GNRO-2002/00030, License Amendment Request, Corrections and Clarifications to Certain Requirerments and Information

License Amendment Request, Corrections and Clarifications to Certain Requirerments and Information

ML022700163
Person / Time
Site:	Grand Gulf
Issue date:	09/18/2002
From:	Eaton W Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GNRO-2002/00030, LDC 2002-035
Download: ML022700163 (26)

Text

Entergy Operations, Inc.

P 0 Box 756 Entergy Port Gibson, MS 39150 Tel 601 437 6409 Fax 601 437 2795 William A. Eaton Vice President, Operations Grand Gulf Nuclear Stationi GNRO-2002/00030 September 18, 2002 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Grand Gulf Nuclear Station, Unit 1 Docket 50-416 License Amendment Request Corrections and Clarifications to Certain Requirements and Information (LDC 2002-035)

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) hereby requests the following amendment for Grand Gulf Nuclear Station, Unit 1 (GGNS). Entergy request that several Technical Specifications (TS) Limiting Conditions for Operations (LCO) and Administrative sections be revised to correct or clarify certain requirements and information.

The proposed changes have been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that this change involves no significant hazards consideration. The bases for these determinations are included in the attached submittal.

The proposed changes do not include any new commitments.

Entergy requests approval of the proposed amendment within one year of the date of this letter.

Once approved, the amendment shall be implemented within 60 days. Although this request is neither exigent nor emergency, your prompt review is requested.

September 18, 2002 GNRO-2002/00030 Page 2 of 2 If you have any questions or require additional information, please contact Ron Byrd at extension (601) 368-5792.

I declare under penalty of perjury that the foregoing is true and correct. Executed on September 18, 2002.

Sincerely, WAEIRWB/amt Attachments:

1. Analysis of Proposed Technical Specification Change

2. Proposed Technical Specification Changes (mark-up)

3. Changes to TS Bases pages cc: U. S. Nuclear Regulatory Commission ATTN: Mr. Ellis W. Merschoff Regional Administrator, Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 U. S. Nuclear Regulatory Commission ATTN: Mr. David H. Jaffe NRR/DLPM (w/2)

ATTN: FOR ADDRESSEE ONLY ATTN: U.S. Postal Delivery Address Only Mail Stop OWFN/7D-1 Washington, D.C. 20555-0001 Mr. T. L. Hoeg, GGNS Senior Resident Mr. D. E. Levanway (Wise Carter)

Mr. L. J. Smith (Wise Carter)

Mr. N. S. Reynolds Mr. H. L. Thomas Dr. E. F. Thompson (w/a)

State Health Officer State Board of Health P. 0. Box 1700 Jackson, Mississippi 39205

Attachment 1 GNRO-2002/00030 Analysis of Proposed Technical Specification Change to GNRO-2002/00030 Page 1 of 8

1.0 DESCRIPTION

This letter is a request to amend Operating License NPF-29 for Grand Gulf Nuclear Station, Unit 1 (GGNS).

The proposed change will revise the Technical Specifications (TS) to correct or clarify certain requirements and information. Except for one technical change that corrects a limiting value, most of the proposed changes are administrative in nature. One of the changes is being proposed at NRC staff request as discussed in a letter from the NRC dated April 26, 2001, "Grand Gulf Nuclear Station, Unit 1-Issuance of Amendment RE: Revision of the Minimum Critical Power Ratio Safety Limit for Cycle 12 Operation (TAC No. MB0514)".

2.0 PROPOSED CHANGE

Entergy proposes the following changes to the GGNS TS:

"* TS Limiting Conditions for Operations (LCO) 3.3.7.1 to correct a wording omission introduced in Amendment No. 145 by letter from W. A. Eaton to the USNRC dated January 21, 2000, "GGNS Pilot Full-Scope Application of NUREG-1465 Alternative Source Term Insights, LDC 1999-082,"

"* TS LCO 3.6.5.4 to correct a specified value identified as being non-conservative during instrument uncertainty and setpoint reviews,

"* TS 5.2 subsections 5.2.1 and 5.2.2 to reflect current position specific terminology,

"* TS 5.3.1 to make the requirement consistent with the Quality Assurance Program Manual (QAPM),

