ML19280B306
| ML19280B306 | |
| Person / Time | |
|---|---|
| Issue date: | 11/04/1981 |
| From: | Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | |
| Shared Package | |
| ML19280B307 | List: |
| References | |
| REF-10CFR9.7, TASK-RIA, TASK-SE SECY-81-631, NUDOCS 8112040050 | |
| Download: ML19280B306 (91) | |
Text
e n
4,b M dhaa$h4.n a,.m& h M /eh.A M e&kr.hs4.Leh deJ5ei M E.~e haMideaO M hb4!
A a# de d'Au esA4bi.sAJE JihMk 65.subdSaadssmMLui
.Z s44 d
4 l
SECY-81-631 November 4, 1981
,. Q,,
r
~
FlUEEMAKING ISSUE
,m
( Af firmation)
' '@ ~ '
% N x
ij-D\\
e For:
The Comissioners 7
[
"u.
c M rg~9,* (7 3, o
From:
William J. Dircks, Executive Director 5
for Operations J
\\/
'A 4.'/
Subject:
FINAL RULE FOR PENDING CP/ML APPLICATIONS s
Purpose:
To obtain Cc raission approval of changes to the subject rule as adopted by the Comission on August 27, 1981.
Discussion:
Subsequent to the Comission's approval of the proposed final rule for pending CP/ML applications, the staff had additional discussions with Comissioners' Assistants and OGC that resulted in a redraf t of the Federal Register Notice of the rule which was forwarded to the secretary by my memorandum of September 25, 1981. Following further staff discussions with Chairmar.
Palladino, Comissioner Ahearne and Comissioners' Assistants on October 13, 1981, it was agreed that tne staff would prepare certain clarifying additional modifications to the rule for Comission consideration.
Enclosed is a revised version of the rule that contains the follcwing changes from the version for-warded by my memorandum of September 25.
1.
A new paragraph (1)(xii) has been added that requires applicants to submit, within two years of CP issuance, a com-parative evaluation of the options for providing hydroge'1 control measures that have been considered.
This is con-sidered necessary because of the relatively unproven effec-tiveness, at this time, of the known viable options other than pre-inerting (i.e., distributed ignition systems and post-accident inerting systems).
2.
For essentially the same reason as in 1) above, para-graph (3)(v)(D) has been revised such that all containment designs must have the capability to safely accommodate the the pressure resulting from inadvertent actuation of a post-
Contact:
Robert A. Purple, DL 492-7980 hh0050811104 CF wpmw
~3rmmemme7nmmr-m----vmumemyn7., + g.
mw,mt, mtmwww
^ '
....a.
a-am ell
'% accident inerting system. ltis requirement will ensure that the post-accident inerting method of hydrogen contol remains a viable option. The Mark III applicants who have not tenta-tively selected the post-accident inerting system have advised the staff that their current containment designs can accom-mc,date this added pressure. The ice condenser applicant (for the manufacturing license) has indicated that the design modi-fications comitted to by letter of September 17, 1981, will accomodate this requirement.
As an additional change in this section, the statement that a pressure test at this level is required has been modified to require only that the containment design be capable of safely accomodating such a test. Details about actually conducting the test -- e.g. specifying the test frequency -- are more appropriately considered during the operating license review stage.
3.
Certain requirements in paragraph (3)(v) -- viz. parts (B), (C), (D)(3), and (E) -- are uniquely related to charac-teristics of the selected hydrogen control system rather than requireircnts that directly affect containment structural design (as are the other parts of paragraph (3)(v)). Since the requirement to provide a hydrogen control system is in Section (2) of the rule, it is more appropriate for these specific requirements now in paragraph (3)(v) to be moved to Section (2) as subparts to the present itec requiring a hydrogen control system (paragraph (2)(ix)). The enclosed rule reflects this chaage.
4.
In its continuing review of the advance draf t rule for-warded to the Comission on September 25, the Staff identified two places in the Response to Coments Section (on Pages 21 and 31) where revisions were necessary to conform to the Comission's decision regarding the hydrogen control issue for tho manufacturing license application.
In addition, the supplementary information section of the Federal Register Notice has been modified where necessary, to conform to the changes described in this paper and to delete Boston Edison Company (Pilgrim Station, Unit 2) as an affected applicant.
. 5.
Places in the proposed rule that have been changed in accordance with the above have been identified by marginal notations. The pages so marked are pp. 8,10, 21, 26, 30, 31, 33, 57, 58, 59, 62, 67, 71, 73, 74, 80, and 81.
In addition to the changes described above, the staff suggests the Commission consider the desirability of further modifying section (3)(v)(B)(1) on page 81 to require that instability be considered in designing the containment to withstand inadvertent inerting.
Such a requirement would be more consistent with conventional design practice. Taking instability into account also would result in a higher ultimate containment capability for the smaller steel containments covered by the rule (e.g.,
80 psig vs 55 psig in the OPS design).
This result is both consistent with and supportive of Commission policy concerning this rulemaking as stated on page 9 of the enclosed Statement of Considerations- "... the Commission has adopted a policy
[in this rule] of allowing construction to proceed while minimizing foreclosure of plant modifications in the structural design area that may result from the rulemaking proceeding on degraded core accidents."l/
A more detailed discussion of the basis for such a change is presented in Enclosure 2.
Recommendation:
That the Commission:
1.
Approve the enclosed rule for publication in the Federal Register _.
2.
Note that, upon Commission approval of the rule, NUREG-
'0718 will be revised to conform to the changes described in this paper.
m
/
p [k William J, Dircks, Executive Director for Operations
Enclosures:
1.
proposed Rule 2.
Staff Memorandum on Containment Instability N ne of the possible results of the degraded core rulemaking proceeding is a O
requirement that containment designs be strengthened beyond that currently required.
_4 Commissioners' comments or consent should be provided directly to the Office of the Secretary by c.o.b. Friday, November 20, 1981.
Commission Scaff Office comments, if any, should be submitted to the Commissioners NLT November 13, 1981, with an information copy to the Office of the Secretary.
If the paper is of such a nature that it requires additional time for analytical review and comment, the Commissioners and the Secretariat should be apprised of when comments may be expected.
This paper is tentatively scheduled for affirmation at an open meeting during the week of November 30, 1981.
Please refer to the appropriate Weekly Commiss,ic.n Schedule, when published, for a specific date and time.
DISTRIBUTION Commissioners Commission Staff Offices Exec Die for Operations Exec Legal Director ACRS ASL99 ASLAP Secretariat t
ENCt.030RE 1 b
NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 Licensing Requirements for Pending Construction Permit and Manufacturing License Applications AGENCY:
Nuclear Regulatory Commission.
ACTION:
Final rule.
SUMMARY
The Nuclear Regulatory Commission is adding to its power reactor safety regulations a set of licensing requirements applicable only to construc-tion permit and manufacturing license applications pending at the effective date of this rule. The requirements stem from the Commission's ongoing effort to apply the lessons learned from the accident at Three Mile Island to power plant licensing. Each applicent covercu by this rule must meet these require-ments in order to obtain a permit or manufacturing license.
EFFECTIVE DATE:
(Insert date 30 days af ar publication in the FEDERAL REGISTER.)
FOR FURTHER INFORMATION CONTACT: Robert A. eurple, Deputy Director, Division of Licensing, Office of Nuclear Reector Regulation, U. S. Nuclear Degulatory Commission, Washington, D. C.
20555. Telephone:
(301) 492-7980.
SUPPLEMENTARY INFORMATION:
Background of the Rulemaking The events leading up to the promulgation of this rule were discussed in detail in the Notice of Proposed Rulemaking, which appeared in the FEDERAL REGISTER on October 2,1980, at pages 65247-65248.
In that notice, the Commission reviewed
. some of the actions it had already taken in response to the accident at Three Mile Island and outlined the options it was considering with regard to the review of construction permit and manufacturing license applications. The Commission proposed to resume licensing using pre-TM1 requirements Lagmented as necessary by new requirements identified in the Commission's TMI Action Plan. NUREG-0660.
In connection with a request for public comments on these new requirements, the Commission noted that final rules might be issued on some or all of the matters discussed in that notice.
The Commission held a series of meetings regarding this proposed rule in January, February, and March of 1981. At its March 12 meeting the Commission decided that a further brief period of public comment was desirable prior to promulgation of a final rule to ensure that all interested persons have an opportunity to review the contents of the proposed rule and, in particular, have the opportunity to comment on the applicability of t'.e proposed rule to the pending manufacturing license application. The additional comment period was discussed and noticed in tne FEDERAL P.EGISTER on March 23, 1981, at pages 18045-18049.
The Commission particularly desired comment on whether or not the pending manufac-turing license application, filed by Offshore Power Systems, Inc., should be covered by the proposed rule.
At issue is whether the rule's requirements for the capacity of containments to withstand the effects of accident-generated hydrogen are sufficient when applied to floatin] nuclear power plants.
Analysis of Public Comments The comments that were received and the Commission's responses are presented below in two parts. The first part addresses the comments received in response to the
. FEDERAL REGISTER Nctice of October 2,1980, regarding the proposed requirements set forth in draft NUREG-0718. The second part addresses comments responding to the March 23, 1981 notice containing the proposed requirements, as modified af ter consideration of comments, in the form of a proposed rule.
I.
Comments to FR Notice of October 2,1980 Comments are received from:
C. W. Rowley, Sand Springs, Oklahoma (Rowley)
Department of the Interior (USDI)
Marvin I. Lewis, Philadelphia, Pennsylvania (Lewis)
Bechtel Power Corporation, San Francisco, California (Bechtel)
Lowenstein, Newman, Reis, Axelrad & Toll (Lowenstein)
Offshore Power Systems (OPS)
Public Service Company of Oklahoma (PS0)
Boston Edison Company (BEC)
Gene-al Electric Company (GE)
Westinghouse Electric Corporation (W)
Portland General Electric Company (PGE)
Duke Power Company (Duke)
Combustion Engineering (CE)
The Commiss ion's consideration of the comments received are reflected in part by revised text in the pertinent sections of NUREG-0718 and in part by the following discussion. The comments are grouped in five areas as indicated below and are referenced by the use of the abbreviations indicated above.
. Comments on Proposed Requirements in NUREG-0718 The following is a discussion of comments received on specific NUREG 0660 items for which draf t NUREG-0718 proposed requirements applicable to the pending applications.
I.B.1.1 - Organization and Management Long Term Improvements (PS0)
II.J.3.1 - Management for Design and Construction (PS0)
The commentor notes that there is an industry-wide effort related to these activities.
Discussion The Commission is not entirely certain to what specific activity the commentor is referri ng. Liaison is maintained with the Institute for Nuclear Power Operations (INPO) which is in the process of conducting utility management audits using its own guidelines.
The classification of Action Plan Item I.B.1.1 has been changed to Category 2 (i.e., an item that is to be addressed at the operating license review stage rather than at the construction permit review stage) since it deals with operations management.
The discussion that follows addresses the comments with respect to guidance availability.
Although the NRC is developing guidelines for utility organization and manage-ment for operations (I.B.1.1), and design and construction (II.J.3.1), the NRC is still required to make a finding on management and organizational capability prior to issuance of a construction permit or operating license, even if approved
. guidelines are not available. Therefore, as has always been the case, applicants are required to describe their organizational structure and management for design and construction, regc.rdless of whether or rat an industry approach is available or is being developed. For example, in the NRC reviews of utility management and organization for recently issued operating licenses, each one has been evaluated on a case-by-case basis.
In conducting these reviews, the draf t document " Guide-lines for Utilities Management Structure and Technical Resources," NUREG-0731, which has been issued for public comment, was used.
The connentor also stated that NRC has ignored design and construction manage-ment guidance in response to Action Plan II.J.3.1.
This is not the case. Draft guidelines for this task were prepared and have been circulated for internal comment. The guidance will be included in the final version of NUREG-0731 or in a separate document.
