ML20081D041
| ML20081D041 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 03/15/1995 |
| From: | Denton R BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20081D045 | List: |
| References | |
| NUDOCS 9503200105 | |
| Download: ML20081D041 (41) | |
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RosEar E. DENTON Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant y#
Vice President Nuclear Energy 1650 Calvert Cliffs Parkway f
Lusby, Maryland 20657 3,
410 586-2200 Ext.4455 local Q:
410 260-4455 Baltimore
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,..V s' i-9 March 15,1995
}r IJ. S. Nuclear Regulatory Commission h
Washington,DC 20555 4
- y ATTENTION
Document Control Desk ys
SUBJECT:
Calvert Cliffs Nuclear Power Plant 9
Unit Nos. ! & 2; Docket Nos. 50-317 & 50-318 License Amendment Request: Administrative Controls Section 6.0 Upgrade and k
Ouality Assurance Policy Chance i.'
3, 7
Pursuant to 10 CFR 50.90, the Baltimore Gas and Electric Company (BGE) hereby requests an
~.
Amendment to Operatmg License Nos. DPR-53 and DPR-69 by the incorporation of the changes described A
below into the Technical Specifications for Calven Cliffs Unit Nos. I and 2.
Pursuant to l~
10 CFR 50.54(a)(3), BGE also requests approval of proposed cLanges to the Quality Assurance (QA)
Policy for Calven Cliffs Unit Nos. I and 2.
A.;;
- F..
DESCRIPTION
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This amendment proposes to reformat the current Administrative Controls Section of the Calvert Cliffs h,
Nuclear Power Plant Technical Specifications consistent with NUREG-1432, " Standard Technical Z
Specifications, Combustion Engineering Plants," (Reference a) and to relocate several requirements to 1
other documents and programs as recommended in References (b), (c), and (d). On February 6,1995, the p
NRC issued a License Amendment for the R. E. Ginna Nuclear Power Plant which is similar to this R
request.
~.'
Q Some of the Technical Specification requirements (e.g., Review and Audit Function) are relocated to the
?
QA Policy which implements 10 CFR 50.54 and 10 CFR Part 50, Appendix B. The proposed relocations X
will retain an appropriate level of regulatory authority and control. In addition, we are proposing changes 1
to some of the items relocated from the Technical Specifications to the QA Policy. These changes would 0.:
nxluce the frequency of some audits and would take an exception to the biennial review of plant procedures O
required by ANSI N18.71976.
Nuclear Regulatory Commission approval is required prior to 7
implementing these changes since they constitute a reduction to a QA Policy commitment previously h..
9503200105 950315 2.7 PDR ADDCK 05000317 I kt M
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Document Control Desk March 1% 1995 Page 2 '
accepted A similar r~uaa to the QA Program audit frequencies was granted to Detroit Edison Company on August 4,1994, and approval for using alternate programs and activities to accomplish the bienmal procedure review was granted to North - Atlantic Energy Sc vices Corporation ' on November 10,1994. Attachment (7) contams the basis for concluding that the revised Policy continues to satisfy Appendix B of 10 CFR Part 50 and that the scope of the Policy has not been reduced.
Anathar iL=ne Amendment Request is being made under separate letter to modify the Radiological Effluent Techmcal Specifications (i.e., Sections 3/4.11 and 3/4.12) in accordance with Generic Latter 89-01 (Reference c). We have incorporated the is-... =4= dons of Generic letter 89-01 into the Adnunistrative Controls Section contained herein with the exception of the Radiological Environmental Monitoring Program discussion which we propose to relocate to the Offsite Dose Calculation Manual consistent with the i.:+==.dations of Reference (b). Additional changes to Section 6 involve replacing plant-specific personnel titles with genene titles in accordance with Reference (f).
BACKGROUND
'Ihc "NRC Interun Policy Statement on Technical Specification Improvements for Nuclear Power Reactors," published on February 6,1987 (52 FR 3788), proposed criteria for defining the scope of the Technical Specifications. 'Ihe policy envisioned that many existing Technical Specification requirements could be relocated to other more appropriate programs and documents which would then be cont.olled by present regulations (e.g.,10 CFR 50.59 and 10 CFR 50.54). The relocation would result in more concise Technical Specifications focusing on the most safety significant requirements.
Followmg issuance of the interirr Policy Statement, the utility Owners Groups and the NRC staff developed improved standard technical specifications (STS) based on the criteria in the interim Policy Statement for each of the four Nuclear Steam Supply System vendors. In September 1992, five NUREGs were published which contained the improved STS for the four vendors (two NUREGs were for boiling water reactors). NUREG-1432 contains the improved STS for Combustion Engineering plants. The NUREGs placed emphasis on human factors principles to clarify and streamline the STS.
On July 22,1993, the Commission issued a " Final Policy Statement on Technical Specification improvements for Nuclear Power Reactors" (58 FR 39132). 'Ihe Commission's policy statement described the safety benefits of the improved STS and encouraged licensees to use the improved STS as the basis for plant-specific Technical Specifications.
