Text

Technical Specifications Amendment Request to Relocate Administrative Titles and Responsibilities and Other Administrative Changes
	ML043560279
Person / Time
Site:	Pilgrim
Issue date:	12/14/2004
From:	Balduzzi M; Entergy Nuclear Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NUREG-1433
	Download: ML043560279 (45)
	v • d • e

I II U-

':~ En tergy Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 Michael A. Balduzzi Site Vice President December 14, 2004 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001

SUBJECT:

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station Docket No. 50-293 License No. DPR-35 Technical Specifications Amendment Request to Relocate Administrative Titles and Responsibilities and Other Administrative Changes

REFERENCE:

NUREG-1 433, Standard Technical Specifications for General Electric Plants, BWR/4, Revision 3 LETTER NUMBER: 2.04.083

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Nuclear Operations, Inc. (Entergy) hereby proposes to amend its Facility Operating License, DPR-35. These changes are consistent with the content in Standard Technical Specifications (NUREG-1433, Revision 3) and changes previously approved by the NRC for other facilities. Entergy has reviewed the proposed amendment in accordance with 10 CFR 50.92 and concludes it does not involve a significant hazards consideration.

Commitments made by the licensee in this letter are listed in Attachment 2.

Entergy requests approval of the proposed amendment by December 30, 2005. Once approved, the amendment shall be implemented within 60 days.

CG 204083

Entergy Nuclear Operations, Inc. Letter Number: 2.04.083 Pilgrim Nuclear Power Station Page 2 If you have any questions or require additional information, please contact Bryan Ford at (508) 830-8403.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the

//' day of December 2004.

Sincerely, Michael A. Balduzzi ES/dm

Enclosure:

Evaluation of the proposed change - 11 pages Attachments: 1. Proposed Technical Specification and Bases Changes (mark-up) -

27 pages

2. List of Regulatory Commitments - 1 page cc: Mr. Robert Fretz, Project Manager Ms Cristine McCombs, Director Office of Nuclear Reactor Regulation Mass. Emergency Management Mail Stop: 0-8B-1 Agency U.S. Nuclear Regulatory Commission 400 Worcester Road 1 White Flint North Framingham, MA 01702 11555 Rockville Pike Rockville, MD 20852 Senior Resident Inspector Mr. Robert Walker, Director Pilgrim Nuclear Power Station Massachusetts Department of Public Health Radiation Control Program 90 Washington Street Dorchester, MA 02121 U.S. Nuclear Regulatory Commission Region 1 475 Allendale Road King of Prussia, PA 19408 204083

ENCLOSURE EVALUATION OF THE PROPOSED CHANGE

ENCLOSURE Evaluation of the Proposed Change

Subject:

Technical Specifications Amendment Request to Relocate Administrative Titles and Responsibilities and Other Administrative Changes

1. DESCRIPTION

2. PROPOSED CHANGES

3. BACKGROUND

4. TECHNICAL ANALYSIS

5. REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration 5.2 Environmental Consideration

6. PRECEDENTS

7. REFERENCES

Letter 2.04.083 Enclosure Page 1 of 11

1. Description Entergy Nuclear Operations, Inc. (Entergy) is requesting to amend Operating License DPR-35 for Pilgrim Nuclear Power Station (PNPS). The proposed changes would revise the Operating License, Technical Specifications (TS):

(1) To eliminate certain administrative requirements for Safety Limit violations that are adequately addressed in 10 CFR 50.36(c)(1)(i)(A), 10 CFR 50.72, 10 CFR 50.73, and by procedures. Elimination of duplicative regulatory reporting requirements will avoid future, and eliminate existing, inconsistent or conflicting regulatory requirements.

(2) To replace plant-specific titles with generic titles. Actual plant-specific titles that fulfill the generic titles will be relocated from the TS to the Final Safety Analysis Report (FSAR),

which will facilitate future Pilgrim organization title changes.

(3) To remove the remaining responsibilities of the Operations Review Committee (ORC).

(4) To replace descriptive details specified in TS 3.13.A.1 associated with 10 CFR 50.55a(f),

"Inservice testing requirements," with reference to the "Inservice Code Testing Program."

Similar detail from TS 4.13.A.1 replaced with editorial re-wording to more closely match presentation in NUREG-1433 Specification 5.5.7, "Inservice Testing Program."

(5) To make administrative changes to TS 5.5.4, "Radioactive Effluent Controls Program," to more closely match presentation in NUREG-1433 Specification 5.5.4.

(6) To make editorial corrections and clarifications.

These proposed changes are considered administrative and will enhance consistency with the BWR/4 Standard Technical Specifications, NUREG-1 433.

Entergy requests approval of the proposed amendment by December 30, 2005. Once approved, the amendment shall be implemented within 60 days.

2. Proposed Changes 2.1 Delete the following administrative reporting and restart authorization requirements that apply in the event of a Safety Limit violation, and editorial rewording to reflect these deletions:

1. TS Section 2.2.1, 'Within one hour notify the NRC Operations Center in accordance with 10CFR50.72."

2. TS Section 2.2.3, 'The Station Director and Senior Vice President - Nuclear and the Nuclear Safety Review and Audit Committee (NSRAC) shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ."

3. TS Section 2.2.4, "A Licensee Event Report shall be prepared pursuant to 10CFR50.73.

The Licensee Event Report shall be submitted to the Commission, the Operations Review Committee (ORC), the NSRAC and the Station Director and Senior Vice President -

Nuclear within 30 days of the violation."

4. TS Section 2.2.5, "Critical operation of the unit shall not be resumed until authorized by the Commission."

Letter 2.04.083 Enclosure Page 2 of 11 2.2 Replace the following plant-specific titles with generic titles as shown, and include TS 5.2.1 requirement to retain specific titles of those personnel fulfilling the responsibilities in the Final Safety Analysis Report (FSAR):

1. TS 5.1.1, "Station Director" is replaced with "plant manager" (in two locations).

2. TS 5.1.2, "Nuclear Operations Supervisor (NOS)" and "NOS" are replaced with "control room supervisor (CRS)" and "CRS" (three locations).

3. TS 5.2.1.a, last sentence, is revised to state: "These requirements, including the plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, shall be documented in the Pilgrim Station Final Safety Analysis Report (FSAR)."

4. TS 5.2.1.b, "Station Director" is replaced with "plant manager."

5. TS 5.2.1.c, "The Vice President - Operations" is replaced with "A specified corporate officer."

6. TS 5.2.2.h, "Operations Department Manager" is replaced with "operations manager or assistant operations manager" and the specific position titles "Nuclear Watch Engineers,"

Nuclear Operations Supervisors," and "Nuclear Plant Operators" are removed.

7. TS 5.2.2.i, "The Shift Control Room Engineer (SCRE)" is replaced with "An individual' (similar change in three locations). "Nuclear Operations Supervisor (NOS)" is replaced with "unit operations shift crew." Other editorial changes are made for consistency.

8. TS 5.5.1, "the approval of the Chemistry and Radiological Department Managers" is replaced with "the approval of the plant manager."

9. TS 5.7.1, "Health Physics personnel" is replaced with "radiation protection personnel."

10. TS 5.7.1.c, "Radiation Protection Manager" is replaced with "radiation protection manager."

11. TS 5.7.2, "the Nuclear Watch Engineer on duty" is replaced with 'an SRO on duty."

12. TS 5.7.2, "health physics supervision" is replaced with "radiation protection supervision."

2.3 Remove Operations Review Committee (ORC) responsibilities as indicated:

1. TS 3.7.A.2.b, Footnote *, remove "ORC approved" criterion for the stated administrative control.

2. TS 5.5.1, remove "review and acceptance by the Operations Review Committee and."

2.4 Modify the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) inservice testing requirements as indicated:

1. For the definition of REFUELING INTERVAL, replace "ASME Code,Section XI IWP and IWV' with "Inservice Code Testing Program."

Letter 2.04.083 Enclosure Page 3 of 11

2. For TS 3.13 and for 4.13 Applicability, delete "or equivalent". For TS 3.13 Objective, delete

"(safety related) or equivalent (important to safety)". For TS 3.13.A.1 and for 4.13 Objective, replace "safety and safety related" with UASME Code Class 1, 2, and 3."

3. For TS 3.13.A.1 and for 4.13.A.5, insert "Inservice Code Testing Program" in place of "ASME Boiler and Pressure Vessel Code." Also for TS 3.13.A.1 delete the follow on descriptive detail: "Section Xl 'Rules for Inservice Testing of Nuclear Power Plant Components" Subsections IWP and IWV as required by 10CFR50.55a(f), except where specific relief has been granted by the NRC pursuant to 10CFR50.55a(f)(6)(i)."