1.C.9 - Long-Term Program Plan for Upgrading of Procedures (PS0)
A conmentor noted that it would be difficult to describe in any significant detail, until af ter January 1982, the extent to which that commentor's program will be coordinated with INP0 activities.
Discussion In consideration of the comment the Commission has modified this requirement, which called for applicants to describe how their program would be coordinated with INP0 activities. The modification requires that applicants ensure coordi-nation, to the extent possible, of their program with INPO and other inoustry efforts.
. I.D.2 - Plant Safety Parameter Display Console (Bechtel)
The commentor suggested adding a reference to the document where
^
.inent staff criteria can be found.
Discussion Reference to NUREG-0696 has been incorporated in NUREG-0718 as suggested.
I.D.4 - Control Room Design Standard (Bechtel, BEC)
The commentor noted that the IEEE standard reference in the requirement is not yet available.
Discussion The Commission has reconsidered this proposed requirement and has placed this Action item in Category 1 (i.e., an item that is not applicable to the construction permit review). However, the need was found to strengthen the I.D.1 requirement governing control room design revisions.
1.D.1 places general requirements on the ML and CP applicants.
1.E.4 - Coordination of Licensee, Industry and Regulatory Programs (PS0)
The commentor objected to describing, prior to issuance of a CP, efforts to evaluate and factor in applicable experience at similar plants on the grounds that the Nuclear Safety Analysis Center (NSAC) is developing a generic industry plan and that a separate response by the utility could undermine the generic industry program.
Discussion The Commission considers it important that those responsible for the design and
. construction of nuclear plants have a program in place prior to issuance of a CP or ML (even if that program is later superceded by an industry program) that assures an early awareness of safety problem areas and areas of safety improve-ments that arise e? iewhere. The Commission would have no objection if a utility were to improve such a plan at a later date by adopting a plan worked out generi-cally between the industry and the NRC staff. The requirements of I.E.4 are covered by I.C.5.
II.A.2 - Site Evaluation of Existing Facilities (USDI, Lewis, Bechtel, Lowenstein, Pa0, BEC, CE)
Siting was one of the four areas that the Commission identified in the October 2, 1980 notice of proposed rulemaking as deserving special attention.
Several comments (Bechtel, Lowenstein, PS0 and BEC) cited Section 108(b) of PL 96-295 (NRC FY 80 Authorization) and express or imply concern that the proposed requirements under II.A.2 are not consistent with exemption from future regulations that are to be proru1 gated under Section 108.
Discussion, The Commission believes that the proposed requirements wculd not have been incon-sistent with Section 108. However, based on preliminary staff evaluation of the sites involved, as well as the requirement added in II.B.8 for each CP applicant to perform a plant / site specific probabilistic risk analysis, the Commission has reclassified II.A.2 to Category 1.
The USDI and Lewis comments are addressed elsewhere in this document under the discussion of comments on the methods of implementing t,%e requirements.
. II.B.1 - Reactor Coolant System Yents (Bechtel)
The commentor suggested that this item be removed since II.B.8 requires appli-cants to describe the degree of design conformance with the proposed interim requirements.
Discussion Since the proposed interim rule, related to hydrogen control and degraded core considerations, as published in the FEDERAL REGISTER (45 FR 65466, October 2, 1980), did not include a requirement to demonstrate by analysis that direct venting will not result in violations of combustible gas concentration limits, II.B.1 has been revised to eliminate the requirement.
II.B.8 - Rulemaking Proceeding on Degraded Core Accider.ts (Bechtel; BEC; Lewis; Lowenstein; OP5; P50; W; CE)
Most comments received opposed requiring any concrete actions in the area of accom-modating degraded-core accidents on the part of the applicants prior to completion of the rulemaking process.
Several commentors noted that the requirement in this area, as expressed in the draft NUREG-0718, was too openended and did not clearly set forth acceptance criteria.
Discussion Degraded core ruicinaking was another of the four areas the Commission identified in the October 2,1980, Federal Register notice as deserving special attention.
As the rule was draf ted in that notice, the applicants would have been required to describe the extent to which their designs conform to the proposed interim hydrogen control rule and to provide reasonable assurance that issuance of a CP or ML would not foreclose the ability to accommodate potential requirements resulting from the
_g.
rulemaking proceedings. The Comission also liste, some features as potential requirements and proposed that the applicants submit an evaluation of the pre-ventive and mitigative features having a potential for significant risk reduc-tions that they would propose to include at their facilities.
In view of the coments and upon further consideration, the Comission has revised this requirement. The principal objective in the revision has been to take advan-tage of the fact that, for a plant that has not yet begun construction, it should be relatively easier to avoid foreclosing design tvadifications resulting from the rule-making. For some of the potential 'iesign requirements that might be required by the final rule, it ic relatively easy to ensure that they can be accomodated at any stage of construction (e.g., by providing large containment penetrations to accom-modate a filtered vented containment concept). However, to extend this approach to every conceivable rule requirement could ehsily lead to major redesigns of these plants, for which considerable design has been completed, possibly causing unneces-sary delays in their construction. On the other hand, to do nothing at this time would very likely result in foreclosure of the practical implementation of some of the future requirements.
Taking into account the fact t'4at the plants represented by the pending applications are of the most recent design and that the proposed sites are comparatively good sites, the Comission has adopted a policy of allowing construction to proceed while minimizing foreclosure of plant modifications in the structural design area that may result from the rulemaking proceeding on degraded core acciderits. Specifically, as reflected in II.B.8, prior to issuance of a CP or ML, the applicants would be
. required to commit to (1) performing a site / plant probabilistic risk assessment (This risk study would encompass many of the other concerns related to siting, systems reliability, and degraded core accidents), (2) making provisions for one or more containment penetrations for possibly venting the containment, (3) pro-viding hydrogen control measures, and (4) providing preliminary design information sufficient to demonstrate, given a 100% fuel clad metal-water reacti.n accompanied by either hydrogen burning or post-accident inerting, that (a) containment integ-rity will be maintained at an internal pressure of at least 45 psig, (b) systems necessary to insure containment integrity will perform their intended function, (c) facility design wil' provide reasonable assurance that uniformly distributed hydrogen concentrations cannot exceed 10% (controlled burning) or, in the alterna-tive, the post-accident atmosphere will not support hydrogen combustion, (d) f acil-ity design will provide reasonable assurance that hydrogen will not collect in areas where locali ed concentrations could unintentionally burn or detonate and result in loss of containment integrity or loss of appropriate mitigating features, and (e) inadvertent operation (based on C0 ) post-accident inerting hydrogen control 2
system can be safely accommodated during plant operation.
II.C.4 - Reliability Engineering (Bechtel; Lowenstein; PS0; W; Duke)
Reliability engineering was one of the four areas that the Connission identified in the October 2, 1980 notice of proposed rule making as deserving special attention.
The commentors generally expressed the view that reliability engineering is an important tool in designing for safety, but felt that, because the nethodology is not well developed, it would be irapprooriate to require extensive analysis as a prerequisite for a construction permit. Most commentors believed that a commitment
. to incorporate reliability engineering during final design, after CP issuance, would be appropriate. However, one commentor argued that no requirement in this area should be specified until the degraded core rulemaking is completed.
Discussion The requirement under II.B.8 in the revised NUREG-0718 to perform an overall plant /
site risk study will, in effect, encompass and go beyond the simplified relia-bility analyses called for in the draf t NUREG-0718. The comprehensive risk study is expected to achieve a more thorough evaluation of plant safety and will pro-vide a sounder technical basis for making decisions regarding potential plant imp rovements. Accordingly, the more limited effort called for in the draft NUREG-0718 has been replaced by the risk study requirement of II.B.8.
II.D.2 - Research on Relief and Safety Valve Test Requirements (Bechtel, BEC) that the two entries shown for this item should either be com-The comment;. nuceu b1ned or one entry deleted.
Discussion Action Item II.D.2 has been placed in Category 1 since it deals with research on generic tests. Action Item II.D.1 has been expanded to include the information presently shown in II.D.2.
II.F.3 - Instrumentation for Morito ing Accident Conditions _
(Regulatory Guide 1.97) (P50)
The commentor expressed concern that since Regulatory Guide 1.97 has not been issued. it will be difficult for the utilities to meet the NUREG-0718 require-ments in a timely manner.
. Discursion Revision 2 to Regult
, Guide 1.97 was issued on Dececoer 24, 1980.
III.A.1 - Improve Licensee Emergency Preparedness - Short Term (BEC, PS0)
III.A.2 - Improve Licensee Emergency Preparedness - Long Term (BEC, PS0)
The commentors suggested that the requirements in these two items be combined and neted that the requirements should only represent information submitted at the CP review stage.
Discussion Item III.A.1.1 in the TMI Action Plan was intended to apply only to operating reactors and certain operating license applicants, not to CP and ML applicants. For CP and ML applicants, the long term item III.A.2 called for licensees to participate in the development of guidant.e and criteria, which has now been completed. The Commission has issued new regulations to upgrade emergency preparedness planning for NRC-licensed facilities. These new regulations were issued on August 19, 1980, and becane effective on November 3,1980. Since item III.A.2 is now covered by the regulations, it has been removed from NUREG-0718.
Item III.A.1.2 has been revised to prov1dc clearer guidance by specific reference to NUREG-0696.
SPECIAL CONSIDERATION AREAS OF SITING. DEGRADED _ CORE RULEMAKING, RELIABILITY ENGINEERING. AND EMERGENCY PREPAREDNESS (See the discussion above under II.A.2, II.B.8, II.C.4, and III.A.1,.2.).
. DEVIATIONS FROM THE STTNDARD P.EVIEW PLA!i Several of the responses commented on the proposed requirements to document deviations from the Standard Review Plan. On October 9, 1980, another Notice of Proposed Rulemaking was published in the Federal Register (45 FR 67009) which also detailed requirements for documenting deviations from the SRP. This second notice not only reiterated the documentation requirements of the first notice, but also extended the requirements to operating plants rnd construction permit holders. A comprehensive final rule which will also include action for the pending CF and ML applications is under consideration in connection with 45 FR 67009. Accordingly, no special requirement nn this subject will be included in this rule.
COMMEMTS ON INSTRUCTION TO ATOMIC SAFETY AND LICENSING AND APPEAL BOARDS (Lowenstein; P50; BEC)
The notice of proposed rulemaking also requested comments on the extent to which judgements reached by the Commission on siting, emergency preparedness, reliability engineering, degraded core rulemaking, and the requirements of NUREG-0718 should form the basis for instructions to licensing and appeal boards in the CP and ML proceedings.
One commentor (Lowenstein) suggested that the licensing boards should be instructed that strict time schedules are to be imposed and enforced for completion of litiga-tion. The Conmission anticipates that licensing boards would, under present author-ity, impose and enforce appropriete schedules.
With r 3pect to siting, this commentor recommends that the licensing boards be per-mitted to entertain contentions that any part of additional requirements proposed by the NRC staff as a result of the proposed rule on siting are unnecessary or that
. such proposed requirements are not being complied with, but that requirements beyond those proposed by the staff may not be entertained and that boards' authority to raise issues sua-sponte should be subject to the same limitations. Also, this com-mentor would have the boards instructed not to entertain contentions that alternate sites be considered due to demograt.11c considerations in view of the provisions of Section 108(b) of the NRC appropriation authorization for Fiscal Year 1980, discussed under item II. A.2 above.
With respect to degraded core rulemaking, the above commentor would have the licensing boards instructed to limit the litigation in a fashion similar to that proposed by this commentor on the siting issue, namely by restricting contentions to the NUREG-0718 requirements applicable to the CP review stage, including the requirement to consider certain preventive and mitigative features.
With respect to reliability engineering, the above commentor would have the licensing boards instructed that they may only entertain contentions on the nature, method of conduct, and completion dates of the studies and the program to assure that the results are reflected in the final design.
Here also, this connentor recommends that the authority of licensing boards to raise issues sua-sponte be subject to these same limitations.
