ML20058J515
| ML20058J515 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 08/10/1982 |
| From: | Mauck J, Rossi C Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| ISSUANCES-OL, NUDOCS 8208110062 | |
| Download: ML20058J515 (56) | |
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
- BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of LONG ISLAND LIGHTING COMPANY Docket No. 50-322 (Shoreham Nuclear Power Station Unit 1)
NRC STAFF REPLY TESTIMONY OF CHARLES E. ROSSI AND JERRY L. MAIJCK ON INSTRUMENTATION FOR POST-ACCIDENT MONITORING (SCCONTENTION27)
(50C CONTENTION 3) l l
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DESIGNATED ORIGINAIJ E88"AA88R*o*388"a82 cart m .a 31 <> [/4 j T PDR /
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OUTLINE OF TESTIMONY ,
i Concurrent with the filing of Staff's original testimony on SC 27/ SOC 3, Applicant purported in its testimony to show compliance with Regulatory Guide 1.97, Rev. 2, or tu provide justifications for ap-propriate alternates. Subsequently, the Commission approved Staff's proposal regarding implementation of Emergency Response Capability in-cluding the Regulatory Guide in question. Staff is unable to state at present whether Applicant is in compliance with the Regulatory Guide.
The Staff's review at Shoreham will follow the guidance approved by the Commission. Such review may be after the beginning of plant operation.
However, it is the Staff's position that Shoreham can, in the interim, be operated without undue risk to the health and safety of the public.
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD ,
~
In the Matter of Docket No. 50-322 LONG ISLAND LIGHTING COMPANY )
) (0L)
(ShorehamNuclearPowerStation )
Unit 1) )
NRC STAFF REPLY TESTIMONY OF CHARLES E. ROSSI AND JERRY L. MAUCK ON INSTRUMENTATION FOR POST-ACCIDENT MONITORING 1.Q Please state your names and positions with the NRC.
1.A My name is Jerry L. Mauck. I am employed by the U.S. Nuclear Regulatory Commission as a Reactor Engineer in the Instrumentation and Control Systems Branch (ICSB) of the Division of Systems Integration (DSI). A copy of my professional qualifications is attached to the original Staff testimony filed in response to SC Contention 27 and SOC Contention 3.
{
A. My name is Charles E. Rossi, I am employed by the U.S. Nuclear Regulatory Commission as a Section Leader in ICSB, DSI. A copy of my professional qualifications is attached.
2.Q Please state the sections of the Shoreham SER and supplements which are relevant to SC 27/ SOC 3.
' 1 2.A Regulatory Gpide 1.97, Rev. 2 is discussed in Section 7.5 of the Shoreham SER dated April, 1981. However, it should be
,noted that some items related to this contention, SC 27/ SOC 3, ,
- appear in Chapter ?? of the Shoreham Supplemental SER's due to their inclusion in NUREG-0737.
3.0 Please state the background events which lead to the preparation of this reply testimony.
3.A In May of this year the Staff filed testimony responding to the contentions in question which stated that:
- 1. Regulatory Guide 1.97, Rev. 2 has not yet been implemented and, accordingly, licensees and applicants have not yet been required to address it.
- 2. The Shoreham applicant will be expected to comply with the Regulatory Guide or to provide appropriate technical justifications for any citernatives when the Regulatory Guide is implemented.
Concurrent with the filing of the above-referenced Staff testimony, the Applicant also filed testimony in response to the contentions. The Applicant's testimony, among other things, purported to show compliance with the intent of Regulatory Guide 1.97, Rev. 2.
Subsequent to the filing of the above referenced testimonies, the Commission considered the Stafs proposed requirements for emergency response capability (including requirements for post .
~
accident monitoring) contained in SECY-82-111. " Requirements
s for Emergency Response Capability" dated March 11, 1982. The Commission approved SECY-82-111 on July 16, 1982. The Staff
, was informed of the Consnission's decision in a memorandum from ,
, Samuel J. Chilk dated July 20, 1982. Copies of these documents, which are attached, were served on the Board and parties on July 27, 1982.
On July 30, 1982, the Board at Tr. 8551-53, requested the Staff to address several matters set out below.
4.Q Are you prepared to state whether or not the Applicant meets the guidelines set out in Regulatory Guide 1.97, Rev.2?
4.A No. There is not sufficient information available at the present time for the Staff to make a decision on the ultimate acceptability of the Applicant's position with respect to the specific items listed in SC Contention 27 and S0C Contention 3.
. Before a proper review can be made, additional details with l
regard to instrument ranges, locations, pcwer supplies, design criteria, and qualification criteria will be necessary (see page 14 of the enclosure to SECY-82-111). Furthennore, it would be imprudent of the Staff to make piecemeal decisions with respect to Regulatory Guide 1.97, Rev. 2 on one specific plant, which might set a precedent for other plants, without benefit of a careful, orderly review. ,
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5.Q Please state.how the Staff plans to review Shoreham for compliance with the Regulatory Guide in question.
5.A ,The Staff plans to perform an audit review of the Shoreham
, plant, to ascertain conformance with R.G. 1.97, Rev. 2, in conjunction with the Staff's review of emergency response capability. This audit review is not a prerequisite for implementation of R.G.1.97, Rev. 2 (see page 14 of the enclosure to SECY 82-111). As noted above, proposed require-ments were sent to the Comission in SECY-82-111 " Requirements for Emergency Response Capability," March 11, 1982. The Comission has approved SECY-82-111 subject to modifications.
The Staff was informed of the Comission's decision by memorandum, S. Chilk to W. Dircks, " Staff Requirements - Affinnation Session
. . , , " July 20, 1982. Enclosure C to the July 20, 1982, memorandum is a " Statement of Policy: Further Comission Guidance on Emergency Response Capability". This statement clearly relates (page 2) the applicability of SECY-82-111 to operating license proceedings including implementation schedules.
The schedule for implenenting basic requirements for Emergency Response Capability is shown on page 1 of the enclosure to SECY-82-111. There it is stated that:
When the basic requirements for emergency response capabilities and facilities are finalized, they should be transmitted to ,
licensees by a generic letter from NRR, ,
promulgated to NRC Staff, and incorporated as .
i l
regulatory requirements (e.g., in the Standard Review Plan or by regulation or Order as appropriate). The letter to licensees should request that licensees submit a proposed schedule for completing actions to comply with the basic ,
requirements. Each licensee's proposed schedules would then be reviewed by the assigned NRC Project Manager, who would discuss the subject with the licensee and mutually agree on schedules and completion dates. The implementation dates would then be formalized into an enforceable document.
Use of existing documentation is addressed on pg. 3 of the enclosure to SECY-82-111:
The NRC Staff recommends that the following NUREG documents are intended to be used as sources of guidance and information, and the Regulatory Guides are to be considered as guidance or as an acceptable approach to meeting formal requirements. The items by virtue of their inclusion in these documents shall not be misconstrued as requirements to be levied on licensees or as inflexible criteria to be used by NRC Staff reviewers.
R.G.1.97 is included in the list of documents. Furthermore, pages 13 and 14 of the enclosure to SECY-82-111 discuss im-plementation of R.G. 1.97. Documentation and NRC Review is addressed on pg.14 where it is stated that:
Deviations from the guidance in Regulatory Guide 1.97 (Rev. 2) should be explicitly shown, and supporting justification or alternatives should be presented.
6.Q Is it possible that the review and implementation of Regulatory Guide 1.97, Rev. 2, may not be completed prior to the licensing of Shoreham?
6.A Yes. As provided for in SECY 82-111, the review is not a
- prerequisite for Applicant irplementation of Regulatory Guide 1.97, Rev. 2 guidance. Further, the date of final
implementati.on as agreed upon between the Staff and the Applicant will be fonnalized as provided in SECY 82-111.
7.Q On what basis does the Staff propose to proceed with the licensing of Shoreham in the absence of the completion of a review of the Applicant's response to the Regulatory Guide?
7.A It should be understood that the Staff has not completed review of the response to Regulatory Guide 1.97, Rev. 2 for any plant - neither for any licensed plant nor for any plant under licensing review. The Shorehsm plant, as well as other plants for which the licensing review has been completed, was reviewed in accordance with the Standard Review Plan to insure that sufficient indications are available for the operator to cope with Design Basis Events. In addition, the Emergency Procedures for Shoreham at the time of initial plant operation will make use only of the indications available at that time. It should be noted that in addition to indications needed to follow the course of design basis events, Regulatory Guide 1.97, Rev. 2, recommends indications (and ranges) for severe accident scenarios that may go beyond the design basis events.
8.Q Please summarize the Staff's position with regard to the contention in question. ,
8.A Although the Applicant has attempted to show compliance with the intent of Regulatory Guide 1.97, Rev. 2, the Staff is i
i
unable to state whether the Applicant has in fact adequately complied with the Regulatory Guide. It is unlikely that
, compliance can be demonstrated prior to final licensing of ,
, Shoreham. By Commission vote the Staff has been instructed to use SECY-82-111 in implementation of Emergency Response Facilities.
SECY-82-111 implementation includes the recommendations of R.G.