Following publication of the NUREGs, the NRC proposed in Reference (b) that significant reductions to the Adnunistrative Controls Section of the STS be done to climinate duplication with other regulatory requirements. 'Ihe proposed changes were evaluated by the Owners Groups and revisions to the Administrative Controls Section of the STS were recv..wi.cr.ded in Reference (c) to complement the NRC's proposal. Baltimore Gas and Electric Company has evaluated the Owners Groups proposed changes and has decided to submit this License Arr.cr.dir.cr.t Request which implements the improved STS format for the Administrative Controls Section and relocates some present Technical Specification requirements consistent with Reference (c).
Document Control Det.k March 15,1995 Page 3 REOUESTED CliANGE Detailed descriptions of the requested changes are contained in Attachment (1). A sumnuuy of where requirements will be located is given in Attachment (2). Attachment (3) classifies the proposed changes into categories. %c marked-up pages to the Technical Specifications for Units 1 and 2 are contained in Attachments (4) and (5), respectively. The change number is shown in the left hand margin of these attachments. Due to the complexity of this amendment request, a clean copy of Unit l's proposed Technical Specifications is contained in Attachment (6) to help visualize the changes. He Unit 2 proposed Technical Specifications are the same and are not attached. Attachment (7) discusses changes to the QA Policy along with thejustification for reducing commitments.
SAFETY ANALYSIS The proposed changes will provide significant human factors improvement for the Technical Specifications by: 1) locating similar requirements within the same section; 2) deleting unnecessary requirements;
- 3) relocating requirements to other BGE controlled documents; and 4) adding new requirements consistent with the standard format contained in NUREG-1432. The proposed changes have been grouped according to these four categories and are discussed in Attachment (3). Strictly editorial changes are discussed under Category 5, and other changes falling outside these categories are discussed under Category 6.
The kachment also provides where cach change belongs relative to these categories. The end result is Technical Specifications which are easier to use due to these hur an factor improvements.
Two significant changes involve removing the Minimum Shift Crew Composition Tabic, and Resiew and Audit Function Section from the Technical Specifications. The Minimum Shift Crew Composition is adequately controlled by regulations and plant procedures where change control is provided by 10 CFR 50.59. The Review and Audit Function is comprised oficquirements for the Plant Operations and Safety Review Committee, Procedure Review Committee, Qualified Reviewers and the Offsite Safety Review Committee. These requirements are relocated to the QA Policy where change control is prosided by 10 CFR 50.54(a) requiring NRC approval for any commitment reduction. Attachment (7) contains the QA Policy and provides the necessary justification for reducint; audit f.equencies and taking an exception to the biennial review of plant procedures required by ANSIN18.7-1976. Since sufficient regulatory controls exist for the QA Policy, relocating these requirements from the Technical Specifications to the QA Policy provides an appropriate level of control so that the proposed relocations will not impact plant safety. Similarly, requirements which are contained in plant procedures controlled by the 10 CFR 50.59 review proen will provide adequate control to ensure plant safety. Other relocated requirements will be controlled by established programs, for which changes are also controlled by regulatory requirements; these programs include the Offsite Dose Calculation Manual, Updated Final Safety Analysis Report, Emergency Plan, and Security Plan. The complete justification for these relocations along with justifications for all proposed changes are given in Attachment (1).
A new provision was added to provide additional administrative flexibility in scheduling control room personnel. The proposed change would allow a qualified on-shift Senior Reactor Operator license holder to serve a dual role as a Shift Technical Advisor (STA). The dual role STA is consistent with the Commission's Policy Statement on Engineering Expertise on Shift and is the preferred option for improsing the ability of shift operating personnel to recognize, diagnose, and effectively deal with plant transients or
i Document Control Desk March 15,1995 Page 4 j
i other abnormal conditions. Another change was made to establish a Technical Specification Bases Control Program. The purpose of the program is to specifically delineate the appropriate methods and reviews required for a change to the Bases. Both proposed changes enhance the Technical Specifications and are consistent with NUREG-1432.
i Two reports were determined to present unwarranted administrative burden and their submission to NRC are proposed for deletion. 'Ihe elimination of the requirement to submit the Startup Report does not affect plant safeiy since it is required to be submitted 90 days following completion of the testing. The report provided a mechanism for NRC to review the appropriatenese of BGE activities after-the-fact, but contained no requirement for NRC approval. Since the Updated Final Safety Analysis Report and plant procalures provide assurance that the prescribed activities are adequately performed and corrective actions taken when required, the climination of this reporting requirement removes an admmistrative burden not affecting plant safety. Similarjustification is provided for eliminating the Special Report on iodine activity levels. Since the plant can cordnue operating under those conditions requiring this Special Report and NRC approval is not necessary for continued operation, the report clearly was not intended to assure safe operation of Calvert Cliffs. Considering that the regulations provide reporting requirements for plant shutdown when iodine activity levels exceed limits and given the low safety significance of reporting operational iodine activity levels, this Special Report is likewise an unwarranted administrative burden.
An adattional change is proposed which replace the film badge with the electronic personal dosimeter as a personnel monitoring device option. This device provides a more effective, efficient, state-of-the-art option for estimating dose, and accordingly, enhances personnel safety. The proposed change will not affect plant or personnel safety.