Another commentor (P50) believes that the Commission should issue a rule directing licensing boards to resume licensing proceedings in accordance with Option 1 (which the commentor believes would entail further notice and opportunity to comment before implementa tion).
(The options are described in the following section.)
If, however,
. Option 3 is adopted by the Commission, then this commentor would propose that the rule should be issuev and made effective within 30 days af ter publication in the Federal Register.
The third commentor (BEC), who also favors Option 1, would have the licensing boards instructed that they may entertain contentions that one or more NUREG-0718 requirements applicable to the CP review stage are not complied with but may not entertain contentions thr.t requirements beyond these are necessary. This commentor would also have the licensing boards' authority to raise issues sua-sponte subject to these same limitations.
Discussion The Commission has decided that Opti n 3 should be embodied as a rule, to be effec-o tive 30 days after publication of the notice in the Federal Register. The rule, like other Commission regulations, may be challenged in accordance with 10 CFR 2.758.
COMMENTS ON THE METHOD OF IMPLEMENTING THE REQUIREMENTS In the notice of proposed rulemaking, three options for resuming licensing on the pending CP/ML applications were presented. Briefly, they were as follows:
OPTION 1 Resume licensing u-ing the pre-TMI requirements augmented by the applicable requirements identified in the Commission's June 16, 1980 Statement of Policy regarding operating licenses.
. OPTION 2 Take no further licensing action until the rulemaking actions described in the Action Plan, NUREG-0660, have been completed.
ODTION 3 Resume licensing as indicated under Option 1 above, but also require certain additional measures or commitments in selected areas (e.g., those that will be the subject of rulemaking.)
A majority of those commenting favor Option 1 which, with respect to the TMI Action Plan, would, in effect, treat the pending applications as if they were the last of the present generation of nuclear power plants.
The applicants for these plants would not under this option, be required to address the four special areas cited in the notice.
Reasons cited for selecting that option include:
Option 3 could significantly delay CP licensing process (Bechtel, PGE)
Option 3 constitutes excessive and unnecessary regulation (Lowenstein) pending CP applicants should be treated like present CP holders (P50)
" additional measures" of Option 3 would be inordinately costly (BSE)
Option 3 proposes a different and esca~tated set of TMI-related requirements (GE)
Option 3 adds uncertainty to the review process by requiring com-mitments to future events (CE) sufficient "in the interim" and can be implemented in a realistic and cost ef fective manner (W) reduce dependence on foreign oil (Rowley)
One comentor (OPS) suggested that either Option 1 or Option 3 would provide a reasonable basis for resuming licensing.
One comentor (Duke) proposed its affected units (Perkins) be exempted from the rulemaking altogether because those units are intended to be identical to other units (Cherokee) already granted CP's.
One comentor (USDI) recomended that no construction permits be issued until the siting rulemaking has been completed.
While it is true that a siting rule is being formulated, it is not expected to be so drastically different from the present guidelines as co make these previously evaluated sites grossly deficient.
The Comission therefore declines as a matter of policy to delay consideration of the pending applications for conclusion of the siting rulemaking.
One comentor (Lewis) asserted that any action at this time is unnecessary and/or p rematu re. Among other things the commentor stated that there is no demand or "need for power" from new plants at this time. The Comission finds that those considerations are outside the scope of this rulemaking.
Need for power and related issues have been or will be addressed in the individual CP or ML pro-ceedings by the licensing boards. This comentor also stated that many new requirements will eventually be developed in answer to the accident at TMI-2.
Included are proposed rule changes on population density, and consideration of
" Class 9" acc idents.
In his view, concurrent consideration of several rule-makings at one time makes for duplicative efforts. However, the comments in this regard overlook the fact that ongoing licensing proceedings are always subject to matters in rulemaking and that applications are in any event judged against current licensing requirements.
. On balance, the Comission continues to believe that Option 3, as modified by revisions to II.A.2, II.B.8, and II.C.4, is the most suitable course of action to take.
II.
Comments to FR Notice of March 23, 1981 Coments were received from:
1.
J. D. Sloan, Charlotte, North Carolina (Sloan) 2.
Southern Company Services, Inc, Birmingham, Alabama (SCS) 3.
Minnesota Pollution Control Agency, Roseville, Minnesota (MPCA) 4.
Offshore Power Systems (OPS) 5.
Baltimore Gas and Electric Company (BG&E) 6.
Boston Edisu 'ompany (Boston Edison) 7.
Gilbert Asst tes, Inc., Reading, Pennsylvania (Giloert) 8.
Town of Hampton Falls, New Hampshire (Hampton Falls /
9.
Marty Casella, Sun Valley, California (Casella)
- 10. Jane J. Estes, Blacksburg, Virginia (Estes)
- 11. Stone & Webster Engineering Corporation, Boston, Massachusetts (S&W)
- 12. Atomic Industrial Forum, Washington, D.C. (AIF)
- 13. Edison Electric Institute, Washington, D.C. (EEI)
- 15. Combustien Engineering, Inc., Windsor, Connecticut (CE)
- 16. Marvin I. Lewis, Philadelphia, Pennsylvania (Lewis)
- 17. Robert Alexander, Houston, Texas,' Alexander)
- 18. Comittee on Nuclear Quality Assurance, American Society of Mechanical Engineers (NQA)
. 19. Bechtel Power Corporation, San Francisco, California (Bechtel)
- 20. Consolidated Edison Company of New York (Con Ed)
- 21. General Electric Company, San Jose, California (GE)
- 22. Carolina Power & Light Company (CP&L)
- 25. Lowenstein, Newman, Reis & Alexrad (Lowenstein) on behalf of Houston Light & Power Company and Puget Sound Power and Light Company
- 26. Conrnonwealth of Massachusetts (Massachusetts)
- 27. Tampa Electric Company (TEC)
- 28. Business and Professional People for the Public Interest, Chicago, Illinois (BPI)
- 30. Westinghouse Electric Corporation, Pittsburgh, Pennsylvania (W)
- 31. Public Service Company of Oklahoma (PS0)
- 33. Portland General Electric Company (PGE)
- 34. Commonwealth Edison Company (CEC)
- 35. Middle South Services, Inc., New Orleans, Louisiana (MSS)
- 37. Central Power and Light Company (Central P&L)
- 40. Ebasco Services, Inc., New York, N.Y. (Ebasco)
- 43. D. Marrack, Be!', aire, Texas (Marrack)
(Letters numbered 23, 29, 32, 38 and 41 are not listed because they are duplicates of the letters numbered 6, 24, 21, 32 and 11, respectively. The letters numbered 1, 8, 9,10 and 26 contain no comments on the proposed rule).
. The staff's consideration of the pertinent comments received is provided in the following discussion. The comments are grouped as indicated belew, with the source of the coments referenced by use of the abbreviations indicated above.
1.
INCLUSION OF THE ML APPLICATION The following is a discussion of the comments received on including the applica-tion for a Manufacturing License (ML) in the rule for licensing requirements for pending applications for Construction Permits and Manufacturing Licenses.
One comentor (Lewis) clearly favors outright exclusion of the ML from the rule.
The basis for exclusion presented by tne commentor is that Offshore Power Systems lacks a customer for the Floating Nuclear Plant (FNP).
A majority (16) of the (20) commenting letters that address the issue strongly favor including the ML in the rule. Three others (Boston Edison, EEI, Lowenstein) believe the ML should be included, but not if this results in a delay in promul-gation of the rule for the CP applications. Some of the reasons given for this support are the standardized plant concept (BG&E, OPS, VEPCO, CGN ED, CP&L, FPC),
conservation of resources, " diversity of fuel supplies", and " innovation" (BG&E).
Also, the considerable expenditure of dollars, expert engineering man-years, and support facility construction are noted.
OFS, particularly, states that exclusion of the ML from the rule would "... greatly damage the concept of standardization and would cast substantial doubt on whether the incentives perceived to result from standardization in fact exist." OPS further submits that the investment in the FNP was made "...in reliance on our understanding that the standards to be applied to the Manufacturing License are the same as those which apply to Construction Permits, with only such distinc-tions as are set out in 10 CFR Part 50, Appendix M" and that to segregate them now would "... insert...a comercial requirement completely at odds with the Manufacturing License concept and the Comission's prior licensing philosophy."
OPS asserts that the requirements in Subsection (3)(v) of the proposed rule are "... entirely appropriate for application to Floating Nuclear Plants", and that "[D]esign features required by the rule can and will be incorporated into the Floating Nuclear Plant design...".
OPS also notes that "[M]any of the Near-Term Construction Permit plants utilize containments with volumes and design pressures comparable to the ice condenser containment emplo'fod in the Floating Nuclear Plant", and that "... information reported at March 1,1981 ACRS meetings... indicate (sic) that the capability to increase containment strength is very nearly the same fcr the Near-Term Construction Permit plants and the Floating Nuclear Plant..."
Discussion for Inclusion of the Manufacturing License in the Rule The Comission generally agrees with the comments that favor inclusion of the ML application in the rule and has, therefore, included it.
2.
COMMENT PERIOD T00 SHORT One commentor (Gilbert) stated that,
" Based upon the numerous criteria contained in this proposal, and the potential monumental impact cf those requirements, the 20-day comment period is too short and restrictive for public rulemaking in spite of the NRC's rationalization of this time interval."
. Discussion The 20-day comment period provided in the notice printed in the Federal Register on March 23,1981 (46 FR 18045) was considered by the Commission to be suffi-cient, considering the 45-day comment period provided in a previous notice on October 2,1980 (45 FR 65247).
Promulgation of the rule will provide the affect-ed parties with a firm basis for responding to TMI-related requirements, thereby eliminating the present uncertainty and its attendant potential for unnecessary delay.
3.
APPLICATION OF THE PROPOSED RULE TO PRESENT cps AND OL APPLICATIONS One commentor (BPI) submits that "the new rule, if. enacted, should be made appli-cable to present holders of construction permits, as well as to applicants for construction permits and manufacturing licenses. To decline to so apply the amendment, especially to plants which are in the very early stages of construction, suggests that the Commission is not seriously attempting to implement the needed upgrading of safety for all nuclear plants." Another commentor (Marrack) argues that all plants not yet operating should meet the minimum improved standards.
Discussion Holders of construction permits have alre1dy been informed by letter that they must meet the TMI-related requirements contained in NUREG-0737. There is an ongoing rulemaking to codify thesi.oquirements in the Commission's regula-tions. This action will ensure that,1e bulk of the requirements that are contained in this new rule for pending CP/ML applicants will be made appli-cable to all holders of construction permits. For those areas in this new rule that go beyond the requirements of NUREG-0737 (such as those related to containment strengthening and other hydrogen control measures), the Comission, in the near future, intends to consider their applicability to present CP holders on a case-by-case basis.
4.
IMPOSITION OF NEW REQUIREMENTS One comentor (FPC) urges "the Comission to impose new licensing requirements on plants during +he licensing process only after a cost / benefit evaluation has been completed utilizing identified safety benefit compared to financial requirements to implement i.e. containment strength. We have a concern that without such evaluations licensing requirements may be imposed with minimal increase or perhaps no increase in overall safety at significant costs. This will quickly erase the nuclear alternative as viable and severely limit our energy resources." Another comentor (CE) also recommends that any major modifications should undergo complete cost / benefit assessment.
In addition, the commentor urges "that this requirement should be coordinated with other rulemaking proceedings in progress, specifically the development of an overall safety goal."
Another comentor (Lowenstein) said, "we also think it essential that the Com-mission recognize that in nany instances applicants have already completed designs, procured equipment, or comitted to fabrication of equipment on much of the proposed plants. The Comission should make clear to the NRC staff that the new requirements should be interpreted to minimize extensive redesign and procurement of new equipment to replace that already purchased."
. Discussion The Comission agrees that new requirements should be t,ased on favorable f.ost/
benefit evaluations, but this is not possible, in quantifiable terms, at present due to the lack of a specified safety goal. The Comission and its staff recog-nize that unnecessary extensive redesign and procurement of new equipment should be avoided.