1.97, Rev. 2. The Staff has been instructed to be flexible in the schedule and implementation of the provisions of R.G. 1.97, Rev. 2, under the premise that an orderly, well planned effort will result in greater safety than would a rapid, possibly fragmented attempt to modify existing designs. The Staff's position with respect to the specific items listed in SC 27/ SOC 3 is that the Shoreham plant can be operated without undue risk to the health and safety of the public until a Staff review and l
decision is made with respect to the applicant's overall compliance with Regulatory Guide 1.97, Rev. 2 l
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- IN RESPONSE REFER
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UNITED STATES -
TO M820716
,i y % % NUCLEAR REGULATORY COMMISSION ,
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- W ASHINGTON, D.C.20555 ACTION - NRR/IE/RES/ ELD
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July 20, 1982 Cys: Dircks Retu S D[* **** C#1 / Stello cFFICE OF THE ..-
SECRETARY MEMORANDUM FOR: William J. Dircks, Executive DiNector for Operations FROM: - - Samuel J. Chilk,- Secretar STAFFREQUIREMENTS-AFFh' " 1982, ION SESSION,
~
SUBJECT:
.. 11:50 A.M., FRIDAY, JULY 16 COMMISSIONERS' CONFERENCE R OM, DC OFFICE
- ' (OPEN TO PUBLIC ATTENDANCE)
SECY-82-lli - YEatTirements-for Emergency Response Capability.
The Commission, by a vote of 4-l* (Commissioner Gilinsky disapproving) approved SECY-82-lll subject to the following:
- 1. The staff should provide an information paper which responds to the ACRS' recommendations (May lef 1982 letter to the Chairman) for additional staff attention to emergency operating procedures', the Safety Parameter Display System, and Control Room Design Reviews. (NRR) -
- 2. Item 8 on page 23 of the Enclosure to SECY 8'2-111 should be revised to read: "8. Staffed using Table as a 2.
(previous guidance approved by the Commission)
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goal. Reasonable exceptions to goals for the number of additional staff personnel and response times for their arrival should be justified and will be considered by NRC staff." (IE)
- 3. Add the following to the first paragraph on page 1 of the enclosure to SECY 82-111:
"It is also not intended that either the guidance documents or the fundamental requirements 'are to be considered binding legal requirements at this time. As indicated below," however, the fundamental requirements will be translated into binding legal requirements in the manner specified."
- Section 201 of the Energy Reorganiration Act, 42 U.S.C. 55841, provides that acticn of the Commission shall be determined by
- a " majority vote of the members present." Commissioner Asselstine was not present when this item was affirmed, but had previously indicated that he would approve. Commissioner Gilinsky was also not present when this item was affirmed, but had previously indicated that he would disapprove. Had Commissioners Gilinsky and Asselstine been present, they would have affirmed their prior votes. Accordingly, the formal vote of the Commission was 3-0 in favor of the decision.
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2 _
- 4. The statement under No. 1, on page 4 should be modified so that the last sentence reads:
- "While the NRC does net plan 'to impose additional . -
requirements on licensees rbgarding SPDS, the NRC will work with the industry to assure the development of
- . appropriate industry standards for SPDS systems. (NRR) 5, -The description provided to the ACRS,,on May 7, 1982 by .
'_. Hugh Thompson on NRC intentions regarding control ~oom r design reviiw-shon1 A he_, incorporated into SECY 82-111.
- (See also Commissioner Ahearne's vote sheet of 6/10/82, and pages 4-5 of the attachments to it.) (NRR)
- 6. A statement referring to the need for operating crew training to cover handling accident conditions both with and without the SPDS should be included in the enclosure to SECY 82-111. (NRR)
- 7. The enclosure to SECY-82-111 should be published as a NUREG-0737 Supplement. You should make the. appropriate modifications to the enclosure to make it suitable for such publication. In particular, you need to include the language proposed by OGC in their July 15, 1992 nemorandum as modified in enclosure A. You also should reflect that the document is Commission direction to licensees rather than a proposal for Commission review. (NRR)
- 8. Page . to the enclosure of SECY-82-lll should be modified as attached (Enclosure B) . (NRR)
! 9. The Commission has agreed that a Policy Statement be issued reflecting the Commission's approval of the enclosure to SECY-82-lll as a Supplement to NUREG-0737.
The proposed Policy Statement by OGC (July 15, 1982 memorandum) should be revised a's attached (Enclosure C)-
with appropriate additional changes.to reflect that l
SECY-82-lll is to be issued as a Supplement to NUREG-0737. (ELD)
- 10. NUREG-0696 should be revised, reviewgd by CRGR and '
approved and issued as a Regulatory Guide. The Commission should be advised of the progress and the final version should be sent to the Commission for review (under l negative consent procedures) prior to issuance.
(200$ (SECY. Suspense: July 15, 1983) -
(RES/IE) 9
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- 11. Prior to sending 50.54 (f) letters to licensees, you should provide the Commission'with a draft for approval.
(me) (SECY Suspense: September 1, 1982)
(NRR) ,
cc: Chairman Palladino Commissioner Gilinsky - . -.
- Commissioner Ahearne ' '
... Commissioner Roberts Commissioner Asselstine
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Commission Staff offices
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- DCS - 016 Phillips
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- ENCLOSURE A e ,
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/~ Insert to SECY-82-lli Report ~ I r-Add to page 2 of "NRC Staff Recommendations on the Requirements for Emergency Response Capability" (Enclosure to SECY-82-Ill) the following: -
i 'The recommended requirements set forth in this document have
. been re, viewed by the Codmission ahd, at a meeting held June 21,
' 1982, were approved by the Commission as appropriately clarifying
~ ~ .
- and providing greater detail with respect _to related TMI Action Plan requirements contained -in: NUREG-0737 for all operating
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license acclican'ts. These recommended requirements are, therefore, to be ac~ corded the status of approved NUREG-0737, items as . set forth in the Commission's " Statement of Policy: Further
' Commission Guidance for Power Reactor Operating Licenses" (45 Fed.Pec. 85236, Dec. ,24, 19Bk) . In this connection, the provisions for scheduling set forth herein supersede any schedules with respect to such items contained in NUREG-0737. ,
Accordingly,' the recommended requirem'ents should be used by the staff and by adjudicatory boards as appropriate clarifications and interpretation of the related NUREG-0737 items.
l l
! The recommended requirements set forth in this document are believed to be consistent with the requirements regarding related ,
I .
items for construction permits and manufacturing licenses contained in 10 CFR '50.34 (f) and 10 CFR Pa'rt 50., Appendix E.
"it'-r, Accordingly, no change in such regulation is required.
the raccr.:nded requircrant: cent:i :d in thir dr-curcnt rherif be i
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ENCLOSURE e, .. : .- . 7 5.: e. . :.c ,
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,- : EM5RGENCY RESPONSE' CAPABILITY -
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'This repor.t ., .
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result of Al review by the Committee to
.' - . Review Generic Requirements '(CRGR). The recommendations herein have been developed by the program offices and are supported by CRER. 'The re'pbrt represents the staff's attempt-to-distill-the fundamental .requiren
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for nuclear plant Emergenc9. Response C'apability from the wide range of
- . g0idance documents that NRC has. issued. It is not intended that these
~
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guidance docuinents '(NUREG reports and Regulatory Guides) be ignored; they hre still useful, sources of guidance for licensees and KRC staff
_. . '. regarding acceatabl.e means for meeting the fundamental requirements - ~ ~
contained-in tiis document. .
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These' fundamental requirements are further specification of the general
- guidance specified previossly by the Commission in its regulations, orde'r's and policy statements on emergency planning and TMI issues. It
~ is . intended that these~ fundamental requirements would be applicable to
. jlicensees of operating 6uclear power plants and holders of construction permits.for nuclear power plants. For applicants for a construction
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permit {Cp) or manufacturing license (ML),'the requirements described in rovisions in the .
this docum'ent must rdle specifying licensing requirements for p,ending ,CP, be supplemented with the specific p'and ML appl
.Th0s'," compliance with requirements in this document may not be sufficient t O
j meet the related requirements in 10 CFR 50.34 (f) and Appendix E.
In" E.fis regard, it is expect'ed that the staff would review CP and ML
, applications against the cuidance in the : current Standard ,
Review Plan. ~
m -
)(whichincludes'theprovisionsofNUREG0718) ~
ano,inis mi j this'docume.ght. nt, ii) lead to more detailed, requirements tha Based on discussions with licensees; the staff has learned that many of the Commission approved schedules for emergency response facilities probably'will not be met. In recognition of this fact and the difficult.
of i.mplementing generic deadlines, the staff propeses that plant-specifi schedules be established which take into account the unique status of each plant.
- is proposed. The. ,following . . . 7, , . , sequence.