4. For TS 4.13.A.1 and 4.13.A.2 combine as TS 4.13.A.1. Delete the text of TS 4.13.A.1 and replace the 4.13.A.2 introduction "Test Frequencies for Code" with 'The ASME OM Code" (Note that the uCode Terminology" and associated "Frequencies" Table remains unchanged).

5. Similarly, Bases detail is corrected and draft changes are provided for information.

2.5 Make the following administrative changes to TS 5.5.4, "Radioactive Effluent Controls Program":

1. TS 5.5.4.b, replace "10 CFR 20, Appendix B, Table 2, Column 2" with "ten times the concentration values in Appendix B, Table 2, Column 2, to 10 CFR 20.1001 - 20.2402."

2. TS 5.5.4.e, delete "and projected dose" from the current sentence and add second sentence "Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days."

3. TS 5.5.4.g, reword "effluents to areas beyond" adding clarifying phrases to read "effluents from the site boundary to areas at or beyond."

4. TS 5.5.4.j, add "beyond the site boundary" after "member of the public."

2.6 Make the following editorial changes, corrections, or clarifications:

1. Remove the ** note from TS 3.7.A.5.

2. TS 3.5.A.5 misspelled word "rector" is corrected to "reactor."

3. TS 3.8.2, on page 3/4.8-2, delete one of two periods in the APPLICABILITY statement.

Additionally, the misspelled word "CHANNEl" in TS 4.8.2.3 is corrected to "CHANNEL."

4. TS 4.9.A.c, on page 3/4.9-2, has a typo in the last sentence reference 4.9.A.1.b.1, which is corrected to 4.9.A.1.b.2.

5. TS 3.9.B.2 last sentence on page 3/4.9-4, "and the NRC is notified within one (1) hour as required by 10 CFR 50.72" is deleted.

6. TS 4.9.A.4.b, on page 3/4.9-4, correct the abbreviation for the unit hertz to "Hz" (without subscripting the "z").

7. TS 3.11.C.2, on page 3/4/11-3, correct the typographical reference to Table "3.3-1" from "3.3.1."

Letter 2.04.083 Enclosure Page 4 of 11

3. Background These proposed changes are consistent with the latest revision of the BWR/4 Standard Technical Specifications, NUREG-1433 (Revision 3, dated 3/31/2004). The proposed changes are also consistent with specific changes that have been made to NUREG-1433 since its initial issuance as Revision 0, dated 9/28192.

1. Removal of the administrative reporting and restart authorization requirements, that apply in the event of a Safety Limit violation, have been specifically addressed by the Technical Specification Task Force (TSTF) in change TSTF-5, which was NRC approved on June 11, 1996. The basis for this change was that the requirements are addressed in 10 CFR 50.36(c)(1)(i)(A), which requires notification and reporting in accordance with 10 CFR 50.72 and 50.73, and Commission approval for resuming operation.

2. Use of generic titles and removal of plant-specific titles has been specifically addressed in change TSTF-65, which was NRC approved on December 2,1997. The acceptance of this change was based on the commitment to relocate and control plant specific titles in the FSAR. This change did not eliminato any qualifications, responsibilities or requirements for these positions.

3. Removing responsibility details of the ORC was generally endorsed by the NRC in a letter from William T. Russell (NRC) dated October 25,1993. The acceptance of this change was based on concluding that specific requirements were not necessary to be included in TS to meet 10 CFR 50.36(c)(5), which states: 'Administrative controls are the provisions related to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.' ORC responsibilities are maintained within plant procedures, consistent with the Quality Assurance Program commitments, adequately assures safe operation.

4. Section 50.55a of 10 CFR requires that IST of certain ASME Code Class 1, 2, and 3 pumps and valves be performed in accordance with the ASME OM Code and applicable addenda, except where alternatives have been authorized or relief has been requested by the licensee and granted by the Commission pursuant to paragraphs (a)(3)(i), (a)(3)(ii),or (f)(6)(i) of 10 CFR 50.55a. In accordance with 10 CFR 50.55a(f)(4)(ii), licensees are required to comply with the requirements of the latest edition and addenda of the ASME Code incorporated by reference in the regulations 12 months prior to the start of the subsequent 120-month IST program intervals. Accordingly, licensees whose subsequent 120-month (10-year) IST program interval began after November 22, 2000, are required to comply with the 1995 Edition with the 1996 Addenda of the ASME OM Code. Similarly, licensees whose 120-month (10-year) IST program interval began after October 28, 2003, are required to comply with the 1998 Edition through 2000 Addenda of the ASME OM Code. In accordance with 10 CFR 50.55a(f)(4)(iv), licensees may use portions of subsequent editions and addenda provided that all related requirements of the respective edition and addenda are met.

In proposing alternatives or requesting relief, the licensee must demonstrate that: (1) the proposed alternatives provide an acceptable level of quality and safety; (2) compliance would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety; or (3) conformance is impractical for the facility. Section 50.55a of 10 CFR authorizes the Commission to approve alternatives and to grant relief from ASME Code requirements upon making necessary findings. NRC guidance contained in Generic Letter (GL) 89-04, "Guidance on Developing Acceptable Inservice Testing Programs,"

Letter 2.04.083 Enclosure Page 5 of 11 provides acceptable alternatives to ASME Code requirements. Further guidance is given in GL 89-04, Supplement 1, and NUREG-1482, "Guidance for Inservice Testing at Nuclear Power Plants."

Amendment 149 to PNPS TS, dated September 28, 1993, created Specification 3/4.13, "Inservice Code Testing." This specification was based on similarity to the previous BWR/5 Standard TS (NUREG-01 23) Section 4.0.5.b. Since then, various administrative and editorial changes have been made to the Standard TS for the Inservice Testing Program that reflect changes to the regulations and guidance found acceptable to the Commission. Based on minimizing duplication with the Regulations, the current BWR/4 Standard TS, NUREG-1433, Revision 3, Specification 5.5.7 has eliminated explicit reference to ASME subsections and explicit reference to regulations governing relief from the Code.

5. The administrative clarifications proposed for TS 5.5.4, "Radioactive Effluent Controls Program," were specifically addressed as part of change TSTF-285, which was NRC approved on June 29,1999.

Each of these changes have been incorporated into the most recent issued revision of the Standard TS NUREG-1433, Revision 3, and have been NRC reviewed and approved on other dockets as acceptable administrative changes with no adverse impact on the health and safety of the public.

4. Technical Analysis The proposed changes (1) to remove the administrative reporting and restart authorization requirements that apply in the event of a Safety Limit violation; (2) to replace plant-specific titles with generic titles; (3) to remove responsibilities of the ORC; (4) to delete regulatory detail for the Inservice Code Testing Program; and (5) other administrative corrections and clarifications; are administrative with no technical change in requirements. As such, no specific regulatory requirements or guidance applies. Additionally, the changes are consistent with the latest revision of the BWR/4 Standard Technical Specifications, NUREG-1433 (Revision 3, dated 3/31/2004).

4.1 TS Section 2.2 provides notification, reporting, and restart requirements to be met in the event of a Safety Limit violation. TS Sections 2.2.1, 2.2.4, and 2.2.5, which are proposed for deletion, are addressed by the requirements of 10 CFR 50.36(c)(1)(i)(A). Furthermore, TS Section 2.2.1 is addressed by 10 CFR 50.72 and TS 2.2.4 is addressed by 10 CFR 50.73; however, the TS 2.2.4 30-day requirement to submit the Licensee Event Report (LER) is no longer consistent with the latest provisions of 10 CFR 50.73, which allow 60-day reporting. This change will correct that inconsistency.

Also proposed for deletion is TS 2.2.3, which directs notification of the Station Director, Vice President - Nuclear, and the Nuclear Safety Review and Audit Committee within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Assurance of these administrative notifications is adequately controlled by plant procedures.

TS 2.2.2.A and 2.2.2.B will be renumbered to 2.2.1 and 2.2.2 because of the deletion of the above TS sections.

Removal of duplicative reporting requirements from the Technical Specifications results in simplification of the Technical Specifications and Bases and less administrative burden to track duplicative reporting requirements. Adequate administrative controls exist in administrative

Letter 2.04.083 Enclosure Page 6 of 11 programs at Pilgrim for the identification and necessary reporting of safety limit violations in accordance with 10 CFR 50.36, 10 CFR 50.72 and 10 CFR 50.73.