However, in its extensive deliberations concerning TMI-related requirements, the Comission has decided that the requirements in the new rule are necessary for protection of the public and that their r. cst are not exorbi-tant. Acceptable alternative methods of meeting the requirements stated in the rule will be considered.
5.
IMPOSING REQUIREMENTS NOW UNDER RULEMAKING Several comentors (S&W, CEC, Lewis, Ebasco) oppose the imposition of require-ments subject to cther rulemaking proceedings, particularly relative to degraded core conditions, as premature.
Another comentor (W) said that "in light of the ongoing generic NRC proceedings with respect to safety goals and methodology, degraded core cooling, siting and emergency planning, the Comission should make it clear that the final rule when adopted is an interim rule to be applied pending the outcome of these proceedings and the risk assessments required by the rule."
" Paragraphs (e)(1)(xv), (e)(3)
(iii), (e)(3)(iv) (B thru D): Each of these items are either premature impositions of requirements not yet authorized by the NRC or are clearly the subject of cur-rent ongoing rulemaking e.g. hydrogen control and degraded core rulemaking. To impose these requirements at the CP stage precludes the full airing of these issues prior to assumption by the applicant of construction costs," stated one comentor (CEC).
. Discussion This rule does include some requirements which are subjects of other ongoing rule-making proceedings. The purpose of including these requirements in this rule is to ensure that future requirements are not rendered impractical because construction has been allowed to proceed on these plants without having made provisions for them.
6.
NUREG-0718 IS PREMATURE. LIMITED AND MISLEADING One commentor (Lewis) states that "the staff guidance in NUREG-0718...
is so limited and so misleading that it will probably be a matter of civil suit between NRC and Licensee's. Many licensee's will be able to argue that the staff guidance mislead them into believing that new requirements would be easy-to-meet and low cost." The commentor therefore, suggested that NUREG-0718 be eliminated.
Discussion The Commission is not aware of specific additionel guidance the commentor would have it provide at this time. The staff will provide applicants with additional guidance as the need arises.
Eliminating NUREG-0718 at this time would remove all guidance and could lead to more instability in the review process.
7.
OBJECTIONS TO DETAIL OF THE CP/ML RULE Two commentors (Gilbert, CEC) object to the regimentation, " great detail", and
" specificity" of placing such a rule in the Code of Federal Regulations. They support the use of Regulatory Guides, Standard Review Plans, and/or various NUREG documents. One commenter (Gilbert) goes on to state: "The current pro-posal applies to but seven pending applications, yet proposes to more than
. double the volume of 10 CFR 50.34. Furthermore, a number of the individual requirements are so design specific as to preclude the possibility of alternate designs or solutions in the future. We thus see these new proposed regulations as in conflict with both President Reagan's directive for both simplified regula-tory requirements, as well as his stated beliefs that new nuclear plants should not be unduly regulated into oblivion... We believe that the general goals and objectives of proposing the new 10 CFR 50.34(e) can be obtained through means othe than the new regulations (as has been done on plants undergoing OL review) on a case-by-case or even a generic basis, and that imposing these requirements by use of a new 10 CFR 50.34(e) is unwarranted and without justification."
Discussion The regulatory authority provided by a rule ensures a clear and concrete way to impose the necessary requirements in the wake of lessons learned from the TMI-2 accident.
Separate rules for the CP/ML applicants and the OL applicants will clarify the specific requirements the Comission considers necessary for plants at these stages in the licensing process. Excessive details have been removed l
from the proposed rule; where details are specified, the Comission has decided they are necessary to ensure the safety of the public.
8.
COMMENTS ON THE METHOD OF IMPLEMENTING THE REQUIREMENTS _
One commentor (PS0) provided coments objecting to Option 3* on the basis of
- timing, "1.e.,
this option requires the completion of a myriad of time con-
- 0ption 3 requires certain measures or comitments in selected areas (e.g.,
those that will be the subject of rulemaking) in addition to those imposed by Option 1.
Option 1 is to resume licensing using the pre-TMI requirements augmented by the applicable requirements identified in the Comission's June 16,1980 Statement of Policy (now replaced by the December 18, 1980 Statement of Policy) regarding operating licenses.
. suming enginaering activities and analyses beft.re issuance of construction permits. On the other hand, Option 1 would have required only that an appli-cant make necessary conmitments, including reasonable implementation schedules, befcre issuance of the construction permits."
Anot5er commentor (TVA) expressed the belief that the major issues in the pro-posed rule have not been reioTved sufficiently to process final rule changes at this time. TVA suggested the following approach as a more effective means of accomplishing the changes in licensing requirements:
"1.
Require that all pending construction permit and manufacturing license applicants commit to implenent the final ruins that 3
grow out of the many pending post-TMI rulemakit.gs, such as prob-abilistic risk assessment methodology, safety goal, siting, degraded core, etc.
"2.
Implement only those changes in the proposed rule which have been prouulgated and issued for use by the near term operating license plants. For other changes, retain tne existing rules pending completion of t'1e post-TMI rulemakings."
Discussion T1.c Commission has adopted Option 3, which will ensare that approved action items in the TMI Action Plan are appliti te the new cps and ML and will pro-vide for early consideration of these added safety measures so as to minimize the costs of incurporating them into the design of the facility.
. 9.
COMMENTS ON PROMPT ADOPTION OF THE RULE Many of the commentors ( AIF, EEI, Lowenstein, etc.) expressed strong support for the prompt adoption of the rule.
One commentor (Boston Edison) submitted "that the Commission would be shi king its vital responsibility in this area r
if it did not issue a rule such' as this and if this rule were not intended as binding upon the Commission's subsinf ary boards." Another stated, "C-E agrees with the Commission's intent of defining the set of TMI-related requireuents that are both necessary and sufficient to resume NRC review and approval of sending and ML applications. These requirenects (as modified to reflect public comments) should therefore be issued expeditiously in conjunction with a clear enunciation of the sufficiency of those requin2nents, to that NRC staff action on pending applications can recommence."
Discussion The Commission believes that issuance of the final rule is the proper response to these comments.
- 10. BASIS FOR COMPLIANCE WITH THE RULE A.
One commentor (Bechtel) noted that most of tre items contained in the pro-posed r ale reference action plan items i" N!! REG-0718 and NUREG-0660 and recom-mended that where the referenced paragraph in these NUREGs amplifies the requirements of the rule, it should be recognized as an acceptable means of cocpliance. Another comm.entor (Ebasco) aisc pcinted out that the proposed rule imposes new rey;;1ements in areas where final NRC acceptance criteria have not been finalized and that NRC policy relative to implementation of those criteria
. must be flexible because of the different types of requirements. To expedite the CP hearing process, Ebasco suggested that " compliance with NUREG-0718 be considered prima facie evidence that TMI requirements have been met."
Discussion The Commission agrees with the comments. The Commission has reviewed NUREG-0718 and hus concluded that the positions contained therein can provide a basis for responding to the TMI-2 accident. Applicants may, of course, propose to satisfy the rule's requirements by a method other than detai~
in NUREG-0718, but in such cases must provide a basis for determining that the requirements of the rule have been met.
NRC acceptance criteria will be suf-ficiently flexible to permit appropriate alternative methods of meeting the requirements.
B.
Two commentors (Boston Edison, Lowenstein) noted that "Some of the pro-visions of the propoced rule require the applicant to conduct studies and submit them to the NRC for review and appropriate action. Boston Edison pointed out that "these studies will Se completed af ter issuance of the con-struction permit, in some instances several years later. We believe it is necessary to make clear that the construction permit licensing boards or appeal boards do not retain jurisdiction or supervisory authority over the applicant and NRC staff for the purpose of reviewing the completed studies.
This would extend the construction permit proceeding far beyond the actual issuance of the permit and continue needless uncertainty.
Issues con-cerning the required studies are appropriate matters for the operating license stage review." Another commentor (Ebasco) noted that NRC will have
. received the studies, in some instances, prior to SER issuance for cps since some of these study requirements were applicable to operating plants and are generic in nature.
Ebasco suggested that the studies be excluded f rom the (CP) hearings.
Discussion The Commission does not expect its adjudicatory boards to retain jurisdiction or supervisory authority over fulfillment of those requirements for studies to be completed subsequent to issuance of the CP.
However, the Consission does expect the staff to review such studies in a timely manner and to take appro-priate action.
Regarding the Ebasco comment, one of the study requirements has been deleted for the reason suggested.
C.
Another commentor (Lowenstein) stated, "It is essential that the Commission make clear that this regulation, along with existing regulations, establishes an adequate and sufficient response to the Commission's post-TMI requirements.
While the notice intimates this on page 18046 (of the FR notice), we urge that it be explicitly stated in the rule."
Discussion in the Notice of Rulemaking (46 FR 18045) published on March 23,1981, under Substance of the Rule, the Commission stated, "It is the Commission's vit
' hat this new rule, together with the existing regulations, form a set of regula-tions, conformance with which meets the requirements of the Commission for issuance of a constr xtion permit or manuf acturing license." The Commission reaffirms this view with the exception of hydrogen control measures for the 1
. manuf acturing license, and, to eliminate any ambiguity regarding its intent, is amending its special review procedures in 10 CFR 2.764 to delete the state-ment in paragraph (e) that compliance with existing regulations may turn out to no longer warrant approval of a license application. However, it should be noted that the Commission also indicated in that notice that some elements in the TMI Action Plen have not been acted upon and thus may be required on the basis of future rulemaking.
11.
ADDITIONAL TMI-RELATED REQUIREMENTS One commentor (MPCA) suggested that additional items of the TMI Action Plan should be incorporated into the rule as CP/ML licensing requirements. The specific items in NUREG-0718 and NUREG-0660 suggested for inclusion in the rule are:
1.A.4.1 Initial Simulator Improvement I.C.1 Short Term Accident Analyses and Procedures Revision II.B.4 Training for Mitigating Core Damage II.B.6 Risk Reduction for Operating Reactors at Sites with High Population Densities II.B.7 Analysis of Hydrogen Control II.E.2.1 Reliance on ECCS II.E.2.3 Uncertainties in Performance Predictions II.E.3.2 Systems Reliability II.E.3.3 Coordinated Study of Shutdown Heat Removal II.J.1.1 Establish a Priority System for Conducting Vendor Inspections 111.D.1.2 Radioactive Gas Management III.D.1.3 Ventilation System and Radionuclide Adsorber Criteria III.D.1.4 Radwaste System Design Features to Aid in Accident Recovery and Decontamination III.D.2.1 Radiological Monitoring of Effluents III.D.2.3 Liquid Pathway Radiological Control III.D.2.4 Offsite Dose Measurerents
. Discussion The Commission has etnsidered incorporating each Lf these requirements into the proposed rule, but for the reasons stated talow it has determined that none of these should be added.
Items II.E.2.3, III.D.1.2-4, III.D.2.1 and III.D.2.3-4 have been judged lower priority TMI issues as reflected by task initiation dates of FY82 or later.
Because of their relative low priority, the Commission believes their incor-poration into the CP/ML rule is unnecessary. However, the results and con-clusions of thes* tasks will be appropriately considered during the OL review.
A second group L
/Jggested items is Covered in other TMI action tasks that are included as equirements in t e proposed rule.
Items II.B.6 and II.E.3.2.3 are intended to be included in 50.34(f)(1)(i), the required plant / site specific probabilistic risk assessment.
Item II.B.7 is covered by 50.34(f)(2)(ix) and (3)(v).
Items I.A.4.1 and I.C.1 are applied to operating plants and the sub-stance is included in 50.34(f)(2)(1) and (ii), respectively, for these CP/ML applications.
Another group of items is not applicable for various reasons.
I tem II.J.l.1 applies to NRC and not to CP/ML applicants.