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for., developing implementation sc
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k' hen'the basic requirements for emergency response capabilities and' facilities are finalized, they should be transmitted to licensees by a pr
.regulatory generic letter requirements from NRR,(e.omulgated g., in the Standard, to NRC Review staff, and Plan incorporated or by regulat as or Order, a's appropriate). The letter to licensees should request that licensees submit a proposed schedule for completing actions to comply with the basic requirements. Each licensee's proposed schedules would then be reviewed by the assi ned 5 HRC Project Manager, who would discuss the subject with the licensee and mutually agree on schedules and cosph h 36tvdb5bmmha 4*vumM3AcaEalaan5LLwMLieuwm&m
. ENCLOSURE C- Z~
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. STATbMENT OF POLICY: FURTHER COMMISSION GUIDANCE :- $.-
ON EMERGENCY RESPONSE CAPABILITY EGR.
=.CT= c:=:=C ::=:::
AGENCY: Nuclear Regulatory Commission ACTION: - Publication of Policy Statement . -
SUMMARY
- On December 24, 1980, the Commission published a Statement of Polipy: "Further Commission Guidance to Power
-Reactor- Operating Licenses" (45 FR 85f36)$ modifying an earlier
~ Policy Statement on the same subject (45 FR 41736, June 23, 1980). The Commission's Policy Statement discussed the background of efforts to improve safety requirements in light of experience resulting from the Three Mile Island accident. The Commission in'dicated that operating license applications would be measured by the Commission's regulations, as augmented by the requirements reflected in NUREG-0737, " Clarification of Action Plan Requirements". The Commission.f.urther noted that it will continue to monitor developments with regard to litigation of i
action plan requirements and continue to offer guidance where appropriate.
l Since that time, the NRC staff has developed a number of l
NUREG documents and other guidance docume.nts which provide information and guidance as to methods of implementation and other details concerning certain NUREG-0737 items relating to ,
emergency response capability. The more important elements of these various staff documents have been identified in "NRC Staff Recommendations on the Requirements for Emergency Response 1
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Capability", which the NRC staff has recommended be adopted by the NRC in order to provide guidance clarifying and amplifying the NUREG-0737 items relating to emergency response capability.
P The Commission has considered the staff
- . . -- Jo e secommendation and
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approved the requirements recommended in "NRC Staff Recommenda-
~ tions on the Requirements for Emergency Response Capability" as appropriately clarifying and providing greater detail with
, respect to the TMI ActioTr Plan Requirements contained in NUREG-0737 relating to emergency response capability. The provisions for scheduling set 'forth in these recommended requirements supersede the schedules with respect to related NUREG-0737 items.
Accordingly, the Commission has concluded that these recom-mended requirements s,hould be used by the staff and by adjudica-tory. boards as appropriate clarification %s and interpretations of related NUREG-0737 items and should be accorded the status of ,
approved NUREG-0737 items as set forth in the December 24, 1980 Statement of Policy. Litigation of the recommended requirements l
set forth in NRC. S,taff Recommendations on the Requirements for Emerge $cy Response Capability should be permitted in operating license proceedings under the same c6nditions as those applicable to NUREG-0737 items in accordance with the December 24, 1980 .
, Statement of Policy. In this regard, it should be understood i
that the Commission's December 24, 1980 Statement of Policy is applicable to all operating license applications and that therefore this new guidance 'on emergency response cacability is l cpplicable to all operating license applications.
1
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March 11, 1982 SECY-82-111 w.....,f POLICY ISSUE (Notation Vote)
For: The Commissioners From: William J. Dircks Executive Director for Operations
Subject:
REQUIREMENTS FOR EMERGENCY RESPONSE CAPABILITY
Purpose:
To request Commission approval of a set of basic re-quirements for emergency response capability and approval for the staff to work with licensees to develop plant-specific implementation schedules.
Discussion: One of the first issues reviewed by the Comittee to Review Generic Requirements (CRGR) was the broad area of emergency response facilities and capabilities at nuclear plants. The Committee found that imple-
- mentation schedules were not being coordinated within the NRC. In addition, existing NRC documents published as guidance to licensees were sometimes being used as fim requirements. Discussions with industry representatives and the staff indicated that some licensees had slowed down on work in this area pending NRC clarification of its requirements.
- Some utilities have virtually stopped work on some of the items, while others have proceeded and, in some cases, completed some of the items. The Committee recommended that steps be taken by the Office Directors involved to clarify the requirements and implementation schedules for the Safety Parameter Display System (SPDS), Control Room Design Review, upgraded Emergency Operating procedures, Regulatory Guide 1.97 Technical Support Center (TSC), Operational
. Support Center (OSC), and Emergency Operations Facility (EOF). In my memo to the Commission dated December 31, 1981, I noted that the DEDROGR staff would work with the program offices to clarify the basic requirements in this area and establish a ,
revised implementation plan. '
. Enclosed are the staff's recommendations for the requirements in the broad area of emergency response facilities and capabilities outlined above. The requirements were developed by the program offices ,
Contact:
V. Stello, Jr., DEDROGR 49-29704
. i j The Comissioners r and are supported by CRGR. The enclasure represents a distillation of fundamental requirements from the -
broad range of guidance documents that NRC has issued (principally NUREG reports and Regulatory Guides). The staff intends that the guidance documents referred to in the enclosure not be used to impose requirements on licensees, but rather that they be used as sources of guidance for NRC reviewers and licensees regarding acceptable means for meeting the fundamental requirements proposed.
In discussions with owners' groups and individual licensees, the staff has learned that the Commission approved schedule of October 1,1982, for implementation of the TSC and EOF probably cannot be met. In recognition of this fact and the difficulty of implementing generic deadlines, the staff is proposing that plant-specific schedules be established which take into account the unique status of each plant.
Each licensee would be requested to submit a proposed schedule for completing the actions to comply with the fundamental requirements. The NRC Project Manager for each plant should be knowledgeable of the overall work effort going on at a plant and, based on guidance received from NRC management, could reach agreement with licensees on schedules which optimize use of. utility and NRC resources. The agreed upon comp.letion dates would be formalized in an order. By this approach, future staff coordination problems regarding implementation schedules will be avoided.
Resource The costs to licensees to implement the requirements Estimates: proposed in the enclosure were included in the estimates set out in NUREG-0660.
Recommendation: That the Commission:
- 1. Approve the fundamental requirements described in the enclosure.
- 2. Approve the issuance of the requirements in the '
enclosura by 50.54f letters as a revision to NUREG 0737,
- 3. Approve the method for establishing plant- i specific implementation schedules described in the enclosure.
. s
. i The Commissioners 4. Approve the implementation of these requirements through plant-specific orders.
- 5. Note that the staff intends to use the previously issued NUREG reports and Regulatory Guides as guidance documents only. .
Scheduling: Licensees are currently required to establish a TSC and EOF by October 1. Prompt action on this paper is required in order to provide guidance to licensees.
William J. Dircks Executive Director for Operations
Enclosure:
NRC Staff Recommendation on the Requirements for Emergency Response Capability l
Commissioners' comments should be provided directly to the Office of the Secretary by c.o.b. Monday, March 29, 1982.
- Commission Staff Office comments, if any, should be submitted to the Commissioners NLT Monday, March 22, 1982, with an infornation 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.
DISTRIBUTION
- Commissioners Commission Staff Offices Exec Dir for Operations Exec Legal Director ACRS ASLBP ASLAP ,
Secretariat
O O E I
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ENCLOSURE 9
NRC STAFF RECOMMENDATIONS
. ON THE REQUIREMENTS FOR
. EMERGENCY RESPONSE CAPABILITY a - . . . .
Ma.r.ch _10,1982 9
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e 9
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,, , D- - - - -.- - - - - - - -
O r .
CONTENTS Pa21
.' 1. I NT R O D U CTI O N . . . . . . . . . . . . . . . . . . . . . . . . .,. . . . . . . . . . . . . . . . . . . . . . . . . 1
- 2. USE OF EXISTING DOCUMENTATION................................. ,
3
- 3. COORDINATION AND IRTEGRATION OF INITI ATIVES. . . . . . . . . . . . . . . . . . . 4
- 4. SAFETY PARAMETER DISP LAY SYSTEM (5P05). . . . . . . . . . . . . . . . . . . . . . . . 7
- Current Regulatory Requirements
- Functional Statement Recommended Requirements
- Integration -
Reference Documents
- 5. DETAILED CONTROL ROOM DESIGN REVIEW........................... 10
- Current Regulatory Requirements .
- Functional Statement
. Recommended Requirements Documentation and NRC Review Integration
. - Reference Documents .
- 6. REGULATORY GUIDE 1.97 - APPLICATION TO EMERGENCY RESPONSE -
FACILITIES.................................................... 13
' Current Regulatory Requirements Functional Statement Recommended Requirements Documentation and NRC Review t
l 7. UPGRADE EMERGENCY OPERATING PROCEDURES (E0Ps)................. 15
! Current Regulatory Requirements
- Functional Statement .
- Recommended Requirements Documentation and NRC Review Reference Documents
- 8. EMERGENCY RESPONSE FACILITIES................................. .
17 Current Regulatory Requirements Technical Support Center.................................... 19 l -
l Functional Statement Recommended Requirements l -
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. I CONTENTS (Continued) 21 Operati onal Support Center. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Functional St'atement
- Recommended Requirements 22 Emergency Operations Facility..............................