With the exception of the elimination of the two reporting requirements which are justifiable, the proposed changes do not decrease requirements and provide an appropriate level of change control for relocated requirements so that plant safety remams unaffected. All requirements relocated from the Technical Specifications have been evaluated with respect to the four criteria of the NRC " Final Policy Statement On Technical Specification Improvements," and found to meet none of the criteria for inclusion in the Technical Specifications. The proposed changes result in Technical Soccifications which focus on the most important requirements without sacrificing safety.
DETERMINATION OF SIGNIFICANT HAZARDS The proposed changes have been evaluated against the standards in 10 CFR 50.92 and have been deterrnined to not involve a significant hazard 2 consideration, in that operation of the facility in accordance with the proposed amendments:
1.
Would not involve a signifcant increase in the probability or consequences of an accident previously evaluated.
Relocating existing requirements, climinating requirements which duplicate regulations and makmg administrative improvements provide Technical Specifications which are easier to use. Because existing requirements are relocated to established programs where changes to those programs are controlled by regulatory requirements, there is no reduction in commitment and adequate control is
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Document Control Desk l
March 15,1995 Page 5 still maintained. Likewise, the elimination of requirements which duplicate regulations enh mm the usability of the Technical Specifications without reducing commitments. Tne admmistrative improvements being proposed neither add nor delete requirements, but merely clarify and improve the readability and understanding of the Technical Specifications. Since the requirements remain the same, these changes only affect the method of presentation and are considered admmistrative, and as such, would not affect possible initiating events for accidents previously evaluated or any system functianni requirement. Therefore, the proposed changes would not involve a significant increase in the probability or consequences of an accident previously evaluated.
Since the Shift Technical Adviser (STA) is not considered an initiator to any previously evaluated accident nor considered in the accident's response, the use of a dual role STA would not increase the probability or consequences of any previously evaluated accident.
The Technical Specification Bases Control Program provides controls which ossure appropriate reviews of changes to the Bases. Because NRC approval is still needed for changes to the Bases which affect the Technical Specifications, the proposed Program would not affect the probability or consequences of a previously evaluated accident.
Eliminating the requirement for submitting two reports which place unwarranted admini=trative burden on both Baltimore Gas and Electric Company and the NRC has no affect on the probability or consequences of an accident previously evaluated. Therefore, the proposed changes would not invohr a significant increase in the probability or consequences of an accident previously evaluated.
Replacing the film badge with the electronic personal dosimeter provides a more effective, efficient, state-of-the-art option for estunating dose and would not impact accidents previously evaluated.
Therefore, the proposed change would not involve a significant increase in the probability or consequences of an accident previously evaluated.
2.
Would no* create the possibility of a new or dferent type of accident from any accident previously evaluated.
As discussed previously, relocating existing requirements, eliminating requirements which duplicate regulations and making administrative improvements are all changes that are administrative in nature. The changes will not afrect any plant system or structure, nor will they it affect any system functional or operability requirements. Consequently, no new failure modes are introduced as a result of the proposed changes. 'Iherefore, these type of changes would not create the possibility of a new or different type of accident from any accident previously evaluated.
Because the STA does not perform equipmen* design or equipment manipulation, the use of a dual role STA would not create the possibility of a new or different type of accident from any accident previously evaluated. Since the Technical Specification Bases Control Program represents an administrative function designed to improve control over changes, it too would not create the possibility of a new or different type of accident from any previously evaluated.
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D===# Control Desk j
March 15,1995 4 Page 6 -
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- i A purely administrative 6mr*ian such as report submittals would not' change the configuration or -
l operation of the plant.' Co-Ply, the ehmination of the i4.a.: to subnut the Stantup i
Report and the Special Report deshag with iodme activity levels, would not create the possdnhty j
of a new or different type of wiamt from any wi&mr piduely evaluated.
l Since the operation or configuration of the plant is not changed by the type of personal dosimeter, this change would not create the possibility of a new or different type of accident from any accident l
previously evaluated.
i Derefore, the proposed changes would not create the possibility of a new or different type of -
j accident from any accident previously evaluated.
i 3.
Would not involve a signipcant reduction in a margin ofsafety.
Since relocation of existing requirements, climmation of requirements which duplicate regulations, j
and improvements made to clarify the Technical Specifications would not affect requirements, the.
~l proposed changes would not affect the Updated Final Safety Analysis Report design bases, accident analysis assumptions or Technical Specification Bases. In -Maa, these changes do not
-l affect effluent release limits, momtoring equipment or practices. Therefore, the proposed changes
'l would not involve a significant reduction in a margin of safety.
t The use of an STA should provide an additional margin of safety in the accident response function i
oflicensed operators beyond that considered in the accident analysis. Since the STA is required to
.l have the same traimng and educational qualifications in either the individual or dual role, the use of a dual role STA should have minimal impact. Consequently, the proposed change would not involve a significant reduction in a margin of safety. De Technical Specification Bases Control Program is an administrative change controlli g how Technical Specification basis information is i
reviewed and incorporated. Derefore, this change would not involve a significant rhiaa in a i
margm of safety.
]
Activities described in the Startup Report will continue to be performed and corrective action taken when required. Sindlarly, iodine activity levels will continue to be monitored and actions taken
)
including the issuance of a Licensee Event Report when conditions warrant Considering the above, climmation of the two reporting requirements would have no impact on the margin of saty.
