety Review (OSR) staff, which is part of the Nuclear Safety Department.
- 18. The changes are administrative in nature and are consistent with Generic Letter 83-43 where applicable.
References to the initial submittal of the Annual report are being deleted since the reports have already been submitted for each Salem unit.
- 19. "Self reading dosimeter" is more accurate than "film badge" when describing dosimetry used at Salem.
In addition to Thermoluminescent dosimeters, Radiation Workers at Salem use either SRD, ALNOR or digital alarming dosimetry when appropriate.
See change number 4 above.
High primary coolant specific activity is no longer classified as a REPORTABLE OCCURRENCE; it will be reported on an annual basis consistent with Generic Letter 85-19.
- 20. 10CFR50.73(g) states that "The requirements contained in this section replace all existing requirements for Licensees to report REPORTABLE OCCURRENCES as defined in individual plant Technical Specifications."
Therefore, since Specifications 6.9.1.7, 6.9.1.8 and 6.9.1.9 are superseded, they are being deleted.
NLR-N90113 21. This revision provides more specific information on radiological sampling location maps.
Dedicating one map to locations near the site boundary and another to include all locations assures that the maps will be scaled such that they are legible.
This change also provides an editorial correction by deleting a footnote that was inadvertently included in Specification 6.9.1.7.
- 22. The Radial Peaking Factor (F
) Limit Report is currently required to be sub~rtted to the NRC sixty days prior to the date the limit would become effective. In order to provide schedular flexibility in the fuel cycle licensing process, th1s change* proposes allowing PSE&G to submit the report upon issuance.
"Upon Issuance" is defined as not later than the first application of the revised Fxy limits, which is generally coincident with taking the core flux map at 25% power.
Surveillances performed on F confirm the acceptability of the corresponding heat flU~ hot channel factor (F0 )
values.
The F limits are determined prior to restart (i.e., in conjfiiction with the Reload Safety Evaluation).
Furthermore, since the report is submitted on an information only basis and not for NRC approval, changing the reporting schedule constitutes a purely administrative change.
- 23. "Reportable Event" is phraseology consistent with Generic Letter 83-43 and 10CFR50.73.
Currently, records of reactor tests and experiments are addressed in both Section 6.10.1 and 6.10.2.
Removing it from 6.10.1 eliminates redundancy and does not affect the commitment to retain the records for the duration of the plant operating license.
10CFR50.59(b) (3) defines the requirements for retaining records of 10CFR50.59 reviews.
Deleting Specification 6.10.2(j) and adding 6.10.l(j) is consistent with 10CFR50.59(b) (3).
- 24.
110SR 11 is the applicable group in the current organizational configuration; the Nuclear Review Board no longer exists.
- 25. Basing an area's radiation level on measurement 18 inches from the radiation source provides a uniform means of measurement consistent with that of the Hope Creek station.
"Radiation Protection" personnel and "Senior Nuclear Shift Supervisor" are the applicable titles in the current organizational configuration.
NLR-N90113 26. "Updated Final Safety Analysis Report" (UFSAR) is the document relevant to Section 6.15.1.
- 27. Salem Unit l's compliance with Environmental Qualification commitments is documented via docketed correspondence, and references to past deadlines for compliance are not necessary.
III. Justification for the Proposed Change The changes proposed in this LCR may be categorized and justified as follows:
- a.
Changes to comply with Generic Letter 83-43 (item nos.
1,3,4,6,18,20,23).
These changes are consistent with the guidance provided in the Generic Letter.
They are intended to reflect compliance with the universally adopted reporting requirements of 10CFR50.72 and 10CFR50.73, and are administrative in nature.
- b.
Changes related to PSE&G's organizational configuration, which reflect titular changes, changes in management responsibility, changes to the SORC and those which supplement LCR 88-07 (item nos. 8,9,10,11,12,13,17,24).
This category is administrative in nature and does not adversely impact management attentiveness to safe operation of the Salem Generating Station.
The recent reorganization of PSE&G's Nuclear Department is intended to increase overall management effectiveness, in some cases by consolidating oversight of related activities.
For example, fire brigade training and implementation of the fire protection program in general will now be administered by the same manager.
Also, quality assurance and nuclear safety review and audit activities will both be under the direction of one general manager.
The Manager - Nuclear Safety will assume management oversight of the Onsite Safety Review staff, thereby improving communication within the Nuclear Safety organization.