Item II.B.4, pertaining to crew training, is more appropriate as an OL item. Finally, II.E.2.1 requires the assessment of ECCS data by operating plant licensees and is not applicable to CP/ML applicants.
In summary, the Commission has reviewed and considered all of the additional requirements suggested by MPCA and has determined that they are either covered by provisions of the proposed rule or are not applicable or appropriate for construction permit and manufacturing license applications.
12.
COMMENTS ON CERTAIN RUL. REOUIREMENTS The following discussion responds to the comments received on the specific items of 10 CFR 50.34(f) listed below:
(1)(1) - Plant / Site Specific PRA Study A.
Two comentors (S&W, CEC) point out that the NRC has not yet defined the methodology to be used in the PRA study.
Discussion The Comission notes that a PRA Procedures Guide was issued as a draft for dis-cussion by an IEEE technical symposiun in October 1981, and will be issued in proposed final form for consideration at an ANS conference in April 1982.
It is expected that the Guide will be published soon af ter the ANS conference.
Meanwhile, plans for a PRA study, and the actual conduct of the study, need not wait until the safety goal and degraded core cooling rulemakings are resolved.
During a meeting with the CP/ML applicants on April 8,1981, the NRC staff made available a PRA program outline which should serve as a guideline for CP/ML applications. The program outline addresses issues such as the scope of the PRA study, how the PRA study should be performed, what should be considered in setting up a schedsle and, most importantly, how the results of the risk study should be factored into the design, fabrication and eventual operation of the plant to improve the reliability of core and containment heat removal system.
It is reasonable to expect that an applicant can utilize the staff guidelines to develop its own program for performing a meaningful PRA study. Consequently, the Comission will retain this requirement.
. B.
Another commentor (GE) cxpressed the belief that " completion of the PRA studies and comparison to a reasonable safety goal will demonstrate that the Boiling Water Reactor includes design features which ensures that the public health and safety is protected.
If, on the other hand, the results of the studies...show that further risk reduction is appropriate, plant modifications...
should be considered".
Discussion Based on the risk studies performed to date, accident sequences relating to core and containment heat removal systems contribute substantially to overall accident risk. To reduce such risk, alternate system designs for core and con-tainment heat removal systems should be considered and PRA studies should be performed in comparison with the PRA study for the original design. The outcome of the compdrison should be selection of a system design from among several design alternatives that incorporates significant improvements in the relia-bility of core and containment heat removal systems.
C.
Two commentors (TVA, B&W) suggested that the improvements that may result from the risk assessment should be lose that are significant with respect to public health and safety, not just generally significant and practical.
Discussion The aim of the probabilistic risk assessment, as expressed in t';e requirement, is to seek such improvements in the reliability of core and containment heat removal systems as are practical and do not impact excessively on the plant.
The Cor.nission be',' aves that such improvements in reliability would also be
. significant with respect to public health and safety. Accordingly, the Com-mission does not consider it necessary to change the language o' the requirement.
(1)(11) - Auxiliary Feedwater System Evaluation Two comentors (CEC, TVA) argued that the existence of paragraph (1)(i) regarding performance of a probabilistic risk assessment (PRA) makes paragraph (1)(ii) superfluous, since a PRA study would include the analyses and reviews discussed in (1)(11) and in paragraphs (1)(iii)-(xii).
Discussion The Commission does not agree with this comment.
It is not at all certain that the PRA would necessarily include all parts of the evaluation called for in paragraph (1)(11). The result might be non-uniform and incomplete sub-mittals by the applicants, with consequent time-consuming reiterations.
It is, therefore, important that the three parts of the auxiliary feedwater system evaluation be specified. However, if an applicant's PRA does, in fact, include all parts of the evaluation called for in paragraph (1)(ii), then this require-ment will be satisfied.
(1)(iii) - Coolant Pump Seal Damage Evaluation One comentor (CEC) states that paragraph (1)(iii) is superfluous, given the requirement for a plant / site specific probabilistic risk assessment (PRA) as specified in paragraph (1)(i).
Discussion The rule requires applicants to evaluate reactor coolant pump seal damage and consequential addd loss-of-coolant, following a small-break LOCA with loss
. of offsite power. The PRA might consider this area only peripherally, if at all, since its thrust is in the improvement of the reliability of core and con-tainment heat removal systems. Accordingly, no change has been made in para-graph (1)(111). However, this requirement will be satisfied if cr. 4pplicant's PRA includes the evaluation called for in paragraph (1)(iii).
(1)(iv) - SBLOCA Probability Due to a Stuck-Open PORV One comentor (CEC) argued that the PRA analyses required by paragraph (1)(1) would also include the analysis discussed in (1)(iv) in terms of the probability of small LOCA events. The comentor said, "the criteria for judging whether or not an improvement is to be made should, however, not rest with LOCA probabili-ties but rather with overall risk contribution and ultimately with the compar-ison of plant risk to a uniform safety goal."
Discussion The WASH-1400 analysis for a PWR indicated that SBLOCAs contribute significantly to core melt probability. Furthermore, the TM1 experience and subsequent an. lysis have shown th4t the likelihood of a SBLOCA due to a stuck-open PORV is greater than that assumed in WASH-1400. The purpose of this requirenent is to determine whether this probability contributes substantially to the SBLOCA probability from all causes.
If it does, an evaluation should be performed to ensure that this probability will be reduced by incorporating an automatic PORY isolation system, which will give assurance that the public health and safety is protected in the event of a stuck-open PORV. The Comission will retain this requi rement. However, the requirement will be met if an applicant's PRA incluc'es the analysis called for in (1)(iv).
. (1)(v through xii) - Additional Studies A.
One comentor (CEC) states that all topics discussed in these paragraphs "could readily be considered in the PRA discussed in paragraph,.)(1)".
Further, the commentor states that "it appears that many of the studies and criteria have a basis only in NRC staff judgment".
Lestly, the commentor states that these studies, which are additional to the PRA discussed in para-graph (1)(1), "should be required only for those cases where the basic systems and related questions involved are shown to have a significant contribution to risk--in order to prioritize the work to be done and to conserve industry and NRC resources."
Discussion In response to the first comment regardihg paragraphs (1)(v through xii), it is noted that the specific paragraphs requiring study or evaluation by the applicant resulted from recomendations by the Bulletins and Orders Task Force. This Task Force conducted generic reviews of loss-of-feedwater and small break loss-of-coolant events on operating PWRs designed by B&W, Westinghouse and Combustion Engineering, and on operating BWRs.
These items were not explicitly included in the PRA in (1)(i) to casure that the areas are specifically addressed.
In some cases, the generalizec DRA may not be extended to cover the required area, for example: paragraph (1)(ii), study to identify practicable system modifications to reduce challenges to and failure of re'ief valves in BWRs. However, if an applicant's PP,A does, in fact, include the items called for ' n paragraphs (1)(v through xvii), then these requirementr will be satisfied.
. With regard to the second comment, it is the judgement of the Commission that potentially significant increases in plant safety could evolve from these studies and evaluations. At this time, the Commission is awaiting results of these studies and evaluations to determine whether certain plant modifications are warranted to improve plant safety.
In response to the last question regarding paragraphs (1)(v through xii), the Commission considers a risk assessment one of many tools which may be used to evaluate plant modifications and improvements. Direct evaluation, as considered in these paragraphs, is an equally valid tool.
In view of the foregoing discussion, no changes have been made in paragraphs (1)(v through xii) as a result of this comment.
However, the Commission has made changes in wording to clarify the intent of paragraphs (1)(vii), (viii) and (ix). Proposed paragraph (1)(xi) has been deleted since a generic study applicable to all the affected applicants has been submitted for Commission review.
B.
Another commentor (GE) noted that the NRC staff has agreed that the require-ments specified in II.K.3.24 of NUREG-0718 should apply only to loss of off-site alternating current power.
Discussion The Conmission concurs and has revised paragraph (1)(ix) as follows to clarify its intent:
" Perform a study to determine the need for additional space cooling to crisure reliable long-term operation of the reactor core isolation cooling
. (RCIC) and high prt, Jre coolant injection (HPCI)* systems, following a complete loss of offsite power to the plant for at least two (2) hours.
(applicable to BWRs only) (II.K.3.24)"
- For plants with high pressure core spray systems in lieu of high pressure coolant injection systems, substitute the words, "high pressure core spray" for "high pressure coolant injection" and "HPCS" for "HPCI."
(2)(111) - Control Room Design One comentor (PS0) states that the text conflicts with the predicate given in 50.34(e)(2) and suggests rewording (2)(iii) to read:
" Provide a control room design that applies state-of-the-art human factor principles (I.D.1)."
Two other comentors (SRW, CEC) suggestcd that the design be submitted for NRC " review" instead of " approval" since the latter has specific legal con-notations in the engineering area. The suggestion was also made that "the rule should stipulate that the control room design consider state-of-the-art humar factor principles, since direct application of all such principles may conflict with existing regulations."
Discussion In response to the first coment, it should be noted that section (2) does not require a control room design prior to the granting of a CP, only sufficient information to ensure that an appropriate design will be submitted prior to f abrication or revision of panels and layouts. The Comission agrees with the other coments and has amended the text to read as follows:
" Provide, for Comission review, a control room design that reflects state-of-the-art human factor principles prior to comitting to fabri-cation or revision of fabricated control room panels and layouts.
(I.D.1)"
. (2)(vi) - Reactor Coolant System Vents The commentor (CEC) notes that it may be well to review this requirement carefully on a plant specific basis to determine if any core cooling benefit can be identified; for some plants, reactor coolant system vents may offer no real benefit.
Discussion The rcactor coolant system high point vent requirement was developed to provide a means to eliminate gases that could inhibit core cooling. Since all plants have a potential to release non-condensible gases, this requirement applies to all plants. Although events in which gas venting would be required are highly unlikely, there does not appaar to be an acceptable substitute at this time for those cases where venting may be needed.
Consequently, the Commission is retaining this requirement, but has madr a minor wording change for clarifi-cation.
The paragraph now reads:
" Provide the capability of high point venting of noncondensible..."
(2)(vii) - Radiation and Shielding Design One commentor (PS0) suggested inserting the words " Provide a plan and submit a schedule to" at the beginning of the text to clarify its intent.
Discussion The Commission does not believe this change is necessary since the language under (f)(2) clearly indicates that only sufficient information is required prior to granting a CP to demonstrate that the requirements, e.g.,
(2)(vii),
will be met by the operating license stage.
However, the Commission has sub-
. stituted the word " materials" for " fluids" in the text since not only fluids are involved, and the words " TID 14844 source term" have been substituted for
" highly" f or clarification.
(2)(ix) - Hydrogen Control System A.
One comentor (OPS) requests clarification of the word " handling" in the requirement, " Provide a system for hydrogen control capable of handling hydrogen generated by the equivalent of a 100% fuel-clad metal water reaction."
Discussion The Comission has substituted the words "that can safely accomodate" for
" capable of handling" to clarify the intent.
B.
Several comentors (OPS, Bechtel, GE, W, CEC, TVA) asserted that the 100%
metal / water reaction requirement is too stringent and inconsistent with the value of 75% metal / water reaction in the proposed interim rule on hydrogen.
Discussion While it is true that the TMI-2 accident produced less hydrogen than that assumed in the rule, and that the 100% requirement is greater than the 75%
requirement in the proposed interim rule, the Comission finds that 100% is appropriate as a conservative bound for the design of plants not yet under con-struction. More specifically, the amount of hydrogen should not be tied to a given accident sequence (e.g., TMI-2), but rather a class of accidents which produce a large amount of hydrogen but hold promise of being recoverable, that is, for cooling to be re-established prior to what would otherwise be a sub-stantial core melt-down.
The proposed interim ruk will be limited to accidents f or which no or limited core melting takes place. The CP/ML rule considers potential accidents that are more severe than those considered in the interim rule. These severe accidents will be the subject of the degraded core rule-making.
C.
