ML20013H952

Informs That Commission Authorized Release of Encl Draft Secy Paper, Combined License Review Process, for Public Comments.Frn Requesting Public Comments on Paper Also Encl

ML20013H952
Person / Time
Issue date:	05/01/1998
From:	Joshua Wilson NRC (Affiliation Not Assigned)
To:	NRC
Shared Package
ML20013H953	List: ML20013H952 ML20247B669
References
NUDOCS 9805070068
Download: ML20013H952 (15)

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Note to:

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From:

Jerry als,NRR 301

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SUBJECT:

COMBINED LICENSE REVIEW PROCESS

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The Commission has authorized the release of the attached Draft SECY paper,

• Combined j

License Review Process,". for public comments. I have attached the Federal Register Notice j

that requests public comments on this paper. The FR notice states that the subject paper is available in the NRC's Public Document Room.' The attached version of the paper, dated May 1,1953, is the version to be released to the public. I have also attached a disk which contains the electronic version of this document for your use.

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EOR The Commissioners

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FROM L. Joseph Callan Executive Director for Operations SUBJECT COMBINED LICENSE REVIEW PROCESS PURPOSE To request the Commission's approval of the NRC staffs positions on a number of policy issues relating to the combined license (COL) review process under 10 CFR Part 52, i

including some emergency planning issues, in accordance with COMSECY-95-028, dated l

September 14,1995, and COMSECY-98-004, dated April 10,1998.

SUMMAILY: This paper responds to direction-setting issue (DSI) #10 and presents the NRC staffs positions on the contents of a COL application; COL inspections, tests, analyses, and acceptance criteria (ITAAC); ITAAC for emergency plans; verificatiun of ITAAC; role of the quality assurance program; and emergency plans for early site permits. Although an application i

for a COL might not be submitted for many years, many of the policy decisions discussed in this paper will affect the COL review process.

BACKGROUND Subpart C of Part 52 sets up a process forissuing COLs for nuclear power facilities. A COL is a license authorizing construction und conditional operation of a nuclear i

power facility. The COL includes ITAAC whose purpose is to provide reasonable assurance that the facility has been constructed and will be operated in conformity with the license, the provi-sions of the Ato.sJc Energy Act ind the Commission's regulations applicable to the facility.

The staff has previously prepared a number of Commission papers that discussed various e

issues associated with the implementation of 10 CFR Part 52. In SECY-90-377, " Requirements j

for Design Certification Under 10 CFR Part 52," the staff made recommendations to the Commission on the level of detail required for a nuclear power facility design in a design certification application. In SECY-91041,"Early Site Permit Review Readiness," the staff made recommendations to the Commission on the actions needed to prepare for the review and acceptance of an early site permit (ESP) application.

The following Commission papers discussed the concept and development of ITAAC and design acceptance criteria (DAC): SECY-91-178, " Inspections, Tests, Analyses, and Acceptance Criteria for Design Certification and Combined Licenses"; SECY-91-210. " Inspections, Tests, Analyses, and Acceptance Criteria Requirements for Design Reviews and issuance of a Final Design Approval"; SECY-92-053, "Use of Design Acceptance Criteria During 10 CFR Part 52 CONTACT:

Jerry N. Wilson, NRR 301-415-3145 May 1,1998 l

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Design Certification Reviews"; SECY-92-196, " Development of Design Acceptance Criteria for i

the Advanced Boiling Water Reactor"; SECY-92-214, " Development of Inspections. Tests,

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' Analyses, and Acceptance Criteria for Design Certificahon"; SECY-92-299, " Development of h'

Desgn Acceptance Criteria for the Advanced Boiling Water Reactorin the Areas of I

[J Instrumentation and Controls and Control Room Design"; SECY-92-327, "Rev'ews of j

inspections, Tests, Ana'yses, and Acceptance Criteria for the General E.lectric Advanced Boiling i

1

' Water Reactor (ABWR)"; and SECY-95-090, " Emergency Planning Under 10 CFR Part 52."

In'SECY-92-287 and SECY-92-287A, " Form and Content for a Design Certification Rule," the staff descnbod aspects of the relationship between a design certification rule (DCR) and a COL j

' that references the DCR." in SECY-92-368, " Final Rule Amending 10 CFR Part 52," the staff 1

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proposed changes to 10 CFR Part 52 to conform to the Energy Policy Act (57 FR 60975). In

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SECY-94-294, " Construction inspection and ITAAC Verification," and the " Draft Report on the

Revised Construction inspection Program," dated October 1996, the staff described its plans for i(

incorporating a " sign-as-you-go" process and 10 CFR Part 52 rwquirements into the NRC's i

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constructior, inspection program. In SECY-96-077 " Certification of Two Evolutionary Designs,"

the staff proposed final desen certification rules for the ABWR and System 80+ designs.

