ML19292A794
| ML19292A794 | |
| Person / Time | |
|---|---|
| Issue date: | 06/27/1983 |
| From: | Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Asselstine J NRC COMMISSION (OCM) |
| Shared Package | |
| ML19292A795 | List: |
| References | |
| TASK-A-01, TASK-A-1, TASK-A-11, TASK-A-30, TASK-A-43, TASK-OR NUDOCS 8307110190 | |
| Download: ML19292A794 (11) | |
Text
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JUN 2 71983 MEMORANDUM FOR:
Commissioner Asselstine FROM:
William J. Dircks Executive Director for Operations
SUBJECT:
DISTINCTION BETWEEN CURRENT AND FUTURE PLANTS IN CONSIDERING C-ENERIC REQUIREMENTS This is in response to your request of June 2, 1983, for information concerning the extent to which the staff, in considering new generic requirements, makes a distinction between the current generation of plants and future generations.
The NRC Regulatory Analysis Guidelines (NUREG/BR-0058) provide (through a checkUst in Appendix C) for differentiated consideration of application of proposed 'equirements to operating facilities and to facilities in an early construct'.an stage and in a late construction stage.
The CRGR charter (at IVB iv and v) also call, for differentiating among new plants, existing plants, and plants at various stages of the licensing process.
The staff distinguishes betweea existing and future plants in regulatory analysis of proposed new or changed requirements and also in prioritization of proposed generic safety issues.
While analyses of the merits offpotential generic safety requirements for future and existing plants are consistent in principle and method, the con-clusions are often affected by plant status.
On the basis of the analyses, the staff, in addressing generic safety issues, notably those classified as Unresolved Safety Issues (USIs), generally aims at achieving a permanent resolution by (a) establishment of guidance fdr future plants and (b) implementation of any cost-effective changes on existing plants.
The consistent analytical approach brings about such differentiated results by mechanisms such as those described below in the several phases of regalatory (and prioritization) analyses.
With some simplification, such analyses can be considered as involving three phases:
(a) a quantitative estimate of safety importance (i.e., of the potential risk reduction), (b) a quantitative estimate of the value-impact (or benefit-cost) relation, and (c) other considerations that may modify the indication from these calculations.
f so For an operating reactor the total net risk reduction that can be achieved
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may be limited because of limitations on changes that it is possible or J
lU i.f,- hj practical to make in effecting a fix, by limitations on what can be changed q
in related systems whose function may be affected, and by the reactor's lesser remaining operating life.
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The value-impact ratio for operating reactors can be substantially less favorable than for future reactors partly because of the indicated limits on safety improvement and partly because costs of implementation are likely to be higher:
they may involve hardware change costs (as contrasted with design-approach changes for future plants), the possible extra burden of work in a radiation environment, and possible need for costly plant shutdown.
Other factors, generally considered separately from the numerical results on risk reduction and value-impact ratio, include possible occupational exposure in effecting an operating-reactor change, system-interaction uncertainties, and the certainty of implementation costs as compared with possibly highly uncertain safety benefits.
Some examples of results for existing and future reactors from recent and current analyses are enclosed.
It should be noted that the examples reflect varying degrees of work completion; some of the items have not yet reached CRGR review, while some others have been released for public comment.
We trust that this information is responsive to your inquiry.
(SignesWilliam J.Dircks William J. Dircks Executive Director for Operations
Enclosure:
Distinction Between Current and Future Plants in Considering Generic Requirements:
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. The value impa ratio for operating reactors can be substa tially less onsafetyin.prov(ementandpartlybecausecostsofimple favorable +.han or future reactors partly because of the ndicated limits ntation are likely to be higher:
they may involve hardware change costs (' s contrasted with design-approach changes for future plants), the possible extra burden of work in a radiatio'q environment, and possible need f,or costly plant shutdown.
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Other factors, gener lly considered separately f. rom the numerical results on risk reduction and v que-impact ratio, include possible occupational exposure ineffectinganoperating-reactorchange, system-interactionuncertainties, and the certainty of implementation costs as gompared with possibly highly uncertain safety benefits.
Some examples of results or existing and uture reactors from recent and current analyses are enclohed.
We trust that this informat n is responsive to your inquiry.
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William J. Dircks
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Executive Director for Operations
Enclosure:
DistinctionBetweenCurrent/
and Future Plants in Considering Generic Requirements:, Examples.
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ENCLOSURE DISTINCTION BETWEEN CURRENT AND FUTURE PLANTS IN CONSIDERING GENERAL REQUIREMENTS EXAMPLES r
1.
USI A-1:
WATER HAMMER Many new plants in the early and mid-1970's were reporting water hammer occurrences, in particular steam generat*r water hammer (SGWH).
There was concern that water hammer loads could disable safety systems and break pressure boundaries.
Th'e staff reviewed water hammer occurrences in nuclear power plants (from 1969 to present), underlying causes, and corrective actions taken, and found that frequency of occurrence and related damage (which for the most part was to pipe hangers, snubbers, and equipment supports) indicated a low safety risk.