Another comentor (B&W) suggested that a maximum rate of hydrogen generation should be provided for the hydrogen control system.
Discussion The hydrogen generation rates and release rates into the containment are a function of the reactor type, the accident sequence being considered, and the recovery (of cooling) schemes employed.
Further, the effects of hydrogen generation rates and release rates (in terms of burning or detonation) are dependent on blowdown and steam-inerting characteristics in the containment.
Thus, one maximum rate would be inappropriate and possibly overly conservative.
Not having a maximum rate does not necessarily mean that the Comission expects detailed mechanistic analyses of hydrogen generation and release for a variety of sequences. Parametric analysis that adequately scopes the physical pro-cesses for the sequences under consideration would be acceptable.
(2)(x) - Relief and Safety Valves Two comentors (Bechtel, B&W) pointed out that this requirement appears to elevate ATWS to the status of a design basis event.
Discussion This is not intended, as the Comission is presently reviewing a proposed ATWS rule. Apprcpriate valve qualification requirements for ATWS can only be finalized after the Comission issues a final ATWS rule or decides that plants do not have to be designed to withstand an ATWS event. To clarify the intent of this requirement, it has been revised tc read as follcws:
" Provide a test program and associated model development and conduct tests to qualify reactor coolant system relief and safety valws and, for PWR's, PORV block valves for all fluid conditions expected c.' der operating conditions, transients and accidents.
Consideration of anticipated transients without scram (ATWS) conditions shall be includea in the test program. Actual testing under ATWS conditions need not be carried out until subsequent phases of the test program are developed."
(2)(xii) - Auxiliary Feedwater System A commentor (CE) suggests that the requirement to " provide an analysis of the effect on containment integrity and return to reactor power of automatic AFW system initiation with a postulated main steam line leak inside containment" be deleted since it would institute a regulatory requirement for an analysis of a condition normally assessed during the design of a safety-grade system, e.g., the auxiliary feedwater system. The commentor maintains that it is un-necessary to require this specific analysis in the rule.
Discussion The Comission agrees with the comment and has deleted this part of the rcquire-ment because the regulations already require analyses of such systems (10 CFR 50.34(a)(4)).
In addition, the term " safety-grade" has been deleted because that term is not explicitly defined in the regulations. With these changes, (2)(xii) now reads as f ollows:
. " Provide automs. tic and manual auxiliary feedwater (AFW) system initiation, and provide auxiliary feedwater system flow indication in the control room.
( Applicable to PWRs only) (II.E.1.2.)"
(2)(xvii) - Primary System Sensitivity to Transients A commertor (Gilbert), referring to this requirement, said "some statements of design criteria are so general as to be nebulous".
Another commentor (B&W) objected to " sensitivity" and " reduce" in this requirement as not well-defined terms, making it difficult to know what features must be provided. A third commentor (PGE) indicated that the reference to NUREG-0718 action plan item II.E.5.2 appears incorrect.
Discussion The requirements in 10 CFR 50.34(f) are intended to be general enoi.gh to allow a reasonable amount of flexibility in their interpretation. However, the Com-mission has deleted this requirement because it has not yet becn sufficiently defined.
Af ter further study, appropriate action on this subject will be imple-mented.
(2)(xix) - Indication of Inadequate Core Cooling A commentor (PGE) suggested the u',E of "and/or" instead of "and" in the last sentence since the present wordino implies that all of the instruments must be provided. Another commer.or (B&W) suggested deleting the examples of instru-mentation that may be required.
Discussion The commentor's reference to the "last sentence" is not clear since (2)(xix) has only one sentence. The Comission believes that the words "such as" clearly indicates that what follows are examples of instrumentation that may be required.
However, the words " exit" and " core coolant flow rate" have been eliminated to better reflect the design requirements. As revised and renumbered (2)(xviii),
the paragraph now reads as follows:
" Provide instruments that provide in the control room an unambiguous indication of inadequate core cooling, such as primary coolant saturation meters in PWR's, 'and a suitable combination of signals from indicators of coolant level in the reactor vessel and in-core thermocouples in PWR's and BWR's.
(II.F.2)"
(2)(xxi) - Power Supplies A commentor (PGE) noted "the requirement that motive and control components be designed to safety grade criteria is,nconsistent with the applicable requirement of NUREG-0737 (which is referenced in NUREG-0718)."
Discussion Paragraph (2)(xxi) has been renumbered (2)(xx) and part (B) has been revised to read:
- Motive and control power connections to the emergency power sources are through devices qualified in accordance with requirements applicable to systems important to safety."
(2)(xxii) - Auxiliary Heat Removal Systems A comentor (PGE) noted that the reference to NUREG-0718 action plan item II.K.l.2 is incorrect.
. Discussion This reference has been corrected to II.K.1.22, and the paragraph has been renumbered (2)(xxi).
(2)(xxiv) - Anticipatory Reactor Trip One comentor (B&W) indicates that a hard-wired, safety-grade reactor trip on loss of feedwater will be incorporated into the design of B&W plants; however, "B&W believes that the reactor trip upon turbine trip is disadvantageous." B&W states that " plants utilizing a once-through steam generator have the capability to run back on turbine trip without a reactor trip" and the " avoiding of a reactor trip for this event results in smaller perturbations in the primary system."
Discussion Prior to the accident at TMI-2, B&W operating plants utilized a runback feature to avoid a reactor trip upon turbine trip. However, for each of these events, the PORV was opened to relieve reactor coolant system pressure. As part of the post-TM1-2 fixes to minimize challenges to the PORV, B&W-designed plants were required to lower the high pressure reactor trip setpoint from 2355 psig to 2300 psig and raise the PORV setpoint from 2255 psig to 2450 psig. These actions removed the runback capability for turbine trip events.
In addition, B&W plants were required to install anticipatory reactor trips for loss of feedwater and turbine trip.
On applications currently undergoing OL review, such as Midland, the applicant has proposed certain design modifications that may rtduce the probability of a small break loss-of-coolant accident (SBLOCA) caused by a stuck-open PORY.
These modifications include:
(1) a fully walified safety-grade PORV; (2) safety-grade indication of PORV position; (3) dual safety-grade PORY block valves, capable of being automatically closed if a PORY maiiur.ction occurs; (4) a test program to demonstrate PORY operability; (5) installation of a safety-grade reactor trip on total loss of feed-water; and (6) resetting the PORY and high pressure reactor trip setpoints to their original values of 2255 psig and 2355 psig, respectively.
Should these modifications be found acceptable by the staff, the necessity of installing an anticipatory reactor trip upon turbine trip may be negated.
However, until these or similar modifications are proposed and found acceptable by the Comission, the plant design must incorporate anticipatory reactor trips for both loss of feedwater and turbine trip.
No change has been made in paragraph (2)(xxiv) because of the coments.
However, the Comission has modified the wording for cisrification and deleted the words " safety grade" oecause this term has not been defined in the regu-lations. The paragraph has been renumbered (2)(xxiii) and modified to read as follows:
" Provide, as part of the reactor protection system, an anticipatory reactor trip that would be actuated on loss of main feedwater and on turbine trip.
( Applicable to B&W-designed plants only) (ll.K.2.10)"
g)(xxvi)-RecordingReactorVesselWaterLevel One comentor (GE) stated that this requirement should be deleted because task II.K.3.23 was not included in NUREG-0737.
Discussion The TMI action plan, Table C.3, NUREG-0660, indicates that this issue is being covered in connection with TMI action plan item I.D.2, plant safety parameter display console; this latter item is identified in NUREG-0737.
Specific con-sole requirements for operating reactor licensees and OL applicants are under consideration by the Commission at the present time. The Comission considers that central water level recording is necessary for BWRs, and that it is appro-priate to address such capability in a preliminary manner during the CP safety review.
Consequently, this requirement will be maintained.
However, the Com-mission has noted that the range over which the reactor vessel water level must be recorded as specified in the proposed rule is inconsistent with that spect-fied in Regulatory Guide 1.97.
Since either range is acceptable for the plants covered by the rule, the Comission has modified the requirement to allow that flexibility in its implementation. This paragraph has been renumbered (2)(xxiv) and changed to read as follows:
" Provide the capability to record reactor vessel water level in 03e location on recorders that meet normal post-accident recording require-ments. ( Applicable to BWR's enly)
(II.K.3.23)"
(2)(xxviii) - ALARA Exposures A comentor (Bechtel) noted that this requirement applies to the design basis of systems outside containment likely to contain radioactive material, rather than the development of leakage control and detection provisions intended by NUREG-0718, Item 111.0.1.1.
Discussion The Comission has isnumbered the paragraph (2)(xxvi) and, for clarification, replaced the requirement with the following:
" Provide for leakage control and detection in the design of systems outside containment that contain (or might contain) TID 14844 source term radioactive materials following an accident. Applicants shall submit a leakage control program including an initial test program, a schedule for re-testing these systems, and the actions to be taken for minimizing leakage from such systems. The goal is to minimize potenthi exposures to workers and public, and to provide reasonable assurance that excessive leakage will not prevent the use of systems needed in an emergency. (III.D.1.1)"
(3)(1), (ii). (iii) - Administrative Procedures and Quality Assurance A.
A comentor (Gilbert) stated that these requirements are a restatement of present 10 CFR requirements.
Discussion Item (3)(i) has not been a previous requirement for CP reviews (recently, this has been identified as a requirement for OLs as Item I.C.5, NUREG-0737) nor have Items (3)(11) and (iii), as stated in the proposed rule, been previous CP requir+ments.
B.
Three commentors (S&W, NQA, TVA) noted that the inference of section (3)(iii) is that Appendix B of 10 CFR 50 is not sufficiently definitive.
If this is the case, the proper place to provide such clarification or additional requirements is through Appendix B.
It is the recomendation of the HQA Com-mittee that parat;raphs 50.34(f)(3)(ii) and (iii) be deleted from the proposed addition to the regulations because they do not clarify Appendix B and can only add confusion.
Discussion 10 CFR 50 Appendix B does set forth basic QA criteria from which to develop a QA program.
10 CFR 50.34(a)(7) requires that the applicant describe its QA program in the PSAR and include a discussion of how the applicab'e reouirements of Appendix B will be satisfied.
Regulatory Guide 1.70 and the W., dard Review Plan provide additional guidance on the extent to which this QA program should be described. The controls described in 50.34(f)(3)(ii) cnd (iii) provide additional detailed criteria for proper implementation of Appendix B requirements.
C.
Two commentors (NQA, Bechtel) noted that existing regulations contain pro-visions for the independence (separation) of those individuals who perform functions of attaining quality objectives from those individuals who verify compliance with requirements. Regulatory Guide 1.64 contains additional expla-nation for the intended independence for design verification purposes. Tra proposed addition to 10 CFR Part 50 goes beyond other regulations and regulatory guides and suggests the emphasis be placed on organizational independence rather than independre.ce of personnel for objectivity and proficiency.
Discussion The Comission agrees that Regulatory Guide 1.64 contains sufficient guidelines for independent verification of designs. Of particular concerr. to the Comis-sion is the lack of sufficient independence of the organization responsible for performing checks, verifications, and inspections. Therefore, this aspect of an effective QA program is emphasized in the rule.
D.
A comentor (NQA) also noted that (3)(iii)(B) "would require the entire body of quality assuring activities to be performed at the construction site. This would require massive upheaval and relocation to the construction site of not only top management, but also all support organizations."
Discussion The objective of item B is to ensure that sufficient quality assurance and quality control activities are performed at the site rather than at corporate offices to prc/de closer management oversight and communication. To clarify i.he Comission's intent, (3)(iii)(B) has been modified to read:
"(B) performing quality assurance / quality control functions at con-sh uttion sites to the maximum feasible extent;"
E.
The comentors (NOA, Bechtel) noted that (3)(iii)(C) is not clear whether quality assurance personnel should be involved in development of the procedures or should be assigned actions through the procedures.
Discussion The Comission agrees that this item needs clarification to ensure a better understanding of the intent.