" Although related to the eventual issuance of a COL, the major focus of these papers was~on the j.

design review and issuance of a DCR. This paper discusses'some of these same topics in the F

context of a COL.

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DISCUSSION. Subpart C of 10 CFR Part 52 delineates the requirements and procedures applicable to the issuance of a COL for nuclear power facilities. An application for a COL may,,

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but is not required to, reference a DCR or an ESP, or both. As discussed in Section 52.79, the i

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contents of a COL application depend on whether or ne!Jeference is made to a DCR or an ESP.

j This paper analyzes the case in which an applicant referenna a DCR. If an ESP is not l

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. referenced in a COL application, all siting issues (including environmental protection, site safety, and emergency plannirg) must be addressed in the COL application.-

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F Contents of an Mcation i

A COL applicant will be responsible for submitting all of the information that would be required

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n for a 10 CFR Part 50 operating license plus the additional information required for issuance of a i

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COL under Subpart C of Part 52 as discussed below. Sections 52.77 and 52.79 require COL i;

t applicants to submit relevant information required of applicants for construction permits (cps) i i

Jand operating licenses (OLs). in' addition, Section 52.78 requires an application to demonstrate

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compliance with the requirements for training programs established in'10 CFR 50.120. Since n

the COL combines both a CP and OL, the staff will need, as part of the COL application, all of i

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(the information required to make the necessary findings under 10 CFR 50.40, 50.42, 50.43, L

50.47,50.50, and Part 51, including those conceming the financial / technical qualification of the A

applicant and the Price-Anderson Act. A COL applicant should also submit a plant-specific

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probabilistic risk auessment. The NRC staff has discussed its views on this subject in i 6

' SECY-94-182, "Probabilistic Risk Assessment (PRA) Beyond Design Certification," dated 3

i j July 11,1994'and will develop an " Operational PRA Rule."'

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in addition to the required siting information, a COL applicant that references a DCR must, under

. Section 52.79, (1) demonst* ate compliance with site parameters and inte: face requirements, (2) provide site-specific design information, including ITAAC, (3) provide any outstanding l

information regarding emergency plans, and (4) have available -for audit-procurement specifi-cations, as well as construction and installation specifications, in accordance with Sections I

52.47(a)(2) and 52.63(c). Also, the applicant must submit a plant-spucific design control document (DCD), as modified and supplemented by the applicant's exemptions and departures, including proprietary and safeguards informat.on; plant specific technical specifications; and j

information that addresses the COL action items (refer to Section IV of design certification rules; j

e.g. Appendix A to 10 CFR Part 52). The generic DCD consists of the design-related information that war approved during the staff's review of the design certification application.

The NRC staff prepared Standard Review Flans (SRPs) for use in future COL applications.

Guidance for reviewing ITAAC and explanations of the information in a DCD is provided in SRP Section 14.3," Inspections, Tests, Analyses, and Acceptance Criteria in Certification,"

l (NUREG-0800). Guidance for reviewing antitrust and financial qualifice a is provided in NUREG-1574 and NUREG-1577, respectively. Guioacce for reviewing environmental reports for ESPs and COLs is provided in the hnvironmental SRP (NUREG-1555).

The licensee bears the responsibility for developing and performing ITAAC. The NRC will verify

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through its inspection program that the licensee has performed ITAAC in an acceptable manner, thereby ensuring there is reasonable assurance that the facility has been built and will operate in accordance with the license and applicable regulations. As discussed further in this paper and i

in SECY-94-294, close coordination will be required between the licensee and the NRC staff during the construction process to ensure that essential inspections, tests and analyses are verified in a timely manner. To facilitate this coordination, the staff will request a detailed construction plan, including construction sequence and schedule, along with, or shortly after, the COL application. The staff believes that applicants will be willing to provide this information, especially if, as industry representatives have suggested, they want to pursue an aggressive construction schedule. Although this information is not required to be submitted, the consequences of not providing it could include diminished coordination between the licensee and NRC staff, which could result in difficulty in scheduling inspections.

COL Form and Content i

Although Subpart C of 10 CFR Part 52 does not specifically discuss the form or content of a combined license (COL), Section 52.97(b)(1) requires that ITAAC be identified within the combined license. The staff prepared a generic COL (refer to the attachment) based upon recently issued operating licenses and the requirements of Part 52. The generic COL contains requirements such as the environmental protection plan, antitrust ccnditions, pre-and post-fuel load license conditions, and plant-specific ITAAC and technical specifications.

