ML20040C515
| ML20040C515 | |
| Person / Time | |
|---|---|
| Issue date: | 01/08/1982 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| Shared Package | |
| ML20040C507 | List: |
| References | |
| NUDOCS 8201280033 | |
| Download: ML20040C515 (40) | |
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UNITED STATES (9
g NUCLEAR REGULATORY COMMISSION
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January 8, 1982 i
i MEMOPANDUM FOR: Victor Stello, Jr.
Deputy Executive Director for Regional Operations and Generic Requirements FROM:
Harold R. Denton, Director Office of Nuclear Reactor Regulation
SUBJECT:
EMERGENCY RESPONSE CAPABILITY AND FACILITIES This is in response to your memo on this subject dated December 29, 1981.
In general, we agree that during the nearly three years since the TMI accident there has been considerable uncertainty and confusion surrounding these facilities and requirements. We agree that it is imperative that we move promptly to begin to accomplish the long sought goals. We endorse the approach reflected in the draft document that confines the statements of requirements to the essential elements, thereby allowing. greater flexibility in implementation.
While we recognize that a number of these items have proven controversial and that it has been difficult to define the " actual requirement" in these areas, we support the approach of generally removing the staff from the traditional pre-implementation review approach.
In general, we have proposed post-implementation reviews in order to identify any inadeqyate performance by a utility.
Our criteria for NRC action as a result of post-implementation reviews would be the test for backfitting in Section 50.109 of our regulations. To help clarify our intent with respect to staff review, we have added new sections entitled, " Required Documentation" and "NRC Review." These new sections identify what is expected to be submitted for staff review, when it is to be submitted, and the nature of the' staff review.
In our review of the draft document, we have compared the requirements statements with those in the TMI Action Plan (NUREG-0660), as approved by the Commission, to ensure that no significant features have been overlooked.
i With the incorporation of our specific comments attached, we believe the i
requirements statements are consistent with previously approved policy.
CONTACT:
D. Eisenhut,- X27672 l
8201280033 820112 PDR REVCP NRCCRGR PDR l
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Victor Stello We call to your attention several items that we have changed that we believe are significant:
a.
We should prohibit the E0F from being expanded to include additional detailed technical plant data that might detract from the principal mission of the EOF.
b.
We propose that the EOF location should remain as previously approvedby)theCommission(afternumerousmeetingsand discussions.
c.
The requirements regarding meteorological data to be supplied to the CR, TSC, and EOF should be quite simple requiring only one meteorological tower with an availability of 0.9 or greater.
d.
The NDL discussion should be deleted from the package at this time.
We have removed all reference to RG 1.23 (Rev. 1) from the requirements document and recommend that this guide be revised to present a clear statement of requirements and guidance on how to meet the requirements. After issuance of the revised guide, we would perfonn only a post-implementation review with any modifications resulting from that review subject to the criteria of 150.109.
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Harold R. Denton, Director Office of Nuclear Reactor Regulation
Enclosure:
As stated
EMERGENCY' RESPONSE CAPABIt.ITY 5tudies thati followed the accident at TMI identified the need to impro,ve the on-site a,nd off-site mana'gement capability for responding to accidents.
The fundamental weakness revealed during these studies w4s the lack of We attention devoted to the " man" in the " man.:- machine equation."
must not detract from this finding. Well trained operating staff with clearly de,fNed emergency roles is't'he cornerstone to accident rAsponse.
Preplanning by utility, industry and goverrmental representatiyes is Well thought out and practiced emergency procedures both on
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n.ecessary.
site ahd o.ff, site are required.
Following the accident at THI, the President directed that the off-site responsibility for emergency response
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_.o.uld be under_the cognizance of the Federal Emergency Management Ayency.
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Significant progress has been made in'this area.
All operating plants,
were required to have fully implemented emergency plans by April 1, 1981, and our review and evaluation of these plans completed about 1 year from then.
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There Progress for improving the on-site capability has been slower.
bas been confusion w'ithin and outside the NRC regar' ding additional These.
features and equipment needed a's part of this emergency response.
include the Safety Para' meter Display System, the Tschnical Support Center, Emergen[y Operatio'ns Facility, the. 0perations Support z
Cent r, revised emergency procedures, control room reviews, the use of Regulatory Guides 1.97 and 1.23.
Consideration of these features, if done in a fragmented and uncoordin'at4d manner, can weaken accomplishing the improvements cited in the preceding paragraph. Therefore, review and ir} corporation of these facilities as aids to various personnel that
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respond to emergenciss must be accomplished and integrated into the overall emergency plan.. Flexibility of equipment and facility arrangement i's mandatory and must reinain asin, overriding,' principle to assure a timely and*fulhjoordinated emergency res;io,nse capability.
Bas'ed on a review of existing reqdirements and guidance on inul.tiple aspects of emergency response capabilities anIf facilities, the CRGR
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believes that the need exists to' articulate the' substantive (basic) e5 The information presented in requirements for power reactor licensees.
several' CRGR meetings ard discussion, underscored the need within the NRC I
1 staff for more effective management, coordination and integration of K
, ' incentatives related to emergency response.-
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Humerous comments received from nuclear industry groups,and individual NRC licensees of operating reactors reflect widespread uncertainty S
r.egarding the extent to which information and guidance publis'hed by NRC The CRGR believes that.-
are being, applied as regulatory requirements.,
the differences between require'ments and information/ guidance and how.
they are applied, need to be reiterated to the industry,and NRC staff.
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When the basic requirements for emergency response capabilities and facilities are finalized, they should be transmitted to ' licensees via a generic letter from NRR, promulgated to NRC staff and incorporated in the Standard Review Plan.
The 1,etter to licensees should request that licensees submit s proposed, schedule for completing actions to comply,
with the bas,ic requirements.
Each licensee's proposed schedules would x + / e_.
be reviewed by g NRC Project Manager, who would discuss them g MDP'-
.x with the licensee a'nd mutually agree upoh schedules and completion dates..The implementation dat'es would then be fonnalized into an enforceable e
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, j The following sectioni describe CRGR recommendations for basic' requirements.
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5 their interrelationships and NRC actions to improve management of emergency response regulation.
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Use of Existing Documentation _"-
The CRGR recommends that,NRC issue a policy' statement such as:
The following NUREG ' documents are,to be used as information only, and
' l.
the P[egulatory Guides are to be considered as guidance or a possible approach to meeting formal reciuireme~ ts. Under no circumstances should.
n the items in these documents be misconstrued as requirements to be-leiied on licensees or,as inflexible criteria to be used by HRC staff reviewers.
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NUREGs 0654 - Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants
- 0696 - Functional Criteria for Emergency Response Facilities 0700 - Guidelin.es for Control. Ro.om Design Reviews f
0799 - Draft Criteria for Preparation of Emergency Operating Procedures l
0801 - Evaluation for Control Room Design Reviews
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0814.- Methodology.for Evaluation of Emergency Response Facilities 0835 - Human Factors Acceptance Criteria for SPDS Reg. Guides 1.23 - Meteorological Measurement Program for Nuclear Power Plants b
1.97 Instrumentation for 1.ight-Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During
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and Following an Accident s
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. Coordination and Integration of Initiatives i
Recomendations 1.
The design of the SPDS, the design of Regulatory Guide 1.97 instrument-displays, the control room design review, and development of symptom oriented.emeNjency operating procedures should be integrated with respect to the overall enhancement of oper.a' tor ability to comprehend plant conditions.
Assessment of infopnation needs, and display. fomats and locations, should be performed by individual licensees. The SPDS influences other control 4-room improvements that lic'ensees'may consider and in some plants may obviat'e V
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t'he need for large scale control room modifications. However, hardware
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,?F procurement for the SPDS should not be contingent upon completion of the
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control room design review.
2.
Implementation of part or all of Regulatory Guides 1.97 and 1.23 represents a control room improvement. The implementation of these guides and any other control room improvements are not to be dependent va TSd, EOF, OSC, and NDL requirements, either in terms of content or sequence of imp 1ementation.
3.
Emergency response facilities (TSC and E0'F) are related in terms of commun-ication and instrumentation needs among the TSC, EOF and control room. TSC and EOF structures are independent of each other. The OSC is independent of TSC and EOF.
4.
The three groups of initiatives discussed above (1-SPDS, 2-control room improvements and 3-emergency response facilities) are independent of each other except for 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 influences other control room improvements that licensees may consider and to an extent could obviate the need for extensive modifications to control rooms.
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t (h) New instrumentation that may'be added to control rooms should be j
i considered for inclusion'in the design of the TSC and EOF only to the extent thaj such instrumentation is essential to TSC and
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EOF functions,.
(c[) The SPDS and control room improvem'ents are es,sential elements in operatoYtraining programs and the final plant-specific emergency
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operating procedures'.
5.
Specific implementation plans and reasonable, achievable sched.ules should be defined by mutual agreement between NRC Project Manager and each
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individual licensee.j,The,NRC,0,ffice_esponsible for implementing.gach_____ }!'
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. initiative should develop procedures identifyjg:
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f (a) The respective roles of NRR and IE Headquarters and Regions in -
checking. licensee rate of progress and verifying cbmpliance, I
, includ'ing the extent to which,NRC approval (review and inspection) i is necessary during implementatfori.
'(b)
Procedural methods and enforcement measures that could be used to
. assure NRC staff and licensee attention to meeting mutually
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' agreed upon schedules without significant del.'ays and extensions.
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I Safety Parameter Display System ('SPDS)
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Current Regulato_ry-Requirements None Functiorial Statement j
The SPDS should provide a concise display of critical plant variables to
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the control, room operators in order to aid them in rapidly a'nd reliably
. determining the. safety status of' the plant. Although the SPDS will be operated during normal operations and all classes of emergencies, the primary purpose of the display is to aid the operators in monitoring the
' safety s.tatus of the plant during anticipated trartsients and the initial
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phase of accidents.
Recommended Reouirements 1.
Each operating reactor shall be provided with a Safety Parameter
' Display System conveniently located within the control room.' This system will serve to concentrate and display a minimum quantity of information from which the plant safety status can be readily cr:
and reliably assessed by control room personnel who are responsible g
v#c for the avoidance of degraded and damaged core events. The display p7 x
shall include a full range;of important plant parameters, and data
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trend. s'which reveal whether process (,l.imits.are approached or exceeded.
The principal purpose and funNion of the SPDS will be.to aid 2.
control rocin personnel in timely detection and assessment of abnormal conditions for the reactor which must be subsequently controlled and corrected through human actions to avoid damage to the reactor
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core.
3.
a.
The SPDS need not meet requirements of the single failure criterion and it need not be qualified per Classf equirements if suitably isolated from equipment and sensors of safety systems.
b.
The SPDS may interface with equipment not yet shown to be environ-mentalry qualified, (such as, sensors, 'and analogue and d.igital g
signal processing equipment) until such equipment is qualified in accordance with the forthcoming Commission rulemaking on environmental
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qualification of safety equipment.
k'here possible, those sensors /
instruments which are now environmentally qualified should be used as sources of infomation to the SPDS.
-c.
The SPDS need not be seismically qualified. However,. the control room operator should be provided with sufficient seismically qualified m
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infomation to permit the assessment of the safety status of the plant
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and performance of an orderly shutdown following a seismic event. Tbi L[ifc.&,.
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) minimal infomati' n set might, a' the licensee's ' discretion, be signifi-o t
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cantly smaller than the information displayed by'the SPDS. The 4 '/
parameter set and equipment qualifications requirements of R. G. l.97 g} <n y
'were chosen in part to achieve this purpose with respect to indication p
available to the operator in the control room.
d.. Highly reliable information which the operator can trust with confidence should be provided. To achieve this objective, a fomal des;gn control s
F3 and review program (i.e., Verification'amj Validation program) should be utilized.
4.
The imp $rti.nt plant functions relevant to the infomation display of the SPDS to be monitored should include, but not be limited to:
o Reactivity control E*
o Reactor core cooling
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o Reactor coolant system integrity
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Containment [i_ntegrity) 4.
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o Radioactivity containment f
5.
A display, designed to human factors principles, of elemental and derived
.information,' readily perceived and comprehended by the operator should be,
provided.
Basic Reference Documents NUREG - 0600
-- Need for SPDS identified NUREG - 0737
-- Specified SPDS NUREG-0696
-- Functional criteria for SPDS
-- Specific acceptance critreia keyed to 0696 RG 1.97
-- Support document for variables to be used on SPDS Required Documentation and NRC Review NRC review will be post-implementation, that is not a pre-requisite for installation of an SPDS.
Staff reviews will be in the form of an audit.
~lo facilitate staff review, the licensee shall submit a report describing how SPDS meets the requirements stated herein.'
The submittal should include:
- 1) An overall sistem functional block diagram.