"* TS 5.5.8.b to correct an existing typographical error,

"* TS 5.5.11 to reflect changes in terminology used in the new 10 CFR 50.59 Rule,

"* TS 5.6.5.a.5 to clarify references as stipulated in a letter dated April 26, 2001 from the USNRC to Mr. W. A. Eaton " Grand Gulf Nuclear Station, Unit I - Issuance of Amendment RE: Revision of the Minimum Critical Power Ratio Safety Limit for Cycle 12 Operation (TAC No. MB0514)", and

"* TS 5.7 subsections 5.7.1, 5.7.2 and 5.7.3 to reflect current job titles/activities and clarify guard posting requirements.

In summary, the proposed changes to TS 3.3.7.1, 5.2, 5.5.8.b, 5.6.5.a.5, and 5.7 are editorial and maintain compliance with the intent of regulatory requirements. The proposed changes to TS 5.3.1 are intended to provide consistency between the TS and the Quality Assurance Program Manual as approved by the NRC staff. The proposed change to TS 5.5.11 is intended to maintain consistency between the TS and 10 CFR 50.59. The proposed change to TS 3.6.5.4 corrects the bounding value to be consistent with the current FSAR analyses.

to GNRO-2002/00030 Page 2 of 8

3.0 BACKGROUND

TS 3.3.7.1, "Control Room Fresh Air (CFRA) System Instrumentation," was amended as part of the Alternative Source Term Full-Scope application. The proposed change was approved by Amendment No. 145 to the GGNS Operating License (TAC No. MA8065). During this process, the change to Condition B was issued as submitted but was not worded per the Nuclear Energy Institute's writers guide. This error was noted during the implementation phase of this amendment but was determined to not be correctable without formal request from the licensee.

The proposed change corrects this error by adding the words "not met" to Condition B.

TS 3.6.5.4, "Drywell Pressure," establishes limits on the drywell-to-primary containment differential pressure. This parameter is an assumed initial condition in the analyses that determine the primary containment thermal hydraulic and dynamic loads during a postulated Loss of Coolant Accident (LOCA). The limit on negative drywell-to-primary containment differential pressure ensures that changes in calculated peak LOCA drywell pressures due to differences in water level of the suppression pool and the drywell weir annulus are negligible. It also ensures that the possibility of weir wall overflow after an inadvertent upper pool dump is minimized. The limit on positive drywell-to-primary containment differential pressure helps ensure that the horizontal vents are not cleared with normal weir annulus water level. The current value for this parameter was reviewed during a recent instrument uncertainty and setpoint calculation review effort. This review determined that the current value was not bounding in all situations and should be changed. This discrepancy was documented in the plant's Corrective Action Program.

TS 5.2, subsections 5.2.1 and 5.2.2, currently contain job terminology, which is no longer desired for use at the site. The current terminology is also not consistent with that used in the latest revision of ANSI/ANS 3.1 to which the site is committed to. The specific functions related to this'proposed change are in the radiation monitoring and controls area.

TS 5.3.1, "Unit Staff Qualifications," currently commits the site to the 1971 version of ANSI N18.1 for members of the Unit Staff. This version has been superceded by a later version entitled ANSI/ANS-3.1-1978. This later version was endorsed in the new common Quality Assurance Program Manual for all Entergy Sites. This was approved by letter from John N.

Hannon (NRC) to Michael R. Kansler (EOI), "Consolidation of Quality Assurance (QA) Programs into One Quality Assurance Program Manual for all Entergy Sites-Arkansas Nuclear One, Grand Gulf Nuclear Station, River Bend Station and Waterford Steam Electric Station (TAC M97893)",

dated November 6, 1998. The proposed change will not affect any special restrictions contained in TS 5.3.1.

TS 5.5.8.b contains requirements for an Explosive Gas and Storage Tank Radioactivity Monitoring Program. A typographical error was made in this TS during the initial conversion to the Improved Technical Specifications. This was issued as Amendment No. 120 in a NRC letter to Mr. C. Randy Hutchinson dated February 21, 1995, Issuance of Amendment No. 120 to Facility Operating License No. NPF-29 Grand Gulf Nuclear Station, Unit 1 (TAC No. M88101).