Plant operating pru-.c.crs are not affected by the type of personnel monitormg device used and as q
a consequence, would not impact a margin of safety. Since the rep!- dosimeter provides a i
more effective mechanism for estimating dose, there is no degradation in personal safety levels.
Ca= gently, the proposed change would not involve a significant reduction in a margm of safety.
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Document Control Desk l
March 15,1995 t
Page 7 ENVIRONMENTAL ASSESSMENT
'Ihe proposed =*t would change record keeping, reporting, and administrative procedures and requirements. Therefore, the proposed =de meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(10). Pursuant to 10 CFR 51.22(b), no enviru..w-.1 impact statement or environmental assessment is needed in connection with the approval of the proposed amendment.
SCHEDULE
'Ihis change is requested to be approved and issued by September 1,1995. However, issuance of this amende is not currently identified as having an impact on outage completion or continued plant operation. We request that upoa NRC approval of the license amendment or QA Policy, whichever is later, it be implementud within 60 days.
}
SAFETY COMMITTEE REVIEW These proposed changes to the Technical Specifications ami our deternunation of significant hazards have been reviewed by our Plant Operations and Safety Review Committee and Offsite Safety Review Committee. 'Ihey have concluded that implementation of these changes will not result in an undue risk to l
the health and safety of the public.
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Document Control Desk-March 15,1995 i
- Page 8 Should you have questions regarding this matter, we will be pleased to discuss them with you.
Very truly yours,.
/
t i
t STATE OF MARYLAND
- TO WIT:
COUNTY OF CALVERT I
9 I hereby certify that on the /
day of dheo b
.1995 before me, the subscriber, a Notary Public of the State of Maryland in and for
/k /v,+ t "
- L,
. personally appm.:cd Robert E. Denton, being duly sworn, and states that he is Vice Prc +Jent'of the Baltimore Gas and j
Electric Company, a corporation of the State of Maryland; that he provides the foregoing response for the purposes therein set forth; that the statements made are true and correct to the best of his knowledge, information, and belief; and that he was authorized to provide the response on behalf of said Corporaten i
WITNESS my Hand and Notarial Seal:
81 o A/h l/V4.
[ Notaiy Public i
/!i :g a m:;.
67 6.9.1.5.a/6.6.1 Deleted film badge measurements as an option to estunate dose and replaced it with electronic personal dosimeter. (6)
Justification:
Replacing the film badge measurement with the electronic personal dosimeter as a means for estimating dose, provides a more effective, efficient, state-of-the-art option for estimating dose and is i
accordinglyjustified.
.q:. +%:Pr.u:c 68 6.9.1.5.a/6.6.1 Replaced the word "shall" with "should" in the followmg sentence.
In the aggregate, at least 80% of the total whole body dose received from external sources "shall" be assigned to specific mapr work functions. (4)
Justification:
The replacement of the word "shall" with "should" provides some relaxation from the requirement to assign whole body doses received from extemal sources to major west functions. Since the dose assignment is an administrative function not impacting accident initiators or consequences and the change is consistent with NUREG 1432 and Regulaf.ory Guide 1.16, the slight relaxation in the requirement will not impact plant safety and is justified.
v, 21
ATTACHMENT (1)
DESCRIPTION OF CIIANGES WITH JUSTIFICATION SPECIFICATION DESCRIPTION OF CHANGE CHANGE OLD/NEW (ATTACHMENT 3.TUSUFICADON CATEGORY) 69 6.9.1.5.b/-
Deleted entire section discussing reporting requirements of the steam generator tube insenice inspection since the discussion is contamed in Specification 4.4.5.5.b. (1)
Justification:
The detailed description ofitems to be reported annually for the steam generator tube inenar'tian is contained in i
Specification 4.4.5.5.b.
Since this Specification contains other reporting requirements for the steam generator tube inspartinae, retaining the information in that location is logical.
70 6.9.1.5.c/6.6.6 His section concerning failures and challenges to pressurizer power-operated relief valves (PORVs) or sdety vJves was renumbered from 6.9.1.5.c to 6.6.6 and given the title, " Pressurizer PORV and Safety Valve Report." Editorial changes were made and the following sentence was added "A report shall be submitted prior to March 1 of each year hnenting all failures and challenges to the pressurizer PORVs or safety valves." His requirement is in the current Specification 6.9.1.4. (5) 71 6.9.1.6/6.6.4, Monthly This section was renumbered from 6.9.1.6 to 6.6.4 and the title is no Operating Report longer all capitalized. Deleted informaton as to where to send the report and copies since this item is adequately covered by 10 CFR 50.4. (5) m 72 6.9.1.7/6.6.2, Annual This section was renumbered from 6.9.1.7 to 6.6.2 and the title is no Radiological Environmental longer all capitalized. The word " Routine" was deleted in describing Operating Report this report. The entire report title was deleted after the first mention of the entire title in the text. Additional editorial changes include changing " reports" to " report," " analysis" to " analyses," and " report period" to "reposting period." The reference to Table and Figures in the ODCM is no longer capitalized. (5) 73 6.9.1.7/6.6.2 Changed the wording from "the radiological emironmental surveillance activities" to "the Radiological Etwironmental Monitoring Program." The wording is consistent with the Standard Technical Specifications and the guidance contained in Generic Letter 89-01. (4)
Justification:
Changing the " radiological environmental surveillance actisities" to the " Radiological Emironmental Monitoring Program" is consistent with NUREG-1432 and Generic letter 89-01, which requires a Radiological Environmental Monitormg Program. The Monitoring Program introduces no new surveillance reqairement, but according to Generic Letter 89-01 will provide the programmatic controls necessarv for the simplifica'. ion of the RETS.