The specified duration for the generic COL is forty years from the date of issuance. This is a change from the earlier version of this paper and is necessary to comply with section 103.c of I

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The Commissioners -

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the Atomic Energy Act of 1954, as amended (1954 Act), which provides that "[e]ach

[ commercial] license shall be issued for a specified penod, as determined by the Commission, -

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_ depending on the type of activity to be licensed, but not exceeding forty years [.]" Section 185.b :

of the 1954 Act, enacted in the Energy Policy Act of 1992 (1992 Act), does not refer to the

't duration of a COL or the duration of the period of operation under the COL, but doec define the

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- COL as e %bined construction and operating license." Prior to operation under the COL, the

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Commissbr, -nakes findings pursuant to $$ 52.99 and 52.103(g), but these findings do not result '

' in the issuance of a new license under the 1992 Act and 10 CFR Part 52. Because a COL is i-clearly a license for the purposes of section 103.c, the duration of a COL is limited to forty years

from the date of COL issuance. Accordingly, $ 52.83, which requires "... that the initial duration 4

of the [ COL] may not exceed 40 years from the.date on which the Commission makes the

' findings required under 6 52.99[,]" appears to be inconsistent with the above interpretation of the 1992 Act.( However, a combined license issued under Part 52, with a duration beginning on ne -

l' date of issuance, would provide a term of operation less than the 40 year term of an operati,w

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' license issued unoer Part 50.- Therefore, the NRC has requested a change to the Atomic

Energy Act to clarify that the duration of a COL may not exceed 40 years from the date on which h

the Commission finds, prior to operahon of the facility, that the acceptance entena required by sechon 185.b have been met.' This would clarify that the term of operation under a COL is 40

- years.' On April 29,1998, the House Commerce Committee ordered reported H.R. 3532, which -

contains such a clarification. The genenc COL, however, was wntten to wwJoriri with the current statutory provision.

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COL ITAAC

.l f-l Section 52.97(b)(1), requires that ITAAC for the COL be necessary and sufficient to

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. demonstrate, before facility operation, that the facility has been built and will operate in.

f' confermity with the license, the 1954 Act, and the Commission's rules and regulations. In L

- addition, pursuant to Section 52.103(g), the Commission must find that all acceptance criteria.

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specified in the license are met before facility operation. Because ITAAC are the sole source of j

acceptance criteria, it ir usential that the COL ITAAC include all significant issues that require L

satisfactory resolution trofore fuel loading. The COL ITAAC consist of the !TAAC from the j

. referenced DCR plus the plant-specific ITAAC resulting from the COL proceeding. The plant-d i

specific ITAAC consists of the ITAAC associated with the site-specific design information and

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the regulations related to a COL applicant.

The staff has reviewed the regulations referenced in Part 52, in addition to the reguletions

currently applicable to Part 50 licensees. The purpose of this review was to determine which regulations would be applicable to a COL holder and when they would be applicable. On the

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basis of this review, the staff has developed the following representative listing of areas to be covered by the plant-specific ITAAC:

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.e emergency plans e

environmental protection plan e'

fitness-for-duty program -

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e personnoi 'elifications e

physical security plan e

plant procedures / administrative' controls

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. technical specifications training programs i

r Certain verification activities (e.g., low-power and power ascension testing) necessary to..

demonstrate proper facility construction and readiness for long-term ope: Mion can only be

. performed after fuel loading begins. Therefore, the staff will incorporate these post-fual-load

~ - verification activities into the COL as license conditions.

ITAAC for Emergency Plans An applicahon for a COL must contain emergency plans that provide reasonable assurance that

- I adequate protective measures can and will be tsken in the event of a radiologier!.,mergency at

~ the' site and these emergency plans must be venfied with ITAAC [Sochons 52.79(c) and (d)].

The NRC sWf outlined the process for reviewing ano approving emergency plans, including the j

use of ITAAC, in SECY-95-090, " Emergency Planning Under 10 CFR Part 52," dated April 11,

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1995.; it was recognized that the NRC will need to make findings on the COL ITAAC before all of

.1 the emergency response requirements are completed. The ITAAC facilitate this process bt enabling the NRC, in consultation with the Federal Emergency Management Agency (FCMA), *a

' l make a predictive regulatory finding of reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. The staff provided generic ITAAC for emergency plans in SECY-95-C30 that addressed 2 of the 16 planning standards in.

10 CFR 50A7(b). These genenc ITAAC are sufficere to demonstrate that ITAAC for emergency plans can be prepared and used to verify the acceptability of COL emergency plans. TLa further refinement of ITAAC for emergency plans and coordination with FEMA will be performed upon receipt of an indication that a COL or ESP application is expected.