In summary, the necessary notification, reporting, and restart requirements to be met in the event of a Safety Limit violation are adequately addressed by existing regulations and plant procedures. As such, these changes are administrative with no technical change in requirements.

4.2 Replacing plant-specific titles with generic titles, and including a TS commitment (in TS 5.2.1) to retain specific titles in the Final Safety Analysis Report (FSAR) of those personnel fulfilling the responsibilities does not eliminate any qualifications, responsibilities or requirements for these positions. Members of the plant staff assigned to these positions shall continue to meet or exceed the minimum qualifications required by TS 5.3, "Unit Staff Qualifications."

Any change of the relocated specifications in the FSAR will be strictly controlled in accordance with the provisions of 10 CFR 50.59, "Changes, tests, and experiments" to determine if the proposed changes will require prior NRC review and approval. Additionally, reporting of any changes to the NRC will be made in accordance with 10 CFR 50.71 (e), "Maintenance of records, making of reports."

Additional administrative plant staff position clarifications outlined in Section 2, Proposed Changes, are also consistent with NUREG-1433, and are discussed below.

In TS 5.2.2.h, the requirement for an Operations Department management position to hold a senior reactor operator license is clarified to include the flexibility of the "operations manager or assistant operations manager." Since both positions are responsible for directing the licensed activities of licensed operators, there is no adverse impact to safe plant operations due to this change.

In TS 5.2.2.h, the discussion of the specific position titles of "Nuclear Watch Engineers" and "Nuclear Operations Supervisors" holding a senior reactor operator (SRO) license, and the "Nuclear Plant Operators" holding a reactor operator (RO) license is also eliminated. The generic requirements for SRO and RO on-shift positions are adequately addressed in TSs 5.2.2.b, 5.2.2.c, and 5.2.2.e, as well as 10 CFR 50.54(k), 50.54(l), and 50.54(m).

Elimination of these plant-specific titles from this Section is consistent with the intent of replacing plant-specific titles with generic titles.

In TS 5.5.1, the required management level for approval of the changes to the ODCM is made more restrictive by replacing the "Chemistry and Radiological Department Managers" with "the plant manager." This change is made for consistency with NUREG-1 433, replaces plant-specific titles with generic titles, and does not preclude the continued approvals of the Chemistry and Radiological Department Managers. As such, there is no adverse impact to safe plant operations due to this change.

In TS 5.7.1 and 5.7.2 reference to "health physics" personnel / supervision is replaced with "radiation protection" to more appropriately reflect the departmental responsibilities. The title case presentation of the "Radiation Protection Manager" in 5.7.1.c is made a generic (i.e., lower case) title "radiation protection manager" consistent with other changes to generic titles.

In TS 5.7.2, "the Nuclear Watch Engineer on duty" is replaced with "an SRO on duty." The NUREG-1433 presentation of "shift supervisor" suggests the equivalent Pilgrim position of NOS.

The proposed change allows maintaining the current requirement for Nuclear Watch Engineer

Letter 2.04.083 Enclosure Page 7 of 11 (i.e., shift manager) to retain this responsibility, but also allows for future procedure revision to assign this responsibility to the NOS if desired. Since there is no actual change to existing requirements, and the possible allowed future change is consistent with the standard TS, there is no adverse impact to safe plant operations due to this change.

In summary, the necessary qualifications, responsibilities or requirements for these positions are adequately addressed by existing regulations and regulatory controls imposed for future changes to the FSAR. As such, these administrative changes do not adversely impact the public health and safety.

4.3 The Operations Review Committee (ORC) responsibilities were relocated from the Pilgrim TS in Amendment 177 on July 31,1998. However, two references to ORC review and approval responsibilities were overlooked for concurrent relocation.

TS 3.7.A.2.b, Footnote *, references ORC approval of the administrative controls used to intermittently open primary containment isolation valves closed to satisfy TS required actions.

The corresponding allowance in NUREG-1433, TS 3.6.1.3, Actions Note 1, does not include any reference to approval authority for the administrative control. Also, TS 5.5.1.b specifies requirements for implementing licensee-initiated changes to the Offsite Dose Calculation Manual (ODCM), which include "review and acceptance by the Operations Review Committee."

The corresponding requirement in NUREG-1433, Specification 5.5.1.b, does not specify the details of programmatic review(s) - only the final approval required by the plant manager.

In summary, the necessary ORC responsibilities are adequately addressed in licensee controlled documents, without explicit TS requirements, as previously approved by the NRC. As such, these changes are administrative with no technical change in requirements.

4.4 The proposed changes to the REFUELING INTERVAL definition and to TS 3/4.13.A reflect administrative changes only. References to ASME Section Xl are revised to reflect current regulations and ASME OM Code and the PNPS specific program name, "Inservice Code Testing Program." The administrative reference to usafety and safety related" and "or equivalent" is corrected to match the regulation, which specifically addresses Code Class 1, 2, and 3 pumps and valves. Since the regulations of 10 CFR 50.55a already adequately enforce the requirements, eliminating detailed reference to the regulation, and explicit reference to regulations governing relief from the Code, the proposed Specification retains only the specific performance frequency definitions for Code terminology. These changes result in a Specification essentially equivalent to the BWR/4 Standard TS (NUREG-1 433) Specification 5.5.7, "Inservice Testing Program." This change does not impact the April 30, 2004 NRC review of the PNPS 4th 10-Year Inservice Code Testing Program, which remains the basis for the current program implementation.

4.5 The following administrative changes involve no technical change and serve to enhance the consistency of the PNPS TS with the NUREG-1 433 Standard TS for consistent use and application for the PNPS operating staff and NRC regulator:

1. TS 5.5.4.b, replace "10 CFR 20, Appendix B, Table 2, Column 2" with "ten times the concentration values in Appendix B, Table 2, Column 2, to 10 CFR 20.1001 - 20.2402."

These values provide reasonable assurance that the levels of radioactive materials in bodies of water in unrestricted areas will result in exposures within (1) the Section IL.A design objectives of appendix I to 10 CFR Part 50 and (2) restrictions authorized by 10 CFR 20.1301 (e).

Letter 2.04.083 Enclosure Page 8 of 11 The existing PNPS TS 5.5.4.b, references the old Part 20.1 - 20.602, Appendix B, Table II (typographically presented as 'Table 2"), as allowed by 10 CFR 20.1008.

Current requirements for the content of TS concerning radioactive effluents are contained in 10 CFR 50.36a. 10 CFR 50.36a requires licensees to maintain control over radioactive material in gaseous and liquid effluents to unrestricted areas, produced during normal reactor operations, including expected occurrences, to levels that are as low as reasonably achievable (ALARA). For power reactors, Appendix I to 10 CFR Part 50 contains the numerical guidance to meet the ALARA requirement. The dose values specified in Appendix I of 10 CFR Part 50 are small percentages of the implicit limits in the old 10 CFR 20.106 and the explicit limits in 10 CFR 20.1301. As secondary controls, the instantaneous concentration release rates required by this TS were chosen by the NRC to help maintain annual average releases of radioactive material in gaseous and liquid effluents to within the dose values specified in Appendix I of 10 CFR Part 50. For the purposes of STS 5.5.4.b, 10 CFR Part 20 is used as a source of reference values only. These TS requirements allow operational flexibility, compatible with considerations of health and safety, which may temporarily result in release rates which, if continued for the calendar quarter, would result in radiation doses higher than specified in Appendix I of 10 CFR Part 50. However, these releases are within the implicit limits in the old 10 CFR Part 20.106 and the explicit limits in 10 CFR Part 20.1302, which references 10 CFR Part 20, Appendix B, concentrations. These referenced concentrations in the old 10 CFR Part 20 are specific values, which relate to an annual dose of 500 mrem.

The liquid effluent radioactive effluent concentration limits given in Appendix B, Table 2, Column 2 to 10 CFR 20.1001 - 20.2402 are based on an annual dose of 50 mrem total effective dose equivalent. Since an instantaneous release concentration corresponding to a dose rate of 500 mrem/year has been acceptable as a TS limit for liquid effluents, which applies at all times to assure that the values in Appendix I of 10 CFR Part 50 are not likely to be exceeded, it is not necessary to reduce this limit by a factor of 10.

The use of effluent concentration values that are 10 times those listed in Appendix B, Table 2, Column 2 to 10 CFR 20.1001 - 20.2402 will not have a negative impact on the ability to continue to operate within the design objectives in Appendix I to 10 CFR 50.