Functional Statement
- Recommended Requirements -
Documentation and NRC Review
- Reference Documents 25 Table 1 - Emergency Operations Facility Location Options..........
Table 2 - Minimum Staffing Requirements for' NRC Licensees for ' 26 Nuclear Power Plant Emergenci es. . . . . . . . . . . . . . . . . . . . . . . . .
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EMERGENCY RESPONSE CAPABILITY
- 1. INTRODUCTION . ,
> This report was prepared as a result of a review by the Comittee to Review Generic Requirements (CRGR). The recommendations herein have been developed by the program offices and are supported by CRGR. The report represents the staff's attempt to distill the fundamental requirements for nuclear plant Emergency Response Capability from the wide range of guidance documents that NRC has issued. It is not intended that these guidance docuinents (NUREG reports and Regulatory Guides) be ignored; they are still useful sources of guidance for licensees and NRC staff I .
regardint acceptable means for meeting the fundamental requirements
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containet in this document. -
j These fundamental requirements are further specification of the general guidance specified previously by the Commission in its regulations, It I
orders and policy statements on emergency planning and TMI issues.
is intended that these fundamental requirements would be applicable to l licensees of operating nuclear power plants and holders of construction permits for nuclear power plants. For applicants for a construction permit {CP) or manufacturing license (ML), the requirements described in this document must be supplemented with the specific provisions.in the rule specifying licensing requirements for pending CP and ML applications.
In this regard, it is expected that the staff would review CP and ML -
applications against the guidance in the current Standard Review Plan.
and this mi.ght lead to more detailed requi,rements than prescribed in this document.
Based on discussions with licensees, the staff has learned that many of the Commission approved schedules for emergency response facilities probably will not be met. In recognition of this fact and the difficulty of implementing generic deadlines, the staff proposes that plant-specific schedules be established which take into account the unique status of each plant. The following sequence for developing implementation schedules is proposed.
When the basic requirements for emergency response capabilities and i facilities are finalized, they should be transmitted to licensees by a generic regulatoryletter from NRR,(promulgated to NRC staff, and incorporated as requirements or Order, as appropriate). The letter to ifcensees should request that licensees submit a proposed schedule for completing actions to comply with the basic- requirements. Each licensee's proposed schedules would then be reviewed by the assigned NRC Project Manager, who would discuss the subject with the licensee and mutually agree on schedules and completion dates. The implementation dates would then b'e formalized into an enforceable document..
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The basic requirements in this ' document do not alter previously issued guidance, which remains in effect. This document does attempt to place that guidance in perspective by identifying the elements that the NRC staff believes to be essential to upgraded emergency response capabilities.
The proposal to formalize implementation dates in an enforceable document reflects the level of importance which the NRC staff attributes to these basic requirements. The NRC staff does not recomend that existing guidance be imposed in this manner, but rather that it be used as guidance to be considered in upgrading emergency response capabilities. This indicates the distinction which the staff believes should be made between
- the basic requirements and guidance.
The following sections describe NRC staff recomendations on basic re-quirements, their interrelationships, and NRC actions to improve manage-ment of emergency response regulation. Reference documents are cited with a description of content as it relates to specific initiatives.
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- 2. USE OF EXISTING DOCUMENTATI0N' .
The NRC staff recommends that the following NUREG documents are intended to be .
used as sources of guidance and information, and the Regulatory Guides are to be considered as guidance or as an acceptable approach to meeting formal requirements. The items by ' virtue of their inclusion in these documents shall not be misconstrued as requirements to be levied on licensees or as inflexible criteria to be used by NRC staff reviewers.
NUREG Report Titles 0695 - Functional Criteria for Emergency Response Facilities 0700 - Guidelines for Control Room Design Reviews Draft Criteria for Preparation of Emergency Operating Procedures 0799 -
0801 - Evaluation Criteria for Control Room Design Reviews 0S14 - Methodology for Evaluation of Emergency Response Facilities 0818 - Emergency Action Levels for Light Water Reactors 0835 . Human Factors Acceptance Criteria for SPDS .
Regulatory Guides , ,
1.23 (Rev. 1)
- Meteorological Measurement Program for Nuclear Power Plants 1.97 (Rev. ;2) - Instrumentation for Light-Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident l ,
1.101 (Rev. 2) - Emergency Planning for Nuclear Power Plants 1.47 - Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems t
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- 3. COORDINATION AND INTEGRATION OF INITIATIVES
- 1. The design of the Safety Parameter Display S'ystem (SPDS), design of instrument displays based on Regulatory Guide 1.97 guidance, control room design review, development of symptom oriented emergency operating proce-dures, and operating staff training should be integrated with respect to the overall enhancement of operator ability to comprehend plant conditions and cope with emergencies. Assessment of information needs and display formats and locations should be performed by individual licensees. The SPDS could affect other control room improvements that licensees may consider. In some cases, a good SPDS may obviate the need for large-scale control room modifications. However, installation of the SPDS should not be delayed by slower progress on other initiatives. The SPDS should not be contingent on completion of the control room design review. NRC does not plan to impose additional requirements on licensees regarding SPDS.
- 2. Implementation of part or all of Regulatory Guide 1.97 (Rev. 2) represents a control room improvement. The implementation of control room improve-ments is not contingent on implementing Technical Support' Center (TSC) and Emergency Operations Facility (EOF) requirements.
- 3. The Technical Support Center (TSC) and Emergency Operations Facility (EOF) are dependent on control room improvements in,tems of communication and instrumentation needs among the TSC, EOF, and control room. TSC and EOF '
facilities are not necessarily dependent on each other. The Operational-Support Center (OSC) is independent of TSC and EOF.
- 4. The three groups of initiatives--SPDS, control room improvements, and ,
emergency response facilities (TSC, EOF, OSC)--should have the following interrelationships: -
. a. The SPDS is an improvement in the control room ~ because it enhances operator ability to comprehend plant conditions and interact in situations that require human intervention. The SPDS could affect other control room improvements that licensees may consider. In some cases, a good SPDS could obviate the need for extensive modifications to control rooms.
- b. New instrumentation that may be added to the control room should be considered a requirement for inclusion in the design of the TSC and
. EOF only to the extent that such instrumentation is essential to the performance of TSC and EOF functions.
- c. The SPDS and control room improvements are essential elements in
. operator training programs and the upgraded plant-specific emergency operating r ocedures.
- d. Acquisition, processing, and. management of data for SPDS, control-room improvements, and emergency response facilities should be coordinated but need not be centralized. .
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Specific implementation plans and reasonable, achievab[e be established by agreement between the NRC Project Ma .
individual licensee.
requirement should develop procedures identifying the following:
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The respective roles of NRR, IE, and Regional Offices in managing implementation, checking licensee rate of progress, and verifying -
compliance, including the extent to which NRC review and inspec is necessary during implementation.
- b. Procedural methods and enforcement measure upon schedules without significant delays and extensions.
6.
The NRC Project Manager for each nuclear power plant is assigned prog management responsibility for NRC staff actions associated with i -
menting emergency response initiatives.
- principal contact for the licensee regarding these initiatives.
7.
NRC will make allowances for work already doneFor byeach licensees case in a faith effort to meet requirements as they understand them.
in which a licensee would have to remove or rip out emergency respons facilities or equipment that was installed in good fai,th to meet p guidance in order to meet the basic requirements described in ment, the Director of the Office of Nuclear Reactor Regulation or tion and Enforcement.will review the circumstances removal is necessary or existing facilities or equipment acceptable alternative. removal or major modification of existing .
equipment requires the specific approval of the Office Director.
8.
NRC recognizes that acceptable alternative methods Each licensee of phasing grating emergency response activities may be developed.
needs flexibility in integrating these activities, taking into acco d varying degree to which the licensee has implemented past re An example of a way in which these activities could be inte-guidance. Other' methods of integration proposed by grated is discussed below. licensees would be reviewed conside l
i, nitiative. '
- a. SPDS '
Review the functions of the nuclear power plant operatin (1) that are necessary to recognize and cope with rare events
.- (a) posa significant contributions to risk, .(b).could ca operators to make cognitive errors in diagnosing them, a (c) are not included in routine operator training programs.
(2)
Combine the results of this review with be incorporated in the SPDS.
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(3) Design, build, and install the SPDS in the control room and train its users.
- b. To be done parallel without delaying SPDS, complete emergency opera-ting procedure technical guidelines that will be used to develop plant-specific emergency operating procedures.
- c. Using these E0P technical guidelines, the SPDS design, and accepted human factors principles, conduct a review of the control room design. Apply the results of this review to:
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(1) Verify SPDS paramete'r selection, data display, and functions.
(2) Develop plant-specific E0Ps.
(3) Design control room modifications that correct conditions adverse to safety (reduce significant contributions to risk),
and add additional instrumentation that may be necessary to implement Regulatory Guide 1.97.
(4) Train and qualify plant operating staff regarding E0Ps and modifications.