Item (3)(iii)(C) has been modified to read as f ollows:
" including QA personnel in the documented review of and concurrence in quality-related procedures associated with design, construction, and instal-lation."
F.
A comentor (NQA) noted that (3)(iii)(D) is "not clear in what is meant by 0A requirements.
If this refers to the requirements for quality assurance pro-gramatic activities, the statement is acceptable; if it refers to requirements for the physical characteristics for classes of equipment, the statement is inap-p rop ri ate. "
Discussion The Comission agrees that this requirement should be clarified.
(3)(iii)(D) has been revised to read:
"estaDlishing criteria for determining QA programatic requirements;"
G.
A comenter (NOA) noted that " existing regulations now require the estab-lishing of qualification requirements for personnel performing quality assurance activities. Regulatory Guides such as 1.58 and 1.146 add additional clari-fication concerning personnel who perform quality verification activities. It is not at all clear what additicnal requirements are intended" by Section (3)(iii)(E).
Discussion The Comission acknowledges that the existing regulations do require, although not explicitly, tne establishment of such qualification requirements. However, the Comission is retaining the requirements stater'. in (3)(iii)(E) to ensure that they are considered in the QA program. The word " minimum" has been deleted from this section to be consistent with Appendix B to 10 CFR Part 50.
H.
The comentor (NQA) notes "that existing regulations would require staffing the quality assurance unit of the organization commensurate with its duties and responsibilities.
It is not at all clear how the organization is staffed com-mensurate with its 'importance to safety'. Ordinarily, duties and responsibil-itiec reflect the importance of the activity to be performed." Part (3)(iii)(F)
"is not clear what is intended by the addition of "importance to safety"."
Discussion To clarify the intent, (3)(iii)(F) has been modified by deleting the phrase "importance to safety". Existing regulations do not specifically address the numbers of QA/QC individuals required for the design arid construction activities associated with building a nuclear power plant. The size of the QA/QC organiza-tion should be dependent upon the quantity and type of quality-related activities that are on-going or projected during the design and construction of the nuclear facility.
I.
The comentor (NQA) notes, relative to (3)(iii)(G), "that existing regulations contain requirements for preparation and maintenance of documentation including
'as-built' documentation. The problem concerning procedures may lie not in the requirements for them or their establishment, but in their implementation; i.e.,
procedures are available, but they may not be being followed."
Discussion Existing regulations (i.e., Criterion VI, " Document Control" of Appendix B to 10 CFR 50) establish QA requirements for "... instructions, procedures, and drawings..." btit do not address "as-built" documentation (e.g., as-built drawings). Because the controls imposed upon as-built drawings, which accurately reflect the actual plant design, have been abused in the past, it is the Comis-sion's position that as-built documentation be addressed specifically by the QA requirements contained in the design and construction QA program. Therefore, (3)(111)(G) has not been mcdified.
J.
Three comentors (S&W, NQA, Bechtel) assert that the intent of (3)(iii)(H) is not clear. The NQA said that "if intent is to place quality assurance personnel on the design and analysis team, their independence may be compromised.
Appendix B now requires that during design, the activities of design control and design verification are to be identified, defined, perforned in accordance with written procedures by persons having proper capabilities and sufficiently inde-pendent of those who produced the design, so as to eliminate any conflict of interest. This being true, it is not at all clear what is intended by the pro-posed addition."
Discussion The Comission agrees that existing regulations (i.e., Criterion 111, " Design Control" of Appendix B to '.0 CFR 50) alrea(v establish the requirements that verification of the adequacy of design be "... performed by individuals er groups other than those who performed the original design..." However, it is the Comis-sion's intent that design documents (e.g., drawings, specifications, etc.) also be reviewed by individuals knowledgable and qualified in OA/QC techniques to ensure that the documents contain the necessary QA/QC requirements (e.g., inspection and test requirements).
For this reason, (3)(iii)(H) has not been changed.
(3)(iv) - Containment Penetratien Several comentors (OPS, Gilbert, L CEC, TVA) centered on the asserted arbitrari-ness of the requirement for a 3-foot diameter penetration, the lack of technical
. justification, and the possibility that containment venting provisions may not provide a significant contribution to safety.
Discussion The containment penetration size was selectad sa that it would be consistent with mitigation features designed to accomodate :4ed;um+ ar.d slow-rate pressure rises in containments that would otherwise have failed.
Aneug the features ccqsidered were filtered vented containment systems and passive containment cooling systems.
Rapid-rate pressure rises from hydrogen burns, for example, werd excluded from consideration. The 3-foot penetration was determined to be a conservative penstra-tion size that would not preclude the eventual installation of one of the afore-mentioned features. Of course, there is the possibility that such penetrations will not be needed, but that will be known only af ter the completion of the degraded core rulemaking. Therefore, the Commission has retained this requirement so as not to preclude later installation of containment venting systems, if required.
(3)(v) - Containment Design A.
One commentor (OPS) interprets the. information requested on post-inerting and ignition systems as not allowips pre-inerting as a hydrogen control r4asure.
Another connentor (CE) s tat nd to the related action plan items in NUREG 0718 and NUREG 0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident." They are provided herein for information only.
- causes, provide a description and evaluation of the effect on small break LOCA probability of an automatic PORV isolation system that would operate when the reactor coolant system pressure falls af ter the PORV has opened.
(Applicable to PWR's only).
(II.K.3.2)
(v)
Perform an evaluation of the safety effectiveness of providing for separation of high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) system initiation levels so that the the RCIC system initiates at a higher water level than the HPCI system, and of providing that both systems restart on icw water level.
(For plants with high pressure core spray systems in lieu of high pressure coolant injection systems, sub-stitute the words, "high pressure core spray" for "high pressure coolant injection" and "HPCS" for "HPCI") (Applicable to BWR's only ).
(II.K.3.13)
(vi)
Perform a study to identify practicable system modifications that would reduce challenges and failures of relief valves, without compromising the performance of the valves or other systems.
(Applicable to BWR's only).
(II.K.3.16)
(vii)
Perform a feasibility and risk assessment study to determine the optimum automatic depressurization system (ADS) design modifications that would eliminate the need for manual activation to ensure adequate core cooting.
( Applicable to BWR's only).
(II.K.3.18)
(viii) Perform a study of the effect on all core-cooling modes under accident conditions of designing the core spray and low pres-sure coolant injection systems to ensure that the systems will automatically restart on loss of water level, after having been manually stopped, if an initiation signal is still present.
( Applicable to BWR's only).
( I I.K. 3.21 )
(ix)
Perform a study to determine the need for ado tional space cooling to ensure reliable long-term operation of the reactor core isolation cooling (RCIC) and high-pressure coolant injection (HPCI) systems, following a complete loss of offsite power to the plant for at least two (2) hours. (For plants with high pressure core spray systems in lieu of high pressure coolant injection systems, substitute the words, "high pressure core spray" for "high pressure coolant injection" and "HPCS for "HPCI") ( Applicable to BWR's only).
(II.K.3.24)
(x)
Perform a study to ensure that the Automatic Depressurization System, valves, accumulators, and associated equipn.ent and instrumentation will be capable of performing their intended functions during and following an accident situation, taking no credit for non-safety related equipment or instru sentation, and accounting for normal expected air (or nitrogen) leslage through valves.
(Applicable to BWR's only).
(II.K.3.28)
(xi)
Provide an evaluation of depressurization methods, other than by full actuation of the automatic depressurizaiton system, that would reduce the possibility of exceeding vessei integrity limits during rapid cooldown.
(Applicable to BWR's only)
(II.K.3.45)
. (xii)
Perform an evaluation of alternative hydrogen control systems that would satisfy the requirements of paragraph (2)(x11) of this section.
As e minimum include consideration of a hydrogen ignition and post-accident inerting system. The evaluation shall include:
(A)
A comparison of costs and benefits of the alternative systems considered.
(B) For the selected system, analyses and test data to verify compliance with tne requirements of (2)(ix) of this section.
(C) For the selected system, preliminary design descriptions of equipment, function, and layout.
(2) To satisfy the following requirements, the application shall provide suf-ficient information to demonstrate that the required actions will be satis-f actorily completed by the operating license stage. This information is of the type customarily required to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues.
(i)
Provide simulator capability that correctly models the control room and includes the capability to simulate small-break LOCA's.
(Applicable to construction permit applicants only)
(I.A.4.2.)
(ii)
Establish a program, to begin during construction and follow into operation, for integrating and expanding current efforts to improve plant procedures. The scope of the program shall include emergency procedures, reliability analyses, human factors engineering, crisis management, operator training, and coordination with INP0 and other industry efforts.
(Applicable to construction permit applicants only )
(I.C.9)
. (iii)
Provide, for Commission review, a control room design that reflects state-of-the-art human factor principles prior to committing to fabrication or revision of fabricated control room panels and lay-outs.
(I.D.I)
(iv)
Provide a plant safety parameter display console that will display to operators a minimum set of parameters defining the safety status of the plant, capable of displaying a full range of important plant parameters and data trends on demand, and capable of indicating when process limit are being approached or exceeded.
(I.D.2)
(v)
Provide for automatic indic6 tion of the bypassed and operable status of safety systems.
(I.D.3)
(vi)
Provide the capability of high point venting of noncondensible gases f rom the reactor coolant system, and other systems that may be required to maintain adequate core cooling.
Systems to achieve this capability shall be capable of being operated from the control room and their operation shall not lead to an unacceptable increase in the probability of loss-of-coolant accident or an unacceptable challenge to containment integrity.
(II.B.1)
(vii)
Perform radiation and shielding design reviews of spaces around systems that may, as a result of an accident, contain TID 14844 source term radioactive materials, and design as necessar) to permit adequate access to important areas and to protect safety equipment from the radiation environment.
(II.B.2)
. (viii) Provide a capability to promptly obtain and analyze samples from the reactor coolant system and containment that may contain TID 14844 source term radioactive materials without radiation exposures to any individual exceeding 5 rem to the whole-body or 75 rem to the extremities. Materials to be analyzed and quantified include certain radionuclides that are indicators of the degree of core damage (e.g., noble gases, iodines and ceshms, and non-volatile isotopes), hydrogen in the containment atmospt.ere, dissolved gases, chloride, and beron concentrations.
(II.B.3)
(ix)
Provide a system for hydrogen control that can safely accorrnodate hydrogen generated by the equivalent of a 100% fuel-clad metal water reacti on. Preliminary design information on the tentatively preferred system option of those being evaluated in paragraph (1)(xii) of this section is sufficient at the construction permit stage. The hydrogen control system and associated systems shall provide, with reasonable assurance, that: (II.B.8)
(A) Uniformly distributed hydrogen concentrations in the containment do not exceed 10% during and following an accident that releases an equivalent amount of hydrogen as would be generated from a 100%
fuel clad metal-water reaction, or that the post-accident atmo-sphere will not support hydrogen combustion.
(B) Combustible concentrations of hydrogen will not collect in areas where unintended combustion or detonation could cause loss of con-tainment integrity or loss of appropriate mitigating features.
. (C)
Equipment necessary for achieving and maintaining safe shut-jown of the plant and maintaining containment integrity will perform its safety function during and af ter being exposed to the environmental conditions attendant with the release of hydrogen generated by the equivalent of a 100% fuel-clad metal water reaction including the environmental conditions created by activation of the hydrogen control system.
(D)
If the method chosen for hydrogen control is a post-accident inerting system, inadvertent actuation of the system can be safely accommodated during plant operation.
(x)
Provide a test progrcm and associated model development and conduct tests to qualify reactor coolant system relief and safety valves and, for PWR's, PORY block vdives, for all fluid conditions expected under operating conditions, transients and accidents.
Consideration of anticipated transients without scram (ATWS) conditions shall be included in the test program. Actual testing under ATWS conditions need not be carried out until subsequent phases of the test program are developed.
(II.D.1)
(xi)
Provide direct indication cf relief and safety valve position (open or closed) in the control room.