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- 2) A description of the SPDS including: parameters, display capabilities, format l
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and location in the control room and a list, for information only, of those l
components used in the SPDS whictt are or will eventually be environmentally and/or seismically qualified.
d7 3)
Should a' backup S PDS be employed, the information requested in item 2 ly
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Confirmation only that appropriate documentation of)he' e
m rification and Validation p,rogram 1s available for audit.
I Docuinehtation should include the licensee's program plan,
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functional design criteria and acceptance criteria, detailed design analysis, reliability analysis, test plan and test resul'ts, programmed licensee review points, and identif.ication by the licensee of appropriate NRC audit points.
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Thestaffreviewofthe.SPDS,R.G.1.97(Rev.2) displays',and l
control room design will be an integrated effort
- with, the licensees integrated effort. This review will consider potential equipment installation / modification' tradeoffs the licensee may elect.
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. Control" Room Design Rev'iew
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, Actual Reculatory Recuirements As specified in item I.D.1 in NUREG-0737, implementation schedule to be developed.
Functional Statement s
To reduce human error in the control roon of a nuclear plant.
Reccmmended Requirements
. 1.
Conduct a control room design review to identify human engineering
' discrepancies.
The review shall consist of:
a.
The establishment of a qualified multidisciplinary review team and a review program incorporating accepted human engineering principles.
b.
A function and task analysis to identify control rotun operator tasks and information and con' trol requirements.
- c. ' A ccmparison of the display and control requi'rements 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 audio and visual alarm sys'tems, information recording and recall capability, and lighting.
2.
For those human engineering discrepancies which could result in operator performance i
errors with potential plant safety consequences, select and implement design improvements.that will correct each of those discrepancies.
I 3.
Verify that each selected design improvement will provide the necessary correction, i
and can be introduced in the control room without creating any additional human engineering discrepancies or unreviewed safety issues.
Documentation and NRC leview The licensee shall submit a program plan as soon as available for NRC infoimation.
1.
NRC review and approval will be post-implementation for operating reactors, j
that is, not a prerequisite for the completion of modifications resulting from the CR review. For operating license applicants, the approval will be 3
prior to the license issuance.for licenses issued after 18 months from th'e date of this requirement.
2.
To facilitate staff, review, the licen,s~ee shall submit a report describing how bif'CR r.eview meets the requirements stated herein, which includes the following:
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I Describe ali design improvements that have been implemented and that are,
a.
planned to be implemented.
b.
Describe any human engineering discrepancies with potential safety con-lt sequences that will not be corrected, or which will only be partially ~correc- {
l ce EE ted. Justify each decision not to provide full correction.
Flexibility is considered in the review, because certain control board f
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a descrepancies can be overcome by techniques not involving control room changes. These techniques should include improved procedures, improved training, or the SpDS.
Basic Reference Documents t
NUREG-0585.' Statement of licensees should conduct review.
NUREG-0660 States that NRR will require for ors, OLs.
l NUREG-0737 States that requirement was issued 6/80, final guidance not yet issued.
NUREG-0700 Final guidelines for DCRDR.
NUREG-0801 Draft for comment; Staff Evaluation Criteria.
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' Regulatory Guide's 1.9
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' lication to Emergency Response Facilities -
(w pp ccg Current Reculatnrv Renoframente None Functional Statement.
Provides data to assist control room operators in preventing and. mitigating
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'the consequences of reactor accidents.
Recommended Recuirements, Control Room RG 1.97 (Rev. 2) - Provide measurements and indication of those variables listed in Types A, B, C, D. E.
Individual licensees may take exceptions based on plant specific design features. BWR incore,thermocouples and continuous.offsite dose monitors are nbt
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E required pending their further development and consideration as re-quirements.
It is acceptable to rely on currently installed eauip-ment. if it will measure over the range indicated in R.G.1.97(Rev. 2) everE if the equipment is not now environmentally qualified.
Eventually, all the equipment required to monitor the course of an accident shall
&.a a e d m m 4 llJ be environmentally qualified by a date to be specified in the
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pending Commission rule on environmental qualification. Application of the other quality standards cited in the guide such as seismic quali-g.
fication. quality assurance, power supply reliability, and redun-2 b
dancy, is required.
A schedule for confonnance may be established
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LM on a case by case basis but should.not exceed the environmental
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l f-qualification schedule.
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i Provide reliable indication of the meteorological variables specified in R.G.1.97 (Rev. 2) for site meteorology namely, wind direction, wind peed,
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1 and atmospheric stability.
The meteorol'ogical system must.have an availability of approximately 0.9 t
or greater.
TMs system should be capable of phregional data regarding
'these, parameters.
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T'echnical Support Center (TSC)_
z o RG 1.97 (Rev. 2) - Provide indication of the A, B, C, D, E variables,in o
the guide, with exceptions based on plant specific features.
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o BWR incore thermocouples and continuous offsite dose monitors are not required pending their further development and consideration at requirements.
o The indicators and associated circuitry shall be of reliable design.
but need not be provided with Standby Power sources (Class IE) and I
need not be specially qualified for earth' quakes or severe environments.
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. i Emergency Operations Facility (EOF)
RG 1.97 (Rev. 2') is acceptable guidance but NRC requires that those primary I
o indicators needed.to monitor contaiment integrity and releases of radioactivity from the plant be provided in the EOF.
o The EOF data indications and associated circuitry shall be of reliable i
design but need not. be provided with' Standby Power sources (Class 1E) and need not be specially qualified for earthquakes or severe environments.
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1 Requir'ed Documentation and NRC Review
.'[l.~~ Reg.luide1[$-
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NRC review will be post-implementation, that is, not a pre-requisite for implementation.
Staff review will be in the fom of an audit and ir}clude a review of the licensee proposed exceptions to RG 1.97j Rev.2) guidance t
with the supporting technical justifications.
f To facilitate staff review, the licensee shall submit a report' describing JA' s
l how.they meet, these requirements. The submittal should include documentation j which may be in the form of a Table, that-includes the following infomation for each Type A, B, C, D, E variable shown in R.G.1.97 I
(Rev.2):
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(a) instrument iange V
(b) environmental qualification (as stipulated in guide or state criteria)
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r (c) seismic qualification (as stipulated in guide or state criteria)
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h (d) quality assurance (as stipulated in guide or state criteria)
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(e) redundancy and sensor (s) location (s)'
j (f) power supply (e.g.,1E, non lE, battery backed)
(g) display (e.g., control room board, SPDS, Chem Lab)
(h) schedule (for installation or upgrade)
Deviations from the guidance shown in R. G.1.97 (Rev. 2) should be i
explicitly shown and supporting justification for exceptions presented.
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Installation of environmentally qualified equipment will be based on a schedule g
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Consistent with th Consnission's forthcoming rulemaking i
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Upgrade of Emergency Operating Procedures (EOPs) i Actual Regulatory Requirements 10 CFR 50.34(b) (6)(v) i I
NUREG-0737, Item I.C.1 Functional Statement To improve human relfability and the abklity to mitigate the consequences of a broad range of initiating events and subsequent multiple failures or' operator P
errors.
Recommended Reg'uirements l '.
In accordance with NUREG-0737. Item [.C,1, reanalyze transients and accidents and prepare Technical Guidelines. Submit Technical Guidelines to NFC for review.
2.
Licensees revise E0Ps to be consistent with Technical Guidelines and an t
' appropriate procedure Writer's Guide.
References NUREG-0660 Items I.C.1, I.C.8, I.C.9 NUREG-0799 Documentation Required and NRC Review A.
1.
The NRC review will be a Pre-Implementation Review for the Technical Guidelines and the Writer's Guide.
i 2.
A post-implementation review of the implementation package and E0Ps t
will be performed on an audit basis. The details and extent of the post-implementation review will be determined at a later' date based on problens identified in previous reviews and audits conducted by
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9 B. 1.
In order to facilitate NRC review the licensee shall submit the Technical Guidelines developed from the reanalysis of transiects and accidents.
Licenseesar(
encouraged to reference generic Technical Guidelines being developed by Owner's Groups. The staff will review and approve generic riter's Guides and. Technical Guidelines and advise the licensees of the acceptability.
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2.
For information only, each licensee shall submit an implementation
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package at least three nonths prior to the date they plan to begin formal operator training on the revised procedures. The implementation h
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package shall includel a.
Technical Guidelines - either plant specific guidelines ~ or gg generic guidelines with plant specific information identified l m:
as necessary.
b.
A Writer's Guide that details the, specific methods to be useS by the licensee in preparing E0Ps based on the Technical Guidelines.
c.,
A description of the program for validation of the E0Ps.
d.
A brief description of the training program for the new E0Ps.
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10 'CFR.50.'47(b).-(for.0Cs)
-ERegilirement' f.or. emergency facilities and) equip..j. j.~.
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ment.to suppo,rt emergency reyponse.
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principa).responsejorganizatiqns to emergency..
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pergonnel and to the public.-
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Requirement that. adequate. methods,.sys.tems and.
equipment for assessing and monitoring acttral or.
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potentiaT'offsite conshquences. of. 'a' radiological' T'
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.e emergency con i ion are in use..
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E 10 CFR 5b.54(9) (for ors) i.Same requi.rements as 10 CFR 50.47b p10s App
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- 10. CFR Par..t 50, ' Appendix E Requirement for:
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'Eq'uipment for determining the magnitudeof.
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t Fac.ilities an'.d supplies at the site for de-
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- contai. nation of onsite individuals;
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injure' individual..sl, or the site to spei:i c:,..,,....
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fially ideritified treatment fac,i.lities out -
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Arrangemen.ts'foi treatment of individuals
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on the site.at treatment facilities outside-the site boun'dary;
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A lice 6se'es' onsite technica1' support center and a licensee near-site.. emergency operations
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facility from which effective direction can 8
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tions system; each system shall have a backup
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ments for emergencies,. including titles and alt.ernate.s'for those in charge at both ends i
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. ' the plume' exposure pathway EPI.
Such communications shall be. tested monthly.
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Provisi.on for c.ommuni. cations. with Federal emergency response' organizations..Such -
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communications systems shall be' tested y,.; :'..
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annual 1y...
,. s Provision for communications among t e c.
, nuclear power reactor control. room, the.
a' onsite technical support center, a'nd the hear-site emer'genc'y operations facility;.,
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Provisions for commun'ications by the
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eme.rgency operations facility.'e.;' such.'.. '
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communications she,)1 be tested monthly..
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. Clarification of requirements and implementation 4 "-
Denton Lett.ey 10/30/79
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. Eisenhut Letter 4/25/80 s.
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Eisenhut Letter 2/18/81..
- . Descript~ ion of locatioC habitability and staf,f f'.....
' (previousTy" deleted from required for emergency re'sponse fa'ilities
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. ' NOREG-0737)
Requestind deitdline'for submit.tal of ' con-ceptdal design of emergency response facilities,
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When activated, the TSC.will be the onsite technica.1'. operations centet.for i
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'.and senior licensee.managesent personnel; *..
u,, prekesignated:. echnicali.,.engin'eering %.'n.,
.. redesignated persennel.
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. Once a'ctivated',' the, TSC' will ~oper, ate ' uninterrupted to perform.the following
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r.. functions until it..is deactivated:.
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' Provide plant management and technical support to p1' ant operations
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e Relievg.the reactor, operators of peripheral duties ~and communicablers not
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di,rectly 'related to reactor system maniphlation's.. -
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Perform, EOF., f' unctions for. the Alert Emergency class and. for the site Area Emefgency. lass and General Emergency cla'ss until,the EOF is fu:.
nctional.
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Provide' Technical Support to the EOF.
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Recommended 'Reo'uirements
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- 1..' Be located within the site protected area.to facilitate necessary interaction
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Be sufficient to accommodate and support NRC and licensee predesignated i
' personnel; equipment and documentation in tihe center.
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Be structurally built in acebrdance With.khe National Uniform.5#
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Building Code.
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Be en'vironmentally controlled to provide.. normal roo@ air temper'ature e
humidity and cleanliness.
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Have available radio. logical pr.otection in..accordance wit.h 10 CF.R.. * -
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.Part 20 " Standards for P'ratection Against Radiation" for p*ersonnel
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coming to, leaving from or located in the centef..
- 6.. Be capable of uninterrupted voice; data and h'ard ' copy communications with CR and. EOF.and uninterrupted voice ccmmunication with OSC and ' '
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Y-display. and' communication sufficient.to1 determine site and regional
- L
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status, determine changes in status, forecast status and take,... #
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The following variables shall be available in l
the TSC:
(a). the ' variables in the appropriate.,Talile 1 or 2,'of R'G.l.97..,.,
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Revision 21 except those variables not required; and.
l (b) the meteorological variables in'RG 1.97 for site locale and r'egion.as accurate as is indicated in RG 1.23, Revision 1.,
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Principally tifose data must be avhilable that would. enabi,e evaluating incide'nt se"quence, determining mitigaEing actions, evaluating
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damages and determining planii status duri,ng recovery operations.