- d. Verify, prior to finalization of designs for modifications and of procedures and training, that the functions of control room operators in emergencies can be accomplished (i.e., that the individual initi.a-tives have been integrated sufficiently to meet the needs of control room operators and provide adequate emergency response capabilities).
- e. Implement E0Ps and install control room modifications coincident with scheduled outages as necessary, and train operators in advance of '
these changes as they are phased into operation.
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Current Regulatory Requirements
." No licensee action is required.
Functional Statement The SPDS should provide a concise display of critical plant variables to the control room operators to aid them in rapidly and reliably determin safety status of the plant.
operations as well as during abnorm'al conditions, the emergency conditions in determining the safety status of the plant and in assessing whether abnormal conditions warrant corrective action by operators This can be particularly important during anticipated avoid a degraded core.
transients and the 1.nitial phase of an accident.
Recommended Requirements
- 1. Each operating reactor shall be provided with a Safety Parameter This Display System that is located convenient to the control room operators.
system will continuously display information from wh
- are. responsible for the avoidance of degraded and damaged core events.
2.
The control room instrumentation required (see General Design Criteria 13 and 19 of Appendix A to 10 CFR 50) forms the basic safety components required for safe reactor operation under normal, transient, and accid ,
The SPDS is used in addition to the basic components and
- conditions. Thus, requirements applicable serves to aid and augment these components.
to control room instrumentation are not needed for this augmentation (e.g., GDC 2, 3, 4 in Appendix A; 10 CFR Part 100; single-failure requ L
ments).. The SPDS need not meet requirements of the single-failure The criteria and it need not be qualified to meet Class 1E requirements.
SPDS shall be suitably isolated from electrical or electronic interference The SPDS with equipment and sensors that are in use for safety s indication is not required for the sole purpose of being a backup for SPDS.
After the SPDS has been installed. operating procedures should be available that will allow timely and correct safety status assessment whe the SPDS is not available.
- 3. There is a wide range of useful information that can be pr vari'ous systems.
Prompt implementation of an SPDS can provide an important contri The selection of specific-information that should be plant safety.
provided for a particular plant shall be based on engineering judg l individual plant licensees, taking into account the importance of prompt
' implementation.
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- 4. The SPDS display shall be designed to incorporate accepted human factors principles so that the displayed information can be readily perceived and .
comprehended by SPDS users.
- 5. Minimum information to be provided shall be sufficient to provide infoma-tion to plant operators about:
- a. Reactivity control >
- b. Reactor core cooling and heat removal from the primary system '
- c. Reactor coolant system integrity
- d. Radioactivity control ,
- e. Containment conditions The specific parameters to be displayed shall be determined by the licensee. ,
- 6. The licensee shall prepare a written safety analysis describing the basis
- on which the selected parameters are sufficient to assess the safety status of each identified function for a wide range of events, which include symptoms of severe accidents. Such analysis, along with the specific implementation plan for SPDS shall be reviewed as described below. .
- 7. The licensee's ' proposed implementation of an SPDS system shall be reviewed I -
in accordance with the licensee's technical specifications to determine .
whether the changes involve an unreviewed safety question or i:hange of .
technical specifications. If they do, they shall be processed in the normal fashion with prior NRC review. If the changes do not involve an .
unreviewed safety question or a change in the technical specifications,
.' the licensee may implement such change's without prior approval by NRC.
However, the licensee's analysis shall be submitted to NRC promptly on completion of review by the licensee's offsite committee. Based on the results of NRC review, the Director of IE or the Director of NRR may request or direct the licensee to cease implementation if a serious safety -
question is posed.by the licensee's proposed system, or if the licensee's analysis is seriously inadequate. . . ..
Integration Prompt implementation of an SPDS is a design goal and of primary importance.
The schedule for implementing SPDS should not be impacted by schedules for the control room design review aind develcpment of symptom-oriented .
emergency operating procedures. For this reason, licensees should develop and propose an integrated schedule for implementation in which If reasonable, the SPDS design is an input to the other initiatives. -
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- this schedule should be accepted by NRC. -
Reference Documents .
I NUREG-0660 -- Need for SPDS identified ,
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NUREG-0737 Specified SPDS Functional critaria for SPDS NUREG-0696 .
NUREG-0835 Specific acceptance criteria keyed to 0696 Reg. Guide 1.97 (Rev. 2) -- Instrumentation for Light-Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident e
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- 5. DETAILED CONTROL ROOM DESIGN REVIEW Current Regulatory Requireme6ts As specified in Item I.D.1 in NUREG-0737, the ' implementation scheduie is still to be developed.
Functional Statement The objective of the control room design review is to " improve the ability of nuclear power plant control room operators to prevent accidents or cope with i - accidents if they occur by improving the information provided to them" (from NUREG-0660, Item I.D.1). As a complement to improvements of plant operating staff capabilities in response to transients and other abnormal c:nditions that will result from implementation of the SPDS and from upgraded emergency opera-t ting procedures, this design review will identify any modifications of control room configurations that would contribute to a significant reduction of risk l
and enhancement in the safety of operation. Decisions to modify the control room would include consideration of long-term risk reduction a'nd any potential temporary decline in safety after modifications This resulting shouldfrom be the need to carefully relearn ma'intenance and operating procedures.
reviewed by persons competent in human factors engineering and risk analysis.
Recommended Requirements
- 1. Conduct a control room design review to identify human engineering dis- .
crepancies. The review shall consist of: .
- a. The establishment of a qualified multidisciplinary review team and a review program incorporating accepted human engineering principles.
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- b. The use of function and task analysis (that had been used as the basis for developing emergency operating procedure Technical Guide-lines) to identify control room operator tasks and information and c'ontrol requirements during emergency operations. This analysis has multiple purposes and should also serve as the basis for developing training and staffing needs and verifying SPDS parameters. .
- c. A comparison of the display and control requirements with a control room inventory to identify missing and surplus (distracting) displays and controls,
- d. ' A control room survey to identify deviations from accepted human factors principles. This survey will include, among other things, .
assessment of control room layout, the usefulness of audible and
- visual alarm systems, information recording and recall capability, and control room environment.
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- 2. Asses's which human engineering discrepancies are significant and should be corrected. Select design improvements that will correct those discrep-ancies. Improvements that can be accomplished with an enhancement program
'. (paint-tape-label) should be done promptly.
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- 3. Verify that each selected design improvement will provide .
unacceptable human engineering discrepancies Improvements that because in which a temporary reduction in safety could occur.
are introduced should be coordinated with changes resulting from other improvement programs such as SPDS, operator training, new instrumentation (Reg. Guide 1.97, Rev. 2), and upgraded emergency operating procedures.
I Documentation and NRC Review 1.
All licensees shall submit a p'rogram plan within two months of the start of the control room review that describes how items 1, 2 and 3 above will l be accomplished. NRC approval is not required before licensees conduct l~ their reviews.
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Selected licensees will underge an in progress audit by the NRR human -
' factors staff b~ased on the program plans and advice from, resident inspectors.
- 3. All licensees shall submit a summary report outlining proposed control room changes. The report will also provide a summary justification for human engineering discrepancies with safety significance to be left uncorrected or partially corrected.
- 4. . Within two weeks after receipt of the licensee's summary report, the NRC will inform the licensee whether it will conduct a pre-implementation onsite audit. The decision will be based on the content of the program The plan, summary report, and results of NRR in progress licensees selected for in progress audits under paragraph 2.
5.
For control rooms selected for pre-implementation onsite audit, within one month after receipt of the summary report, the NRC will conduct:
- a. A pre-implementation audit of proposed modifications (e.g., equipmen additions, deletions and relocations, and proposed modifications).
- b. An audit of the justification for those human engineering discrep-l ancies of safety significance to be left uncorrected or only partially corrected.
The audit will consist of a review of licensee's record of the room reviews, discussions with the licensee review team, and usually a '
Within a month after this onsite audit, NRC will cont'rol room visit.
- issue its safety evaluation report (SER).
6.
For control rooms for which NRC does not perform a pre-implementation onsite audit, NRC will conduct a review and issue its SER within two
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months after receipt of the licensee's summary report. The review shall be similar to that conducted for pre-implementation plants under para-graph 5 above, except that it may or may not include a specific audit. -
The SER shall indicate whether, based on the review carried out, changes in the licensee's modification plan are needed to assure operational safety. Flexibility is considered in the control room review, because certain control board discrepancies can be overcome by techniques not involving control board changes. These techniques could include improved procedures, improved training, or the SPDS. .
- 7. The following approach will be used fo'r OL review. For OL applications with SSER dates prior to June 1983, licensing may be based on either a Preliminary Design Assessment or a Control Room Design Review (CRDR) at the applicant's option. However, applicants who choose the Preliminary Design Assessment option are required to perform a CRDR after licensing.
For applications with SSER dates after June 1983, Control Room Design Review will be required prior to licensing. -
Integration Prompt im'lementation p of an SPDS is a design goal and of primary importance.
The schedule for. implementing SPDS should not be impacted by schedules
- for the control room design review and development of symptom-oriented .
emergency operating procedures. For this reason, licensees should .