TS 5.5.11 Technical Specifications (TS) Bases Control Program contains wording that is no longer consistent with that used in the revised 10 CFR 50.59 Rule. The Federal Register Notice publishing the revised R6le contained guidance that allowed continued use of current programs without any issues. Revised wording was proposed via the Industry/TSTF Standard Technical to GNRO-2002/00030 Page 3 of 8 Specification Change Traveler Process as TSTF-364, which was subsequently approved by the NRC. Entergy is including this change to remain consistent with NUREG 1434, Rev. 2 and 10 CFR 50.59.

TS 5.6.5.a.5 Core Operating Limits Report (COLR) was revised by Amendment No. 146 as issued by letter from the NRC dated April 26, 2001, "Grand Gulf Nuclear Station, Unit I Issuance of Amendment RE: Revision of the Minimum Critical Power Ratio Safety Limit for Cycle 12 Operation (TAC No. MB0514)". In the Safety Evaluation for this amendment, the NRC documented a need for the licensee to revise the TS to clarify certain references contained therein. This change request provides the requested clarification.

TS 5.7, subsections 5.7,1, 5.7.2 and 5.7.3, establish administrative controls for high radiation area entry, posting, and guarding. Sections 5.7.1 and 5.7.2 currently refer to job titles/functions that are no longer desired for use at the station. In the case of section 5.7.3, the current wording has led to some confusion as to when a guard must be posted. The proposed change is intended to clarify the intent by either having a guard posted or meeting the other stipulations of the specification at all times. This will eliminate the need to determine if a guard can be posted and demonstrating this fact to other parties.

4.0 TECHNICAL ANALYSIS

4.1 Technical Specification 3.3.7.1 The format and content of the TS is controlled through the utilization of a standard writer's guide. The guidance currently being utilized is NEI 01-03, Writer's Guide for the lmproded Standard Technical Specifications. The guidance in section 4.1.6.i.5 stipulates that certain Conditions end with the phrase "not met". This phrase was omitted from Condition B of this specification in the amendment to incorporate the Alternative Source Term Full-Scope application. This change is strictly administrative in nature and has no impact on the operation of the plant. The requested change will bring the TS back into compliance with the writer's guide and with the Standard Technical Specification (STS), NUREG 1434 Rev. 2.

4.2 Technical Specification 3.6.5.4 Drywell-to-primary containment differential pressure is a key parameter for ensuring over all containment function under design basis accident conditions. The parameter is controlled through the setpoint for the normal drywell vacuum breaker.

The maximum allowed (analytical) negative drywell-to-primary containment differential pressure is limited to prevent an excessive increased drywell side suppression pool water level due to the differential pressure. A higher water level in the suppression pool on the drywell side has two negative effects. The first is a higher resultant drywell accident pressure, as it will take longer for suppression pool vent clearing during a Loss of Coolant Accident (LOCA). The second is the potential for equipment damage in the drywell if weir wall overflow occurred during an inadvertent upper containment pool dump.

The current Technical Specification (TS) value of -0.26 psid was used in the calculations for weir wall overflow. The evaluation of weir wall overflow is discussed in a letter from L. F. Dale 1'

Attachment I to GNRO-2002/00030 Page 4 of 8 to the U. S. Nuclear Regulatory Commission (USNRC) dated August 19, 1982, AECM 82/353.

The results from these calculations determined that there were no adverse consequences from weir wall overflow at this value.

The analytical value used for the vent clearing analysis was -0.25 psid and not -0.26 psid as discovered during the aforementioned review effort. The vent clearing analysis is described in subsection 6.2.1.1.6 of the Updated Final Safety Analysis Report (FSAR) and was also discussed in a letter from L. F. Dale to the USNRC dated October 22, 1982, AECM-82/497.

Therefore, the current TS value does not bound the value needed to support the vent clearing analysis. This discrepancy was documented in the plant's Corrective Action Program. The bounding limit of -0.25 psid is currently being controlled under administrative controls until such time as the TS is amended.

4.3 Technical Specification 5.2 The proposed changes to sections 5.2.1 and 5.2.2 are administrative in nature. The changes reflect the current terminology and titles for positions performing the related activities at the site.