The changes to SORC composition and voting quorum requirements will increase the level of assurance that the various Station departments are adequately represented in the SORC quorum.
By placing a two vote per department limit on the quorum and restricting the use of alternates, at least three departments will be represented in the quorum of each SORC meeting
- This category includes the proposed change to require SORC review for procedure changes and changes to the Security Plan, Emergency Plan and Fire Protection
NLR-N90113 Program Plan only if a 10CFR50.59 safety evaluation is involved.
This approach will consolidate the screening and review process for procedure changes by doing away with the significant safety issue determination currently in use at the Salem Generating Station.
Screening for safety significance will be performed by determining 10CFR50.59 applicability (plus 10CFR50.54(p) for Security Plan and 10CFR50.54(q) for Emergency Plan changes), which is consistent with NRC regulations regarding procedure changes (10CFR50.59 and 10CFR50.54).
- c.
Changes consistent with Generic Letter 85-19, which help streamline reporting requirements by allowing high primary coolant activity to be reported annually rather than on a short term basis (item nos. 5,19).
This category is consistent with the model Technical Specifications appended to Generic Letter 85-19.
- d.
Change in the schedule for submitting the Radial Peaking Factor Limit Report (F
), which would prevent the report from becoming ax~onstraint on the restart schedule without affecting the development of the F limits or their use in assuring that F0 provides th~y margin of safety required by the reloaa safety evaluation (item nos. 2,22).
- e.
Changes related to Radiation Protection and Radiological Monitoring (item nos. 7,21,25).
PSE&G's current practice of preparing and submitting the Annual Radiological Environmental Operating Report includes the two sampling location maps explicitly described by the proposed change.
This LCR clarifies the current practice, which is acceptable to the NRC, in order to remove any ambiguity from the Technical Specifications.
Radiation measurement will be made at 18 inches from the source in order to utilize a consistent means for defining a radiation area, high radiation area, etc.
This approach is currently used at the Hope Creek Generating Station and is based on the minimum expected distance between the radiation source and the radiation worker, excluding extremities.
Consistent with the practices used to keep exposures ALARA, identification of hot spots, shielding and minimization of exposure time (especially at close distances) are used to reduce undue exposures.
Changes to reporting levels of radioactivity concentrations and lower limits of detection (LLD) for the radiological environmental monitoring program are consistent with NUREG 0472, "Standard Radiological Environmental Technical Specifications for Pressurized water Reactors", Revision 3, draft.
These changes
NLR-N90113 propose allowing higher reporting levels and lower limits of detection for tritium and I-131 if a drinking water pathway is not potentially affected by the effluent being monitored.
The provisions of 40CFR141 will still be complied with where applicable.
10CFR20.106 requires that effluents released to unrestricted areas are maintained within the limits of Appendix B, Table6II of 10CFR20.
The Table II limit for tritium is 3 x 10 pCi/l.
The proposed reporting level and LLD for tritium are respectively 1% and one-tenth of 1% of the Table II limit.
Table II specifies a limit of 300 pCi/l for soluble I-131 and 60,000 pCi/l for the insoluble form.
Using 300 pCi/l for comparison purposes, the proposed reporting level and LLD for I-131 are respectively 6.7% and 3.3% of the Table II limits.
Therefore, the proposed changes will not affect
- compliance with 10CFR20.106 and will not allow for an increase an radiation dose to any member of the public.
- f.
Changes to the description of Nuclear Safety Review (NSR) responsibilities (item nos. 13,14,15,16,17).
These changes include consolidating the management of the Offsite and Onsite Safety Review Groups and revising the description of NSR activities to increase
- specificity and eliminate redundancy, and relaxing schedular constraints for internal distribution of audit reports.
Changes in this category will not lessen the scope of NSR activities and will increase the effectiveness of the nuclear safety review function by consolidating the SRG and the OSR staff under a single manager.
This organizational arrangement will provide for improved communication and more effective utilization of resources.
None of the changes in this category will reduce the effectiveness of NSR review and audit functions.
The relaxation of schedular requirements for internal distribution of audit reports will increase the time allotted for report preparation and review.
Management oversight of the audit process will not be reduced, especially since any potentially significant findings requiring immediate attention may be brought to the attention of the Vice President and Chief Nuclear Officer prior to formal issuance of the audit report.
- g.