Thus, the change to STS 5.5.4.b maintains the same overall level of liquid effluent control while retaining the operational flexibility that exists with TS under the previous 10 CFR Part 20. This limitation (i.e., less than 10 times the concentration values ...)

provides reasonable assurance that the levels of radioactive materials in bodies of water in Unrestricted Areas will result in exposures within (1) the Section Il.A design objectives of Appendix I to 10 CFR 50 and (2) restrictions authorized by 10 CFR 20.1301(e).

2. TS 5.5.4.e, delete "and projected dose" from the current sentence and add second sentence "Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days." This change is an administrative clarification approved by the NRC in TSTF-308, and presented in NUREG-1433, Revision 3. This avoids possible misinterpretation that projecting doses for the current calendar quarter, as well as for the current calendar year, are required every 31 days. This clarification does not reflect any change in requirements or procedures.

3. TS 5.5.4.g, reword "effluents to areas beyond" adding clarifying phrases to read aeffluents from the site boundary to areas at or beyond." This change is an administrative clarification approved by the NRC in TSTF-258, and presented in NUREG-1433, Revision 3. This clarification does not reflect any change in requirements or procedures.

Letter 2.04.083 Enclosure Page 9 of 11

4. TS 5.5.4j, add "beyond the site boundary" after "member of the public." This change is an administrative clarification approved by the NRC in TSTF-258, and presented in NUREG-1433, Revision 3. This clarification does not reflect any change in requirements or procedures.

4.6 The following editorial changes, corrections, or clarifications involve no technical change and serve to clarify the use and application of TS for the operating staff:

1. Remove the ** note from TS 3.7.A.5 since it was only applicable through 1998.

2. TS 3.5.A.5 misspelled word "rector" is corrected to "reactor." During the processing of License Amendment 200, dated April 22, 2003, "reactor" was misspelled in TS 3.5.A.5.

3. TS 3.8.2, on page 3/4.8-2, has two periods in the APPLICABILITY statement. The extra period at the end of the sentence is removed and the misspelled word "CHANNEl" in TS 4.8.2.3 is corrected to "CHANNEL." These typographical errors were inadvertently introduced during License Amendment 177, dated July 31, 1998.

4. TS 4.9.A.c, on page 3/4.9-2, has a typo in the last sentence reference to U4.9.A.1.b.1,"

which is corrected to "4.9.A.1.b.2." This typographical error makes incorrect reference to the Specification, which was inadvertently introduced during the License Amendment 179, dated December 18,1998.

5. TS 3.9.B.2 last sentence on page 3/4.9-4, "and the NRC is notified within one (1) hour as required by 10 CFR 50.72" is deleted. This is adequately required by the 10 CFR 50.72 and applicable plant procedure implementation of the regulation.

6. TS 4.9.A.4.b, on page 3/4.9-4 correct the abbreviation for the unit hertz to "Hz" (without subscripting the "z"). This was a typographical error only.

7. TS 3.11.C.2, on page 3/4/11-3, reference to 'Table 3.3.1" is revised to correctly reference "Table 3.3-1." This was a typographical error only.

Elimination of duplicative regulatory reporting requirements will avoid future, and eliminate existing, inconsistent or conflicting regulatory requirements. The proposed use of generic personnel titles will allow Pilgrim the flexibility to revise position titles while still meeting the appropriate personnel qualifications required by TS 5.3, "Unit Staff Qualifications." Additionally, the use of generic personnel titles will reduce and/or eliminate the need for future license amendments related to revised position titles. Administrative corrections and enhancements serve to clarify the use and application of TS for the operating staff.

These proposed changes are considered administrative with no adverse impact on the public health and safety.

5. Regulatory Safety Analysis 5.1 No Significant Hazards Consideration Entergy Nuclear Operations, Inc. (Entergy) is proposing to modify the Pilgrim Technical Specifications (TS): (1) to remove the administrative reporting and restart authorization requirements that apply in the event of a Safety Limit violation; (2) to replace plant-

Letter 2.04.083 Enclosure Page 10 of 11 specific titles with generic titles; (3) to remove the remaining responsibilities of the Operations Review Committee (ORC); (4) to delete regulatory detail for the Inservice Code Testing Program; and (5) to make other administrative corrections or clarifications.

Entergy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No. The proposed change is administrative in nature and does not involve the modification of any plant equipment or affect basic plant operation.

There is no impact to any accident previously evaluated.

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

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

Response: No. The proposed change does not involve any physical alteration of plant equipment and does not change the method by which any safety-related system performs its function. As such, no new or different types of equipment will be installed, and the basic operation of installed equipment is unchanged.

The methods governing plant operation and testing remain consistent with current safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No. The proposed change represents the relocation of specific Technical Specification requirements, based on regulatory guidance and previously approved changes for other stations or deletion of detail redundant to regulations or no longer applicable (i.e., expired one-time exceptions). The proposed change is administrative in nature, does not negate or revise any existing requirement, and does not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by requirements that are retained, but relocated from the Technical Specifications. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

5.2 Environmental Consideration A review has determined that the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in

Letter 2.04.083 Enclosure Page 11 of 11 individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

6. Precedents The NRC has approved similar changes (e.g., changes adopting TSTF-5, TSTF-65, TSTF-258, and TSTF-308) in a number of amendments. Examples include Waterford Steam Electric Station, Unit 3, amendment No. 188 dated April 3, 2003, Cooper Nuclear Station amendment No. 200 dated July 15, 2003, Callaway Plant, Unit 1, amendment No. 155 dated June 3, 2003, and Calvert Cliffs Nuclear Power Plant, amendment No. 259, dated July 15, 2003.

7. References

1. NUREG-1433, Rev. 3, "Standard Technical Specifications, General Electric Plants, BWR/4."

2. 10 CFR 50.36, "Technical specifications."

3. Pilgrim Nuclear Power Station amendment No. 177, dated July 31, 1998.

4. Waterford Steam Electric Station, Unit 3, amendment No. 188 dated April 3, 2003.

5. Cooper Nuclear Station amendment No. 200 dated July 15, 2003.

6. Callaway Plant, Unit 1, amendment No. 155 dated June 3, 2003.

7. Calvert Cliffs Nuclear Power Plant, amendment No. 259, dated July 15, 2003.

ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION AND BASES CHANGES (MARK-UP)

1.0 DEFINITIONS (continued)

REFUELING INTERVAL REFUELING INTERVAL applies onfy to AOME Code, Ceefior. XI 4WP..i'd'W,% surveillance tests. For the purpose of designating frequency of these code tests, a REFUELING INTERVAL shall mean at least once every 24 months.

REFUELING OUTAGE REFUELING OUTAGE is the period of time between the shutdown of the unit prior to a refueling and the startup of the plant after that refueling. For the purpose of designating frequency of testing and surveillance, a REFUELING OUTAGE shall mean a regularly scheduled outage; however, where such outages occur within 11 months of completion of the previous REFUELING OUTAGE, the required surveillance testing need not be performed until the next regularly scheduled outage.

SAFETY LIMIT The SAFETY LIMITS are limits below which the reasonable maintenance of the cladding and primary systems are assured.

Exceeding such a limit Is cause for unit shutdown and review by the Nuclear Regulatory Commission before resumption of unit operation. Operation beyond such a limit may not In Itself result in serious consequences, but it indicates an operational deficiency subject to regulatory review.

SECONDARY SECONDARY CONTAINMENT INTEGRITY means that the CONTAINMENT reactor building is intact and the following conditions are met:

INTEGRITY

1. At least one door In each access opening Is dosed.

2. The stanoby gas treatment system is operable.

3. All automatic ventilation system isolation valves are operable or secured In the isolated position.

SIMULATED AUTOMATIC SIMULATED AUTOMATIC ACTUATION means applying a ACTUATION simulated signal to the sensor to actuate the circuit in question.

SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor Is exposed to a radioactive source.

STAGGERED TEST A STAGGERED TEST BASIS shall consist of: (a) a test BASIS schedule ford systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals; (b) the testing of one system, subsystem, train or other designated components at the beginning of each subinterval.

SURVEILLANCE Each Surveillance Requirement shall be performed within the FREQUENCY specified SURVEILLANCE INTERVAL with a maximum allowable extension not to exceed 25 percent of the specified SURVEILLANCE INTERVAL The SURVEILLANCE FREQUENCY establishes the limit for which the specified time interval for Surveillance Requirements may be extended. It permits an allowable extension of the normal surveillance interval to facilitate surveillance schedule and 9

1-5 Amendment Not V

2.0 SAFETY LIMITS 2.1 Safety Limits 2.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% of rated core flow:

THERMAL POWER shall be < 25% of RATED THERMAL POWER.