This proposed change only affects those positions pertaining to radiation protection related work activities. The requirements of TS 5.3 Unit Staff Qualifications will remain applicable to these positions. Therefore, the changes have no effect on the qualifications, functions or organizational freedom of the positions affected.

4.4 Technical Specification 5.3 This proposed change would eliminate the need for maintaining dual qualification/certification records for Grand Gulf Nuclear Station (GGNS) personnel. Currently the TS invoke a different standard than does the Quality Assurance Program Manual (QAPM). This has led to duplicated documentation since the different versions are not exactly the same. This change is considered administrative in nature since the NRC has previously approved the use of ANSI/ANS 3.1-1978 as part of the approval of the current QAPM for Entergy. The technical merits of the two standards were compared and deemed appropriate during the approval process for the QAPM.

Thus, the TS reference to ANSI N18.1, 1971 will be replaced with ANSI/ANS 3.1-1978.

A clarifying statement is also being proposed to allow the use of the QAPM as the single document to control exceptions or clarifications to the standard. This is proposed to minimize the potential for having conflicting upper tier documents. Additionally, this will allow Entergy to only have to change one document if any additional clarifications are needed in the future. The current QAPM already contains some clarifications, which are not currently in the TS.

Special requirements contained in TS 5.3.1 for the radiation protection manager and the STA positions will not be changed by the proposed amendment.

4.5 Technical Specification 5.5.8 b The proposed change to TS 5.5.8.b is an administrative change to correct an obvious typographical error made during the original conversion to the Improved Technical Specifications. The word "or" is being corrected to "for". The proposed change has no bearing on the technical content of the existing specification and is clearly typographical in nature.

to GNRO-2002/00030 Page 5 of 8 4.6 Technical Specification 5.5.11 The NRC amended 10 CFR 50.59 concerning the authority for licensees of production or utilization facilities, to make changes to the facility or procedures, or to conduct tests or experiments, without prior NRC approval. The final rule clarifies the specific types of changes, tests, and experiments 6*onducted at a facility that require evaluation, and revises the criteria licensees must use to determine when NRC approval is needed before such changes, tests, or experiments can be implemented. The final rule also adds definitions for terms that have been subject to differing interpretations, and reorganizes the rule language for clarity. The Bases Control Program required by TS 5.5.11 allows Entergy to make changes to the Bases without NRC approval provided the change does not involve a change to the FSAR or Bases that involves an unreviewed safety question as defined in the old 10 CFR 50.59. With the revision to 10 CFR 50.59, the definition of unreviewed safety question was eliminated. Therefore, the TS should be revised to be consistent with the revision to 10 CFR 50.59. This is considered to be an administrative type change since the NRC has previously approved the change being sought. The change is consistent with the Standard Technical Specification, NUREG 1434, Rev. 2.

4.7 Technical Specification 5.6.5.a.5 The proposed change to TS 5.6.5.a.5 is an administrative change that clarifies a reference contained in TS 5.6.5.a.5. The current reference to LCO 3.3.1.1 does not identify the particular function to which the Core Operating Limits Report (COLR) applies. The proposed change clarifies that the setpoint limits for function 2.d of Table 3.3.1.1-1, the Average Power Range Monitor (APRM) flow biased simulated thermal power - high trip, are to be specified in the COLR. Table 3.3.1.1-1 does not need revision since footnote b already states that the allowable values for this function are specified in the COLR.

The NRC staff identified the desire for this change in the safety evaluation for Amendment No.

146. The evaluation was issued in a letter from the USNRC to William A. Eaton, "Grand Gulf Nuclear Station, Unit 1, Issuance of Amendment RE: Revision of the Minimum Critical Power Ratio Safety Limit for Cycle 12 Operation (TAC No.MB0514)" dated April 26, 2001. In this letter, the staff documented an understanding that Entergy would revise this TS in a reasonable time frame. The safety evaluation stated that this proposed clarification had no direct affect on the determination of the safety limit minimum critical power ratio (SLMCPR).

4.8 Technical Specification 5.7 TS 5.7 contains the administrative controls for high radiation areas. The proposed change involves the title descriptions for personnel performing radiation protection activities. The term "health physics technician" is changed to "health physicists" and "health physics supervision" is changed to "radiation protection supervision". The changes do not change the qualification requirements or the technical capabilities of the assigned individuals currently in the affected positions. These changes are administrative in nature and are intended to reflect the current terminology used in the industry.