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Be Have available accurate, complete and curreht piant records essential '
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Be 's.taffe'd' by predesigncted personnel u'nder the directi,cn 'of a pr'e-
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dehignated senior licensee official and be opiration:1'within approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after act.iv.ation..
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Be designed taking into account good human facto'rs engineering' principles.
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Functional Statement
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s When activated, the'OSC-will.be the.on-site ar.ea separate from the control room where prede.dignated operations', support personnel will
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ass emble.
A pre. designated. licensee official shall be responsible for
' c'oordinating and assigning the personnel to tasks designated by the CR, TSC and EOF.
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Recommended Reauirements_
- 1.. Se located on site to serve as an assembly po' int for suppor.t personnel and to facilitate performance of support, functions and tasks..
2..
Be capable of uninterrupted voice communic.ations with CR,. SC and.
T EOF,.
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The EOF:is a 1,1censee controlled and operated support cer.ter... The E0P....
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_..e Finagement'of overall. licensee emergen6 respo'nse,, ',
Coordination of radiological kn'd environmental assessment,
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Determination of recommended public protective: actions, and Coordination of emergency response activities with Federal, State, and local agencies.
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' K' hen the EOF isi activated, it. shall be staffed by" predesignated emergency-
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p Tofficial will manage licensee activities in the EOF..
. Facilities shall be provided in th.e E'0F for the acquisition, display,'
and evaluation of all radiological, mete'or'ological, and containment These facilities failure data required to determine protective measures.
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.. ill be used to evaluate the magnitude and effects of actual or po en
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Recommended Reouirements_
Located as permitted in Table 1 (as previously written and approved by the 1.-
Commission).
Be sufficient to accommodate and support Federal, State, local and 2.
licensee predesignated persdnnel, equipment and documentation in the EOF.
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- 3.. Be structurally buil,t in accordance with the National Uniform p
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Be environmentally controlled to provide normal room air temperature..
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humidity and cleanliness I
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Have, uninterrupted voice, data and hard copy cpmmunications facilities
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to the TSC and control room., and unin.terruptable voice communication,. :7
.. facil,ities to NRC,. State and local emergency operations centerf.
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The normal communication path between the EOF and the control room.
will be through the TSC.
- 6.. Be, capable of ' uninterrupted collection. storage, analysis, display and commun'ication of data addressing containment failure, radiological I
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release and meteorological and geophysical data sufficient t determine site,and regional s'tatus, determine changes in status, forecast, status and..
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l take appropriate actions. Variables-from the following categories
~
that are essen'tial to ' EOF functi,on shall be available.in the EOF:
('a) variabies from the appropriate Table 1 or 2 of RG 1.97, Revision 2, except those variables not required; and the meteorological variables' in RG 1.97 for site Docale and.
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.r (b) re,gion as accurate as is indicated in RG 1.23, Revision 1.
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Principally those. data must be available that Qould enable evaluation.,
of 1.ncident sequen'e, determinati6n of m.itigating actions, evaluation o f. daina g es.,..and determination,of plane, status during recove'ry operations.
7s Have. ready access to current
- plant records, prqcedures, and emergency
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' plans needed to perform EOF functions.
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Be staffed in accordance with Table 2'(as previously approved by t6e.
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Be prov'ided with indu,strial sicurity when it is activated to.eiclude 9.
unauthorized personnel and when it is fdle to maintain its readiness.,
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Be d.esigne'd t'aking into account' good' human factors' engineering princip'les..
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1 Basic Reference Docume6ts
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v Requirements for emergendy facilities and J.
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10 CFR 50.54(g).and Appendix E Requirements for ' emerge'ncy-facilitie's and '.
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. equipment for ors.
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NUREG-0550 Descriptiort of 'and implementation schedule '
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.. Request for commitment to meet requirements.,.,.
Eisenhut " letter 9/13/79..
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'Denton -l etter 10/30/79' Clarification of requireme.nts and implementa-
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Eisenh'ut letter 4/25/80 Clarification of requirements.
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. Functi.onal criteria dor' emergency. resp.onse
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NUREG-0737 (deleted for document)
Eisenh0t 15tter.2/18/81
. Description of location, habitability and staff required for emergency facilities l
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on of emergency.
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'Suidance:foi_. var.iables.to.be.used..in i :
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'RG 1.23 Guidan,ce.for.. meteorology..
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Required Documentation and NRC Review
. ~.
Documentation has already been provided by licensees. NRC review should be a post-implementation audit on selected facilities.
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TABLE'l
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EMERGENCY OPERATIONS FACILITY -
Option 1 Option 2 Two Facilities One Facilitt A. Close in' primary: Reduced liabitability*
e At or Beyond 10 miles a within 10 miles e No special protection factor o If beyond 20 miles, specific approval e protection factor = 5-required by the Consission, and some e ventilation isolation provision for NRC site team closer to site with IIEl A (no charcoal) e Strongly reconinend location he coordinated 0..
Dackup EOF with offsite authorities 3
e between 10-20 miles e no separate, dedicated facility e arrangements for portable backup equipment strongly reconinend location e
be coordinated with offstte authorities e continuity.of dose projection and decision making capability For both Options:
- located outside security boundary sp, ace for about 10 NRC employees none designed for severe phenomena, e.g., earthquakes liabitabliity requirements are only for that part of the E0F in which dose assessments comunications and decision making take place.
If a utility has begun construction of a new bu(1 ding for an EOF tivit is located within 5 miles, that new facility is acceptable (with lesh than protection factor of 5 and ventilation. f solation with IIEPA) provided that a backup, EOF similar to "B" in Option 1 is provided.
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TABLE 2 i
MINIMUM STAFFING RtQUIREMElTS FOR NRC. LICENSEES
~
FOR HUCLEAR POWER. PLANT: EMERGENCIES
~
Capability for Additions Position Title on Hajar Functional Area Major Tasks or Expertise Shift
- 30 min.
60 min.
Picnt Operations a'nd Shif t supervisor (SRO) 1 Assessment of Shift foreman (SRO) 1 Control-room operators 2
Tperaticnal Aspects Auxiliary operators 2
Eccrgency Direction and Shift technical advisor,
'1*
- Centrol (Emergency shif L' supervisor, or Cterdinator)***
designated. facility manager Hr,tification/
Notify. licensee, state, 1
1 2
Ccmunication**** '
local, and federal
~
personnel & maintain
~
communication 1
Radiolcgical Accident Emergency operations *,
Senior manager-Assesszent and Support facility (EOF) director of Operational Accident Offsite dose Senior health physics 1
Oss0ssment assessment (llP) expertise.
2 2,
offiste surveys 1
1 Onsite (out-of plant)
IIP technicians 1
1 1
Inplant surveys 1
Chemistry / radio-Rad / chem technicians 1
chemistry Holt:
Source of this table is HUREG-0654, "Functiong1 Criteria for Emergency Response Facilities."
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TABLE '.2
- (CONTINUED)
Capability for Additions Position Title On 90ajor Functional Area Major Tasks or Expertise Shift
- 30 min.
60 min.
Plant System Technical support Shift technical advisor 1
1 Engineering, Repair Core / thermal hydraulics 1
and Corrective Actions Electrical 1
Mechanical 1
Repair.and corrective Mechanical maintenance / '
1**
1-actions Radwaste operator Electrical maintenance /
1**
1 1
instrument and control 1
1 (I&C) technician Protective Actions Radiation protection:
IIP technicians 2**
2 2
(In-Plant) a.
Access control
~
b.
IIP Coverage for repair, corrective actions, search and rescue,.
first-aid; & firefighting c.
Personnel monitoring d.
00simetry Fire bri-Local Firefighting grade per support technical specifi-cations 2**
Local Rescue Operations.,
support End First-Ald 4
4
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TADLE 2 -
(CONfINUED)
~
~
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Capability for Additions Position Title on
!!ajer Functional Area Major Tasks or Expertise Shift
- 30 min.
50 min.
Site Access Control S'ecurity, firefighting Security personnel All per and Parsonnel communications, per-security Acccuntability sonnel accountabiiity plan Total 10 '
11 15
~
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i
^ f or each unaffected nuclear unit in operation, maintain at least one shift foreman, one control-room operator,
.and one auxiliary operator except that units sharing a contro1' room may share.a shift foreman if all functions are covered.
- 4 May be provided by shift personnel assigned other functions.
CAA Overall direction of facility 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 technical support center or control room.
l**** Hay be performed by engineering aide to shift supervisor.
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SUMMARY
OF ITEM TO BE ADDRESSED BY CRGR l
IDENTIFICATION: Equipment Qualification Program DESCRIPTION :
Program includes:
o Development of rules, guidance, and review proce-dures for environmental, seismic and dynamic l
qualification of electrical and mechanical equipment important to safety, o Development of accreditation program for test laboratories.
o Independent verfication of equipment qualification.
o Indepth review of industry testing programs.
o Development of an environmental qualification data system.
o Development of methodology, guidance, and review of sensitivity and survivability of equipment exposed f
to hydrogen-burn environments.
o Development of equipment qualification technology.
OBJECTIVE:
To provide a systematic approach for assuring that all equipment important to safety is demonstrated to i
be qualified to perform its safety functions, even.if postulated accident conditions should occur.
BACKGROUND:
Throughout the development of the commercial nuclear power industry there have been evolving changes in equipment qualification requirements and rev1ew l
procedures.
In November 1977, the Union of Concerned Scientists petitioned the Commission to upgrade current standards for the environmental qualification of equipment in operating facilities. This petition ultimately led to the Commission's Memorandum and Order of May 23, 1980 to resolve this matter in an expeditious manner. The impetus for the equipment qualification program under development is to demonstrate and document the quality of safety-related equipment in accordance with current standards.
BASIS:
Staff response to the Commission's Memorandum and Order of May 23, 1980.
SCHEDULE:
The entire program is projected to take 3 to 4 years to implement. Accomplishments to date include:
sb
.. 1 o IEEE agreement to assess and certify testing i
organization's capability to perform qualification tests on certain equipment (Oct. '81).
i o Proposed rule on " Environmental Qualification of Electrical Equipment For Nuclear Power Plants" and proposed revision to Reg. Guide 1.89 " Qualification of Class IE Equipment for Nuclear Power Plants" l
approved by Commission for publication for coment i
(Jan. '82).
i o Independent test verfication of a few selected components Other significant items pending:
o Approval of program plan by Commission.
o Laboratory accreditation rule published for comment.
o Mechanical equipment qualification rule published for comment.
j
~
o Seismic and dynamic qualificition of electrical equipment rule published for comment.
o Indepth review of industry test programs of l
selected critical components.
j CONTACT:
.Zoltan Rosztoczy, NRR ISSUES:
o Environmental qualification of electrical equipment proposed rule was approved for publication but was delayed due to the following types of issues:
(a) implementation date, (b) deferring of seismic and dynamic requirements, (c) licensee justification for continued operation, (d) conditions when " analysis only" is acceptable in lieu testing, (e) NRC notification when equipment fails, (f)' qualification of equipment necessary for " cold shut down", (g)
NRC enforcement action considerations.
o Proposed independent verification test schedule may have to be delayed due to the constraint that equipment must first be qualified by the industry.
o Original plan to have inspection teams review each qualification test phase may have to be revised due to resource constraints.
o Timely completion of value/ impact statement for seismic and dynamic qualification of electrical equipment.
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VIEWPOINTS:
IEEE - Concerned with timely development by NRC of lab accreditation rule to support their commitment and efforts.
DEDR0GR Staff - Timeliness of seismic and dynamic qualifica-tion appears to be hampered by difficulty in i
defining demonstrable requirements.
Since NRC supports the lab accreditation program and NRC review of vendor tests, consideration should be given to redirecting efforts from independent verification of equipment qualification to development of j
standardized qualification criteria and test plans.
RECOMMENDATIONS / COMMENTS TO CRGR:
The program is very diversified and involves many elements of three program offices.
It has a designated overall program coordinator and thus appears to be fairly organized. Timeliness seems to be the most significant factor and items that may delay implementation should be considered during CRGR presentation.
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.s Aucust 20, 1981 POLICY ISSUE SECY-81-504 The Commissione(rsNotation Vote)
For:
From:
William J. Dircks Executive Director for Operations
Subject:
EQUIPMENT QUALIFICATION PROGRAM PLAN
Purpose:
To inform the Commission of the staff's program plan for qualification of equipment important to safety used in nuclear power facilities and to obtain the Commission's consent to implement the program.
Discussion:
Background
In November 1977, the Union of Concerned Scientists (U3,C) petitioned the Commission to upgrade the environmental qualification of equipment in operating facilities to current standards.