I develop and propose an integrated schedule for implementation in which tne SPDS design is an input to the other initiatives. . If reasonable.
- this schedule should be accepted by NRC.
Reference Documents _.
NUREG-0585 --
States that licensees should conduct review.
NUREG-0660, Rev. 1 -- States that NRR will require reviews for operating reactors and operating licensee applicants.
NUREG-0700 Final guidelines for CRDR.
-- States that requirement was issued June,1980, final '.
guidance not yet issued.
' NUREG-0801 -- October 1981 draft for comment; staff evaluation criteria. , ,
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r REGdLATORY GUIDE 1.97
- 6. APPLICATION TO EMERGENCY RESPONSE FACILITIES Current Regulatory Requirements .
No licensee action is required.
Functional Statement Regulatory Guide 1.97 provides data to assist control room operators in pre venting and mitigating the conseque'nces of reactor accidents.
Recommended Requirements
- 1. Control Room
- Proviiie measursments and indication of licensees Individual Type A, may B, C, D,excep-take E variables lis in Regulatory Guide 1.97 (Rev. 2). .BWR incore thermocouples tions based on plant-specific design features.
and continuous offsite dose monitors are not required pending theirIt is acce further development and consideration as requirements.
to rely on currently installed equipment if it will measure over the range indicated in Regulatory Guide 1.97 (Rev. 2), even if the equipment
- presently not environmentally qualified. i t lly
. required to monitor the course of an accident woul
- qualification.
Provide reliable indication of the meteorological variables (wind direc tion, wind speed, and atmospheric stability) specified in RegulatoryN Guide 1.97 (Rev. 2) for site meteorology.
logical monitoring systems are necessary if they have historically t-provided reliable indication of these variables that are represen Information meteorological conditions in the vicinity of the plant site.
on meteorological conditions for the region in which the site is locate l
shall be available via communication with the National W
- 2. Technical Support Center (TSC)_ ,
f The Type A, B, C, D, E variables that are essential for performa functions shall be indicated in the TSC.
a.
BWR incore thermocouples and continuous offsite dose monitor '
required pending their further development and consideration as
- requirements,
- b. The indicators and associated circuitry shall requirements.
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- 3. Emergency Operations Facility (EOF) -
- a. Those primary indicators needed to monitor containment conditions and releases of radioactivity from the plant shall be provided in the EOF.
i b. The EOF data indications and associated circuitry shall be of reliable design but need not meet Class 1E, single-failure or seismic qualification requirements.
Documentation and NRC Review ,
NRC review is not a prerequisite for implementation. Staff review will be in the form of an audit that will include a review of the licensee's method of implementing Regulatory Guide 1.97 (Rev. 2) guidance and the licensee's sup-
. porting technical justification of any proposed alternatives.
The licensee shall submit a report describing how it meets these requirements.
The submittal should include documentation which may be in the' form of a table that includes the following information for each Type A, B, C, D, E variable shown in Regulatory Guide 1.97 (Rev. 2): ,
(a) instrument range ,
(b) environmental qualification (as stipulated in guide or state criteria)
(c) seismic qualification (as stipulated in guide or state criteria)
(d) quality assurance (as stipulated in guide or state criteria)-
(e) redundancy and sensor (s) location (s)
(f) power supply (e.g., Class IE, non-Class IE, battery backed) ,
(g) location of display (e.g. , control room board, SPDS, chemical laboratory) l (h) schedule (for installation or upgrade) - <
. Deviations from the guidance in Regulatory Guide 1.97 (Rev. 2) should be explicitly s'hown, and supporting justification or alternatives should be presented. ,
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- 7. UPGRADE EMERGENCY CPERATING PROCEDURE.S (EOPs)
Current Regulatory Requirements i .' NUREG-0737, Item I.C.1, which has been approved by the Commission for imple-mentation.
Functional Statement .
Symptom-based emergency operating procedures will improve human reliability and the ability to mitigate the consequences of a broad range of initiating events and subsequent multiple failures or operator errors.
Recommended Requirements
- 1. In accordance with NUREG-0737, Item I.C.1, reanelyze transients and accidents and pr,epare Technical Guidelines. These analyses will identify The analyses also -
l operator tasks, and information and control needs.
serve as the basis for integrating upgraded emergency operating procedures and the control room design review and verifying the SPDS design.
- 2. Upgrade E0Ps to be consistent with Technical Guidelines and an appropriate
- procedure Writer's Guide.
- 3. Provide appropriate training of operating per'sonnel on the use of opgraded
. E0Ps prior to implementation of the E0Ps. . .
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- 4. Implement upgraded E0Ps. .
Documentation and NRC Review NRC will perform a pre-
- 1. Submit.Te'chnical Guidelines to NRC for review.
implementation review of the Technical Guidelines and the Writer's Guide.
Within two months of receipt of the Technical Guidelines and. Writer's Guide, NRC will advise the licensees of their acceptability.
- 2. Each licensee shall submit to NRC a procedures generation package at least three months prior to the date it plans to begin formal operator training on the upgraded procedures. NRC approval of the submittal is not The procedures necessary prior to upgrading and implementing the E0Ps.
generation package shall include:
- a. Plant-Specific Technical Guidelines -- plant-specific guidelines For plants using for.
plants not using generic technical guidelines.
generic technical guidelines, a description of the planned method for
- dev. eloping plant specific E0Ps from the generic guidelines, including plant specific information. .
- b. A Writer's Guide that details the specific methods to be used by the
- l. icensee in preparing E0Ps based on the Technical Guidelines.
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- c. A description of the program for validation of the E0Ps.
i d. A brief description of the training program for the upgraded E0Ps.
- 3. All procedures generation packages will be reviewed. On an audit basis for selected facilities, upgraded E0Ps will be reviewed. The details and extent of this review will be based on the quality of the procedures generation packages submitted to NRC. A sampling of ugpraded E0Ps will be reviewed for technical adequacy in conjunction with the NRC Reactor Inspection Program.
Reference Documents NUREG-0660, Item I.C.1,.I.C.8, I.C.9 NUREG-0799 e q e
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- 8. EMERGENCY RESPONSE FACILITIES Current Regulatory Requirements ,
10 CFR 50.47(b)(6) (for Operating License applicants) -- Pequirement for prompt communications among pr'incipal response organizations and to emergency personnel and to the public.
10 CFR 50.47(b)(8) -- Requirement for emergency facilities and equipment to support emergency response.
. 10 CFR 50.47(b)(9) -- Requirerneot th:t adeqnte methods, systems and equipment ~
for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition are in use.
10 CFR 50.54(q) (for Operating Reactors) -- Same requirement as 10 CFR 50.47(b) plus 10 CFP. 50, Appendix E. -
10 CFR 50, Appendix E, Paragraph IV.E Requirement for:
"1. Equipment at the site for personnel monitoring; "2. Equipment for determining the magnitude of and for continuously assessing the impact of the release of radioactive materials to the environment, -
"3. Facilities and supplies at the site for decontamination of
- onsite i.ndividuals; "4. Facilities and medical supplies at the site for appropriate I
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emergency first aid treatment; l
"S. Arrangements for the services of physicians and other medical personnel qualified to handle radiation emergencies on site;
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"6. Arrangements for transportation of contaminated injured individ-uals from the site to specifically identified treatment facili-ties outside the site boundary; -
"7. Arrangements for treatment of individuals injured in support of licensed activities on the site at treatment facilities outside the site boundary; ,
-"8. A licensee'onsite technical support center and a licensee near-site emergency operations facility from which effective direction can be given and effective control can be e::ercised during an emergency; ,
l "9. At ieast one onsite and one offsite communications system; each system shall have a backup power source.
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~ 18 All communication plans shall have arrangements for emergencies, including titles and alternates for those in charge at both ends of the communication links and the primary and backup means of communication. Where consistent with the function of the governmental agency, these arrangements will include:
"a. Provision for communications with contiguous State / local governments within the plume exposure pathway (emergency planning zone) EPZ. Such communications shall be tested monthly. .
"b. Provision for c'ommunications with Federal emergencySuch c response organizations.
be tested annually.
"c. Provision for communications among the nuclear power reactor control room, the onsite technical support center, and the near-site emergency operations facility; and among
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the nuclear facility, the principal State and local emer-gency operations centers, and the field assessment teams.
Such communications systems shall be tested annually.
'd. Provision for communications by the licensee with NRC Headquarters and the appropriate NRC Regional Office Operations Center from the nuclear power reactor control - ,
room, the onsite technical support center, and the near-site emergency operations facility. ' Such communications shall be tested monthly."
- Within this section on emergency response facilities,ethe Technical Suppo Center (TSC), Operational Support Center (OSC) and Emergency Operation Facility (EOF) are addressed separately in terms of their functional stat The subsections on Documentation and NRC Rev and recommended requirements.and Reference Documents that follow t section on' emergency response facilities.
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r Technical Support Center (TSC)
Functional Statement
.' The TSC is the onsite technical support center for emergency response. When activated, the TSC is staffe'd by predesignated technical, engineering, senior management, and other licensee personnel, and five predesignated NRC personnel.