2.1.2 With the reactor steam dome pressure > 785 psig and core flow Ž 10% of rated core flow:

MINIMUM CRITICAL POWER RATIO shall be> 1.06. ..

2.1.3 Whenever the reactor Is in the cold shutdown condition with irradiated fuel In the reactor vessel, the water level shall not be less than 12 Inches above the top of the hormal active fuel zone.

2.1.4 Reactor steam dome pressure shall be < 1325 psig at any time when Irradiated fuel is present In the reactor vessel.

2.2 Safety Limit Violation With any Safety Limit not met he following actions shall be met:

-e [W~itin 0fl3 hour Rudyf~, the NiRC Opzratio;ClQlel in d;U [~

thin tohu Restore compliance with all Safety Limits, and Insert all Insertable control rods.

124. T SAton 17irogl-e% wad RnSon-V.ie President Nuelear &e-#*4kho Ri wzl RaxdaluA 2;;d AucGit Gewvnttoa CMS1AC) hlNnoiod:tirC

-HeemassEvt~ill Retpoii t S! veft"46LJ 61t- t- C~rrisln, the.eperatkcn sh ommltteo Gl~r~ t~NfAG RGt) zerm th le L~ ir~c re c~o

2.2.6r -G~tieel aperagn of ho Unit ehaII me11.W you d 9~~nfiLa UWij-!zo4by thoe Amendment 15, 2-7.,472,1339, 1146, 7AV'2

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 CORE AND CONTAINMENT COOLING 4.5 CORE AND CONTAINMENT COOLING SYSTEMS SYSTEMS A. Core Sorav and LPCI Systems (Cont) A. Core Soray and LPCI Systems (Cont)

4. During Run, Startup, and Hot Shutdown 1. c. Motor As Specified Modes with the LPCI system Inoperable, Operated in 3.13 restore the LPCI system to Operable Valve status within 7 days and maintain both Operability core spray systems and the diesel d. Core Spray Header generators Operable. Otherwise, be in Ap Instrumentatfon at least Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

5. Two low pressure injection/spray Check Once/day subsystems shall be Operable during Calibrate Once/3 Cold Shutdown and Refuel Modes r months unless the is remov ,e spent fuel poo gates are removed, and Test Step Once/3 water level is at greater than or equal to months elevation 114 foot, except as specified in 3.S.A.6. 2. This section intentionally left blank

6. During Cold Shutdown and Refuel Modes unless the reactor head Is 3. LPCI system testing shall be as follows; removed, the spent fuel pool gates are removed, and water level is at greater a. Simulated Once/

than or equal to elevation 114 foot: Automatic Operating

a. With one of the required low pressure Actuation Test Cycle Injection/spray subsystems inoperable, restore the inoperable b. Pump When tested required low pressureinjection/spray Operability. as specified in subsystem to Operable status within 3.13, verify that 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Otherwise, take Immediate each LPCI pump action to suspend activities with delivers potential for draining the reactor 4800 GPM at a vessel. head across the

b. With both of the required low pump of at least pressure injection/spray subsystems 380 ft.

inoperable, take Immediate action to

c. Motor As Specified suspend activities With potential for Operated in 3.13 draining the reactor vessel and Valve restore 1 low pressure Operability injection/spray subsystem to Operable status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Otherwise, take immediate action to restore secondary containment and one standby gas treatment system to Operable status and to restore isolation capability In each required secondary containment penetration flow path not isolated.

Amendment No. 3J4.5-2 3/4,.2,

3.7 CONTAINMENT SYSTEMS (Cont) 4.7 CONTAINMENT SYSTEMS (Cont)

A. Primary Containment (Cont) A. Prmarv Containment (Cont)

5. All containment isolation check 4. Combined main steam lines: 46 valves are operable or at least scfh @ 23 psig.

one containment isolation valvf in each line having an where P. & 45 psig inoperable valve is s£ ured In L= 1.0% by weight of the contained the isolated position>> air 0 45 psig for 24 hrs.

Primary Containment Isolation Valves Prmarv Containment Isolation Valves

2. b. 1. 'The primary containment

2. b.. In the event-any automatic isolation valves surveillance Primary Containment Isolation shall be performed as follows:

Valve becom6s Inoperable, at lleast one containment Isolation a. At least once per operating valve In each line having an cycle the operable primary inoperable valve shall be containment isolation valves deactivated in the isolated that are power operated and condition. (This requirement may automatically Initiated shall be satisfied by deactivating the* be tested for simulated Inoperable valve In the Isolated automatic iniltati6n and

-condition. Deactivation means to closure times. '

electrically or pneumatically b. Test primary containment

disarm, or otherwise secure the isolation valves:

va!ve.)s

1. Verify power operated primary containment isolation valve operability as specified in 3.13.

2. Verify main steam

Isolation valves closed to satisfy these requiremc isolation valve operability may be reopened on an intermittent basis under 9T) as specified in 3.13.

~drnministrative controls.

Checkalve 30-C 32 will be sidered b2erable L itil rev se flow tes g is perform no late ban

th 198 tenance ou e.i\

AmendmentNo. 43-e ._

Amnendment No. 113, 136,119, 160, 167Jl7i(, 314.7-5

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8 PLANT SYSTEMS (CONT) 4.8 PLANT SYSTEMS .(CONT)

2. Mechanical Vacuum Pump Isolation 2. Mechanical Vacuum Pump Isolation Instrumentation Instrumentation LCO 3.8.2 - NOTET When a channel Is placed In an Four channels of the Main Steam Line Inoperable status solely for the Radiation Monitoring System Radiation - performance of required High function forthe mechanical vacuum Surveillances, entry Into associated pump shall be OPERABLE Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months APPLICABILITY: provided the associated Function maintains mechanical vacuum Whenever any main steam isolation pump Isolation capacity.

valve is open with steam flowing.

ACTIONS:

NOTE 1. Perform a CHANNEL CHECK Separate Condition Entry is allowed for every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

each channel.

2. Calibrate the trip units every A. One or more required channels 92 days..

inoperable. 3. Perform a CHA Jv

1. Restore channel to OPERABLE CALIBRATION ev 24 status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . months. The allowable trip A

value shall be 5 5.5 x normal OR background.

2. -- NOTEE Not Applicable If inoperable 4. Perform a LOGIC SYSTEM channel is the result of an FUNCTIONAL TEST Including Inoperable isolation valve. isolation valve actuation every 24 months.

Place channel or associated trip system In trip within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B. Required Action and associated Completion Time of Condition A not met.

OR Mechanical vacuum pump isolation capability not maintained.

1. Isolate mechanical vacuum within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

OR

2. Isolate Main Steam Lines within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

OR

3. Be in HOT SHUTDOWN within

-12 hburs.

._ 3/4.8-2 Amendment NolTh~b

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.9 AUXIUARY ELECTRICAL SYSTEM 4.9 AUXILIARY ELECTRICAL SYSTEM A. Auxillary Electical EIuloment (Cont) A. Auxiliary Electrcal Eguglment Surveillance (Cont)

IA/

7W 1. Vedfying do-energlzaton of the emergency buses and load shedding from the emergency buses.

2. Verng the diesel starts from ambient condition on the auto-start signal, energizes the emergency buses with permanenty connected loads, energizes the auto-connected emergency loads through the load sequence, and operates for' 6 minutes while Its generator Is loaded with the emergency Ioads.

During performance of this surveillance verIfy that HPCI and RCIC Inverters do not trip.

The resuts shall bi logged.

c. Once per operating cycle with the diesel loaded per 4.9A1.b verify that on diesel gnerator trip, secondaWy (offalte) AC power Is automatically conrected Wthln 11.8 to 13.2seconds to the emergency service buses and emergency loads are energized through the-oad sequencer In the same manner as described In 4.9A1sl The results shall be logged PNpSA 3r4.9-2 3/4.9-2 Amendment No-l Amendment

LIMITNG CONDMIONS FOR V:ERATION SURVEEILLANCE REQUIREMENTS 3.9 AUXILIARY ELECTRICA SYSTEM (Cont) 4.9 A,UXILIARY ELECTRICAL SYSTEM (Cont)

B. Operation with Inoperable. EWuloment Auxiliary Electrical Eguipment Surveillance (Cont)

Whenever the reactor Is In Run Mode or Startup Mode with the reactor not In a 8 Emergency 41 60V Buses AS-A6 Cold Condition, the avallability of electric Degraded Voltage Annunclation power shall be as specified In 3.9.8.1, System.