This petition ultimately led to the Commission's Memorandum and Order of May 23, 1980 (CLI-80-21),
which provides guidance and directives to resolve this matter in an expeditious manner.
In July 1980, the Office of Inspection and Enforcement submitted to the Commission'(SECY-80-319) an analysis of alternatives for conducting independent verification testing of environmentally qua,lified equipment.
By a memorandum dated September 16, 1980, the Commissichers approved the recommendations of this paper, subject to prior Commission comments and directives provided in a SECY to EDO memorandum of July 18, 1980 and in the Commission meeting of July 15, 1980.
The enclosed Equipment Qualification Program Plan (EQPP) discusses how' the directives of the Commission's Memorandum and Order and the inde-pendent verification testing program are integrated into the overall program plan, which includes an env'ironmental, seismic, and dynamic qualification testing program, rulemaking activities, and research to 1
be conducted in support of the program.
Objective of the Program Plan This program plan is to provide a systematic approach to ensure that all equipment important to safety in both operating and new facilities is properly qualified to perform its safety functions if subjected to postulated accident conditions or a seismic event.
The program is expected to take about 4 years to complete and by that time to have accomplished:
S1092RO%c R
szteczy, NRR:DE:EQB h
4,*
', s 4 The Commissioners 2
A review of the qualification status of equipment important to safety in operating facilities and identification of inadequately qualified equipment. ~
Enforcement of appropriate corrective actions, including reloca-tion, replacement, or requalification of the equipment.
The development of standardized NRC review procedures for equip-inent qualification to be utilized in the review of new facilities.
The development of,a rule on the qualification of equipment important to safety and the development of regulatory guides in support of the rule.
Development of technology (analytical and experimental) in support of the equipment qualification reviews and the development of the rule on equipment qualification.
Testing and inspection of selected equipment by NRC to independ-ently verify equipment performance under accident
- conditions.
The development and implementation of an accreditation program for test laboratories.
The approximate 4-year schedule to accomplish the above objectives is based on currently projected manpower.
The Commi'ssion action on the recent " Petition for Extension of Deadline for Compliance with CLI 21" may also affect the schedule and staff resources. This effect is not expected to be significant if the staff recommendations are adopted by the Commission.
The following table lists the requirements for the 4-year program which is recommended by the staff:
FY 1981 FY 1982 FY 1983 FY 1984 PY SK PY SK PY SK PY SK NRR 21.3 3250 23.2 3800 20.0 3300 20.7 1900' IE 11.1 600 11.6 900 12.1 1100 11.5 1500 RES 4.8 2450 5.8 2850 5.5 3000 5.5 2000 Total NRC 37.2 6300 40.6 7550 37.6 7400 33.7 5400 Should the Commission decide that this program, as outlined, needs to proceed more rapidly, the staff believes that a maximum of 1 year could be cut from the program; NRC manpower needs will increase about 10 PY per year through FY 1983.
The technical assistance will cost about
$4.5 million less in the accelerated program.
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The Commissioners 3
The resources shown in the projected budgets by the Offices of Inspection and Enforcement (IE) and Research (RES) are not affected whether the program is 3 or 4 years.
The technical assistance funds required by the Office of Nuclear Reactor Regulation (NRR) for the conduct of this program in FY81 is currently available but some reprqgramming within NRR will be necessary.
The above resource requirements have been estimated through FY 1984. We expect that some followup work will be necessary beyond this period; however, it is difficult to make estimates at this time.
Recommendation :
We request that the Commission approve the proposed Equipment Qualification Program Plan.
William J. Dircks Executive Director for Operations
Enclosure:
Proposed Equipment Qualification Program Plan Commissioners' comments should be provided directly to the Office of the Secretary by c.o.b. Tuesday, September 8,1981.
Commission Staff Office comments, if any, should be submitted to the Commissioners NLT August 31, 1981, with an information copy to the
. Office of the Secretary.
If the paper is of such a nature that it requires ad'ditional time for analytical review and comment, the Commissioners j
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 Secretariat
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t CONTENTS Page ABBREVIATIONS..................................................
vii
1.0 BACKGROUND
1 2.0 OBJECTIVE OF THE PROGRAM PLAN.............................
4 3.0 EQUIPMENT QUALIFICATION REVIEW APPR0ACH...................
6 4.0 ENVIRONMENTAL QUALIFICATION REVIEWS AND IMPLEMENTATION....
8 4.1 Introduction.........................................
8 4.2 Task 0bjective.......................................
8 4.3 Task P1an............................................
8 4.4 Schedule.............................................
9 4.5 Resources............................................
11 5.0 SEISMIC AND DYNAMIC QUALIFICATION REVIEWS AND IMPLEMENTATION............................................
13 5.1 Introduction.........................................
13 5.2 Task 0bjective......................................".
14 5.3 Task Plan............................................
14 5.4 Schedu1e'.............................................
15 5.5 Resources............................................
15 6.0 EQUIPMENT QUALIFICATION STANDARDS DEVELOPMENT.............
17 6.1 Introduction.........................................
17 6.2 Task 0bjective.......................................
17 6.3 Task Plan............................................
17 6;4 Schedules............................................
18 6.5 Resources............................................
18 7.0 EQUIPMENT QUALIFICATION TEST PROGRAM REVIEWS AND IMPLEMENTATION............................................
19 7.1 Introduction.........................................
19 7.2 Task 0bjective.......................................
19 7.3 Ta s k P l a n............................................
19 7.4 Schedule.............................................
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- 7. 5 Resources............................................
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- 8. 0 SENSITIVITY AND SURVIVABILITY OF EQUIPMENT EXPOSED TO HYDROGEN-BURN ENVIRONMENTS...................................
26 8.1 Introduction............................................
26 8.2. Task 0bjective..........................................
26 8.3 Task P1an...............................................
26 8.4 Schedule................................................
26 8.5 Resources...............................................
27 9.0
SUMMARY
28 9.1 Introduction.............................................
28 9.2 Summary of Planned Accomplishments and Resource Requirements.............................................
28 9.3 Resource Requirements............i.......................
29 REFERENCES.........................................................
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4 ABBREVIATIONS AEB Accident Evaluation Branch AEC U.S. Atomic Energy Commission ASB Auxiliary Systems Branch ASME American Society of Mechanical Engineers B&W Babcock and Wilcox BWR boiling-water reactor CE Combustion Engineering CEM Chemical Engineering Branch CFR Code of Federal Regulations CSB.
Containment Systems Branch 00R Division of Operating Reactors DSI Division of Systems Integration EQB Equipment Qualification Branch FRC Franklin Research Center GDC General Design Criteria GE General Electric Company GSB Geosciences Branch HELB high-energy-line break ICSB Instrumentation and Control Systems Branch IE Office of Inspection and Enforcement
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IEEE Institute of Electrical and Electronics Engineers LER Licensee Event Report LOCA loss-of-coolant accident MEB Mechanical Engineering Branch MTEB Materials Engineering Branch 4
NRC U.S. Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation NSSS nuclear steam supply system (vendor)
NTOL near-term operating license ORAB Operating Reactors Assessment Branch ORNL Oak Ridge National Laboratory PORV power-operated relief valve PWR pressurized-water reactor QAB Quality Assurance Branch RES Office of Nuclear Regulatory Research RSB Reactor Systems Branch SD Office of Standards Development SEB 5tructural Engineering Branch SEP Systematic Evaluation Program SER Safety Evaluation Report SRP Standard Review Plan
'TMI Three Mile Island TMI-2 Three Mile Island Unit 2 W
Westinghouse Corporation vii
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EQUIPMENT QUALIFICATION PROGRAM PLAN
1.0 BACKGROUND
The need for equipment that could function under postulated accident conditions was recognized very early in the history of atomic energy.
Critical components were specified to be of high quality in accordance with the industrial standards existing at that time.
Consideration was given to potential radiation fields as well as to thermal, pressure, and moisture environments to which the equipment might be exposed.
In some cases, components were tested under research and development programs to verify that they would work under accident conditions.
To quote from the safeguards report for Pathfinder Atomic Power Plant, a facility designed more than 20 years ago:
The seal materials and designs used for electrical cable penetrations through the containment shell were tested prior to application on Pathfinder.
The tests consisted of a series of experiments which exposed full scale mockups of cable seals to steam for periods ranging from 5 to 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> and at pressures between 75 and 100 p*sig and temperatures from 310 to 325 F which are expected during the maximum design accident.
No leakage was detected during these laboratory tests.
Radiation levels at the cable penetrations under conditions due to the maximum design accident are not high enough to produce any deterioration of seal materials.
There are few records of qualification tests performed during this period on other equipment and they probably would be difficult to resurrect.
Although the assumed conditions following a postulated loss-of-coolant accident l
(LOCA) were generally the same in the early days of the atomic age as they are l
now, and although some qualification tests were performed for LOCA environments, the seismic design requirements were generally those found in then-existing building codes and, except for military applications, dynamic tests of equipment were generally not performed.
Equipment was, however, specified and designed to fail in a safe condition, should it fail because of adverse conditions.
Subsequently, both the industry standards and the U.S. Nuclear Regulatory Commission (NRC) (previously the U.S. Atomic Energy Commission (AEC)), require-l ments have become more specific and more demanding.
The current regulations for equipment qualification are embodied in the General Design Criteria (GDC) 1, 2, 4, and 23 of Appendix A and Sections III and XI of Appendix B to 10 CFR Part 50.
More detailed guidance relating to methods, procedures, and guidelines for demonstrating this capability have been set forth in various industry standards and in NRC regulatory guides.
Since the early days of AEC and NRC, the scope and depth of licensing reviews of equipment quality have broadened.
Conclusions reached in early reviews were based primarily on the reviewer's experience and judgment.
Later, the review procedure became codified in the Standard Re. iew Plan (SRP).
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In view of changes in equipment qualification requirements and review procedures, the quality of installed equipment, especially that equipment in older operating facilities, has been questioned.
This does not necessarily mean that the equip-ment is not of good quality, but rather that the quality has not been famon-strated and documented in accordance with current standards.
In November 1977, the Union of Concerned Scientists petitioned the Commission to upgrade current standards for the environmental qualification of equipment in operating facilicies.
This petition ultimately led to the Commission's Memorandum and Order of May 23, 1980 (CLI-80-21) (Ref. 1) which provides guidance and directives to resolve this matter in an expeditious manner.
Therefore, the immediate goal'of the Equipment Qualification Program is t6 assure that equipment important to safety in operating facilities is demonstrated to be qualified to function in a harsh environment such as might result from a LOCA or high energy-line break (HELB).
The figure included with this plan (Fig. 1) depicts the organization and coordination structure within NRC to accomplish the immediate objectives.
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Equipment Qual.
Prcgrmo Cecrdinaticn HRR/DE f
Z.R. Rosztoczy I
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I Environmental Qual.
Seis./Dyn. Qual.
Equipment Qual.
Equip. Qual. Test Equip. Survivability Reviews & Implem.
Reviews & Implem.
Standards Dev.
Program Coord.
in Hydrogen Env.
NRR/EQB NRR/EQB RES/DET NRR/EQB NRR/EQB P.A. DiDenedetto To Be Designated G. Arlotto Z.R. Rosztoczy H.C. Garg Electrical Equip.
Seismic Qual. of Rule for Qual. of Accreditation Development Harsh Environment Electrical Equip.
Electrical Equip.
of Test Labs.
of Methodology NRR/EQB NRR/EQB RES/EEB IE/VSPB NRR/CEB 1
P.A. DiDenedetto A.J. Lee D.F. Sullivan W.R. Rutherford C.I. Parczewski Electrical Equip.
Seismic Qual.
Rule for Qual.
Industry Test Hild Environment of Mech. Equip.
of Mech. Equip.
Program Reviews Test Program NRR/EQB NRR/EQB RES/MSEB IE/VSPB RES/FBRB H.B. Le R.G. LaGrange W.F. Anderson W.R. Rutherford D.A. Hoatson Mechanical Equip.
Rule for Lab.
Review of Test Equipment Qualification Accreditation Reports Survivability Reviews NRR/EQB RES/HFB NRR/EQB NRR/EQB H.C. Garg S.D. Richardson K. Desai H.C. Garg Equip. Qual.
Independent Data System Verification Tests NRR/EQB & FRC IE/VSPB W. Booth W.R. Rutherford Dev. of Equip.
Qual. Technology RES/MSEB & EEB J.E. Richardson D.F. Sullivan Figure 1 Nuclear Regulatory Commission coordination structure for the conduct of the overall Equipment Qualification Program
2.0 OBJECTIVE OF THE PROGRAM PLAN It is the objective of this plan to provide a systematic approach for assuring that all equipment important to safety in both operating and new facilities is properly qualified to perform its safety functions, even if postulated accident conditions should occur.