Venfication of ITAAC

. The licensee documentation requirements for a facility that is licensed under Part 52 are similar to the documentabon requirements under Pa ' 50. The difference is that under Part 52 the

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documentation should be formatted to demomarate the bases for successful completion of ITAAC.' The licensee must certify to the NRC that an ITAAC has been successfully performed and that the acceptance entena have been met. The ITAAC certification letter should identify the specific ITAAC that has been completed; it should identify, in summary form, the bases for the conclusion that the ITAAC has been met; and it should identify the location of any supporting

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documentation that is available for audit. The supporting documentation may include such items.

n J as test reports, engineering analyses, calculations, drawings, vendur c:,inponent tests, inspec-Ltions,' quality assurance records, and other facility records The Nuclear Energy Institute (NEl) m-'en v,

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uI provided a ' preliminary conceptual example of this type of letter in a meeting with the NRC staff L

on March 15,1995,'as documented in a meeting summary dated April 7,1995. However, each

licensee must establish the specific bases fer satisfaction of any particular ITAAC.

l The design descriptions ~and functional system drawings available for review during the design -

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certification and COL application stages were sufficent to perform licensing reviews and make i

final safety determinations but are not adequate for actual construction or construction inspechon activities. Theinfore, before construction begins on any given portion of the facility, the licensee must ensure that the certified design plus site-specific design information in the U

. COL' application, including that required by the design acceptance criteria (DAC), has been i'

translated into detailed, plant-specific, design and construction drawings. -The level of detail in i

_ the certified design and the use of DAC allow for some variation in implementing the certifed design. The applicant or lice".see also has some flexibility in completing the final design, by 4

l-means of ths lange process (refer to the change process in Section Vill of the design j-certificahon rule, e.g., Appendix A to Part 52). The NRC staff will verify completion of ITAAC by L

the licensee, and conformance with the approved design, in part by using these detailed i

da awings. Therefore, the licensee should ensure that the drawings and other documentation p

reflect the final as-built configura* ion of the facility so that they can be used as part of the bases,.

F where appropriate, for completion of the COL ITAAC.

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The NRC staff anticipates that the I consee will, on a voluntary basis, submit penodic i

construction status and completion reports to the NRC, in order to facilitate issuance or Federal Regrafer notices regarding the successful completion of ITAAC (5 52.99).- The NRC staff has

' included a condition in the COL that requires the licensee to stai4, under oath or affirmation, that.

ll the COL ITAAC have been met (refer to item 2.H in the attached generic COL).

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In SECY-94-294, NRR outlined its program to develop a new construction inspection program to accommodate the requirements of future reactors licensed under Part 52 and to incorporate lessons leamed from experience with the current construction inspection program (CIP). The j

. staff completed a draft report on, "The Revised Construction inspection Program," dated j-

. October 1996, and placed it in the NRC's Public Document Room, as reported to the F

Commission on November 5,1996. When implemented, one of the objectives of ;he CIP will be

to inspect the licensee's process for performing ITAAC and to inspect the licensee's program for

- ensuring ITAAC requirements are met. This could include the results of the preoperational test ll

. program, quality assurance program, and various facility construction programs. The staff ex-pects that there will be significant interaction between the licensee and the NRC throughout tt'e

facility construction stage.- Increased NRC onsite staffing, the formal designation of mandatory

verification activities by the COL ITAAC, and the optional implementation of a " sign-as-you-go"

. (SAYGO) inspection program will create a more structured and a more interactive environment.

In addition to an increased NRC onsite presence, NRR will have an active role in the construction and verification activities. NRR will retain program management responsibility (including the functions of interpreting DCR and COL requirements), coordinating inspection program and licensing activities, and issuing periodic FederaIRegister notices.

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The Commissioners ;

l The NRC's inspection program is written to provide general guidance to the inspection staff on a wide range of construction, preoperational, startup, and power operation areas. The inspection staff will adapt the general inspection guidance to develop a site-specific inspection plan tnat incorporates the specifics of the COL ITAAC and license conditions. The NRC's acceptance of ITAAC will be based upon licensee completion reports and independent NRC inspection and design review activities. The inspection program will provide for independent verification of site activities that support ITAAC. A" bough the results of specific NRC inspections will have a direct impact on the staffs conclusions regarding the successful completion of ITAAC, the NRC inspection program will not be limited to verification of specific ITAAC requirements. For example, the NRC inspection program might identify deficiencies in the quality assurance program that are not related to the successful completion of ITAAC but could result in an enforcement action.

I Egle of the Quality Assurance Proaram As discussed above, the NRC staff anticipates that there will be design, construction, and testing activities, related to ITAAC, for which the staff will not be able to refy solely on NRC n pections to verify proper completion. For these activities, the staff must rely on the lics r

's quality assurance (QA) program to provide suitable controls for effective verification. The staff must I

have confidence that the licensee's QA program is adequate and that it is being properly implemented so that design, construction, or testing deficiencies are identified, documented, and corrected. The QA requirements of Appendix B to Part 50 apply to all safety-related activities being conducted by the licensee during the design, construction, and operations phase { post-I design certification design, procurement, fabrication, installation, construction, testing, inspections, and operations), including those safety-related activities performed to satisfy ITAAC. For example, preoperational test program testing performed to demonstrate that safety-related structures, systems, and components (SSCs) will perform satisfactorily in service must be conducted under a program that satisfies Criterion XI,