22
1 ATTACHMENT (1)
DESCRIPTION OF CHANGES WITH JUSTIFICATION SPECIFICATION DESCRIPTION OF CHANGE CHANGE OLD/NEw (ArrACHMENT 3 MSTFICATION CATEGORY) 74 6.9.1.7/6.6.2 Deleted discussion that the report is to include compansons with preoperational studies,' operational contmis, and ~ previous environmental surveillance reports. Also deleted that the report is to include an assessment of observed impacts of operation on the environment Replaced these del *== with, "De material provided in the report is to be consistent with the objectives authnad in the i
ODCM and 10 CFR 50, Appedix I, Sections IV.B.2, IV.B.3, and l
IV.C." He footnote stating that a single submittal may be made was *; = 'si It now reads, "A single subauttal may be made for Calvert Cliffs. The submittal should combine those =cria== that are i
common to both units." (3) j Justification:
References to 10 CFR 50, Appendix I, Section IV.B.2, IV.B.3, and IV.C were added to help support deletion of the report content Report content deleted from the Specification not coured by tir regulations will be incorporated into the ODCM. In addition, i
proposed Specification 6.5.1, "Offsite Dose Calculation Manual (ODCM)," requires the format and content of the Annual Radiological Emironmental Operating Report be included in the ODCM. Since report content is still controlled and content remains the same, the proposed deletions are justified.
75 6.9.1.7/-
Deleted discusion for including description of the radiological environmental monitonng program and sampling location maps, l
including footnote * (on p.6-20) stating that one map shall cover i
stations near the site boundary and a second shall include more distant stations. (3)
Justification:
See Change 74.
76 6.9.1.7/-
Deleted discussions on Interlaboratory Comparison Program, deviations from sampling schedule, non-achievable analyses from Table 4.12-1, cause of the unavailability of samples, and 1
identification of new locations for replac~naar samples. (3) j Justification:
See Chanac 74.
77 6.9.1.8/6.6.3, Semi-Annual his section was renumbered from 6.9.1.8 to 6.6.3 and the title is no Radioactive Effluent Release longer all capitalized. Deleted the word " Semiannual" from the title.
Report The word " routine" was also deleted in describing the report. He entire report title, " Radioactive Effluent Release Report," was deleted after the first mention of the entire title in the text.
Additional editorial changes include changing " reports" to " report,"
and " plant" to " units." (5) 23
ATTACHMENT W DESCRIPTION OF CIIANGES WITII JUSTIFICATION I
SPECIFICATION DESCRIPTION OF CilANGE 3
l CHANGE OLD/NEW (A*ITACHMENT 3 JUSTD1 CATION CATEGORY):
78 6.9.1.8/6.6.3 The dates for submitting the report and the report penod has been deleted and replaced with "in a 4.se with 10 CFR 50.36a (i.e., time berv.a subnuttal of the reports must be no longer than 12 months)." Deleted reference to RPM Guide 1.21 and added i
that the material is to consistent with the objectives outlined in the ODCM and PCP, and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.I. (3)
Justificatica:
{
'Ihe semi-annual report sub,aittal requirement is changed to conform j
with 10 CFR 50.36a which requires an annual report. The 60 day wmdow for submitting the reports after January I and July 1 of each 3
year is no longer retained since 10 CFR 50.36a specifies that the time period between reports must be no longer than 12 months
{
References to 10 CFR 50.36a and 10 CFR 50, Appendix 1, SectionIV.B.1 were added to help support deletion of the report content. Report content deleted from the Specification not covered by regulations will be incorporated into the ODCM. In addition, proposed Specification 6.5.1, "Offsite Dose Calculation Manual (ODCM)," requires that the fonnat and content of the Radioactive Effluent Release Report be included in the ODCM. Since report i
content is still controlled and content remains the same, the proposed i
deletions arejustified.
79 6.18/6.6.3 Added requirement for reporting licensee initiated major changes to the Radioactive Waste Systems along with specific details as to what to include in the report. '1his discussion is contained in current Specification 6.18, which is being deleted. (1)
Justification:
Information to be included in the Radioactive Effluent Release Report should be centrally located under that report heading to facilitate use. The requirement for reporting licensee initiated major changes to the Radioactive Waste Systems along with specific i
report details falls under the requirements of the Pa%ive Effluent Release Report. Consequently, relocating that information is justified and consistent with NUREG-1432.
- a t
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ATTACHMENT m DESCRIPTION OF CHANGES WITH JUSTIFICATION SPECIFICATION DESCRIPTION OF CHANGE CHANGE OLD/NEW (ATTACHMENT 3 JUSTIFICA* DON CATEGORY) 80 6.9.1.8/-
Deleted discussions about meteorological data along with the assessment of the previous year's releases. Deleted e Lw to acceptable methods for calculating the dose contribution from liquid and gaseous emuents. Deleted footncAe allowing Sr89 and Sr90 analyses results to be submitted in a suppWay report, and another footnote allowing the summary of required me,eciological data to be retained onsite rather than suH*ia= with the first half-year Radioactive Effluent Release Report.