It is anticipated that this program, expected to take about 4 years to complete, will by that time have accomplished:
An expedited review of the qualification status of equipment important to safety in operating facilities and identification of unqualified or improperly qualified equipment, Enforcement' of corrective actions, including relocation, replacement, or requalification of inadequately qualified equipment, The development of standardized NRC review procedures for equipment qualifi-cation to be utilized in the review of new facilities, The development of a rule on the qualification of equipment important to safety and the development of supporting regulatory guides, Development of technology (analytical and experimental) in support of the equipment qualification reviews and the development of the rule on equipment qualification, Testing of selected equipment by NRC to independently verify equipment performance under accident conditions, The development and implementation of an accreditation program for test
- laboratories.-
Three NRC offices will be involved in the execution of the program, namely:
the Office of Nuclear Reactor Regulation (NRR), the Office of Nuclear Regulatory Research (RES) which now includes the former Office of Standards Development (SD), and the Office of Insp'ection and Enforcement (IE).
Each NRC line organization will perform its normal function so far as this j
program goes, and normal channels of communication will be utilized for routine l
operation.
Thus, overall responsibility for directing the various program l'
components resides with the appropriate NRC office.
Because of the many interfaces between the activities of the various offices, close coordination and cooperation is envisioned.
l To facilitate the implementation of this program, the Equipment Qualification Branch (EQB) in the Division of Engineering of NRR will provide overall inter-l office coordination.
EQB's routine responsibility is to evaluate tihe capability l
of systems and components important to safety to perform their design functions I
under all normal, abnormal, and accident environment conditions and in the event of seismic occurrences and other pertinent dynamic loads.
4
With coordination of the Equipment Qualification Program provided by NRR, the Office of Inspection and Enforcement will witness selected licensee tests, perform inspections of equipment at the various sites, and direct the activities associated with the accreditation of testing laboratories and independent testing of selected equipment.
The new Office of Nuclear Regulatory Research will be responsible for developing a rule and associated regulatory guides addressing NRC requirements regarding equipment qualification.
This office also will develop and execute research programs to provide pertinent technical information and support for the Equipment Qualification Program.
In addition to coordinating the overall Equipment Qualification Program, NRR will also review licensee submittals, develop an equipment qualification data bank, develop standard qualfication criteria, and perform the necessary licensing activities associated with the program.
NRR and IE will also review and monitor the equipment testing programs conducted by the industry and by testing laboratories on behalf of NRC.
This is to assure that the objectives of the Equipment Qualification Program are being met.
The NRC coordination structure for the conduct of the overall Equipment Quali-fication Program is shown in Figure 1.
As indicated on the figure, the basic equipment qualification program consists of five principal parts:
Environmental qualification reviews and implementation (review of licensee /
applicant submittals and licensing and enforcement actions, as necessary),
- Seismic and dynamic qualification reviews and implementation (review of licensee / applicant submittals and licensing ar.c enforcement actions, as necessary),
Equipment qualification standards development (development of qualification rule and associated regulatory guides),
Equipment qualification test program reviews and implementation (both industry-and NRC-sponsored tests).
Evaluation of the sensitivity and survivability of equipment exposed to hydrogen-burn environments.
These program parts are described in Sections 4.0, 5.0, 6.0, 7.0, and 8.0.
respectively.
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3.0 EQUIPMENT QUALIFICATION REVIEW APPROACH A typical nuclear power plant might have approximately 30 systems important to safety that need to function in an accident environment.
In addition, there are approximately 20 display. instruments important to safety used by the plant operators in the performance of their functions important to safety.
Each of these systems and instruments has many components that must be qualified for i
the expected environment.
It is the responsibility of the licensee / applicant to:
Identify all systems important to safety and display instruments and all components of these systems and instrume'nts, Establish expected environmental, seismic, and dynamic conditions for various parts'(zones) of the plant following postulated accidents or earthquakes, Qualify the systems and components to enviromental, seismic, and dynamic conditions corresponding to the location of the equipment, Submit to NRC the list of equipment important to safety, the expected environmental, seismic, and dynamic conditions, and a summary sheet on the j
qualification of each component type, Maintain an auditable central file of all relevant qualification data for the lifetime of the plant.
Thus, the main effort regarding equipment qualification will be taken by the licensees and applicants.
The reports submitted by the utilities wil.1 contain a large amount of detailed information as well as the licensee's or applicant's conclusions.
To review 4
this information within a reasonable period of time (given the resources
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available), the NRC staff will audit only a portion of it.
To obtain reasonable confidence that each submittal is thorough and factual, NRC will audit about i
20% of the information pertaining to specific 6quipment, equipment tests, and test ecports for that equipment which the licensee or applicant concludes is qua.: 'ied and meets NRC regulations.
In addition, NRC will evaluate the licensee's or applicant's quality assurance procedures used in his indepth review.
If, based on this limited review, NRC findings and conclusions are generally consistent with those in the submittal, no further review will be deemed necessary.
If however, the staff disagrees with the licensee or ap,,licant regarding more than 10% of the audited findings regarding equipment qualification (about 2% of the total), the submittal, i.dentifying the shortcomings found, will be returned to the licensee.
It will be the licensee's responsibility.to review and correct the entire report.
Once the corrected report is resubmitted,
'a new NRC audit will take place.
The NRC will review the licensee's or applicant's conclusions regarding all of the equipment' identified as not meeting the NRC requirements for any reasons such as lack of proper documentation or test limitations or deficiencies.
Detailed guidance and instructions for reviewers are being prepared.
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In reaching their conclusions, the licensees / applicants as well as the staff will have to make certain judgments regarding the qualification of specific equipment.
One problem anticipated is the adequacy of documentation for existing equipment in operating facilities.
Other problems or discrepancies might include deficiencies in the test program or. differences of opinion as to the safety significance of certain equipment.
Recognizing that the judgments of technical experts may vary, the staff will consider the rationale presented in the submittal and resolve outstanding issues.
When the staff has completed its review, a Safety Evaluation Report (SER) will be issued to document the NRC conclusions.
It will identify the equipment found qualified according to NRC regulations as well as the equipment that still needs to be justified, tested further, or replaced.
Licensing activities associated with the Equipment Qualification Program such as issuance of orders and SERs and implementing changes in technical specifica-tions will be handled within the NRR Division of Licensing.
NRR (EQB) will direct and monitor the establishment of an NRC equipment qualifi-cation data bank.
The programming and assimilation of the data base is being performed by Franklin Research Center (FRC).
During FY 1981, this data wil.1 be transferred to the NIH computer system, designated NEW WYLBUR, and will be available to NRC headquarters and regional offices.
The NRC staff _will also keep informed of the status of a similar data bank being developed by Electric Power Research Institute.
The purpose of these two data systems is to provide the current qualification status of equipment important to safety and references to pertinent documents related to their qualification tests.
The data banks can be utilized by licensees, applicants, and NRC staff to support conclusions in submittals regarding equipment qualification as well as the staff's conclusions reported in SERs.
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4.0 ENVIRONMENTAL QUALIFICATION REVIEW AND IMPLEMENTATION 4.1 Introduction Environmental qualification of equipment important to safety concerns both electrical and mechanical equipment and the environments associated with normal, abnormal, and accident conditions.
The normal and abnormal environ-ments are referred,to as "mila."
They are the typical variances one would expect to experience from events not associated with the breach of a high-pressure fluid boundary.
Mild environments result from an uncontrolled change in the environmental conditions because of accidents other than the LOCA or HELB and from anticipated operational occurrences.
As an example, the loss of electrical power could result in a loss of ventilating equipment and change the normal environment to a mild environment.
The " harsh" environment is a result of postulated LOCAs, HELBs, and core damage.
These conditions could affect environmental parameters significantly.
The parameters to be considered for environmental qualification (seismic and dynamic qualification discussed
, in Section 5.0) are:
Temperature Radiation Aging Chemical environment Pressure Humidity Submergence Dust By the Commission Memorandum and Order (CLI-80-21), dated May 23, 1980 (Ref. 1),
the interim criteria associated with environmental qualification of electrical equipment have been established as the DOR Guidelines (Ref. 2) and NUREG-0588
.(Ref. 3).
The establishment of criteria for mechanical equipment environmental qualification is the subject of a subtask listed below.
4.2 Task Objective The objective of this task is to (1) assess the environmental qualification of all equipment important to safety in operating facilities during normal, abnormal, and accident environments; (2) develop interim guidelines and criteria to be used to evaluate the qualification status of equipment important to safety; (3) identify unqualified or insufficiently qualified equipment; (4) enforce appropriate corrective actions, including relocation, replacement, or requalification; and (5) develop a qualification data system compiling all environmental seismic and dynamic qualification data.
4.3 Task Plan
~J Subtask 1--Assess the adequacy of environmental qualification of electrical equipment important to safety exposed to a harsh environment using the interim criteria established by the Commission Memorandum and Order (CLI-80-21), dated May 23, 1980 (Ref. 1).
Review and evaluate licensee submittals relating to 8
unresolved issues, corrective actions', and justifications for continued operation.
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Take necessary enforcement actions.'
Subtask 2--Develop review requirements and issue requests to both licensees 7
and applicants for the review of electrical equipment important to safety in A
mild environments.
The licensees and applicants are to document the extent the equipment-is qualified and propose correcti.ve actions where required.
1 Review and evaluate-licensee / applicant-response.
Review and evaluate licensee submittals relating to unresolved issues and corrective actions.
Subtask 3--Establish requirements for the environmental qualification of mecnanical equipment important to safety.
Issue request to utilities.
The licensees and applicants are to document the extent the equipment is qualified and propose corrective actions where required.
Review and evaluate licensee /
applicant response.
Take necessary enforcement actions.
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Subtask 4--Develop an environmental qualification data system with capabilities to cross-reference qualification data from plant to plant.
Input test report data for use in determining applicability. ' Input replacement program schedules for use in' determining continuous qualification' status of licensees.
Extend the data system to include pertinent seismic and dynamic qualification information.
- 4. 4 Schedule
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s The following milestones have been 'stablishr inp3eting each subtask:
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s Completion Subta_sk 1, Electrical Equipment, Expos'ed to Harsh Enviroments Date Raview licenseC s April 14 and November 1 submittals.
Is.s ue 6/1 5/81 SER for,all operating facilities.
Identify all unqualified or insufficiently qualified equipment 3/1/82 including equipment needed for, cold shutdown and new equip-
. ment installed as a result of 'the lessons learned from the
' Three Mile Island'(lhI) accident; require corrective actions; take enforcement action if submittal was insufficient.
Review licensee's proposed corrective actions, including 3/1/82 requalification program and replacement schedule.
A11 electrical equipment important to safety must conform 6/30/82 to qualification requirements (per CLI-80-21),
T Inspect conformance.with qualification requiremelits; take 6/30/83 enforcement actions.
Review, audit, and ~ evaluate" equipment qualifications Continuous information provided in support of operating licensee, applications; inspect new facilities prior to operation; issue SER 4 months after receipt of complete submittal.
Review and evaluate equipment qualification concerns related Continuous to operating facilities (Licensee Event Reports (U:Rs)).
Inspect licensee's central files'on equipmer;t q alification.
Continuous Approximately 10 facilities per' year'will be inspected.
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Completion Subtask 2, Electrical Equipment Exposed to Mild Environment Date Develop qualification criteria for mild environments; 9/1/81 issue request to licensees and applicants.
Licensees respond to request.
12/1/81 Review, audit, and evaluate licensee response; identify 3/1/82 shortcomings; require necessary corrective actions.
All electrical equipment important to safety must conform 6/30/82 to qualification requirements (per CLI-80-21).
Inspect conformance with qualification requirements; 6/30/83 take enforcement actions.
Completion Subtask 3, Mechanical Equipment Qualification
. Date Establish qualification criteria for mechanical equipment 9/1/81 important to safety to be qualified; issue request to licensees and applicants.
Licensees respond to request.
9/1/82 Review, audit, and evaluate licensee response; identify 5/1/83 shortcomings; require appropriate corrective actions.
All aechanical equipment important to safety must conform
- 5/1/84 to qualification requirements.
Inspect conformance with qualification requirements; 10/1/84 take enforcement actions.
Review, audit, and evaluate equipment qualification information Continuous prov.ided in support of operating licensee applications.
Issue SER 4 months after receipt of complete submittal.
Review and evaluate equipment qualification concerns relating Continuous to operating facilities (LERs).
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Completion Subtask 4, Equipmenc Qualification Data System Date Input information provided in November 1 submittals on 5/1/81 electrical equipment in harsh environment.
Exercise and query data system; issue user's manual.
1 0/1/81 Input information provided in February 1 submittals 12/1/81 on cold shutdown and TMI related, if any.
Input informatien provided in licensee's 12/1/81 submittals.