Changes deleting references to past deadlines and outdated requirements or documents are justified on the basis that they are largely editorial and provide clarification without reducing any commitments (item nos. 10,18,26,27).
NLR-N90113 IV.
Significant Hazards Consideration Analysis The proposed changes to the Technical Specifications:
- 1.
Do not involve a significant increase in the probability or consequences of an accident previously evaluated.
Those proposed changes which are administrative in nature do not impact any accident analyses used to support operation of the Salem Nuclear Generating Stations.
Furthermore, the proposed changes do not involve any reduction in management effectiveness, nor do they adversely affect the design or operation of any systems or components important to safety.
Consequently, the reliability of the performance of plant safety functions is not adversely affected.
The proposed changes to Section 3.4.8 of Unit 1 and 3.4.9 of Unit 2 delete the action statement requiring discontinuation of operation if primary coolant specific activity exceeds 1.0 µCi/g for 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> (or 10% of unit total annual operating time) during any 12 mQnth period.
The activity limits of Figure 3.4-1 and 100/E µCi/g will still apply.
These limits are bounded by the assumptions used in the steam generator tube rupture accident analysis described in UFSAR Section 15.4.4, which bases primary coolant activity on 1% defective fuel cladding.
Therefore, the Technical Specifications will continue to assure that the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> exclusion area dose in the event of a steam generator tube rupture accident analysis would be a small fraction of the 10CFR100 limits.
Furthermore, deletion of the 800 hour0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> limit has been endorsed by the NRC in Generic Letter 85-19.
The NRC's position is predicated on promulgation of 10CFR50.72 reporting requirements and improvements made in fuel management subsequent to initial development of the primary coolant specific activity limits.
Generic Letter 85-19 states that "[l]icensees are expected to continue to monitor activity in the primary coolant and *** maintain it at a reasonably low level (i.e., accumulated time with high iodine activity should not approach 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />)".
PSE&G recognizes the significance of coolant activity and will continue to monitor and control accumulated specific activity levels.
The proposed increases to the reporting levels of radioactivity and lower limits of detection for tritium and I-131 (Section 3/4.12) do not increase the probability or consequences of an accident.
Since the current values are based on 40CFR141 requirements for drinking water supplies, this proposed change revises those levels in cases where no drinking water pathways are potentially affected.
The
NLR-N90113 reporting levels and LLD for tritium and I-131 will continue to be a small fraction of the levels allowed by 10CFR20.
Therefore, it may be concluded that the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2.
Do not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes do not adversely affect the design or operation of any systems or components important to safety.
No physical plant modifications or new operational configurations will result from these proposed changes.
Therefore, it may be concluded that the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3.
Do not involve a significant reduction in a margin of safety.
With the exception of the deletion of the 800 hour0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> limit on primary coolant specific activity and revision of radiological environmental monitoring parameters discussed above, the changes propose herein potentially affecting parameters defining margins of safety affect only their administrative aspects and do not allow for any reduction in a margin of safety.
The proposed changes to Specification 6.9.1.9 regarding F reporting requirements will not reduce the margin of safet~Yassociated with the heat flux hot channel factor (F0 ), since the same methodology will be used to define the acceptable limits.
The proposed change to specific activity limits does not involve a significant reduction in a margin of safety.
The limits of Figure 3.4-1, which are bounded by the assumption of 1% defective fuel clad for the steam generator tube rupture accident analysis, will still apply.
Also, the change has been endorsed by the NRC via Generic Letter 85-19.
The proposed change to the radiological environmental monitoring program does not involve a significant reduction in a margin of safety.
The program will st.ill implement Section IV.B to 10CFR50, Appendix I.
As stated above, the proposed changes do not affect compliance with 10CFR20.
The proposed change deletes radioactivity concentration values associated with 40CFR141 in cases where it is not applicable (i.e., no drinking water pathways are involved).
Therefore, it may be concluded that the proposed changes do not involve a significant reduction in a margin of safety.
NLR-N90113 v.
Conclusions As discussed above, PSE&G has concluded that the proposed changes to the Technical Specifications do not involve a significant hazards consideration since the changes (i) do not involve a significant increase in the probability or consequences of an accident previously evaluated, (ii) do not create the possibility of a new or different kind of accident from any accident previously evaluated, and (iii) do not involve a significant reduction in a margin of safety.
NLR-N90113 ATTACHMENT 2