These findings revealed that:
a) approximately one-half of the occurrences were in the pre-operational phase plus first year of operation, b) about half were attributable to operation and maintenance actions and c) corrective design changes (e.g., J-tubes in PWR top-feedring steam generator design, keep-full systems and vacuum breakers in BWRs) have been implemented over the past 5-10 years.
. Based on these findings, the staff concluded that backfits were not necessary and that plants in the OL review cycle were being reviewed with respect to the water hammer issue.
The staff therefore recommended t hat appropriate sections of the Standard Review Plan (SRP) be modified to ensure that adequate design and review considerations be maintained for future applications.
Thus a forward fit implementation (for cps only) was recommended and furthermore (since operator actions were a major contributor to water hammer occurrences) that operator feedback (through TMI Task Action Plan, Part I. A.2.3) be followed up.
The impact of these proposed regulatory charges is considered minimal, and the proposed resolution has been issued "For Comment" (see SECY-83-188).
This issue has been reviewed with the CRGR and the CRGR concurs with the proposed actions.
e' 3-2.
USI A-11:
REACTOR VESSEL MATERIAL TOUGHNESS This issue concerns improved methods of addressing impairment of the toughness of reactor vessel materials, primarily weld materials, under prolonged exposure to operational radiation conditions.
With future reactors the problem is expected to be forestalled by use of materials that will remain tough enough during the vessel's entire design life of 40 years (32 years effective full power operation).
For existing reactors, the pertinent regulation (10 CFR 50, Appendix C) requires that when toughness degrades below a specified level (Charpy V-notch impact-test upper-shelf energy levels of less than 50 ft.-lb.) a safety analysis be performed to demonstrate continued presence of adequate safety margins under normal, test, upset, emergency, and faulted conditions.
The staff is proposing an improved, more rigorous method of analysis (employing advanced elastic-plastic frv 'ure mechanics concepts) that better reflects toughne:-
characteristics at operating temperatures.
Existing requirement:
in 10 CFR 50, Appendix G) call for design for thermal annealability (to restore an adequate toughness margin) if predicted nil-ductility transition reference temperature exceeds 200 F before end of life.
The USI A-11 staff proposal was sent to CRGR for information.
It was not reviewed by the Committee, on the ground that it provided an alternativt means of meeting an existing regulatory requirement rather than a change in requirements.
4-3.
USI A-43:
CONTAIkMENT EMERGENCY SUMP PERFORMANCE The major issue in USI A-43 is concern that long-term recirculation capability (which relies or the containment emergency sump as the water source) could be impaired by sump screen blockage by fibrous thermal insulation that can come loose in a LOCA.
The specific nature and extent of the problem are i,lant specific.
Resolution of the issues involves calculations for specific plants by methods, based on experimental findings, documented in revisions of the applicable Regulatory Guide (1.82) and Standard Review Plan section (6.2.2),
to show that the LOCA debris will not prevent keeping an adeqJate net positive suction head.
Based on such calculations, new plants are expected to avoid the problem by design at the outset.
Some operating plants, in the staff's opinion, called for backfits, requiring insulation replacement, which can be expensive.
These were addressed in the value-impact analysis (NUREG-0869).
Implementation costs were esti m ed to be relatively small when compared to calcu'ated reduction in potential public dose; i.e., gproximately $450/ person-rem.
The propose.1 regulatory requirements have been issued for comment (see SECY-83-149) and the staff will iaview the comments received prior to developing a final regulatory implementation position.
The proposed
5-implementation actions have been reviewed with the CRGR.
The Committee has agreed with the technical findings and proposed RG 1.12 and SRP 6.22 changes, but recommended further work on the trerits of backfitting, in lighc of a DEDRDGR staff analysis which indicated that the NRR estimates of public dose may have been based on assumptions that were too conservative.
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- 4. ISSUE A-30:
RELIABILITY OF DC POWER SYSTEMS Regulatory analysis in progress indicates that for new plants a DC power supply consisting of four independent divisions, would provide greatly enhanced reliability as compared with currently prevalent two-division systems.
Complete separation of reactor-safety loads from balance-of plant loads would also greatly improve reliability.
For new plants the cost of these improvements would be relatively small.
The resolution being developed for operating reactors involves attempts at achieving most of the new plant safety enhancement with less radical measures, such as strengthened operating requirements and administrative procedures and limited design modifications.
- 5. ISSUE 12:
BWR JET PUMP INTEGRITY Prioritization analysis indicated that the economic impact of an available complete resolution of this issue--use of improved-design hold-down beams--
increases sharply as plant status proceeds from a pre-construction stage (where new beams can be used at essentia'ily no extra cost) to advanced con-struction (when $100,000 change-out cost may be involved) and to operation (when complete change-out of the beams may involve some $3,000,000 in plant outage cost).
For opnrating reactors the prioritization analysis points to consideration of a continuing monitoring program and selective change-out of beams with incipient crack indications, while for future reactors nothing prevents use of the available improved beams.
CONTROL NO.
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