2/1/82
. Input information provided in licensee's 9/1/82 submittals 2/1/83 on environmental qualification of mechanical equipment and seismic and dynamic gr-lification of both mechanical and electrical equipment.
Input information provided by operating license applicants Continuous on equipment qualification (electrical and mechanical).
Input results of test report reviews Continuous 4.5 Resources The manpower and technical assistance required for the Equipment Qualification Branch to complete this program as scheduled are shown below.
Additional manpower from other NRR branches (ASB, CSB, ICSB, RSB, QAB, and AEB) will be required to assure consistency between EQB equipment qualification reviews and other NRR activities.
Assistance from IE will be required throughout the environmental quafification effort to inspect utilities and ensure that appropriate'. enforcement actions are being taken.
Additional assistance from IE, as an inspecti~n effort, will o
be required (1) to verify that replacement schedules are being maintained for the service life of_ the equipment, (2) to assure that the licensee's equipment qualification files have been set up and are maintained; and (3) to ascertain that replacement parts installed by the licensee meet regulatory requirements.
The environmental part of the equipment qualification data bank is being developed at the Franklin Research Center (FRC) on a technical assistance contract.
Similarly, part of the plant audits and reviews are being performed by the Idaho National Engineering Laboratory (INEL) and.FRC.
The NRC manpower estimates assume continuation of the technical assistance contracts as indicated.
The resources required to implement this task are shown below in terms of NRC manpower requirement (person years) and technical assistance requirement ($K):
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l Resource Requirements FY 1981 FY 1982 ^
FY 1983 FY 1984 Organization PY
$K PY
$K PY
$K PY
$K EQB 8.6 1500 7.0 1450 7.0 1250
l TOTAL 16.6 1500 11.0 1450 11.0 1250 10.5 450 O
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5.0 SEISMIC AND DYNAMIC QUALIFICATION REVIEWS AND IMPEMENTATION 5.1 Introduction The criteria and methods for the seismic qualification of mechanical and electrical equipment have changed significantly over the years.
Current licensing requirements are contained in GDC 2 and 4 and further guidance is provided in SRP Sections 3.9.2 and 3.10 and Regulatory Guide 1.100.
A number or operating reactors were licensed before the Standard Review Plans and i
Regulatory Guides were adopted.
Consequently, the margins of safety provided in existing equipment to resist seismic loads and the documentation to support the
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extent to which the equipment is qualified may vary conside: ably.
The seismic qualification of the equipment in operating plants must, therefore, be reassessed to assure that the plant can be brought to a safe shutdown condition when sub-jected to a seismic event.
Since current criteria are also subject to different interpretations, the margins in currently licensed plants may also vary and require reassessment, although not to the same extent as in older operating plants.
The seismic Category I equipment and supports in plants under construction with Ma.rk II and III containments designed by the General Electric Company (GE) are required to be designed and qualified to withstand the effects of hydrodynamic vibratory loads associated with either safety / relief valve discharge or LOCA blowdown into the pressure suppression pool in addition to the effects of dynamic loads arising from earthquakes.
This requirement, coupled with the need to cddress current licensing criteria with respect to seismic design, has resulted in a reevaluation and requalification effort on the part of boiling-water reactor (BWR) near-term operating license (NTOL) plants.
In addition to considering hydrodynamic loads in the suppression pool, other vibrations and accident-induced dynamic loads may have a noticeable effect on the functional capability of safety-related mechanical or electrical equipment.
These dynamic loads may not have been taken into consideration by the industry-in thcir present qualification program.
In the past it has been generally accepted that the seismic qualification of equipment is sufficient to cover the effects of other undefined vibratory loads that may occur during the life i
of'a plant..Information is needed to define the anticipated vibratory environ-ment in various locations of a plant during seismic and accident conditions and to determine whether such environments exceed the design-basis envelope for the installed equipment.
The criteria and methods for demonstrating the operability of active safety-related pumps and valves during transient and accident conditions have also changed over the years.
Current licensing guidance is contained in SRP Section 3.9.3 and Regulatory Guide 1.48.
Current practices vary for the methods and requirements for the design, specification, qualification, and preoperational and surveillance testing of these pumps and valves; often performance requirements are unclear, unspecified, or improperly specified.
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i A few examples are the PORV and block valves used at pressurized-water-reactor (PWR) facilities whose qualifications are now being reevaluated in view of the Three Mile Island Unit 2 (TMI-2) accident, the containment purge and vent system valves at many facilities whose design and operating requirements were in many cases not properly specified, and deep draft pumps at some facilities whose operability assurance programs are suspect because of improper quality assurance and inadequate preoperational testing.
5.2 Task Objective The objectives of this task are:
(1) to assess the qu lification of all equ'p-i ment important to safety to assure the ability of this equipment to function during and after a safe shutdown earthquake and other postulated vibrations and accident-induced dynamic loads, and (2) to assess the operability of active mechanical equipment important to safety including pumps and valves during transient and accident conditions.
This task is applicable to both operating and new facilities.
Equipment will he considered acceptable if it qualifies in accordance with the criteria contained in SRP Sections 3.9.2, 3.9.3, and 3.10 and Regulatory Gu-ide 1.100.
It is recognized that equipment in older facilities may not meet these criteria, but may still be adequate.
Therefore, an additional objective of this task is to develop more definitive crite'ria to be used in judging whether equipment needs to be requalified if found deficient with respect to the criteria con-sidered acceptable today.
5.3 Task Plan Subtask 1--Assess the adequacy of the seismic qualification of all electrical equipment important to safety in operating facilities to assure the ability of this equipment to function during and after a safe shutdown earthquake.
Con-sider normal and accident-induced vibrations and dynamic loads and combinations of the seismic and dynamic loads.
Request pertinent information from licensees.
Review, audit, and evaluate the licensee's submittals.
Take appropriate enforce-ment actions.
Subtask 2--Review the adequacy cf the seismic qualification of mechanical equip-ment important to safety in the same manner as it is described under Subtask 1.
Assess the adequacy of the operability assurance program for. mechanical equipment important to safety, including pumps and valves, to assure the ability of this equipment to function during and after transient and accident conditions.
Con-sider the combination of loading conditions for which the pump or valve is expected to function.
Evaluate the test procedure, the test conditions, and loads which are imposed on the pump or valve and the comparisons showing that
.this test and test loading are representative of those conditions and load combinations specified in the plant-design specifications.
Assess ary analytical methods used in lieu of testing.
Request pertinent information from licensees.
Review, audit, and evaluate the licensees' submittals.
Take appropriate enforce-ment action.
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5.4 Schedule The following milestones have been established for completing both subtasks:
Assess the adequacy of the seismic input level and floor response spectra defined for each operating plant. Where inadequacies are identified, additional effort outside this program will be required before the folicwing tasks can be completed.
Separate schedules will need to be defined for such plants................. 8/1/81 Develop review requirements, including criteria to be used and the equipment to be assessed, and issue request to both applicants and licensees to document the extent to which each piece of equipment is qualified.
Provide additional guidance on the interpretation of GDC 2 as it relates to the combination of accident and earth-quake 1oads....................................................... 9/1/81 Licensees and applicants respond to request.
Equipment identi-fied as not meeting current requirements should be assessed by licensees / applicants and a determination made as S whether this equipment is adequately qualified or if it must be requali-fied or replaced.
The basis for such conclusions should be documented in the response........................................ 9/1/82 Recognizing that the qualification of some equipment may be marginal but still acceptable, develop more definitive criteria to judge the positions that may be taken by the licensees /
applicants........................................................ 9/1/82 Review licensees'/ applicants' submittals and issue SER which descibes NRC evaluation of the licensees'/ applicants' findings relative to the equipment which must be requalified or replaced... 5/1/83 All equipment important to safety must conform to the seismic qual i fi cati on requi rements........................................ 5/1/84 Inspect conformance with qualification" requirements; take e n f o rceme n t act i o ns............................................... 10/1/84 Review, audit, and evaluate seismic and dynamic qualification information, and pump and valve operability data provided in support of operating license applications.
Issue SER 4 months af ter receipt of submitta1................................ Conci nuous Review and evaluate equipment qualification concerns relating to operating facilities (LERs)..................................... Continuous 5.5 Resources The manpower and technical assistance required for the Equipment Qualification Branch to complete this program as scheduled are shown below.
Additional man-power from other branches also will be required as discussed below.
Assistance from the Office of Inspection and Enforcement will be required in FY 1982, FY 1983 and FY 1984.
IE is responsible for inspecting a representative sample of the equipment to provide assurance that the equipment is installed in accordance with the design documents specified by the licensees / applicants.
15
IE will also evaluate LERs and inspect conformance with the qualification requirements specified in the staff SERs, and will take appropriate enforcement actions.
The Geosciences Branch (GSB) and the Structural Engineering Branch (SEB) will provide the assessment of input and floor spectra.
The Mechanical Engineering Branch (MEB) will (1) provida guidance on the interpretation of GDC2 as it relates to the combination of accident and earthquake loads and (2) provide guidance on the definition of the anticipated vibratory environments to be considered in the qualification of the equipment.
QAB is responsible for assessing the extent to which each licensee's equipment qualification program is in compliance with GDC1 and Appendix B to 10 CFR 50.
The resources needed to implement this task are shown below in terms of NRC manpower requirement (person years) and technical assistance requirement ($K):
Resource Requirements FY 81 FY 82 FY 83 FY 84 Organization PY
$(K)
PY
$(K)
PY
$(K)
PY
$(K)
EQB 2.5 1000 6.0 1500 6.0 1550 9.0 950 IE 3.0 3.0 3.0 Other NRR branches 1.0 2.0
- 1. 0 0.5 TOTAL 3.5 1000 11.5 1500 10.0 1550
.12.5 950 f
l l
l l
1 l
l 16
,... s.
CQUIPMENTQUAl.IFICATIONSTANDARDSDEVEl.0PMENT 6.0 6.1 Introduction The Commissioner's Mencrandum and Order of May 2I, 1980 (CLI-80-21) (Ref. 1) directed that the staff initiate a rulemaking on the subject of environmental qualification of electrical equipment important to safety.
In addition, NRR has requested that a broad single rulemaking cction be evaluated such as an amendment to 10 CFR 50 which would include:
(1) the rule-making directed ev the Commission on environmental qualification of electrical equipment, (2) addressing the qualification of both electrical and mechanical equipment for seismic and dynamic loading conditions as well as other environ-mental conditions, and (3) the rulemaking proposed in the. Commission paper from IE (SECY-80-319) (Ref. 4) on the accreditation of testing laboratories.
The proposed broad rule would address requirements for operating as well as for~
new facilities.
6.2 Task Objective The objective of this task is to develop a rule (or rules), possibly as an amendment to 10 CFR 50, to give guidance and requirements for the qualification of equipment important to safety used in nuclear power facilities.
6.3 Task Plan The Office of Nuclear Regulatory Research will initially pursue this effort as three :eparate and concurrent subtasks; it may be consolidated later in the program if circumstances warrant it.
The rationale for this approach is that subtask 1 (described below) may be completed earlier, if pursued separately, than i[ it were incorporated into a single rulemaking effort.
Subtask 1--Develop a rule Wdressing the environmental, seismic, and dynamic qualification of electrical equipment as directed by the Commission's Memorandum and Order (Ref. 1).
A number of standards and regulatory guides are available on this s,ubject and can serve as the bases for this rule.
Subtask 2'-Develop a general rule addressing all mechanical equipment important to safety.
The rule will provide only general criteria and requirements for review of applicants' submittals; guides and standards give detailed guidance.
At present, no regulatory guides are available on environmental, seismic, and dynamic qualification of mechanical equipment.
Consequently, NRR will need to perform detailed review of applicant's submittals for conformance and provide-the necessary specifics for IE.
Hence, the principal effort in support of this subtask will be the development of guides and standards for qualification of this equipment.
Subtask 3--Develop a rule addressing labcratory accreditation as approved by the Commission.
These' efforts will be coordinated with the Office of Inspection and Enforcement task on this subject in Section 7.3.
17 o
6.4 Schedule The following milestones have been established for completing each subtask:
Subtask Completion Date Suotask 1, Rule for Qualification of Electrical Equipment Rule for comment 12/1/81 Issue rule 1/1/83 Subtask 2, Rule for Qualification of Mechanical Equipment Advanced notice for rulemaking +
8/28/81 4
Rule for comment 6/1/82 Issue rule 4/1/83 Subtask 3, Rule for Laboratory Accreditation Rule for comment 12/1/82 Issue rule 9/1/83 Note:
In January 1982 and 1983 a review will be made to determine the practicality of combining the three separate rules into a single rule.
+ Assumes ANR will be issued by the EDO.