Deleted detailed information required for each class of solid waste, rharnesian on description of unplanned releases to unrestricted areas and changes made to the PCP. (3)
}
Justification:
i Report content deleted from the Specification not covered by regulations will be incorporated into the ODCM. In addition, proposed Specification 6.5.1, "Offsite Dose Calculation Manual (ODCM)," requires that the format and content of the Radioactive Emuent Release Report be included in the ODCM. Since report content is still controlled and content remains the same, the proposed deletions are justified.
81 6.9.1.8/6.6.3 Added reference to Specification 6.5.1.c where additional l 1
information is contained concerning Radioactive Emuent Release.
Report content. The following s,.ntence was added
'T.: rge:t shall include changes to the ODCM in accordance with Specification 6.5.1.c."
The reference was added so that all the requirements for the Radioactive Effluent Release Report are centrally located. Since the requirements remain the same, the change is purely editorial. (5) 82 6.9.1.8/-
Deleted reference to Specification 3.12.2 requiring a listing of new locations for dose calculations identified by the land use census. (2)
Justification:
Specification 3.12.2 requires that new locations for dose calculations identified by the land use census be identified in the Annual Radiological Environmental Operating Report.
Specification 3.12.2 does not require us to include that information in the P=Anadive Effluent Release Report. Since the information is contained in the Annual Radiological Environmental Operatmg Report, deletion of that information from the Radioactive Effluent Release Report is justified.
83 6.9.1.9/6.6.5, Core Operating Changed number from 6.9.1.9 to 6.6.5 and added acronym (COLR) r Limit Report to title. (5) 25
ATTACHMENT ft)
DESCRIPTION OF CHANGES WITH JUSTIFICATION SPECIFICATION DESCRIPTION OF CHANGE r
CHANGE OLD/NEw (ATTACHMENr 3 JUS 11F1CA110N CATEGORY) 84 6.9.2/, Special Reports Deleted entare section covenng special report subnuttals to the NRC Regional Office since this information is =dminad in individual l
Section 3 Speci6 cations. (2) l Justification:
'Ihe special report section duplicates the requiranents =dminad in individual Section 3 Specifw=hane. Since these reports are not routmely required, but are necessary to report a special ocamrrence, their location within the individual Specificatian is appropriate. As such, deletion of these requirements from Section 6 is justified.
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2 85 6.10/, Record Retention Deleted entire section covenng what records are to be retamed and l
the period of retention since the requirements are addressed in plant procedures. (3) i i
Justification:
Records will h-the plant's activities in such areas as normal l
startup, operataan and shutdown, abnormal conditions and emergencies, refueling, safety-related==" ~- =, surveillance and l
testmg, and radiation control Record retention provides retrievability for review of compliance with requirements. Such'a review does not directly assure operatica in a safe manner since the activities described in these documents have already been performed.
l Regulations (i.e.,10 CFR 20, Subpast L and 10 CFR 50.71) require j
the retention of certain records. In addition, plant procedures, which are controlled by the 10 CFR 50.59 review process, address the retention requirements for all records proposed for deletion.
j Considenng that appropriate control is maintained by 10 CFR 50.59 l
and that record retendon does not directly assure safe operation, the i
deletion is justified.
l 1
86 6.11/, Radiation Protection Deleted entire section stating that procedures for peri,c. d radiation Program protection shall be prepared consistent with 10 CFR Part 20 and shall be approved, maint=H and adhered to for all operations i
invohing personnel radiation exposure since this discussion is duplicative of 10 CFR Part 20 requirements. (2)
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Justification.
i Since the requirements for a Radation Protection Program are i
contained in 10 CFR 20.1101, elimination of this section is justified.
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1 ATTACHMENT (1)
DESCRIPTION OF CHANGES WITH JUSTIFICATION j
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SPECIFICATION DESCRIPTION OF CHANGE CHANGE OLn/NEw (ArrACHMENT 3 AnmFICAT10N CA1EGORY) 87 6.12/,High Radiation Area Deleted entire section discussing controls used in High R=A=han areas since regulatory requirements are suf5cient to protect plant g i s d. (2)
Justification:
' Ibis section provides access controls for high radiation areas in lieu of those specified in 10 CFR 20.1601. Since plant procedures controlling access to high radiation areas will reference 10 CFR 20.1601, adequate control is inmintainad over the area to
- q ensure safety of nuclear plant permannal. Furthennore, access i
controls have no impact on nuclear safety. In adatian, since plant personnel are not " members of the public," the principal operative standard in Section 182.a of the Atomic Energy Act; " health and safety of the public" does not apply. Based on these considerations, it is not necessary to include these controls in the Spacinentian and the deletion isjustified.
no m
88 6.13/6.5.3, System Integrity Changed number from 6.13 to 6.5.3 and changed title from " System Integrity" to " Primary Coolant Sources Outside C-a *
= ^ "
Moved entire section to proposed Specification 6.5,.3, " Primary
]
i Coolant Sources Outside Containment." (1) l i
Justification:
j
'Ihis change is justified since it locates a program to a common
)
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location consistent with NUREG-1432.