• Test Control," of Appendix B to Part 50 and may also satisfy testing required by the ITAAC process. The scope of the initial test program, however, is not limited to just safety-related SSCs. Specifically, Regulatory Guide (RG) 1.68, " Initial Test Programs for Water-Cooled Nuclear Power Plants," Revision 2 (August 1978), specifies the scope of plant SSCs to be tested to satisfy the requirements of Criterion 1

" Quality standards and records," of Appendix A to Part 50, and Appendix B to Part 50. Although testing is required for all SSCs within the scope of RG 1.68, it is not required that all of them be tested to the same stringent requirements. Accordingly, the administrative requirements that govem the conduct of the test program, e.g., test program objectives, phases, organizational elements, personnel qualification, review, evaluation and approval of test results, test records retention, etc., contain provisions for the application of such administrative controls in a manner commensurate with the safety significance of the SSCs within its scope. Therefore, the staff expects that, because of the special significance of ITAAC in demonstrating conformance of the as-built facility with the approved design, the licensee will implement administrative requirements l

or processes for the verification of ITAAC that are similar to those implemented for the conduct of theinitial test program.

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I The Comissioners <

As discussed in meetings with NEi representatives, there may be deficiencies identif ed by the QA program that are relevant to ITAAC and that must be addressed by the licensee before the l

ITAAC can be completed.- The NRC staff disagrees with any assertion that quality assurance and quality control (QA/QC) deficiencies have no relevance to the NRC determination of whether ITAAC have been successfully completed. Simply confirming that ITAAC had been performed in some manner and a result obtained apparently showing that the acceptance criteria had been met would not be sufficient to support a determination that ITAAC had been successfully completed. The manner in which ITAAC are performed can be relevant and material to the results of the ITAAC. For example, in conducting ITAAC to verif;i a safety-related pump's flow rate, it is necessary, everi if not explicitly specified in the ITAAC, that the gauge or instrument used to verify the pump flow rate be calibrated in accordance with the requirements of Appendix 8 to Part 50 and that the test configuration be representative of the final as-built i

plant conditions (i.e., valve or system lineups, gauge locations, system pressures, or temperatures). Otherwise, the acceptance criteria for pump flow rate could apparently be met wh le the actual flow rate in the system could be different than that required by the approved design.

The staff has determined that a QA/QC deficiency may be considered in determining whether an ITAAC has been successfully comp;eted if: (1) the QA/QC def!Ncy is directly and material!y related to one or more aspects of the relevant ITAAC (or suppc< ting Tier 2 information) and (2) the deficiency (considered by itself, with other deficiencies, or with other information known to the NRC) leads the NRC to question whether there is a reasonable basis for concluding that the relevant aspect of the ITAAC has been successfully completed. This approach is consistent with the NRC's current methods for verifWng initial test programs. The NRC recognizes that there may be programmatic QA/QC deficiencies that are not relevant to one or more aspects of a given ITAAC under review and, therefore, should not be relevant to or considered in the NRC's 1

determination ac te whether that ITAAC nas been successfully completed. Similarly, individual QAiOC deficiencies unrelated to an aspect of the ITAAC in question would not form the basis for an NRC determination that an ITAAC has not been met. Using the ITAAC for pump flow rate example, a specific QA deficiency in the calibration of pump gauges would not preclude an NRC determination of successful ITAAC completion if the licensee could demonstrate that the original deficiency was properly corrected (e.g., analysis, scope of effect, root cause determination, and corrective actions, as appropriate), or that the deficiency could not have materially affected the test in question.

j During the ITAAC development, the design r" tification applicants determined that it was impossible (or extremely burdensome) b fide all details relevant to verifying all aspects of ITAAC (e.g., QA/QC) in Tier 1 or Tier L ' uefore, the staff acceptM the applicants' proposal that top-level design information be stateo n the ITAAC to ensure that it was verified, with an emphasis on venfication of the design and construction details in the "as-built" faci.A To argue that consideration of underlying information, which is relevant anu material to determining whether ITAAC have been successfully completed, ignores this history of ITAAC development in summary, the staff recommends that the Commission conclude that information such as QA/QC deficiencies, which are relevant and material to ITAAC, may be considered by the NRC

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The Commissioners a determining whether the ITAAC have been successfully completed, in addition, there may also be deficiencies identified that are not relevant to ITAAC. These deficiencies still need to be

' addressed by the licensee, and the NRC may take enforcement action. However, these

- deficiencies will not delay a finding on ITAAC completion or plant operation.

Furthermore, although the Tier 1 information was developed to focus on the performance of the structures, systems, aM components of the design, the information contains implicit quality standards. For example, the design descriptions for reactor and fluid systems describe which systems are " safety-related"; important piping systems are classified as " seismic Category I" and identify the ASME Code class; and important electrical and instrumentation and control systems are classified as "Clacs 1E." The use of these terms by the evolutionary plant designers was meant to ensure that the systems would be busit and maintained to the appropriate standards.