(II.D.3)
(xii)
Provide automatic and manual auxiliary feedwater (AFW) system initi-ation, and provide auxiliary feedwater system flow indication in the control room.
(Applicable to PWR's only) (II.E.1.2)
(xiii) Provide pressurizer heater power supply and associated motive and control power interfaces sufficient to establish and
. maintain natural circulation in hot standby conditions with only onsite power available.
(Applicable to PWR's only)
(II.E.3.1)
(xiv)
Provide containment isolation systems that:
(II.E.4.2)
(A) ensure all non-essential systems are isolated automatically by the containment isolation system, (B) for each non-essential penetration (except instrument lines) have two isolation barriers in series, (C) do not result in reopening of the containment isolation valves on resetting of the isolation signal, (D) utilize a containment set point pressure for initiating containment isolation as low as is compatible with normal operation, (E) include automatic closing on a high radiation signal for all systems that provide a path to the environs.
(xv) Provide a capability for containment purging / venting designed to minimize the purging time consistent with ALARA principles for occupational exposure.
Provide and demonstrate high assurance that the purge system will reliably isolate under accident conditions.
(II.E.4.4)
(xvi)
Establish a design criterion for the allowable number of actuation cycles of the emergency core cooling system and reactor protection system consistent with the expected occurrene rates of severe overcooling events (considering both anticipated transients and accidents).
(Applicable to B&W designs only).
( II.E.5.1 )
. (xvii) Provide instrumentation to measure, record and readout in the control ro2a:
( A) containment pressure, (B) containment water level (C) containment hydrogen concentration, (D) containment radiation intensity (high level), and (E) noble gas efiluents at all potential, accident release points.
Provide for con-tinuous sampling of radioactive iodines and particulates in gaseous effluents f rom all potential accident release points, and for onsite capability te analyze 'and measure these samples.
(II.F.1)
(xviii) Provide instruments that provide in the control room an unambiguous indication of inadequate core cooling, such as primary coolant saturation meters in PWR's, and a suitable combination of signals from indicators of coolant level in the reactor vessel and in-core thermocouples in PWR's and BWR's.
(II.F.2)
(xix)
Provide instrumentation adequate for monitoring plant conditions following an accident that includes core damage.
(II.F.3)
(xx)
Provide power supplies for pressurizer relief valves, block valves, and level indicators such that:
( A) level indicators are powered from vital buses; (B) motive and control power con-nections to the emergency power sources are through devices qualified in accordance with requirements applicable to systems inportant to safety and (C) electric power is provided from emergency power sources.
(Aplicable to PWR's only).
(II.G.1)
(xxi)
Design auxiliary heat removal systems such that necessary automatic ated manual actions can be taken to ensure proper functioning when the main feedwater system is not operable.
( Applicable to BWR's only).
(II.K.1.22)
(xxii) Perform a failure modes and effects analysis of the integrated control system (ICS) to include consideration of failures and effects of input and output signals to the ICS.
(Applicable to B&W-designed plants only).
(II.K.2.9)
(xxiii) Provide, as part of the reactor protection system, an anticipatory reactor trip that would be actuated on loss of main feedwater and on turbine trip.
(Applicable to B&W-designed plants only).
(II.K.2.10)
(xxiv) Provide the capability to record reactor vessel water level in one location on recorders that meet normal post-accident recording requirements.
(Applicable to BWR's only).
(II.K.3.23)
(xxv)
Provide an onsite Technical Support Center, an onsite Operational Support Center, and, for construction permit applications only, a nearsite Emergency Operations Facility.
(III.A.1.2)
(xxvi) Provide for leakage control and detection in the design of systems outside containment that contain (or might contain) UD 14844 source term radioactive materials following an accident. xpplicants shall submit a leakage control program, including an initial test program, a schedule f or re-testing these systems, and the actions to be taken for minimizing leakage from such systems. The goal is to minimize potential exposures to workers and public, and to provide reasonable assurance that excessive leakage will not prevent ths use of systems needed in an emergency.
( III.D.1.1 )
(xxvii) Provide for monitoring of inplant radiation and airborne radioactivity as appropriate for a broad range of routine and accident conditions.
(III.D.3.3)
(xxviii) Evaluate potential pathways for radioactivity and radiation that may lead to control room habitability problems under accident conditions resulting in a TID 14844 sotrce term release, and make necessary design provisions to preclude such problems.
(III.D.3.4)
(3)
To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the requirement has been met.
This in.'ormation is of the type customarily required to satisfy 10 CFR 50.34(a)(1) or to address the applicant's technical qualifications and management structure and competence.
(i)
Provide administrative procedures for evaluating operating, design and construction experience and for ensuring that applicable important industry experiences will be provided in a timely manner to those designing and constructing the plant.
(I.C.5)
(ii)
Ensure that the quality assurance (QA) list required by Criterion II, App. B,10 CFR Part 50 includes all structures, systems, and components important to safety.
(I.F.1)
(iii)
Establish a quality assurance (QA) program based on consideration of:
( A) ensuring independence of the organization performing checking functions from the organization responsible for performing the functions; (B) performing quality assurance / quality control functions at construction sites to the maximum feasible extent; (C) inciuu;ag QA personnel in the documented review of and concurrence in quality related procedures associated with design, construction and installatien; (D) establishing criteria for determining QA pro-gramatic requirements; (E) establishing qualification requirements for QA and QC personnel; (F) sizing the QA staff connensurate with its duties and responsibilities; (G) establishing procedures for maintenance of "as-built" documentation; and (H) providing a QA role in design and analysis activities.
(1.F.2)
(iv)
Provide one or more dedicated containment penetrations, equivalent in size to a single 3-foot diameter opening, in order not to preclude future installation of systems to prevent containment f ailure, such as a filtered vented containment system.
(II.B.8)
(v)
Provide preliminary design information at a level of detail con-sistent with that normally required at the construction permit stage of review sufficient to demt,nstrate that:
(II.B.8)
(A) Containment integrity will be maintained (i.e., for steel con-tainmer.ts by meeting the requirements of the ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsubarticle NE-3220, Service Level C Limits, except that evaluation of instability is not required, considering pressure and dead load alone. For concrete containments by meeting the requirements of the ASME Boiler Pressure Yessel Coda,Section III, Division 2 Subsubarticle CC-3720, Factored Load Category, considering pressure and dead load alone) during an accident that releases niydrogen generated from 100% fuel clad metal-water reaction accompanied by either hydrogen burning or the added pressure f rom post-accident inerting assuming carbon dioxide is the inerting agent. As a minimum, the specific code requirements L
set forth above appropriate for each type of containment will be met for a combination of dead load and an internal pressure of 45 psig. Modest deviations from these criteria will be con-sidered by the staff, if good cause is shown by an applicant.
Systems necessary to ensure containment integrity shall also be demonstrated to perform their function under these conditions.
Subarticle NE-3220, Division 1, and subarticle CC-3720, Division 2, of Section III of the July 1,1980 ASME Boiler and Pressure Vessel Code, which are referenced in 50.34(f)(3)(v)(A)(1_) and 50.34(f)(3)(v)(B)(1_), were approved for incorporation by reference by the Director of the Office of the Federal Register, A notice of any changes made to the material incorporated by reference will be published in the FEDERAL REGISTER.
Copies of the ASME Boiler and Pressure Vessel Code may be purchased from the American Society of Mechanical Engineers, United Engineering Center, 345 East 47th St., New York, NY 10017.
It is also available for inspectio at the Nuclear Regulatory Comission's Public Document Rcom,1717 H St. NW, Washington, D. C.
' (B)
Q) Containment structure loadings produced by an inadvertent full actuation of a post-accident inerting hydrogen control system (assuming carbon dioxide), but not including seismic or design basis eccident loadings will not produce stresses in steel con-tainments in excess of the limits set forth in the ASME Boiler and Pressure Yessel Code,Section III, Division 1, Subsubarticle NE-3220, Service Level A Limits, except that evaluation of insta-bility is not required (for concrete containments the loadings specified above will not produce strains in the containment liner in excess of the limits set forth in the ASME Boiler and Pressure Yessel Code,Section III, Division 2, Subsubarticle CC-3720, Service Load Category), (2_) The containment has the capability to safely withstand pressure tests at 1.10 and 1.15 times (for steel and concrete containments, respectively) the pressure cal-culated to result from carbon dioxide inerting.
(vi)
For plant designs with external hydrogen recombiners, provide redundant dedicated containment penetrations so that, assuming a single failure, the recombiner systems can be connected to the containment atmosphere.
(II.E.4.1) tvii)
Provide a description of the management plan for design end construc-tion activities, to include:
(A) the organizational and management structure singularly responsible fcr direction of design and construc-tion of the proposed plant; (B) technical resources director by the applicant; (C) details of the interaction of design and construction within the applicant's organization and the manner by which the appli-
'. cant will ensurt close integration of the architect engineer and the nuclear steam supply vendor; (D) proposed procedures for handling the transition to operation; (E) the degree of top level management over-sight and technical control to be exercised by the applicant during design and construction, including the preparation and implementation of procedures necessary to guide the effort.
(II.J.3.1) 10 CFR PART 2 3.
The Authority citation for Part 2 reads as follows:
AUTHORITY:
Secs. 161p and 181, Pub. L.83-703, 68 Stat. 950 and 953.
(42 U.S.C. 2201(p) and 2231; sec 191, as amended, Pub.L.87-615, 76 Stat.
409 (42 U.S.C. 2241); sec 201, as amended, Pub.L.93-438, 88 Stat. 1242 (42 U.S.C. 5841); 5 U.S.C. 552; unless otherwise noted.
Sections 2.200 -
2.206 also issued under sec.186, Pub.L.83-703, 68 Stat. 955 (42 U.S.C.
2236) and sec. 206, Pub.L.93-438, 88 Stat.1246 (43 U.S.C. 5846).
Sections 2.800 - 2.808 also issued under 5 U.S.C. 553.
4.
Paragraphs (e)(1)(ii) and (e)(3)(iii) of $2.764 are revised to read as follows:
92.764 Immediate effectiver.>s o' certain initial decisions.
a (e)(1)
(ii)
In reaching their decisions the Boards should interpret existing regulations and regulatory policies with due consideration to the impli-cations for those regulations and policies of
. the Three Mile Island accident.
As provided in paragraph (e)(3) of thir section, te additioa to taking generic rulemaking actions, the Com-mission will be providing case-by-case guidance on changes in regulatory policies in conducting its reviews in adjudicatory proceedings. The Boards shall, in turn, apply these revised regu-lations and pnlir.ies in cases then pending before them to the extent that they are applicable. The Commission expects the Licensing Boards to pay par-titular attention in their decisions to analyzing the evidence on those safety and environmentn; issues arising under applicable Commission rpgu-lations and policies which the Boards believi present serious, close questions and which the Boards belies 2 nay be crucial to whether a license stouid becor:e effecitre before full appellate review is cttpleted. Furthermore, the Boards should identify e.ny aspects of the case which in their judgment, present issues on which prompt Com-mission policy guidance is called for. The Boards may request the assi;tance of the parties in iden-tifying such policy issues but, absent specific Com-nission directives, such policy issues' thal ~) not be the subject of discov2ry, examination, or cross-examination.
'. (3i ***
(iii)
In announcing the result of its review of a:ty Appeal Board stay decision, the Comission may allow the proceeding to run its ordinary ccurse or give whatever instructiccu cs to the future handling of the proceading it deems appropriate Ocr example, it may direct the /pseal Board to review the nrf ts of particular issues in expe-dited fashion; furnish policy guidim.ca with respect to particular issues; or decide to review the merits of particular issues itself, bypassing the Appec13rard).
g Dated at Washington, D. C., this day of 1981.
FOR THE NUCLEAR REGULATORY COMMISSION Samuel J. Chilk Se:.retary of the Coranitsion
~
s
4 e
a e
ENCLOSURE 2