Additionally, the phrase "shift superintendent on duty" in section 5.7.2 is revised to "Operations shift management". The current wording causes confusion as to the level of involvement required of the shift superintendent regarding key control. This change clarifies that the control to GNRO-2002/00030 Page 6 of 8 of keys for doors that are locked to prevent unauthorized entry into certain high radiation areas may be delegated to any of the Operations shift management positions. The proposed change does not reduce the effectiveness of controls for prevention of unauthorized entry into the locked high radiation areas.

Additionally, in subsection 5.7.3 there is a discussion concerning use of barricades and postings for areas exceeding 1000 mrem/hr. This subsection requires clarification to remove the ambiguity concerning the posting of guards. The proposed change eliminates the need for determining if a guard can be posted and proposes to either require a guard to be posted or that the area be controlled as required by the technical specification. The proposed change is more simple, straightforward and easier to understand. The requirement that restricts access to areas of high radiation levels will continue to be assured by the proposed change.

5.0 REGULATORY ANALYSIS

5.1 Applicable Regulatory Requirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met. The proposed changes to TS 3.3.7.1, 5.2, 5.5.8.b, 5.6.5.a 5, and 5.7 are editorial and maintain compliance with the intent of regulatory requirements. The proposed changes to TS 5.3.1 are intended to provide consistency between the TS and the Quality Assurance Program Manual as approved by the NRC staff. The proposed change to TS 5.5.11 is intended to maintain consistency between the TS and 10 CFR 50.59. The proposed change to TS 3.6.5 4 corrects the bounding value to be consistent with the current FSAR analyses.

Entergy has determined that the proposed changes do not require any exemptions or relief from regulatory requirements, other than the TS, and do not affect conformance with any GDC differently than described in the FSAR.

5.2 No Significant Hazards Consideration Entergy Operations, Inc. (Entergy) proposes to amend the Grand Gulf Nuclear Station, Unit I (GGNS) Technical Specifications (TS) to correct or clarify certain requirements or information.

The proposed amendment will revise TS 3.3.7.1, "Control Room Fresh Air (CRFA) System Instrumentation"; 5.2, "Organization"; 5.3.1,"Unit Staff Qualification"; 5.5.8.b, "Explosive Gas and Storage Tank Radioactivity Monitoring Program"; 5.6.5.a.5, "Core Operating Limits Report"; and 5.7, "High Radiation Area". Additionally, Entergy is requesting to correct the limiting value specified for TS 3.6.5.4, "Drywell Pressure".

Entergy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment 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.

Attachment I to GNRO-2002/00030 Page 7 of 8 The proposed changes are primarily to correct word omissions, typographical errors, reflect current terminology, and make the TS consistent with other NRC approved documents. These changes are all of an administrative nature and have no effect on any plant equipment or structures. Therefore, these changes do not increase the probability or consequences of an accident previously evaluated.

The proposed amendment also revises the allowed drywell-to-primary containment differential pressure limit. This limit is intended to ensure that containment conditions are consistent with safety analyses. The proposed smaller negative pressure ensures that the design assumptions for the containment will be met if and when a postulated loss of coolant (LOCA) should occur. Moving the limit in a conservative direction will not increase the probability or consequences of previously evaluated accidents.

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 changes do not involve a physical alteration of the plant. No new or different equipment or modes of operation are being introduced by this proposed change. Thus, the changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

The change to the allowed drywell-to-primary containment differential pressure limit does not adversely impact the ability of the containment to perform its intended function. The establishment of a more conservative limit for this parameter ensures that the plant stays within current safety analysis and therefore, can not create the possibility of a new or different kind of accident.

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

Response: No.

Margin of safety is related to the confidence in the ability of the fissioh product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The proposed changes are primarily administrative in nature and can not affect any safety barriers. The proposed change to the allowed drywell-to-primary containment differential pressure limit establishes a more conservative limit for a key parameter for the containment than is currently specified in the TS. The revised differential pressure limit is consistent with'current assumptions of the accident analysis. 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(s) present 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.