6.5 Resources The resources needed to implement this task are shown below in terms of NRC manpower requirement (person years) and technical assistance requirement (SK):
Resource Requirements Organization FY 81 FY 82 FY 83 FY S4 NRR (EQB) 1.0
- 1. 5 l
RES*
3.0 4.0 4.0 4.0 i.
IE 0.1 0.1 0.1 TOTAL 4.1 5.6 4.1 4.0 A
Development and maintenance of regulations, guides and standards will be a continuing effort through FY 84 and beyond.
I l
l 18
7.0 EQUIPMENT QUALIFICATION TEST PROGRAM REVIEW AND IMPLEMENTATION 7.1 Introduction This part of the program consists of five subtasks involving three offices of the NRC.
Coordination of these subtasks is performed by the Equipment Qualifi-cation Branch of NRR; however, direct responsibility for the conduct of the work under each subtask resides with the indicated NRC office.
Review and guidance for most of the subtasks is provided by an interoffice team.
Many of these subtasks were mandated by the Commission's Memorandum and Order of May 23, 1980 (CLI-80-21) (Ref. 1).
7.2 Task Objective NRC requires that samples of equipment that could be subjected to a harsh environment or seismic and dynanic loads be tested in an equivalent environment or the facility owners need to prove by analyses, based on existing test data, that the equipment is qualified.
The objective of this task is to (1) assure that qualification tests are conducted to provide reasonable assurance that equipment important to safety will perform its design functions, and (2) develop the technology for environmental, seismic, and dynamic qualification.
~
7.3 Task Plan Subtask 1:
Accreditation of Test Laboratories--As discussed in Section 6 of this program plan, NRC is preparing a new rule that will address in detail the subject of equipment qualification.
This rule will provide specific guidance and set requiremer.ts for meeting the Commission's present General Design Criteria for nuclear power plants.
One of the requirements will be that future equipment qualification tests be performed in a laboratory accredited for that purpose.
Accredited laboratories could be operated by equipment manufacturers, utilities, independent research and development institutes, universities, and independent testing laboratories.
NRC will work with. either the Institute of Electrical and Electronics Engineers (IEEE) or the American Society of Mechanical Engineers (ASME) to initiate an accreditation program for laboratories.
Initially, the program will review the capabilities of laboratories already conducting tests on equipment for the nuclear industry.
Its purpose is to achieve greater uniformity and consistency in the testing process regardless of the specific organization interest.
Subtask 2:
Indepth Review of Industry Test Programs--As part of its overall equipment qualification program, the NRC will inspect and review the industry test programs of selected critical components.
This will include an NRC review of equipment specifications, test plans, test procedures, and acceptance standards before the industry performs their qualification tests; In addition, the inspections will include observation of the tests, review of test results, and site inspections of equipment installations.
19
This review and inspection of the ongoing qualification tests will afford NRC the opportunity to ensure that necessary changes or adjustments are made before the work is completed.
The inspections will be performed by an inspection team consisting of members from IE, NRR, and Sandia Laboratories.
Under the allotted resources for the inspection program, approximately 10 complete equipment qualification packages will be inspected per year.
The initial inspections are scheduled to begin in December 1980.
The inspection teams will review each qualification test phase and provide comments as to conformance to the test requirements.
Subtask 3:
Review'of Test Reports--A successful c'ompletion of a qualification test culminates in a test report that fully describes the test and the test results.
A large number of such test reports have been and are being generated.
It is the intent of NRC to conduct an audit review of about 20% of these reports initially and then decide at that time whether or not the review of an additional fraction of the reports is warranted.
The review effort is divided as follows:
Nuclear steam supply system (NSSS) vendor reports (B&W,'CE, GE, and W)--
technical assistance undesignated SEP facility-related reports--technical assistance provided by FRC FRC reports--EQB staff Miscellaneous selected reports--NRC staff Reports associated with NRC audit of industry's environmental qualification program--IE/NRR team.
Subtask 4:
Independent Verification Tests--In a memorandum to the Commissioners dated July 1,1980 (SECY-80-319) (Ref. 4) and in a briefing held on July 15, 1980, IE identified the staff program for conducting independent testing and inspection of'the environmental qualifications of equipment important to safety.
Subsequent memoranda provided further specific information requested by the Commission.
In addition to its audit of industry test programs, NRC will conduct independent qualification tests of important equipment to verify the industry results.
To the extent practicable, the NRC tests will be conducted on equipment which has been in use in a nuclear power facility.
When this is not practical, specimens will be obtained from stock designated for a nuclear power facility, artificially aged, and then tested.
Contracts have been let with Sandia Laboratories and Franklin Research Center to conduct independent verification tests of selected equipment previously qualified by the industry.
Equipment for the NRC tests will be selected for a
' variety of reasons such as:
safety significance, volume use in plants, complex-ity of equipment, sensitivity of equipment, age (installed life vs. qualified life), installation concerns, degree of confidence in qualification report, and so forth.
20
.e4 To obtain naturally aged equipment and spare equipment, NRC will need the cooperation of the licensees of operating plants.
A number of licensees will be requested to provide samples at specified time intervals in order to accurately determine the effects of a natural nuclear plant environment on certain components or assemblies important to safety.
Initially, under this program, NRC will conduct four er five independent verification tests each year.
Depending upon results, the number of tests may be increased.
Subtask 5:
Development of Equipment Qualification Techno1 cay--This program consists of two parts:
(1) the development of technology for environmental qualif.ication and (2) the development of technology related to seismic and dynamic qualification.
Although much is known about these subjects, there are still questions left unanswered, particularly as to extrapolating or inter-polating from one set of conditions to another.
Regarding environmental testing, methods available for accelerated aging of various materials will be reviewed.
The role of dust'in the environmental qualification of nuclear power plant equipment will be assessed.
Studies will be made and experiments performed to compare radiation damage from commonly used radiation simulators to radiation sources that might result from an accident.
Guidelines will be developed for the preconditioning df equipment
- prior to its being tested in a harsh environment.
Qualification tests will be performed on selected equipment:
(1) to verify recommended preconditioning and testing procedures, (2) to study failure modes, and (3) to establish j
margins to failure.
Detailed examination of equipment that was subjected to the TMI-2 harsh environment will be performed to the extent practicable and conclusions published to provide information as to the survivability of a wide range of devices under accident conditions.
A program to. study seismic qualification testing criteria and methodology is-
. currently being initiated.
This program will evaluate past and present methods of seismically qualifying the operability of mechanical and electrical components and compare these with current criteria.
A procedure also will be developed to extract fragility data from existing qualification test results for use in system-risk analyses.
l Additional research efforts will include:
k An analytical program with some testing to study the effects of various
(
inputs to determine which wave forms are acceptable for simulation of earthquake excitation.
An evaluation of the influence of component aging and environmental degradation effects on the dynamic qualification of equipment, The performance of fragility tests to identify failure modes and failure levels on selected critical equipment identified by the Seismic Safety Margins Research Progrant, The development of scale modeling guidelines for the dynamic testing of equipment, An evaluation of the pump and valve operability assurance programs currently being conducted by the industry, 21
a.
A program to identify vibrations and accident-induced dynamic loads that may have a significant effect on the functional capability of mechanical and electrical equipment important to safety, An assessment of the reliability and uncertainty of dynamic qualification methods.
7.4 Schedule The following milestones have been established for completing each subtask:
Completion Subtask 1, Accreditation of Test Laboratories Date Reach agreement on accreditation program with the 6/15/81 appropriate professional organization (ASME or IEEE).
Develop accreditation criteria.
3/1/82 Start accreditation of laboratories.
9/1/83 Accredit 5 to 10 laboratories per year.
Continuou's Completion ~
Subtask 2, Review Industry Test Programs Date Obtain from the industry a list of eq'uipment 1/1/81 qualification tests planned for the next 2 years.
Select the 10 tests to be inspected in 1981.
5/1/81 l.
. Issue a tentative program for the review and 7/1/81 inspection of all 10 tests.
Perform the necessary reviews and inspections for 3/1/82 the 10 tests selected for 1981.
Issue final reports and SERs.
Will be done continuously during the year; i
the last SER will be issued by completion date.
Select the 10 tests to be inspected in 1982 and 11/1/81
' issue tentative review and inspection schedule.
Perform the necessary reviews and inspections for 2/1/83 the 10 tests selected for 1982.
Issue final reports and SERs.
Will be done continuously during the year; the last SER will be issued by completion date.
Select the'10 tests to be inspected in 1982 and 11/1/82 issue tentative review and inspection schedule.
Perform the necessary reviews and inspections for 1/1/84 the 10 tests selected for 1983.
Issue final reports and SERs.
Will be done continuously during the year; the last SER will be issued by completion date.
s 22 s
O Completion Subtask 3, Review of Test Reports Date Establish a master list of all tast reports referenced 6/1/81 by licensees in the Nov. 1, 1980 and previous equipment qualification submittals.
Indicate the review status of each report.
Complete the review of approximately 20% of the 10/1/81 referenced test reports.
Review additional test reports as required.
Continuous Select a contractor for the review of NSSS vendor qualification programs and test reports.
9/1/81 Complete the review of all four NSSS vendor
. 3/1/83 programs and test reports.
Completion Subtask 4, Independent Verification Tests Date Select the four to five tests to be conducted 2/1/81 in 1981.
Issue schedule for these tests.
Complete test #1 (sliding link terminal block);
4/15/81 issue test report.
Complete test #2 (cable shop splice test);
9/18/81 issue test report.
Complete test #3 (solenoid valves); issue
'11/I/81 test report.
Complete test #4 (electrical connectors).
9/1/81 Select the four to five tests to be conducted 11/1/81 in 1982.
Issue schedule for these tests.
Select the four to five tests to be conducted 11/1/82 in 1983.
Issue schedule for these tests.
i 1
s 23
i....
Subtask 5, Development of Equipment Qualification Technology Date Prepare a users' need memorandum on additional jnformation
~ 7/1/81 t
needed from RES with respect to the environmenth1 qualifi-of electrical equipment.
Prepare program assumptions (plans and schedule) for FY 82 8/1/81 on the evaluation of equipment qualification, including pumps and valves.
l Interim report on methodology to precondition or age equip-3/1/82 ment for DBA Qualification.
Evaluation of accelerated aging methodology.
Continuous Interim report on evaluation of simulator adequacy for 3/1/82 radiation qualification of safety related equipment.
Evaluation of radiation simulation methodology.
Continuous Conduct evaluation of equipment, including pumps and Continuous valves, removed from TMI-2.
Completed selection of contractor and award contract for the 6/1/81 evaluation of the methodology for the seismic qualification of equipment important to safety.
Issue a detailed schedule and milestones for seismic quali-7/1/81 fication methodology evaluation program.
Prepare a users' request memorandum on additional seismic-7/1/81 and dynamic qualification information needed from RES.
Initiate supplemental program to evaluate seismic and dynamic 12/1/81 qualification criteria and methodologies consistent with the resources shown in the tables to Section 7.5 below.
Issue detailed schedule for supplemental program.
2/1/82 7.5 Resources The resources needed to implement this task are shown below in terms of NRC manpower requirement (person-years) and technical assistance requirement ($K):
Resource Requirements FY 81 FY 82 FY 83 FY 84 Oroanization PY
$(K)
PY
$(K)
PY
$(K)
PY
$(K)
NRR 4.2 500 4.0 500 4.0 500 3.7 500 IE 5.0 600 5.5 900 6.0 1,100 5.5 1,500 RES*
1.5 1.900 1.5 2,600 1.5 3,000 5.5 2,000 TOTAL 10.7 3,000 11.0 4,000 11.5 4,600 10.7 4,000
- The Long Range Research Plan contemplates these research programs will continue beyond FY 84.
24
=
A breakdown of the resource requirements by subtasks is summarized below:
Subtask 1, Accreditation of Test Laboratories:
FY 81 FY 82 FY 83 FY B4 Organization PY
$(K) PY
$(K)
PY
$(K)
PY
$(K)
' O. 5
~
0.5 1.0 1.0 Subtask 2, Review Industry Test Programs:
l FY 81 FY 82 FY 83 FY 84 Organization PY
$(K) PY
$(K)
PY
$(K) PY
$(K) 1 NRR (EQB) 1.1 1.2 1.2 1.2 IE 2.0 100 2.0 150 2.0 150 2.0
- 300, Subtask 3, Review of Test Reports:
FY 81 FY 82-FY 83 FY 84 Organization PY
$(K) PY
$(K) PY
$(K) PY
$(K)
NRR (EQB) 2.4 500 2.3 500 2.3 500 2.0 500 Subtask 4, Independent Verification Tests:
FY 81 FY 82 FY 83 FY 84 Organization PY
$(K) PY
$(K)
PY
$(K)
PY
$(K)
- 2. 5 500 3.0 750 2.5 950 2.5 1200 Subtask 5, Oevelopment of Equipment Qualification Technology:
FY 81 FY 82 FY 83 FY 84 Organization PY
$(K)
PY
$(K)
PY
$(K)
PY
$(K)
NRR (EQB) 0.4 0.2 0.2 0.2 RES (Electrical) 0.5 1,500 0.5 2,000 0.5 2,000 0.5 1,000 RES (Seismic)
- 1. 0 400
- 1. 0 600 1.0 1,000 1.0 1,000 25
8.0 SENSITIVITY AND SURVIVABILITY OF EQUIPMENT EXPOSED TO HYDROGEN-BURN ENVIRONMENTS 8.1 Introduction The TMI accident identified the need for considering the control of hydrogen generated from the metal-water reaction resulting from degraded core crents.