QA deficiencies for these systems would be assessed for their impact on the performance of the ITAAC, on the basis of their safety significance to the system. Because the ITAAC process includes safety-related activities which must be conducted under a QA program that meets the requirements of Appendix B to Part 50, licensees must develop programmatic controls and procedures that delineate how such activities will be implemented. Due to the significance of

ITAAC in the licensing process, the staff recommends that in developing this verification program licensees also include appropriate mechanisms for controlling ITAAC activities that are not safety-related but play a significant role in the verification of the design integrity of the as-built facility.

Emeroency Plans for I:ariv Site Permits

. During tho' preparation of SECY-95-090, the NRC staff identified a cor.com regarding the lack of

. a requirement in 10 CFR Pad 52 for updating emergency planning information and the plans approved as part of an early site permit (ESP). An ESP issued under 10 CFR Part 52 may be salid for up to 20 years. Moreover, as stated in SECY-95-090, it is not clear whether 10 CFR o2.39(a)(2) permits the staff to require a COL applicant who references an ESP to submit updated information on significant changes in the emergency plans that were approved as part of the ESP. Although 10 CFR 52.39(a)(2) offers methods by which members of the public could raise such issues, it does not present any procedure for allowing the NRC to do so. The NRC staff did not propose a resolution for this matter in SECY-95-090 because the industry had not

- yet expressed any interest in ESPs and the matter could be deferred until a " lessons leamed" rulemaking to update 10 CFR Part 52 is conducted pursuant to DSI 10. The staff plans to issue

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. a proposed rulemaking in 1998 to revise Part 52 on the basis of the lessons leamed during the rulemaking rectivities for certification of the evolutionary designs. Among the issues to be considered will be the need for a requirement for holdern of ESPs to periodically update,

' throughout the duration of an ESP, the emergency planning information and plans approved as part of an ESP.

The staff of the NRC and FEMA have jointly drafted emergency planning criteria in a proposed

~ Supplement 2 to NUREG 0654, " Criteria for Emergency Planning in an Er y Site Permit Application," for ESP applications. A Federa/Reaisternotice was issued on May 14.1996 (61 p

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2 The.Comissior.2rs.

EB 24307), announcing the availability of draft Supplement 2 to NUREG-0654 for public review and comment. Two comment letters were received: one from the Nuclear Energy Institute and one from the State of Illinois, Department of Nuclear Safety, both of which will be addressed in the final document. The NRC staff intends to proceed with the development of final Supplement 2 to NUREG-0654 following resolution of the ESP update issue. The comments received in response to the Federa/Reaister notice on draft Supplement 2 will be considered during the development of the final document. The NRC staff will request the Commission's approval before publishing final Supplement 2 to NUREG-0654.

RECOMMENDATION. That the Commission:

1. Approve requests for detailed construction information.

2. Approve the form and content of the generic COL.

3. Approve t',e ITAAC approach for emergency plans.

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4. Approve the inspection program for verification of ITAAC.

5. Approve the recommended role of QA relative to ITAAC.

6. Approve the plan for Supplement 2 to NUREG-0654.

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' L. Joseph Callan Executive Director for Operations

Attachment:

Generic Combined License

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j, p GENERIC COMBINED LICENSE i

INAME OF NUCLEAR FACILITY 1

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[NAME OF NUCLEAR FACILITY OWNER]-

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.1 Docket No. 52-IXXX]

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License No. NPF-[XX]

1.

. The Nuclear Regulatory Commission (the Commission), based on [ reference SER and

. EIS) has found that:

A.

The application for a combined license filed by (name of nuclear facility owner (s))

(the licensee [, which references the (incart DCR title) in Appendix to 10 CFR Part 52,] complies with the standards and requirements of the Atomic Energy Act of -

t 1954, as amended (the Act), and the applicable regulations set forth in 10 CFR Chapter I, and all required notification, te c%r agencies or bodies have been duly made;-

B.

The applicable requirements set forth in 10 CFR 52.77, 52.78, 52.79, 52.81, 52.83, 52.85, 52.87, 52.89, [52.91, if applicable), 52.97 and Appendix _ to 10 CFR Part 52 nave been met;

' C.

There is reasonable assurance that the facility will be constructed and will operate in conformity with the application, as amended, the provisions of the Act, and the 3

applicable regulations set forth in 10 CFR Chapter i except as exempted from compliance in Section 2.E below; D.

There is reasonable assurance: (1) that the activities authorized by this combined license can be conducted without endangering the health and safety of the public and (ii) that such activities will be conducted in compliance with the applicable regulations set forth in 10 CFR Chapter I, except as exempted from compliance in Section 2.E below; E.

The licensee is technically and financially qualified to engage in the activities authorized by this combined license in accordance with the applicable regulations set forth in 10 CFR Chapter I;

' F.