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i 89 6.14/, Iodine Monitoring Deleted entire section discussing the Iodine Monitonng Program since these provisions will be relocated to the UFSAR. (3)
Justification:
'Ihe Iodine Monitonng Program provides controls to ensure the capability to accurately determine the airborne iodme concentration in vital areas under accident conditions. 'Ihe purpose of the program is to mimmize radiation exposure to plant personnel i
following an accident. Since it does not include monitoring process variables that are initial conditions for a design basis tranaiant or accident, nor involve a primary success path to mitigate a deeign basis accident, the program description will be relocated to the UFSAR. Details of the program are contained in plant operation manuals (e.g.,
himy procedures, training instructions, maintenance procedures, and ERPIPs) anxl changes are controlled by 10 CFR 50.59.
Since appropriate control is===i='aM the relocation is justified.
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ATTACHMENT (1)
DESCRIPTION OF CHANGES WITH JUSTIFICATION SPECIFICATION DESCRIPTION OF CHANGE CHANGE OLD/NEW (ATTACHMENT 3 JUSTIFICA'UON CATEGORY) 90 6.15/6.5.2, Post-Accident Changed number from 6.15 to 6.5.2 and the title is no longer all Sampling capitalized. Movul entire section along with its footnote to proposed Specification 6.5.2, " Post-Accident Sampling." (1)
Justification:
This change is justified since it locates a program to a common location consistent with NUREG-1432.
91 6.16/~, Process Control Deleted the requirement that the PCP be approved by the Program (PCP)
Commission prior to implementation since the Program has already been implemented. Deleted the review and approval authority discussion since this requirement is described in a plant procedure.
(3)
Justification:
Procedure RP-2-100 implements and describes the PCP including the review requirements, and ensures compliance with 10 CFR Part 20, Part 61 and Part 71. Since the regulations address the transfer and disposal of licensed material, including the classification of waste and waste characteristics, the regulations provide adequate i
controls over the PCP. In addition, changes to plant procedures are controlled by 10 CFR 50.59. Because the controls are adequate, elimination of the PCP from the Specification is justified.
92 6.16/-
Deleted the rest of the section describing the content of the 4
Radioactive Efiluent Release Report for licensee initiated changes to the PCP since the ODCM will contain this information. (3)
Justification:
Report content deleted from the Specification not covered by regulations will be incorporated into the ODCM. In addition, proposed Specification 6.5.1, "Offsite Dose Calculation Manual (ODCM)," requires that the format and content of the Radioactive Effluent Release Report be included in the ODCM. Since report content is still controlled and content remams the same, the proposed deletions are justified.
93 6.17/, Offsite Dose Deleted title "Offsite Dose Calculation Manual (ODCM)" and Calculation Manual (ODCM)
"6.17." (5) 28
ATTACHMENT (1)
DESCRIPTION OF CHANGES WITH JUSTIFICATION SPECIFICATION DESCRIPTION OF CHANGE CHANGE OLn/NEW (ATTACHMD4T 3 JUSTIFICA~ DON CATEGORY) 94 6.17.1/-
Elinunated section 6.17.1 which states that the ODCM shall be approved by the Commission prior to implementation. (5)
Justification:
Since the ODCM has already been implemented, deletion of the statement that the ODCM shall be approved by the Conunission prior to implementation is a moot point.
+
95 6.17.2/6.5.1 Deleted section number "6.17.2" and moved entim secten to proposed Specification 6.5.1, "Offsite Dose Calculation Manual (ODCM)," after reformatting and addmg more de-tation requirements for licensee initiated changes to the ODCM as discussed in Change 53. (1)
Justification.
The requirement for documenting licensee initiated changes to the ODCM along with discussing its content has been moved to a new Section 6.5, " Program and Manuals." "Ihis change is justified since it locates a manual to a common location consistent with NUREG-1432.
6.18/-
Deleted title " Major Changes to Radioactive Liquid, Gaseous, and Solid Waste Treatment Svstems" (5) 97 6.18.1/6.6.3 Deleted section number "6.18.1" and moved entire section to proposed Specification 6.6.3,
" Radioactive Emuent Release l
Report." Replaced "POSRC" with "onsite resiew function" and deleted " Semiannual" from the title of the Radioactive Emuent Release Report. (1)
Justification:
Information to be incloded in the Radioactive Emuent Release Report should be centrally located under that report headmg to facilitate use. The requirement for reporting licensee initiated major changes to the Radioactive Waste Systems along with specific report details falls under the requirements of the Radioactive Emuent Release Report. Consequently, relocating that information is justified and consistent with NUREG-1432. Replacing reference to "POSRC" with the phrase "onsite review function" is desirable since the specific name may change, whereas a more generic description will not require such a change. The Radioactive Emuent Release Report is no longer required on a semi-annual basis, so deletion of that word from the title is justified.