During periods of activation, the TSC will operate uninterrupted to provide plant management and technical support to plant operations personnel, and to relieve the reactor operators of peripheral duties and communications not directly related to reactor system manipulations. The TSC will perform EOF functions for the Alert Emergency class and for the Site Area Emergency class and General Emergency class until the EOF is functional.
Recommended Requirements The TSC will be: .
- 1. Located within the site protected area so as to facilitate necessary interaction with control room, OSC, EOF and other personnel involved with the emergency. .
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- 2. Sufficient to accommodate and support NRC and licensee predesignated personnel, equipment and documentation in the center.
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- 3. Structurally built in accordance with the National Uniform Building Code.
- 4. Environmentally controlled to provide room air temperature, humidity and '
cleanliness appropriate for personnel and equipment.
- 5. Piovided with radiological protection and monitoring equipment necessary to assure that radiation exposure to any person working in the TSC would
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not exceed 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.
- 6. Provided with reliable voice and data communications with the control room and EOF and reliable voice communciations with the OSC, NRC Operations Centers and state and local operations centers.
- 7. Capable of reliable data collection, storage, analysis, display and communication sufficient to determine site and regional status, determine changes in status, forecast status and take appropriate actions. The following variables shall be available in the TSC: ,
(a) the variables in the appropriate Table 1 or 2 of Regtilatory
- Guide.1.97 (Rev. 2) that are essential for performance of TSC functions; and ,
(b) the meteorological variables in Regulatory Guide 1.97 (Rev. 2) for site vicinity and National Weather Service data available by voice communication for the region in which the plant is located.
20 Principally those data aust be available that wo determining plant status during. recovery operations.
8.
Provided with accurate, complete and current pl accident conditions.
- 9. Staffed by sufficient technical, engineering, and within approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> af,ter activation.
10.
Designed taking into account good human factors engineering p e
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Operational Support Center (OSC)
Functional Statement When activated, the OSC will'be the onsite area separate from the control room where predesignated cperations support personnel will assemble. A predesignated licensee official shall be responsible for coordinating and assigning the personnel to tasks designated by control room, TSC or. EOF ,
personnel.
. Recommended Requirements The OSC will be:
- 1. Located onsite to serve as an assembly point for support personnsi and to facilitate performance of support functions and tasks.
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Emergency Operations Facility (EOF) .
Functional Statement The EOF provides for The EOF is a licensee controlled and operated facility.
ennagement of overall licensee emergency response, i coord and environmental assessment, determination of and local agencies.
When the EOF is activated, it will be staffed by predesignated emergency personnel identified in the emergen'cy plan. official will mana i
Facilities shall be provided in the EOF for. the acquisition, ditions display evaluation of radiological and meteorological data and containment con l
necessary to determine protective measures. evaluate the ma from the plant and to determine dose projections.
Recommende'd Requirements .
T.he EOF will be: ,
Table 1 (previous guidance approved by th
- 1. '
appropriate radiological monitoring systems. .
2.
Sufficient to accommodate i and support Federal, State, local t and documentation in the EOF.
- predesignate'd personnel, equ pmen 3.
Structurally built in accordance with the National Uniform 4.
Environmentally controlled to provide room air temperat cleanliness appropriate for personnel and equipment. TSC 5.- Provided and control room, withand reliable reliable voice and data commun voice communication OSC and .
facilities to to NRC, State and local emergency operations centers. i Capable of reliable collection, storage, analysis, displa d
- 6. tion of information on containment conditions, radiological r determine meteorology sufficient to determine site and regional status, changes in status, forecast status and take appropriate acti .
Variables from the following categories that are essential t functions shall be available in the EOF: 1.97 variables from the appropriate Table 1 or 2 Regulatory Gu (a)
(Rev. 2), and .
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(b) the meteorological variables in Regulatory Guide 1.97 (Rev. 2) foi-site vicinity and regional data available via communication from the National Weather Service. ,
- 7. Provided with up to date plant records (drawings, schematic diagrams, etc.), procedures, emer^gency plans and environmental information (such as ~
geophysical data) needed to perform EOF functions.
- 8. Staffed in accordance with Table 2 (previous guidance approved by the Commission). Reasonable exceptions to the 30-minute and 1-hour tise limits for staffing should be justified and will be considered by NRC
.- statf.
- 9. Provided with industrial security when it is activated to exclude unauthorized personnel and when it is idle to maintain its readiness.
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- 10. Designed taking into account good human factors engineering principles. -
Documentation and NRC Review The conceptual design for emergency response facilities (TSC, OSC, and EOF) have been submitted to NRC for review. In many cases, the lack of detail in Some designs these submittals has precluded an NRC decision of acceptability.
have been disapproved because they clearly did not meet the intent of the applicable regulations. NRC does not intend to approve each design prior to implementation, but rather has provided in this document those " recommended requirements" which should be satisfied. These recommended requirements provided a degree of flexibility within which licensees can exercise management
-prerogatives in designing and building emergency response facilities (ERF)'that satisfy specific needs of each licensee. The foremost consideration regarding ERFs is that they provide adequate capabilities of licensees to respond to emergencies. NUREG guidance on ERFs has been intended to address specific issues which the Commission believes should be considered in achieving improved capabilities.
Licensees should assure that the design of ERFs satisfies these basic, requirements. Exemptions from or alternative methods of implementing these requirements should be discussed with NRC staff and in some cases could require Commission approval. Licensees should continue work on ERFs to complete them acccrding to schedules that will be negotiated on a plant-specific basis. NRC will conduct appraisals of completed facilities to verify that these requirements have been satisfied and that ERFs are capable of performing their intended functions. Licensees need not document their actions on each specific .
- item contained in NUREG-0696 or 0814.
Reference Documents (Emergency Response Facilities) 10 CFR 50.47(b) -- Requirements for emergency facilities and equipment for OLs. ,
10 CFR 50.54(q) and Appendix E, Paragraph IV.E -- Requirements for emergency facilities and equipment for ors.
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NUREG-0660 -- Description of and implementation schedule for TSC, OSC and EOF.
Eisenhut letter to power reactor licensees 9/13/79 -- Request for commitment to meet requirements. ,
Denton' letter to power reactor licensees 10/30/79 -- Clarification of requirements -
and implementation schedule.
Eisenhut letter to power reactor licensees 4/25/80 -- Clarification of requirements. ,
NUREG-0654 -- Radiological Emergency Response Plans NUREG-0696 -- Functional criteria for emergency response facilities.
NUREG-0737 -- Guidance on meteorological monitoring and dose assessment. ,
Eisenhut letter to power reactor license 2/18/81 -- Commission' approved guidance on location, habitability and staff for emergency facilities. Request and deadline for submittal of conceptual design of faci 11 ties.
NdREG-0814 (Draft Report for Comment) -- Methodology for evaluation of -
emergency r.esponse facilities. . - - -
NUREG-0818 (Draft Report for Comment) -- Emergency Action Levels ,
Reg. Guide 1.97 '(Rev. 2) -- Guidance for variables to be used in selected emergency response facilities.
l COMJA-80-37, January 21, 1981 -- Commission approval guidance on EOF location and habitability.
Secretary memorandum 581-19, February 19, 1981 -- Commission approval of NUREG-0696 as general guidance only.
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TABLE 1 -
EMERGENCY OP'ERATIONS' FACILITY Option 1 .
- Option 2 ,
One Facility Two Facilities Cics~2-in Primary: Reduce Habitability
- o At or Beyond 10 miles.
o No special protection factor.
o within 10 miles o If beyond 20 miles, specific o protection factor = 5 '
approval required by the o v2ntilation isolation Commission, and some provi-with HEPA (no charcoal) sion for NRC site team closer to site, ' '
o Strongly recommended location be coordinated with offsite authorities.
B:ckup EOF ,
u o bstween 10-20 miles
- o no separate, dedicated facility o arrtngements for portable bickup equipment .
e strongly recommended location b2 coordinated with offsite authorities o centinuity of dose projection '
and decision making capability For both Options:
- located outside security boundary 9
- - space for about 10 NRC employees
.none designated for severe phenomena, e.g., earthquakes rbittbility requirements are only for the part of the EOF in which dose assessments communications and - '
Scisienmakingtakeplace.
a utility has begun construction of a new building for an EOF that is located with 5 miles, that new yllity is acceptable (with less than protection factor of 5 and ventilation isolation and siEPA) provided tt a brckup EOF siellar to "B" in Option 1 is provided. -
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TABLE 2 ' -
l MINIMUM STAFFING REQUIREMENTS FOR NRC LICENSEES ,
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FOR N0 CLEAR POWER PLANT EMERGENCIES '
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- Capability for Additions Position Title on or Expertise Shift
- 30 min. 60 min.