3.9.B.2, 3.9.1.3,;3.9.6.4, and 3.9.6.

a. Once each operating cycle,

1.. From.and after the date that calibrate the alarm sensor.

incoming power is not available from the startup or shutdown b. Once each 31 days perform a transformer, continued reactor channel functional test on the

. operation Is permissible under this alarm system.

condition for.

c. In the event the alarm system

a. 3 days with the startup Is determined Inoperable transformer Inoperable under 3.b above, commence logging safety related bus or voltage every 30 mInutes until

b. 7 days with the shutdown such time as the alarm Is transformer Inoperable restored to operable status.

During this period, both dlassl 4. RPS Electrical Protection generators and associated Assemblies emergency buses must remain operable. a. Each pair of redundant RPS EPAS shall be determined to

2. From and after the date that be operable at least once per Incoming power is not available 6 months by performance of from both startup and shutdown an Instrument functional test.

. transformers, continued operation Is permissible provided both diesel b. Once per18 months each generators and associated pair of redundant RPS EPAs emergency buses remain operable, shall be determined to be all core and containment cooling operable by performance of systems are operable, reactor an Instrument calibration and power level Is reduced to 25% of by veriing tripping of the desl N notfe circuft breakers upon the simulated conditions for automatic actuation of the protective relays wthn the

3. From and after the date that one of following limIts:

the diesel generators or associated emergency bus Is-made or found to Overvoltage 5 132 volts be Inoperable for any reason, Undervoltage . 108 volts continued reactor operation Is Underfrequency i 57FI) permissible In accordance with SpecifIcations 3.4.1.1, 3.5,F.1, I-- kqz 3.7.B.1.e, 3.7.B.1.e, 3.7.B2.c, and 3.7.B.2.e if Specification 3.9.A1 and 3.9A.2.a are satisfied.

314.94 Amendment No.??9)

LIlITING CONDITIONS FOR OPRATION SIURlVEILAMNCE REqUIMENTS 3.11 REACTOR FUEL ASSEMBLY (Cont) 4.11 REACTOR FUEL ASSEMBLY (Cont)

C. Linimum Critical Power Ration MCPR C. Minimum Critical Power Ration MCP, (Cont'd) (Cont) d)

2. The operating limit 2ICPR values . b. The average scram time to as a function o the r are dropout of Notch 34 is given in Table 3. of the determined as follows:

Core Operating Limits eport where r is given by n specification 4.11.C.2. ? Ni ri rave i=,l n

Ni i=1 Where: an n - number of surveillance tests performed to date -in the cycle.

Ni - number of active control rods measured in the ijh surveillance test.

i- average scram time to dropout of j Nbtch 34 of all rods measured in l the ith surveillance test. )

c. The adjusted analysis mean

scram time (r3) is calculated as'follows:

1l ti N 1 IlM b fBM+I i5I l

.1 Where:

p - mean of the distribution for average scram insertion time to dropout of Notch 34, 0.937 sec. 1 N1 -

a -

total number of active control rods at BOC during the first surveillance test..

standard deviation of the.

I distribution for average-scram .

insertion time to the dropout of Notch 34, 0.021 seconds.

I Amendment No. 243-42/44T,59.-333,-138 3/4-.11-3

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE IRQUIEMENTS 3.13 INSERVICE CODE TESTING 4.13 INSERVICE CODE TESTING Applicability: Apnlicabilitv:

Applies to ASME Code Class 1, 2 and Applies to the periodic testing 3 l-eqrerei&

pumps and valves. requirements of ASME Code Class 1, 2 and 3

. *nt pumps and valves.

Objective:

To assure the operational readiness of ASME Code Class 1, 2, and 3 To assess the operational readiness of

,--a -. _pumps ~ and valves kintpereant pumps and by performance o inse valves. - f~ASME &Ie C/>ss Specification: i, z, A,, 3 Specification:

A. Inservice Code-Testing of'Pump and A. Inservice Code Testing of Pumps Valves and Valves

1. Based on the Facility Commercial Operation Date, Inservice Code AVME Co dP Testing of n ne e I/Z141 dr C/ass

) -44ated pumps and valves shall be performed in accordance with 3~ te g MtoIler a ressure1^

ess Code, Se on'XI "Rule fo0 nserv ce esting of ea JPwr Plant Cmponents' ~sectii IWPad U as requir~db

'lOC`R f af), ex rwere spec Ife relief h been gra by,/ he NRC purs t0o 46 CSO55a(f 6)(i).

rVice Code r12Av Proqravi Code Terminology Weekly 7 Days Monthly 31 Days Quarterly or 92 Days 3 Mths Semiannually/ 184 Days 6 Mths .

9 Months 276 Days Yearly/Annually 366 Days Biannual/2 Yrs 732 Days 3/4.13-1 Amendment No.

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.13 INSERVICE CODE TESTING 4.13 INSERVICE CODE TESTING

3. The provisions in Definitions (1.0) for REFUELING INTERVAL, SURVEILLANCE FREQUENCY, and SURVEILLANCE INTERVAL are applicable to Code testing and to the above frequencies for performing Code testing activities.

4. Performance of Code testing shall be in addition to other specified Surveillance Requirements.

5. Nothing in the A5M ;4inr-ef4

-Prcs~r .o Cal. shall supersede the requirements of Technical Specifications.

3/4.13-2 Amendment No. w-

BASES:

3.13 and 4.13 Inservice Code Testing 2e §s-Hrq Ptjta The Limiting Conditions for Operation establishes the requirement that inservIce testing ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with the sprodieilly updated odion of 8ootion Xl of the ASME Boilor and Pr.ssuro V-&sol CGee and

~Adena rq~krod by4'1CM5 ",.Hti. 58.55aff).

F 5> The detailed procedures for testing of pumps and valves are documented in the PNPS Inservice Testing Program. Oxl This specification Includes a clarification of the frequencies for performing the testing activities required by lee the ASME Doirsnrce Vce Code and applicable Addenda.

enXof This clarification Is provided to ensure consistency in Surveillance Frequencies throughout the Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice testing activities.

Under the terms of this Specification, the more restrictive requirements of the Technical Specifications take precedence over the ASME BoIler amd P eee oo Code and applicable Addenda.

. nical ecificati require omponents X be declaryd operable nor t(

into an erationalode. Th SME B&PV/tode provIon which ows pL and ves to b asted up t one week aftr return to operation Is Iormal Revision B73,/4.43-BS/4.1 3-1

BASES:

314.4 STANDBY LIQUID CONTROL SYSTEM Surveillance Requirements if both SLC subsystems are Inoperable for reasons other than condition 3.4.A, at least one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The allowed completion time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is considered acceptable given the low probability of a DBA or transient occurring concurrent with the failure of the control rods to shut down the reactor.

34.0D ff Any action and associated completion time is not met, the plant must be brought to a MODE in Which the LCO does not apply. To achieve this status, the plant must be brought to Hot Shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed completion time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Is reasonable, based on operating experience, to reach Hot Shutdown from full power conditions In an orderly manner and without challenging plant systems.

4.4.1 Demonstrating that each SLC System pump develops a flow rate of 39 gpm at a minimum system head of 1275 pslg ensures that pump performance Is acceptable during the fuel cycle. This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration requirements, the rate of negative reactivity Insertion from the SLC System wl adequately compensate forthe posiffve reactivlty effects encountered during power reduction, cooldown of the moderator, and xenon decay. This test confirms one point on the pump design

~ curve and Is Indicative of overall performance. Such Insetce Inspections confirm component OPERABILITY, trend performance, and detect incipient failures by Indlcating o o ne. Tet the p s an valves In accordance with the ser na P ram Conde Se naoitces MWP and IWV. excep wflere specitac relief is granted)]y a equatey assesses component operational readiness.

4.4.2; This Surveillance ensures that there Is a functioning flow path from the boron solution storage tank to the RPV, Including the firing of an explosive valve. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of that batch successfully fired. The pump and explosive valve tested shoufd be altemrated such that both complete flow paths are tested every 48 months at alternating 24 month intervals. The Surveillance may be performed Inseparate steps to prevent injecting boron Into the RPV. An acceptable method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and Into the RPV. The 24 month Frequency Is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient If the surveillance were performed at power. Various components of the system are Individually tested periodically, thus making more frequent testing of the entire system unnecessary.