The licensee has satisfied the applicable provisions of 10 CFR Part 140, Financial

,. Protection Requirements and indemnity Agreements."

G.

The issuance of this license will not be inimical to the common defense and security or to the health and safety of the public; Attachment

?

?

l 3

F ;

V, L

H. LThe issuance of this license is in accordance with 10 CFR Part 51 and all applicable requirements have been satisfied; and l

1; The receipt, possession, and use of source, byproduct, and special nuclear material '

as authonzed by this license will be in accordance with the applicable regulations in -

10 CFR Parts 30,40, and 70.

2.

On the basis of the foregoing findings regarding this facility, Combined License No. NPF-

[XX) is hereby issued to [ licensee), to read as follows:

l A.

This license applies to the [Name of Nuclear Facility), a light-water nuclear reactor l

and associated equipment (the facility), owned by the licensee. The facility is -

1 located and is described in the licensee's final safety analysis report, as supplemented and amended, and the licensee's environmental report, as supple-mented and amended i

i

B.

Subject to the conditions and requirements incorporated herein, the Commission hereby licenses:

j l

'(1)

The licensee, pursuant to Sections 103 and 185.b of the Act and 10 CFR j

Part 52,"Early Site Permits; Standard Design Certifications; and Combined 3

' Licenses for Nuclear Power Plants," to construct,' possess, use, and operate the facility at the designated location in accordance with the procedures and limitations set forth in this license; u

(2)

(l) The licensee, pursuant to the Act and 10 CFR Part 70, to receive and I

possess at any time, special nuclear material as reactor fuel, in accordance

.l with the limitations for storage and amounts required for r3 actor operation,

-l and described in the final safety analysis report, as supplemented and -

amended; (ii) The licensee, pursuant to the Act and 10 CFR Part 70, tc use special nuclear material as reactor fuel, after the finding in 2.1 of this license has

. been made, in accordance with the limitations for storage and ainounts required for reactor operation, and described in the final safety analysis report, as supplemented and amended; (3)

. The licensee, pursuant to the Act and 10 CFR Parts 30,40, and 70, to receive, possess, and use, at any time, any byproduct, source, and special nuclear material as sealed neutron sources for reactor rtartup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)-. The licensee, pursuant to the Act and 10 CFR Parts 30,40, and 70, to receive, possess, and use in amounts as required, any byproduct, source, or

~.- - -

i special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive l

apparatus or components; and

.(5) l The licensee, pursuant to the Act and 10 CFR Parts 30,40, and 70, to:

F

~

possess, but not separate, such byproduct and special nucisar materials as may be produced by the' operation of the facility.

n C.

The license is subject to, and the licensee shall comply with, all apHicable -

L provisions of the Act, and the rules, regulations, and orders of the Commission [,

including Appendix _ to 10 CFR Part 52).

D.

The license is subject to, and the licensee shall comply with the conditions set forth in 10 CFR Chapter I, now or hereafter applicable [ consistent with the requirements i~

in Section Vill of Appendix _ to 10 CFR Part 52); and the conditions specified and incorporated below:

(1)

Maximum Power Level The licensee is authorized to operate the facility at reactor steady state core power levels not in excess of [XXXX] megawatts thermal in accordance with t

the conditions specified herein.

2 (2)

Technical Specificahons and Environmental Protection Plan 1

[

The plant-specific Technical Specifications contained in Appendix A of this license and the Environmental Protection Plan contained in Appendix B of -

this license, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environruntal Protection Plan.

.(3)

Antitrust Conditionc The licensee shall comply with the ant.4 rust conditions set forth in Appendix

- C to this license; Appendix C is hereby incorporated into.this license.

E. ' The following exemptions are authorized by law and will not endanger life or property or tho common defense and security. Certain special circumstances are present and these exemptions are otherwise in the public interest. Therefo'e, these exemptions ~are hereby granted.

. [1. LISTING OF EXEMPTIONS FROM DESIGN CERTIFICATICO RULE]

- [2. LISTING OF EXEMPTIONS WHICH ARE OUTSIDE THE SCOPE OF THE DESIGN CERTIFICATION RULE]

1 i.

l

[ F. - The licensee shall report ariy violations oi

equirements set forth in Section

)

2.D(1) of this license within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Initia notification shall be made in 1

accordance with the provisions of 10 CFR 50.72, with written follow-up within 30 days, in accordance with the procedures described in 10 CFR 50.73(b), (c), and (e).

G.

The licensee shall fully implement and maintain in effect all provisions of the physical security, guard training and qualification, safeguards contingency plans,-

and all amendments made pursuant to the authority of 10 CFR 50.90 and 10 CFR j

50.54(p).

1 H.

Before loading fuel, 'he licensee shall state under o.u s or affirmation to the Commis-sion that the acceptance criteria contained in the COL ITAAC[, including the ITAAC in Appendix _,to 10 CFR Part 52,) have been met.

i 1.