Attachment I to GNRO-2002/00030 Page 8 of 8 5.3 Environmental Considerations 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 individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in .,JO CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Attachment 2 GNRO-2002100030 Proposed Technical Specification Changes (mark-up) to GNRO-2002/00030 Page 1 of 10 CRFA System Instrumentation 3.3.7.1 3.3 INSTRUMENTATION 3.3.7.1 Control Room Fresh Air (CRFA) System Instrumentation LCO 3.3.7.1 The CRFA System instrumentation for manual isolation shall be OPERABLE. I APPLICABILITY: Modes 1, 2, and 3 During operations with a potential for draining the reactor vessel (OPDRVs).

I ACTIONS

-.---------------------. -.--.----- --. NOTE------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Place channel in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> inoperable, trip. I B. Required Action and B.1 Close associated 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> I

associated Completion isolation dampers.

Time.

GRAND GULF 3.3-73 Amendment No. +Lao, 145 to GNRO-2002/00030 Page 2 of 10 Drywell Pressure 3.6.5.4 3.6 CONTAINMENT SYSTEMS 3.6.5.4 Drywell Pressure LCO 3.6.5.4 Drywell-tojprimary containment differential pressure shall be . -0.2Bp"bsid and s 2.0 psid.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell-to-primary A.1 Restore drywell-to- I hour containment primary containment differential pressure differential pressure not within limits, to within limits.

B. Required Action and 8.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.4.1 Verify drywell-to-primary containment 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> differential pressure is within limits.

GRAND GULF 3.6-62 GAendment No. 120 to GNRO-2002/00030 Page 3 of 10 Organization 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 corporate 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 cormunication 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 relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. 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 UFSAR;

b. The plant manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant;

c. A specified corporate executive 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, and providtng technical support to the plant to ensure nuclear safety. The specif.ied corporate executive shall be documented in the UFSAR; and

d. The individuals who train the operating staff, carry out 1441&r ...-- h th p-*-e, or perform quality assurance functions may

• 'i1,i -"-i report to the appropriate onsite manager; however, these Individuals shall have sufficient organizational freedom to a c. ensure their independence from operating pressures.

(continued)

GRAND GULF 5.0-2 Amendment No. 120 to GNRO-2002/00030 Page 4 of 10 Organization 5.2 5.2 Organization (continued) 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 on site while the unit is in MODE 1, 2, or 3.

b. At least one licensed RO shall be present in the control room when fuel is in the reactor. In addition, while the unit is in MODE 1, 2, or 3, at least one licensed SRO shall be present in the control room.

c. Shift crew composition may be one less than the minimum requirement of 10CFRSO.54(m)(2)(i) and Specifications 5.2.2.a and 5.2.2.g for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immedlate action is taken to restore the shift crew o n to within the minimum requirements.

S.op

d. A health '*.W.P shall be on site when fuel is in the reactor. The position may be vacant for not more than Z hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

e. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safet related functions (e.g., licensed SROs, licensed ROs, ycJSýAf ýhea , non-licensed operators, and key maintenance personnel).

Adequate shift coverage shall be maintained without routine heavy use of overtime. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed:

1. An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, excluding shift turnover time;

2. An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7 day period, all excluding shift turnover time; (continued)

GRAND GULF 5.0-3 Amendment No. 120 to GNRO-2002/00030 Page 5 of 10 Unit Staff Qualifications 5.3 5.0 ADMINISTRATIVE CONTROLS 5.3 Unit Staff Qualifications 5.3.1 Each member of the Unit staff shall meet or exceed the minimum ifcations of fa1 Ifor comparable positions as modified by Specification 5.2.2.f, except for the radiation protection manager and the STA, who shall meet or exceed the education and experience requirements of ANSI/ANS 3.1-1981 as endorsed by Regulatory Guide 1.8, Revision 2, 1987.

'oq/

c"-' A-

/ tq us,'_ 64 v, 4, GRAND GULF 5.0.-5 Amendment No. 120 to GNRO-2002/00030 Page 6 of 10 Proqrams and Manuals 5.5 5.5 ýrograms and Manuals 5.5.8 Explosive Gas and Storage Tank Radioactivity, Monitorir, Program (continued)

b. A surveillance program to ensure that the quantity of radioactive material coitained in any outside temporary tank not including liners shipping radwaste is < 10 curies, excluding tritium and solved or entrained noble gases.