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1 ATTACHMENT (2)
SUMMARY
OF RELOCATED TECHNICAL SPECIFICATIONS Existing TS Title Relocation Notes 6.2.2.a Minimum Shift Composition 10 CFR 50.54(k),(1)
Table 6.2.1 Minimum Shift Composition and (m) 6.2.2.e SBO Present During Fuel Movements 6.4.1 Training ANSIN18.1-1971 and 10 CFR55 6.5 Review and Audit QA Policy 6.6 Reportable Event QA Policy 6.7 Safety Limit Violation Specifications 2.1.1 and 2.1.2 l
6.8.1.b, c, f Refuel Operation, Surveillance and Test Activitics of Safety-Commitment to Related mpment and Fire Protection Program Implementation Regulatory Guide 1.33 6.8.ld Security Plan Implementation Security Plan 6.8. le Emergency Plan Implementation Emergency Plan 1
6.8.lh Process Control Program Plant Procedure and 6.16 10 CFR Parts 20,61 and 71 6.8.2 Procedure Review Authority QA Policy 6.8.3 Approval Authority 6.8.4 Review Process 6.8.5 Temporary Change Provision 6.9.1 Startup Report UFSAR 6.9.1.5b Steam Generator TubeInspection Specification 4.4.5.5.b 6.9.1.7 Annual Radiological Environmental Operating Report ODCM 6.9.1.8 Semiannual Radioactive Efiluent Release Report ODCM 6.9.2 Special Reports Technical Specification Section 3 6.10 Record Retention Plant Procedures 6.11 Radiation Protection Program 10 CFR Part 20 6.12 High Radiation Area 10 CFR 20.1601 t
6.14 Iodine Monitoring UFSAR i
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ATTACHMENT (3)
I DESCRIPTION OF JUSTIFICATION CATEGORIES The changes bem8 Proposed will provide a significant improvement to human factors for the Calvert Cliffs Nuclear Power Plant Technical Specifications by locating similar requircenents within the same section, unaanaamary requiremcsits, relocatmg requirements to other Baltimore Gas and Electric Company documents, and addmg new requ,irements consistent with the standard format cantamad in NUREG-1432, " Standard Technical Specifications, Combustion C=1--g Plants." 'Ihe changes have been grouped i.cw.Lg to these four categories and are discussed 5elow. Strictly editonal are i
discussed under Category 5, and the few changes falling outside any of these categones are discu under Category 6.
j i-Category 1 - Locatina Similar Requirements within the Same Section Many of the chartges were made to support consolidation of similar requirements. 'Ibese changes will e
facihtate using the Specifications by grouping requirements in a logical manner. Since the reguirements i
remam the same, these changes only affect the method of presentation and are considered admmistrative.
'Ihe following changes were relocated Changes 3,6,13,15,18,19,22,23,24,29,36,38,52,54,55, 69,79,88,90,95 and 97.
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Catenory 2 - Deleting Unnecessary Requirements Several administrative requirements either duplicate regulations or other Specifications. It is desirable to elimmate this type of duplication because the information is adequately controlled by regulations or other sections of the Specification. Since the requirements remain the same, these changes only affect the method of presentation and are considered administrative. The following requirements were removed from Section 6 or revised to elimmate duplication of regulatory or other Specification requirements.
Changes 11,12, 14,20,21,25,28,32,35,42,43,45,82,84,86 and 87.
Relocating Requirements to other Baltimore Gas and Electric Company Controlled Category 3 Documents These changes are being made to reduce the need to request Technical Specification changes for issues which do not directly attect public safety. 'Ihe requirements will be mamtained in documents and programs which have sufficient controls in place to effectively manage the implementation of the requirements and future changes thereto. Controls are established by 10 CFR 50.54(a)(3) and 10 CFR 50.59 to ensure adequate review of future changes. With the exception of our request to reduce commitments contamed in the Quality Assurance (QA) Policy, all requirements remam the same. Since the requirements remam the same and thejustification for the reduction in commitments are contamed in the QA Policy submittal, these changes only affect the method of presentation and are considered admmistrative. The following changes were relocated from Section 6 to the QA Policy, UFSAR, Security Plan, Eswrp=.wy Plan, ODCM and/or i
plant procedures. Changes 9,10, 30, 31, 33, 39, 44, 49, 61, 74, 75, 76, 78, 80, 85, 89, 91, and 92.
Catemory 4 - Adding new Standard Technical Specification Requirements These changes, with the exception of Change 17, merely improve the presentation format and do not' impose any new requirements. Change 17 does permit a dual role Shift Technical Advisor which is no l presently allowed, but is consistent with NRC guidance. 'Ihe change ramains admmistrative in nature not affectmg plant operatmg parameters. 'Ihe following Standard Technical Specification admmistrative changes were added to Section 6 to provide additional efficiencies and controls m accordance with NRC guidance, as discussed in Attachment 1.
Changes 17,46,47,50,51,53,57,58,68 and 73.
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ATTACHMENT (3)
DESCRIPTION OF JUSTIFICATION CATEGORIES Category 5 - Editorial Changes These changes were made to support changes to Specification numbers discussed presiously or to provide clarificatsort 1he following changes are editorial only and do not affect plant operation.
Changes 1,2, 4, 5, 8,16, 26, 27, 34, 37, 40, 41, 48, 56, 59, 60, 63, 64, 65, 66, 70, 71, 72, 77, 81, 83, 93, 94 and %.
Category 6 - Other Changes lhese changes either reduce reporting requirements or add a more effective option for estunating dose. All changes are administrative in that plant operations remam unaffected. The following changes do not fall precisely within any of the previous categones.
Changes 7,62 and 67.
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