~~
One such way of controlling the resultant environment is to intentionally burn the hydrogen as it escapes from the primary system.
However, this method may pose a threat to sensitive equipment important to safety.
8.2 Task Objective The objective of this task is to (1) survey available information and ongoing work, (2) develop a predictive technique for the environmental conditions inside the containment during the hydrogen burn, (3) develop a program for functional testing of selected equipment to determine the sensitivity to a burn environ-ment, (4) provide guidance for the determination of survivability of sensitive equipment exposed to hydrogen-burn environments, and evaluate the licensee's and applicant's submittals on equipment survivability.
8.3 Task Plan Subtask 1--Development of methodology for evaluating environmental conditions and equipment survivability in the hycrogen-burn environment.
- Subtask 2--Perform scoping tests on equipment survivability and perform experi-ments to confirm the analytical model developed in subtask 1.
Subtask 3--Provide guidance to industry review and evaluate survivability of equipment exposed to hydrogen-burn environment proposed by the industry.
8.4 Schedule The following milestones have been established for completing each subtask:
l Completion Subtask 1, Development of Methodology Date Survey information and available work 9/1/81 Develop ana!ytical model 5/1/82 Revise model based on experimental work 9/1/82 1
26 t.
i l
Cumpletion Subtask 2, Experimental Work Date Perform hydrogen burn tests 9/1/81 Perform equipment survivability tests 5/1/82 Simulated containment environmental tests 9/1/82 Completion Subtask 3, Guidance, Review, arid Evaluation Date Provide guidance to industry 10/1/82 Review and evaluate applicant's ice condenser containment 1/31/82
- 0ther plants Continu::us
- Based on the results of the analytical and biperimental work, deci-sion will be made whether a continuous review for other plants is warranted or the program should be terminated.
8.5 Resources The manpower and technical assistance required for the Equipment Qualification Branch to complete this program as scheduled are shown below.
Additional manpower from other branches also will be required as discussed below.
The Chemical' Engineering Branch (CEB) will coordinate subtask 1; this subtask is a joint effort on the part of EQB, CEB, and Sandia Laboratories.
RES and EQB will conduct subtask 2.
EQB will handle subtask 3.
EQB will request assistance from other branches, such as CSB and ICSB, when needed.
The resources needed to implement this task are shown below in terms of NRC manpower requirement (person years) and technical assistance requirement ($K):
Resource Requirements FY 81 FY 82 FY 83 Organization PY SK PY SK PY SK EQB 1.0 1.0 1.0 RES 0.3 550 0.3 250 Other NRR Branches 1.0 250 0.7 350 TOTAL 2.3 800 2.0 600
- 1. 0 t
N 27 O
9.0
SUMMARY
9.1 Introduction The Equipment Qualification Program Plan gives a systematic approach to ensure that' equipment important to safety in operating and new facilities is qualified to satisfactorily perform its safety functions if subjected to postulated acci-dents or a seismic event.
The recommended approach will take four years to complete.
It is possible to accelerate the plan to complete it in three years; this will require reprogramming of the plan and additional NRR staffing of 10 person years per year for each of the three years.
Also, the accelerated program will require industry to invest significantly more resources during the first two years to meet the faster response needed. We-have not evaluated the ability of industry to meet either of these program schedules.
The NRC technical assistance will cost about $4.5 million less in the three year program.
9.2 Summary of Planned Accomplishments and Resource Reouirements The major mi.lestones of the program and the projected dates of completion are summarized below.
Accomplishment Completien Date Recommended Accelerated
- Approach Approach Environmental Qualification of Electrigal Equipment:
Issue SER for each operating facility on the qualification of equipment exposed to harsh environment.
6/15/81 6/15/81 Issue Supplement 1 to SER to cover new, TMI-related equipment, equipment needed for cold shutdown, and other outstanding items.
3/1/82 9/1/81
~
Issue Supplement 2.to SER on licensee's proposed corrective actions and on the qualification of equipment exposed to mild environment.
3/1/82 11/1/81 Complete corrective actions for all electrical equipment important to safety to assure conformance with interim requirements.
6/30/82 6/30/82 Issue new rule on the qualification of electrical equipment and issue revised
1/1/83 1/1/83 Implement those provisions of the new rule which exceed the already enforced interim requirements.
1/1/84 1/1/84 28
Accomplishment Completion Date Recommended Accelerated" Approach Approach Seismic Qualification of All Equipment and Environmental Qualification of Mechanical Equi.pment:
Issue request to licensees.
9/1/81 6/1/81 Licensees and applicants respond to request.
d/1/82 9/1/82 Issue SER for each operating facility.
5/1/83
~10/1/82 Complete corrective actions to assure conformance with interim requirements.
5/1/84 10/1/83 Issue new rule and a new regulatory guide on the qualification of mechanical equipment.
4/1/83 4/1/83 Implement those provisions of the new rule which exceed the interim requirements.
10/1/84 1/1/84
~
Supporting Functions:
Develop and inaintain a computerized data system.
Continuous Continuous Review test reports referenced by the industry.
Continuous Continuous Review in depth, approximately 10 industry-sponsorsd qualification. tests in each year.
Continuous Continuous
~
Perform approximately five independent verification tests per year.
Continuous Continuous Conduct a research program (including qualfication tests) in support of the qualification reviews and of the development of the new rule.
Continuous Continuous Initiate accreditation of test laboratories.
9/1/83 9/1/83
" Approach for accomplishing the objectives of the. program in the shortest reasonable time.
9.3 Resource Recuirements The resources required to implement the program are shown below for each HRC office.
The tabulation shows the resources needed to complete each task, as scheduled in each of the next four fiscal years.
The accelerated plan is shown in brackets and reflects resources needed to complete the plan in three years.
29
NRR Resource Requirements FY 81 FY 82 FY 83 FY 84 Task PY
$(K)
PY
$(K)
PY
$(K)
PY 5(K) 1 10.6 [15.0] 1500 [1000] 8.0 [11.0) 1450 [1000] 8.0 [11.0] 1250
[800] 7.5 - 450 -
2 3.5 [7.0) 1000
[750] 8.0 [15.0] 1500 [1000] 7.0 [10.0] 1550 [1000] 9.5 - 950 -
3
- 1. 0 [1.5]
1.5
[1.5]
[1.5]
4 4.2 [4.9] 500
[500] 4.0
[4.7]
500
[500] 4.0 [4.7) 500
[500] 3.7 - 500 -
5 2.0 [2.0]
250
[250] 1.7
[1.7]
350
[350] 1.0 [1.0)
TOTAL 21.3 [30.4] 3250 [2500] 23.2 [33.9] 3800 [2850] 20.0 [28.2] 3300 [2300] 20.7 - 1900 -
IE Resource Requirements FY 81 FY 82 FY 83 FY 84 Task PY
$(K)
PY
$(K)
PY
$(K)
PY S(K)*
1 6.0 3.0 3.0 3.0 3.0 3.0 2
3.0 0.1 0.1 3
0.1 4
5.0 600 5.5 900 6.0 1100
- 5. 5 1500 Total 11.1 600 11.6 900 12.1 1100 11.5 1500 RES Resource Requirements FY 81
.FY 82 FY 83 FY 84 Task PY
$(K)
PY
$(K)
PY
$(K)
PY S(K) 4.0 4,o. -
4.0 3
3.0 4
1.5 1900 1.5 2600
- 1. 5 3000 1.5 2000 5
0.3 550 0.3 250 TOTAL-4.8 2450 5.8 2850 5.5 3000 5.5 2000 I
30
....e Total NRC Resource Requirements j
r Organi-FY 81 FY 82 FY 83 FY 84 zation PY
$(K)
PY
$(K)
PY
$(K)
PY
$(K)
NRR 21.3 [30.4] 3250 [2500] 23.2 [33.9) 3800 [2850] 20.0 [28.2] 3300 [2300] 20.7 1900 I
IE 11.1 600 11.6 900 12.1 1100 11.5 1500 RES 4.8 2450 5.8 2850 5.5 3000 5.5 2000 1
l TOTAL 37.2 [47.3] 6300 [2500] 40.6 [50.8] 7550 [2850] 37.6 [45.~B] 7400 [2300] 37.7 5400 f
I O
a 9
1 i
1 31
,' * * ?
1 1
REFERENCES 1.'
Memorandum and Order from the Commissioners, NRC, in the matter of Petition for Emergency and Remedial Action, CLI-80-21, May 23, 1980."
l 2.
Memorandum from H. R. Denton, NRC (NRR) to V. Stello, NRC (IE),
Subject:
Guidelines for Evaluating Qualification of Class IE Electrical Equipment in Operating Reactors, November 13, 1979.*
3.
U.S. Nuclear Regulatory Commission, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment, USNRC Report NUREG-0588, November 1979.**
4.
Memorandum from V. Stello, NRC (IE), to the Commissioners, NRC,
Subject:
Analysis of Alternatives for Conducting Independent Verification Testing of Environmentally Qualified Equipment, SECY-80-319, July 1,1980.*
"Available in NRC Public Document Room for inspection and copying for a fee.
The PDR is located at 1717 H Street, N.W., Washington,-D.C.
20555.
- Available for purchase from the NRC/GPO Sales Program, U.S. Nuclear Regulatory' Commission, Washington, D.C.
20555, and/or the National Technical Information Service, Spr.ingfield, Virginia 22161.
32
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SAMUEL J. CHILK, SECRETARY OF THE COMMISSION b )1 e
FROM:
COMMISSIONER AHEARNE
SUBJECT:
SECY-81-504-EQUIPMENTQUALIFICATIONPROGRAMPLAN
.g APPROVED DISAPPROVED ABSTAIN ~
NOT PARTICIPATING ~
REQUEST DISCUSSION d
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COMMENTS:
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A 116NAIURE.
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SECRETARIAT NOTE:
PLEASE ALSO RESPOND TO-AND/OR COMMENT ON OGC/0PE MEMORANDUM IF ONE HAS BEEN ISSUED ON THIS PAPER.
NRC-SECY FORM.DEC. 80
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Mr. Ahearne's coments on Secy-81-504:
I would like the staff to address the following issues:
1.
The bases for undertaking such an endeavor that would clearly consume a great deal of NRC and industry resources. The staff experience with some of our re-evaluation efforts could provide such a basis.
For example:
SEP findings so far for some of the older plants. (I am puzzled by the lack of reference to the SEP.
I wou~id have thought that SEP work would provide a major input for detennining where are the weak spots, e.g., in the seismic area.)
General finding by the SQRT case-by-case audit.
The major finding of the ongoing environmental qualification review program. What lessons have we learned about developing criteria, the time and effort required to develop and to review responses, etc.?
Any re-evaluations of some of the staff guides (e.g., SRP 3.9.2, 3.9.3, 10, R.G. 1.48) to indicate the inadequacies of the,
existing technical bases.
2.
Cost / safety benefit studies for:
Electric equipment qualification for mild environments.
(Do we have a set of conditions that are characterized as " mild environments"?)
Seismic qualification for electric equipment.
Mechanical equipment qualification.
3.
Scope of the re-evaluation: Have we identified what is to be included in " equipment important to safety" or developed criteria for identifying such equipment? Do we have a list of equipment / components that needs to be re-evaluated?
4.
Schedule: The schedule presented in the paper appears to be un-realistic, based on the experience with the ongoing EQ program. Are the issues so important to warrant such a schedule?
If not, can the industry support such a schedule?
The logic in the overall program plan appears to be weak.
There is a good idea imbedded here. However, before taking off on a very ambitious program, let us review the first effort, environmental qualification of electrical equipment, to see what mistakes we made.
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2 I suggest an approach should then be as follows:
(1)
Decide what equipment is important to safety.
(2)
Decidewhataretheconditions(T,p,g,etc.)theseequipment must withstand.
(3)
Decide (agree to) what tests are necessary or acceptable for verification.
(4)
Decide how long licensees should be given for verification.
(5) Then take appropriate action, including giving licensees time to make corrections.
The staff proposal can be interpreted as: proposing to have licensees decide (1) and (2), without criteria from the NRC. Then have the NRC audit. But against what standards? -- the reviewers' subjective feel?
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