The provisions of SR 3.0.21-nd SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance freqiencies.

5.5.9 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks, and acceptabil'ty of stored fuel oil every 92 days, by determining that the fuel oil has:

1. a water and sediment contents within limits, and

2. a kinematic viscosity within limits for ASTM 2D fuel oil;

b. Total particulate concentration of the new fuel is

2 mg/lO0 ml when tested in accordance with ASTM 0-2274-70 within 7 days after addition of the new fuel to the storage tank: and

c. total Darticulate concontratior of the fuel oil in the storage tanks is < 2 mg/lO0 ml when tested every 92 days in accordance with ASTM 0-2274-70.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesei Fuel Oil Testing Program testing frequencies.

(continued)

GRAND GULc 5.0-14 Amendment No. +%0&,142 to GNRO-2002/00030 Page 7 of 10 Programs an4 Manuals 5.5.

5.5 Programs and Manuals (continued) 5.5.11 Technical Specifications iTS) 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 44vio. either of the following:

1. A change in the IS incorporated in the license; or 4Iwe, 4he v ~ 2. A Ghangs to the UFSAR or Bese: that inyavelti-*4A

-~er iewd -saFty-rs IWi 49 RAJoz

c. The Bases Control Program shall contain provisions to ensur "e that the Bases are maintained consistent with the UFSAR.

LAejo,,4 +o d. Proposed changes that do not meet the criteria of either

/OCFA ,5".S Specification 5.5.11.b.1 or Specification 5.5.11.b.2 above rshall 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).

5.5.12 10 CFR 50, Appendix 0. Testinq Program This program establishes the leakage rate testing program of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be implemented in accordance with the Safety Evaluation issued by the Office of Nuclear Reactor Regulation dated April 26, 1995 (GNRI-9S/00087) as modified by the Safety Evaluation issued for Amendment No. 135 to the Operating License. Consistent with standard scheduling practices for Technical Specifications required surviellances, intervals for the recommended surveillance frequency for Type A, B and C testing may be extended by up to 25 percent of the test interval, not to exceed IS months.

GRAND GULF 5.0-16 Amendment No. 424135 to GNRO-2002/00030 Page 8 of 10 Reporting Requi rements 5.6 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report (continued) results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

5.6.3 Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit during the pre4ious calendar ye-ar sh0A1 bte *hett y

May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste'released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and process control program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

5.6.4 Monthly Operating Reports Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the main steam safety/relief valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

5.6.5 Core Operating Limits Report (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented ir.the COLR for the following:

1) LCD 3.2.1, Average Planar Lirear 4eat Generation Rate

2) LCO 3.2.2, Minimum Critical Power Ratio (MCPR),

3) LCO 3.2.3, Linear Heat Generation Rate (LHGP.).

4) LCO 3.2.4, Fraction of Core Boiling Boundary (FCBB).

/ .) LCO 3.3.1.1, RPS Instrumentationand (g h 6) LCO 3.3.i.3, Period Based Detection System (PBDS).

GRAND GULF 5.0-18 Amendment No. +-Lt,141 to GNRO-2002/00030 Page 9 of 10 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(a), each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation is

> 100 mrem/hr but < 1000 mrem/hr, 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., health,,hYfii tcchWici:) or personnel "coninuously escorted by such individuals may be exempt from the

/'ASI1'C '57L5 RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates 10DO0 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 or 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

-L-Ith-physi* supervision in the RWP.

5.7.Z In addition to the requirements of Specification 5.7.1, areas with radiation levels a 1000 mrem/hr shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the

,Pe~wh'13 shift superintandent an duty or h supervision. Doors hishall remain locked except during periods offaccess by personnel

,""mot ee under an approved RWP that sh pecify theldose rate

• _r~,*?,*

• _*.*n )(continued)

GRAND GULF 5.0-22 Amendment No. 120 to GNRO-2002/00030 Page 10 of 10 High Radiation Area 5.7 5.7 High Radiation Area 5.7.2 (continued) 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 personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

5.7.3 In addition to the requirements of Specification 5.7.1, for individual high radiation areas with radiation levels of

> 1000 mrem/hr, accessible to personnel, that are located within large areas such as reactor containment, where no enclosure for purposes of locking, or that eampatontinuously guard and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a warning device.