4.4.3 This Surveillance verifies the continuity of the explosive charges Inthe Injection valves to ensure that proper operation will occur ifrequired. Other administrative controls, such as those that limit the shelf life of the explosive charges, must be followed. The 31 day frequency is based on operating experience and has demonstrated the reliability of the explosive charge continuity.

Revision 44, Amonrnen

-1334.4-4 Flo 1 G

Coe pryIndLPI yse Core Spray and LPCI System B 3/4.5 .A BASES ACTIONS 3.5.A.6.b During Cold Shutdown and Refuel Modes unless the reactor head is removed, the spent fuel pool gates are removed, and water level is at greater than or equal to elevation 114 foot with both of the required ECCS injection/spray subsystems inoperable, all coolant inventory makeup capability may be unavailable. Therefore, actions must immediately be initiated to suspend OPDRVs to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until OPDRVs are suspended. One ECCS injection/spray subsystem must also be restored to Operable status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

If at least one low pressure ECCS Injection/spray subsystem is not restored to Operable status within the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time, additional actions are required to minimize any potential fission product release to the environment. This includes ensuring secondary containment Is Operable; one standby gas treatment subsystem is Operable; and secondary containment isolation capability (i.e., one isolation valve and associated instrumentation are Operable or other acceptable administrative controls to assure isolation capability) in each associated penetration flow path not isolated that is assumed to be isolated to mitigate radioactivity releases. Operability may be verified by an administrative check, or by examining logs or other information, to determine whether the components are out of service for maintenance or other reasons. It Is not necessary to perform the Surveillances needed to demonstrate the Operability of the components. If,however, any required component is inoperable, then it must be restored to Operability status. In this case, Surveillance may need to be performed to restore the component to Operable status.

Actions must continue until all required components are Operable.

The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months completion time to restore at least one low pressure injection/spray subsystem to Operable status ensures that prompt action will be taken to provide the required cooling capacity or to initiate actions to place the plant in a condition that minimizes any potential fission product release to the environment.

SURVEILLANCE The testing interval for the core and containment cooling systems is REQUIREMENTS based on industry practice, quantitative reliability analysis, judgment and practicality. The core cooling systems have not been designed to be fully testable during operation. To Increase the availability of the

-U+e JIt4e(rvIce core and containment cooling systems, the components which make up the system; i.e., instrumentation, pumps, valves, etc., are tested Code 1es+-h'&A frequentiv. The ursand motor operated valves are tested in accordance wiff7WiE )MVC0l, cio!VPndetW, ee speif~-elefan, d)5o assure their operability. The frequency and methods of testing are described in the PNPS 1ST (continued)

Revision - P3/4.5-2b JI

Core Spray and LPCI System B 3[4.5.A BASES

SURVEILLANCE program. The PNPS IST Program is us o assess the operational REQUIREMENTS readiness of pumps and valves that are related orimpont

!Maty t (continued) sefe When components are tested and found inoperable the Impact on system operability is determined, and corrective action or Limiting Conditions of Operation are Initiated. A simulated automatic actuation test once each cycle combined with code Inservice testing of the pumps and valves is deemed to be adequate testing of these systems.

The surveillance requirements provide adequate assurance that the core and containment cooling systems will be operable when required.

I _i 0 Aecndmonl PVoI 7z B3/4.5-3 I

RHR Suppression Pool Cooling B 314.5.6.1 B 314.5 CORE AND CONTAINMENT COOUNG SYSTEMS BASES SURVEILLANCE SR 4.5.5.1.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the RHR suppression pool cooling mode flow path provides assurance that the proper flow path exists for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve Is also allowed to be Inthe nonaccident position provided it can be aligned to the accident position within the time assumed In the accident analysis. This Is acceptable since the RHR suppression pool cooling mode is manualy initiated. This SR does not require any testing or valve manipulation; rather, It involves verification that those valves capable of being mispositloned are in the correct position.

This SR does not apply to va!ves that cannot be Inadvertently misaligned, such as check valves.