Majer FunctionallArea Major Tasks Shift supervisor (SRO) 1 Pltnt Operations and 1 -- --
Ass:ssment of Shift. foreman (SRO) -- -- -
Control-room operators 2 Operational Aspects . Auxiliary operators . 2 Shift technical advisor, 1** -- --
Emergency Direction and
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shift supervisor, or Centrol (Emergency designated facility Coordinator)*** manager l
~2 es l 1 1 Notification / Nofity licensee, state "'
Communication **** local, and federal personnel & maintain I communication ,
Senior manager -- -- 1 Rrdiological Accident Emergency operations Ass:ssment and Support facility (EOF) director 1 -- l Offsite dose Senior health physics --
of Operational Accident -
(HP) expertise Assrssment assessment
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-- 2 2 .
Offsite surveys 1
-- 1 Onsite (out-of plant)- 1 1 Inplant surveys HP technicians . I
-- 1 Chemistry / radio- Rad /ches technicians 1 chemistry ,,
NOTE:
Source of this table is NUREG-0654, " Functional Criteria for Emergency Response Facil.ities." .
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TA8t.E 2 (Centinued) .
Capability for Additions .
. Position Title On
,l Maj r Functional Area Major Tasks or Expertise Shift" 30 min. 60 min.
- Plcnt System Technical support Shift technical advisory 1 --
Engineering . Repair Core / thermal hydraulics --
1 --
cnd Ccrrective Actions Electrical -- --
1 Mechanical -- --
1
'; Repair and corrective Mechanical maintenance / 1** --
1 actions Radwaste operator 1 Electrical maintenance / '
1** 1 1 - . i instrument and control 1
, (I&C) technician 1 --
i Protective Actions Radiation protection: HP technicians 2** 2 2 (In-Plant) Ci
- a. Access control
- b. HP Coverage for repair, correc-tive actions.
search and rescue first-aid, & .
firefighting '
- c. Personnel monitor-ing
- d. Dosimetry -
Firefighting -- -- Fire (ocal brigade support *'
per techni-
. cal * '
speciff-cation ;
t-R2scue Operations -- - -- 2** Local :
(nd First-Aid support I
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TABLE 2 (Continued) r
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- Capability for Additions .
Position Title on Major Tasks or Expertise Shift
- 30 min. 60 min.
Majcr Function'al Area Site Access Control Security; firefighting Security personnel All per communications, per- security and Personnel plan Acccuntability sonnel accountabil,ity ,
Total 10 11 15 "For each unaffected nuclear unit in operation, maintain at least one' shift foreman, one control-room cperator,'and one auxiliary operator except that units sharing a control room may share a shift foreman -
if all functions are covered. .
acMay be provided by shift personnel assigned other' functions.
caroverall direction of fecility response to be assumed by EOF director when all centers are fully manned. Director of minute-to-minute facility operations remains with senior manager in technica~1 support center or control room.
C28CMay be performed by engineering aide to shift supervisor.,
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STATEMENT OF. PRDFESS10t4AL QUAL'!JICAT10ftS
. CHARLES E. ROSSI I have been with the U. S. Nuclear Regulatory Comission (NRC) since -
October 1980. Since Augu'st 1981 I have been a Section Leader in .the .
Instrumentation and Control Systems Branch, Division of Systems Integration.
Office of Nuclear Reactor Regulation. I am responsible for supervising the I
review of nuclear power plant instrumentation and control system designs for compliance with regulatory criteria. From October 1980 to August 1981 I was a Principal Reactor Engineer in the Instrumentation and Control Systems Branch.
I performed the operating license review of the Callaway and Wolf Creek '
instrumentation and control system designs, the review of construction permit applicant responses to Three Mile Island Lessons Learned Items related to instrumentation and control systems, and the review of licensee responses to recommendations made by Babcock and Wilcox resulting from failure modes and effects analyses of the Integrated Control System.
I have a Ph.D degree (1969) and M.E degree (1967) in Applied Physics from Harvard University, a M.S degree (1962) in Physics from George Washington University and a B. A degree Magna cum Laude Highest Honors (1958) in Engineering and Applied Physics from Harvard University. I have a certificate from a six month reactor engineering course given by the Bettis Atomic Power Laboratory (1960). I was elected to Phi Beta Kappa in 1958 and Sigma Xi in 1962.
From June 1958 to July 1962 I served as a comissioned officer in the United States Navy. I was assigned to Naval Reac, tors, U. S. Atomic Energy 1
Commission, where I reviewed and approved test and operating procedures for ,
submarine nuclear power plant fluid systems and reactor instrumentation and control systems designs for the pressurized water reactor at Shippingport, PA.
/
Professional Qualifications Charles E. Rossi From September 1966 to November 1977 I held professional and management positions in.the Nuclear Energy Systems division of the Westinghouse _
ElectricCorgration. As a manager I supervised the preparation of system .
functional design requirements for nuclear reactor plant systems which affect plant control, protection, and transient performance. In addition to reactor control and protection systems, these systems included emergency fee &ater systems, emergency boration systems, and steam dump systems. For four years I was the lead engineer responsible for establishing functional requirements for reactor control and protection systems used in the Westinghouse 3 loop nuclear reactor plants and for perfortning transient and accident analyses of these plants for safety analysis reports submitted to the Atomic Energy Comission.
From November 1977 to October 1980 I was Systems and Civilian Applications Program Manager in the Office of Inertial Fusion at the U. S. Department of Energy. In this position, I provided technical and administrativ.e direction for studies of the comercial applications of inertial confinement fusion.
I am a member of the American Nuclear Society and past member of the IEEE Nuclear Power Engineering Comittee Standards Subcommittee (SC-6) on Safety Related Systems. I have authored or co-authored over ten technical articles for presentation at conferences or publication in journals.
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD I
IntheMitterof LONG ISL'AND LIGHTING COMPANY Docket No. 50-322 (OL)
(Shoreham Nuclear Power Station, Unit 1)
CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF REPLY TESTIMONY OF CHARLES E.
ROSSI AND JERRY L. MAUCK ON INSTRUMENTATION FOR POST-ACCIDENT MONITORING (SC 27/ SOC 3)" in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class, or, as indicated by an asterisk, through deposit in the Nuclear Regulatory Comission's internal mail system, or, as indicated by a double asterisk, by hand delivery, or, as indicated by a triple asterisk, by Express Mail, this 10th day of August, 1982:
Lawrence Brenner, Esq.** Ralph Shapiro, Esq.
Administrative Judge Camer and Shapiro Atomic Safety and Licensing Board 9 East 40th Street U.S. Nuclear Regulatory Comission New York, NY 10016 Washington, D.C. 20555 Dr. James L. Carpenter **
Administrative Judge Howard L. Blau, Esq. ,
Atomic Safety and Licensing Board 217 Newbridge Road U.S. Nuclear Regulatory Comission Hicksville, NY 11801 Washington, DC 20555
.. Dr. Peter A. Morris ** W. Taylor Reveley III, Esq.
Administrative Judge Hunton & Williams Atomic Safety and Licensing Board P.O. Box 1535 U.S. Nuclear Regulatory Conr Mico Richmond, VA 23212 Washington, DC 20555 Cherif Sedkey, Esq.
Matthew J. Kelly, Est, Kirkpatrick, Lockhart, Johnson Staff Counsel & Hutchison New York Public Service Comission 1500 Oliver Building 3 Rockefeller Plaza Pittsburgh, PA 15222 Albany, NY 12223 ,
Stephen B. Latham, Esq.
John F. Shea, III, Esq. Herbert H. Brown, Esq.**
Twomey, Latham & Shea Lawrence Coe Lanpher, Esq.
Attorneys at Law Karla J. Letsche, Esq.
P.O. Box.398 Kirkpatrick, Lockhart, Hill, ,
33 West Second Street Christopher & Phillips Riverhed,NY 11901 1900 M Street, N.W.
8th Floor Washington, D.C. 20036 Atomic Safety and Licensing Board Panel
- Docketing and Service Section*
U.S. Nuclear Regulatory Commission Office of the Secretary Washington, D.C. 20555 U.S. Nuclear Regulatory Comission Washington, D.C. 20555 Atomic Safety and Licensing Appeal Board Panel
- U.S. Nuclear Regulatory Comission Washington, DC 20555 M
Bernard M. Bordenick Counsel for NRC Staff o e
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l COURTESY COPY LIST l
Edward M. Barrett, Esq. Mr. Jeff Smith ;
General Counsel Shoreham Nuclear Power Station Long Island Lighting Company P.O. Box 618 250 Old' County Road North Country Road Mineola, NY 11501 Wading River, NY 11792 Mr. Brian McCaffrey MHB Technical Associates ***
Long Island Lighting Company 1723 Hamilton Avenue 175 East Old Country Road Suite K Hicksville, New York 11801 San Jose, CA 95125 Marc W. Goldsmith Hon. Peter Cohalan Energy Research Group, Inc. Suffolk County Executive 400-1 Totten Pond Road County Executive / Legislative Bldg Waltham, MA 02154 Veteran's Memorial Highway Hauppauge, NY 11788 David H. Gilmartin, Esq.
Suffolk County Attorney Mr. Jay Dunkleberger County Executive / Legislative Bldg. New York State Energy Office Veteran's Memorial Highway Agency Building 2 Hauppauge, NY 11788 Empire State Plaza Albany, New York 12223