Fuel will not be loaded into the reactor vessel until the NRC has found that all of the ITAAC have been met, in accordance with 10 CFR 52.99 and 52.103(g). After this finding has been made, the ITAAC do not constitute regulatory requirements; except for specific ITAAC, which are the subject of a Section 103(a) hearing, their expirstion will occur upon final Commission action in such proceeding.

)

J.

The licensee shall perfo.m and successfully complete the following activities after fuJ load.

I

- power ascension tests

)

- low-power tests K.

The licensee shall have and maintain financia protection of such type and in'such amounts as the Commission shall require in L;cordance with Section 170 of the Atomic Energy Act of 1954, as amersded to cover pQlic liability claims.

1 L

This license is effective as of the dste of issuance and shall expire at midnight on

[the date 40 years from the date of issuance).

t FOR THE NUCLEAR REGULATORY COMMISSION 1

Samuel J. Collins, Director Office of Nuc! ear Reactor Regulation Attachments / Appendices.

1. Appendix A-Technical Specifications L

' Appendix B - Environmental Protection Plan

!~

3. Appendix C - Antitrust Conditions Date of issuance:

i I

_ _. _ _. _.. ~. _

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j

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7590-01 I,

i UNITED STATES NUCLEAR REGUI.ATORY COMMISSION j

i NOTICE OF PUBLICATION OF DRAFT COMMISSION PAPER j

i

" COMBINED LICENSE REVIEW PROCESS" i

i The U.S. Nuclear Regulatory Commission (NRC) has issued a draft version of a j

i Commission paper entitled " Combined License Review Process" and is requesting'public

)

comments on this paper. Subpart C of 10 CFR Part 52 presents a process for issuing combined licenses (COLs) for nudear power facilities. A COL is a single license authorizing cons'_ruction l

i and conditional operation of a nuclear power facility. This draft paper informs the Commission

.l i

about the NRC staffs positions on a number of issues relating to the COL review process,.

j including: contents of a COL application; COL inspections, tests, analyses, and acceptance criteria (ITAAC); ITAAC for emergency plans; verification of ITAAC; role of the quality assurance.

-i j

program in ITAAC; and emergency plans for 7arly site permits.

j i

An earlier version of the draft paper was issued in April 1993. The NRC received comments from the nuclear industry (NUPN RC) on this paper. As a r9sult,'several changes I

were made to the draft paper. The most significant of these changes include; removing a proposed license condition regarding detailed design drawings, removing any mention of hold i

points in the construction inspection process, revising the format of the sample license, and

.l I

shortening the duration of a combined license to conform with the Atomic Energy Act of 1954, as

'l t

amended /- An amendment to the Atomic Energy Act has been proposed to correct the COL 1

duration issue.

j A copy of the draft paper has been placed in NRC's Public Document Room, the Gelman Building,2120 L Street, N.W., Washington, D.C. 20037, for review by interested persons.

'990!490177 990501 U

-CF SUBJ V

08sM-6COM MTG CF Q

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e.

s Questions and comments should be directed to Jerry N. Wilson, Mail Stop O-10 D22, U.S.

Nuclear Regulatory Commission, Washington, D.C. 20555-0001. Email:jnw@nrc. gov or telephone: 301-415-3145. Comments should be submitted within 120 days of the publication of this notice.

Dated at Rockville, Maryland, this '

• st day of May 1998.

FOR THE NUCLEAR REGULATORY COMMISSION original signed by:

{

Theodore R. Quay, Director Standardization Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation DISTRIBUTION:

Central File pra PDST R/F TQuay JNWilson j

JMoore,0-15 B18 i

DOCUMENT NAME A:\\FEDREGNO.JNW To receive a copy of this document, indicate in the box: "C" = Copy without attachment / enclosure "E" = Copy with e ttachment/ enclosure "N" = No copy I

OFFICE SPA:PDST:DRPM D:PDST:DRPM l OGC W l

l NAME Jf) Wilson:O ma)

TRQuay TJXA JMeere gvMuy DATE T /$/98 MfF 6/i 98 Y /20/98 1

V OFFICIAL RECORD COPY 4hogctk<44awh

e e

t'.

+

6 l l Questions and comments should be directed to Jerry N. Wilson, Mail Stop O-10 D22, U.S.

f Nuclear Regulatory Commission, Washington, D.C. 20555-0001, Email:jnw@nrc. gov or telephone: 301-415 3145. Comments should be submitted within 120 days of the publication of this notice.

Dated at Rockville, Maryland, this 1st day of May 1998.

i FOR THE NUCLEAR REGUl.ATORY COMMISSION r

N Theodore R. Quay, Director i

Standardization Project Directorate Division of Reactor Program Management I

Office of Nuclear Reactor Regulation I

'l l

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1

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i