GRAND GULF 5.0-23 Amendment No. 120

Attachment 3 GNRO-2002100030 Changes to Technical Specification Bases Pages (For Information Only) to GNRO-2002/00030 Page 1 of 3 Drywell Pressure B 3.6.5.4 B 3.6 CONTAINMENT SYSTEMS B 3.6.5.4 Orywell Pressure BASES BACKGROUND Drywell-to-primary containment differential pressure is an assumed initial condition In the analyses that determine the primary containment thermal hydraulic and dynamic loads during a postulated loss of coolant accident (LOCA).

If drywell pressure is less than the primary containment airspace pressure, the water level in the weir annulus will increase and, consequently, the liquid inertia above the top vent will increase. This will cause top vent clearing during a postulated LOCA to be delayed, and that would increase the peak drywell pressure. In addition, an inadvertent upper pool dump occurring with a negative drywell-to-primary containment differential pressure could result in overflow over the weir wall.

The limitation on negative drywell-to-primary containment differential pressure ensures that changes in calculated peak LOCA drywell pressures due to differences in water level of the suppression pool and the drywell weir annulus are negligible. It also ensures that the possibility of weir wall overflow after an inadvertent upper pool dump is minimized. The limitation on positive drywell-to-primary containment differential pressure helps ensure that the horizontal vents are not cleared with normal weir annulus water level.

APPLICABLE Primary containment performance is evaluated for the entire SAFETY ANALYSES spectrum of break sizes for postulated LOCAs. Among the inputs to the design basis analysis is-the initial drywell internal pressure (Ref. 1). The initial drywell internal pressure affects the drywell pressure response to a LOCA (Ref. 1) and the suppression pool swell load definition (Ref. 2).

Additional analyses (Refs. 3 and 4) have been performed to show that if initial drywell pressure does not exceed the negative pressure limit, the suppression pool swell and vent clearing loads will not be significantly increased and the probability of weir wall overflow is minimized after an inadvertent upper pool dump.

(continued)

GRAND GULF B 3.6-120 Revision No. 0 to GNRO-2002/00030 Page 2 of 3 Drywell Pressure B 3.6.5.4 BASES APPLICABLE Drywell pressure satisfies Criterion 2 of the NRC Policy SAFETY ANALYSES Statement.

(continued)

LCO A limitation on the drywell-to- mary containment differential pressure of ; -0.2ipsid and s 2.0 psid is required to ensure that suppression pool water is not forced over the weir wall, vent clearing does not occur during normal operation, containment conditions are consistent with the safety analyses, and LOCA drywell pressures and pool swell loads are within design values.

APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to the primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, maintaining the drywell-to-primary containment differential pressure limitation is not required in MODE 4 or 5.

ACTIONS A.1 With drywell-to-primary containment differential pressure not within the limits of the LCO,- it must be restored within I hour. The Required Action is necessary to return operation to within the bounds of the safety analyses. Thd I hour Completion Time is consistent with the ACTIONS of LCO 3.6.5.1, *Drywell,* which requires that the drywell be restored to OPERABLE status within I hour.

B.1 and B.2 If drywell-to-primary containment differential pressure cannot be restored to within limits within the required Completion Time, 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 at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

(continued)

GRAND GULF B 3.6-121 GRevision No. 0 to 'a-GNRO-2002/00030 Page 3 of 3 Drywell Pressure B 3.6.5.4 BASES (continued)

SURVEILLANCE SR 3.6.5.4.1 REQUIREMENTS This SR provides assurance that the limitations on drywell-to-primary containment differential pressure stated in the LCO are met. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />'Frequency of this SR was developed, based on operating experience related to trending of drywell pressure variations during the applicable MODES and to assessing proximity to the specified LCO differential pressure limits. Furthermore, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal drywell pressure condition.

REFERENCES 1.-..UFSAR, Section 6.2.1.

2. UFSAR, Section 3.8.

3. UFSAR, Section 6.2.1.1.6.

4. UFSAR, Section 6.2.7.

GRAND GULF B 3.6-122 Revision No. 0