The frequency of 31 days Is justified because the valves are operated under procedural control, Improper valve position would affect only a single subsystem, the probability of an event requiring Initiation of the system Is low, and the subsystem Is a manually initiated system. This frequency has been shown to be acceptable based on operating experience.

~~~SR 4.5,13.1.2 -L1,p, Verifying that each RHR pumpefctio 314.13.

(Ref. 1) while operating In the suppression pool cooling mode W1 2.o through the associated heat exc se's'8gl _)

mpeernoemance has notde.graded5d-8Fo-tes 151eTn--Tul pup OT rformnren re ~A~ Codd Lec ,n Xl (Red])u m p st confli es oneppit on tepum desigr iu bIn no reut ieidicat, of overl perfo mnce/ Such insinceins~sbos 96fim co ponent OpRAB5Yen perfrmaces he I reqency Q

of this SR is in accordance I

with the Inse e esting Program, Specification 314.13.

REFERENCES 1. FSAR, Section 14.S.v Fa,1 /ueeaehdr

{ At,3_4.5-2.

RHR Containment Spray B 314.5.8.2 B 314.5 CORE AND CONTAINMENT COOLING SYSTEMS BASES REFERENCES 1. FSAR, Section 4.8

2. FSAR, Section 14.5.

3. ASME of-I Code LP B3/4.5-1 0

.i

.. i ., .

HPCI System B 314.S.C B 314.6 CORE AND CONTAINMENT COOLUNG SYSTEMS BASES SURVEILLANCES The testing interval for the core and containment cooling systems is based on industry practice, quantitative reliability analysis, judgment and practicality. The core cooling systems have not been designed to be fully testable during operation. For example, in the case of the HPCI, automatic initiation during power operation would result in pumping cold water into the reactor vessel which is not desirable. To increase the availability of the core and containment cooling systems, the components which make up the system: I.e., instrumentation, pumps, valves, etc., are tested frequently. Thejumps and motor operated valves are tested in accordance with AGMiB&PV-eede assure their operability. The frequency and methods of testing are described in the PNPS [ST program. The PNPS IST Program Is used to assess the operational readiness ofaumps and valves4e-ef-

_ ,ht

_ _, o.When components are tested and found inoperable the impact on system operability is determined, and corrective action or Umiting Conditions of Operation are initiated.

A simulated automatic actuation test once each cycle combined with code inservice testing of the pumps and valves Is deemed to be adequate testing of these systems.

The surveillance requirements pi,ovide adequate assurance that the core and containment cooling systems will be operable when required.

A£'4+Coe:(

.9 And 1 m..t Nu. 17c B314.5-19

RCIC System B 314.5.D B 314.5 CORE AND CONTAINMENT COOUNG SYSTEMS 314.5.D. Reactor Core Isolation Cooling (RCIC) System BASES BACKGROUND The RCIC is designed to provide makeup to the nuclear system as part of the planned operation for periods when the normal heat sink is unavailable. The Station Nuclear Safety Operational Analysis, FSAR Appendix G, shows that RCIC also serves as redundant makeup system on total loss of all offslte power in the event that HPCI is unavailable. In all other postulated accidents and transients, the ADS provides redundancy for the HPCI.

SPECIFICATION The requirement that RCIC be operable when reactor coolant temperature is greater than 3650F is Included in Specification 3.5.D.1 to clarify that RCIC need not be operable during certain testing (e.g.,

reactor vessel hydro testing at high reactor pressure and low reactor coolant temperature). 3651F is approximately equal to the saturation steam temperature at 150 psig.

ACTION Based on this and judgments on the reliability of the HfPCI system, an allowable repair time of 14 days is specified.

SURVEILLANCES The testing interval for the core and containment cooling systems is based on industry practice, quantitative reliability analysis, judgment and practicality. The core cooling systems have not been designed to be fully testable during operation. To Increase the availability of the core and containment cooling systems, the components which make ASME..

- £j; e IJR45tvlc .eA up the system; i.e. Instrumentation, pumrs, valves, etc., are tested uently Xfre Theumps and motor operated valves are tes ed in Code

{ c )

Co4e 1:eS- Xaccordance

,Al wi Isr) Prs ?Am c .-. r)to assure their operability.

'~C~ The frequency and methods of testing are described in the PNPS IST program. The PNPS IST Pro ram is used to assess the operational readiness o umps an v

-. ofet)- When components are tested and found inoperable the impact on system operability is determined, and corrective action or Umiting Conditions of Operation are initiated. A simulated automatic actuation test once each cycle combined with code Inservice testing of the pumps and valves is deemed to be adequate testing of these systems.

The surveillance requirements provide adequate assurance that the core and containment cooling systems will be operable when required.

-Amoxeadmenhs-7 B3/4.5-20 clz_)___

Responsibility 5.1 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility ioriiieeetef shall be responsible for overall unit operation and shall In writing the succession to this responsibility during his absence.

The Station -Diroct or his designee shall approve, prior to implementation, each proposed test, experiment, or modification to systems or equipment that affect nuclear safety.

5.1.2 r ( shal sponsible for the control room command function. Durng any absen of th from the control room while the unit Is in an operational o e other than Cold Shutdown or Refueling, an individual with an active lenlor Reactor Operator (SRO) license shall be designated to assume th rol room command function. During any absence of the rom e control room while the unit is In Cold Shutdown or Refueling, an individual with an active SRO license or Reactor Operator (RO) license shall be designated to assume the control room command function.

.. . .I 5.0-1 Amendment No? An

Organization

5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for unit operation and.

corp orate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.

a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationshis-and job descriptions for ke ersonnel ositions, or in equivalent forms of documentation. These requirement shall be documented in the Pilgrim Station Final Safety Analysis Report (FSAR);

b. The ENi1Fsl lbe responsible for overall safe operation of the plant and shall have control over those onsite activities necessary fd a c of the plant;

c. gor Pilgrim shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, ani providing technical support to the plant to ensure nuclear safety; and I

d. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating' pressures 5.2.2 Unit Staff The unit staff organization shall include the following:

a. A non-licensed operator shall be on site when fuel is in the reactor and an additional non-licensed'operator shall be assigned when the reactor Is in an operational mode other than Cold Shutdown or Refueling.

(continued)

'NrZ 5.0-2 Amendment No.

INSERT page 5.0-2

..., including the plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, ...

Organization 5.2 5.2 Organization 5.2.2 Unit Staff (continued)

b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, while the unit is in an operational mode other than Cold Shutdown or Refueling, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.

c. At least two licensed ROs shall be present in the control room during reactor startup, scheduled reactor shutdown and during recovery from reactor trips .

d. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(1) and 5.2.2.a and 5.2.2.1 for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Inorder to accommodate unexpected absence of on-duty shift crew members provided immediate action Is taken to restore the shift crew composition to within the minimum requirements.

e; Higher grade licensed operators may take the place of lower grade licensed or unlicensed personnel.

f. An individual qualifled.ln radiation protection procedures shall be on site when fuel is in the reactor. The position may be vacant for not more thah 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence, provided immediate action is taken to fill the required position:.

g. - The amount of overtime worked by unit staff members performing

-safety-related functions shall be limited and controlled In accordance with the NRC Policy Statement on working hours (Generic Letter 82-12). ro'-,_a,; 11ieoteo 'v s

h. The 0perations aan nage ,lucicor WotehEneo~s, and.

9pemtionc Supep.4seor shall hold a Senior Reactor Operator Auetea Uvcense. The4Urs a Pl- GAdaF shl 1402M hold arW =

i - Th h Roi'R9 'a ( shalll e technical support to the I_ e in the S jo d.'.' areas of engineering and accident assessment Li4Rddon, th SCRE-shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift. -A6Whf CntrcR Eng§ieef with a Senior Reactor Operator license may simultaneously 5.0-3 Amendment No.

Programs and Manuals 5.5 5.0 ADMIN1STRATIVE CONTROLS 5.5 Programs and Manuals The following programs shall be established, implemented and maintained.

5.5.1 Offsite:Dose Calculation Manual (0D0M)

a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and tzip setpoints, and in the conduct of the radiological environmental monitoring program; and

b. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the Information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release, reports required by Specification 5.6.2 and Specification 5.6.3.

Ucensee Initiated changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

1. sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s); and

2. a determination that the-change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations;

b. Shall become effective after ro'ie' Mcceptnceby

_nd' the eperetis the approval of the-Gherlniet

,gnage' and n alogi-

c. Shall be submitted io the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive EffluentRelease Report for the period of the report In which any change in the ODCM was made. Each change shall be identified by markings In the margin of the affected pages, clearly indicating the area of the page that was changed, and shall Indicate the date (i.e.,

month and year) the change was Implemented.

(continued) 5.0-6 Amendment No.

! il, Jo Programs and Manuals 5.5 5.5vrograms ana Manuals c Ween *C-Cies oe Aeon 5.5.4 Radioactive Effluent ControlsProgram (continued) ,___

__I__.__.

-I-

b. Limitations on the concentrations of radioactive raterial released in liquid effluents to unrestricted areas, conforming to 44GGR 2 '-

Appendi1B,..Table 2,:ColumnI

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters In the ODCM;.

d. Umitatlons on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released to unrestricted areas, conforming to 10 CFR 50, Appendix l;

e. Determination of cumulativeln ectdocontributions from radioactive effluents for thq currentced uarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days' IN5i--

f. ULmitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment conforming to 10 CFR 50, Appendix l; Hi

g. Limitatlons on the dose rate resulting from radioactive material released in gaseous effluents o areas beyond the site boundary conforming to the following:

1. For noble gases: Less than or equal to 500 mrem/yr to the total body and less than or equal to 3000 mrem/yr to the skin, and

2. For iodine-131, lodine-133, Tritium, and all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mremlyr to any organ.

h. Umitations.on.the.annual and quarterly air doses resulting from noble gases released in gaseous effluents to areas beyond the site boundary, conforming to 10 CFR 50, Appendix l; 9

eto m nd (continued)

.0 8A ip Pr 12NI12R 5.0-8 Amendment No. 147_21-

INSERT page 5.0-8

.... Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days;

Program and Manuals 5.5 5.5 Program and Manuals 5.5.4 Radioactive Effluent Controls Program (continued)

}. Limitations on the annual and quarterly doses to a member of the public from lodine-131, Iodine-1 33, Trtium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released to areas beyond the site boundary, conforming to 10 CFR 50, Appendix l; and 1- Lrmitatlons on the annual dose or dose commitment to any member of the publi ue to releases of radioactivity and to radiation from uranium fuel cyclesources, conforming to 0 CFR 190 5.5.5 Component Cyclic or Transient Limit C' bea'cd L lit( s;cbag This program provides controls to track the FSAR Section C.3.4.1, cyclic and transient occurrences to ensure that components are maintained within the design limits:

5.5.6 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. a change in the TS Incorporated In the license; or

2. a change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases-Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

d. Proposed changes that meet the criteria of Specification 5.5.6b above shall be reviewed and approved by the NRC prior to Implementation.

Changes to the Bases Implemented without prior NRC.approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71 (e).

5.0-9 Amendmeht No.

High Radiation Area 5.7 5.0 ADMINISTRATIVE CONTROLS 5.7 High Radiation Area 5.7.1 Pursuant to 10 CFR 20, paragraph 20.1601(c), in lieu of the requirements of 10 CFR 20.1601, each high radiation area, as defined In 10 CFR 20, in which the intensity of radiation is > 100 mrem/hr but < 1000 mremlhr, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Individuals qualified In radiation protection procedures (e.g., r raut"'m +i0' personnel) or personnel continuously escorted by such individuals may (nAecc*-~v be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates < 1000 mrem/hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one o(more of the following:

a. A radiation monitoring device that continuously Indicates the radiation dose rate in the area.

b. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels In the area have been established and personnel are aware of them.

c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the

$adiationrtrotectionfanager in the RWP.

5.7.2 In addition to the requirements of Specification .7.1, areas with radiation levels

> 1000 mrem/hr shall be provided with locked r continuously guarded doors to prevent unauthorized entry and the keys shall e maintained under the administrative control of he~latci+E.geeF-ondutyorhealth-phy supervision. Doors shall remain locked except during periods of access by

( eD+

Jr t-ot ' 2 personnel under an approved RWP that shall specify the dose rate levels in the immediate work areas and the maximum allowable stay times for Individuals in those areas. In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by (Continued) 5.0-14 Amendment No1.S<

ATTACHMENT 2 LIST OF REGULATORY COMMITMENTS

List of Regulatory Commitments The following table identifies those actions committed to by Pilgrim in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

REGULATORY COMMITMENT DUE DATE Relocate to FSAR plant-specific title and Within 60 days of license amendment approval.

indicate relationship to generic titles used in the Specifications.

ML043560279

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