ML20202G032

Submits Comments Re IAEA Draft Rept, Compilation of Generic Safety Issues for Light Water Reacter Npp

ML20202G032
Person / Time
Issue date:	01/06/1998
From:	Knapp M NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:	Stoiber C NRC OFFICE OF INTERNATIONAL PROGRAMS (OIP)
References
REF-GTECI-MI, REF-GTECI-SC NUDOCS 9802200042
Download: ML20202G032 (6)

Text

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MEMORANDIIM TO: C:rit:n R. Stoiber, Director Jariuary b.1996

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Office of Int 21 Progr ms FROM

MOch K 9 ting Director /s/ Bill M. Morris for ory Research fgegigpl

SUBJECT:

Q9 OP'MWbu'BNbA!NING UENERIC SAFETY ISSUES FOR LIGHT WATER TO REACTORlAEA DRAFT RE tluCLEAR POWF.R PLANTS" Your request for review comments by January 12,1998, did not provide adequate time for a complete review of this 373 page draft repori, consequently the " usual format" was not developed for the comments provided . However, several RES and NRR staff knowledgeable in certain safety issues have prepared comments dealing those issues and these are enclosed.

This limited review leads to the following general comments:

1, Some safety issue writeups are out of date and do not present current state-of-knowledge nor resolution positions for the USNRC. This situation is illustrated by the recommended revisions in the attached safety issue reviews.

2. NUREG 0933, *A Priontization of Generic Safety Issues," compiles USNRC safety '

issues, discusses safety significance, describes how those safety issues were resolved and provides relevant referencu. We recommend that the authors or this report utilize the current version of NUREG-0933, as a principal reference for representing USNRC safety issues and extracting USNRC positiens. We note that for some safety issues identified, NUREG-0933 is referenced but for other issues, positions or findings are drawn from other sources, and in some cases do not correctly extract positions set forth in NUREGs or other NRC references cited. The IAEA document included many, but not all genenc safety issues in NUREG-0933,

3. If this report is to be published as an lAEA report, considerable effort will be needed to bring issues into a representation of the current statt; of knowledge, integration of experiences and/or actinns unde,taken by member states and finally subjected to review by all participating parties.

We recornmend that these general comments be provided to the IAEA along with the comments in the enclosures, if we can be of further assistance, please call Al Serkiz I

(415-6563).

Erclosure: As stated Distribution: ~

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4 Comments on Safety issues Ref. IAEA Draft Report

" Compilation of Generic Safety lasues for Light Water Reactor Nuclear Power Plants"

Title:

Water hammer in the feedwater line (PC 7)

(pages 109 to 110)

General Water hammer has occurred in other NPP !ocations and safety systems than Commer't: just steam generator feedrings and feedwater piping. NUREG-0970 Rev.1, +

which is referenced in this section, identifies those systems, underlying causes, safcty significance and corrective measures to minimize water hammer occurrence. Expanding the scope of this section to water hammer occurrences in NPPc and ;ntegrating measures taken by member states into this report wotild <

benefit all participating countries.

I Para. 2, Tables 31 and 3 2, in NUREG-0927, Rev.1, identify water hammer causes Page 109 and preventive measures (both design ar.d plant operation related) for U.S BWR and PWR systems. In response to the recommendation made in this paragraph, this compilation could be used to initiate development of a consensus position for design and operational crsnsiderations for minimizing occurrence of water hammer.

Para. 3 & 4 Tables 1 1 and 12, in NUREG-0970, Rev.1, provide a relative ranking of Page 109 safety significance attached to water hammer occurrences in important systems for BWRs and PWRs. These insights identify potential weaknesses from a deterministic and operatici.at point of view. Development of a consensus position in response to the subjects in these paragraphs would benefit member states.

Page 110 This section would benefit from a summary of the USNRC's evaluation of the I

USl A 1 Water Hammer safety issue and concluding actions taken as described in NUREG-0933, paga 2.A.1-1,6/?O/85.

Title:

ECCS sump screen adequacy (SS1)

(Pages 114 - 116)

General Comment:

The issue description is too narrowly focused. The real safety issue is assurance of ECCS systems design adequacy to provide long term core cooling following LOCA debris generation and transpt *, to sump screens (PWRs) or ECCS suction strainers (BWRs). Thermal insulation debris and other materials present pnor to a LOCA must be considered in judging design

, adequacy. It is recommended that the author (s) of this section become fami!iar with

.J

t, NEA/CSNI/R (95)11,

• Knowledge Base for Emergency Core Cooling System Recirculation Reliability,' February 1996, which was prepared by an international workgroup composed of Swedish (SKl), USNRC, German (GRS), Finnish (STUK), U.S. BWROG and other U.S insulation vendors. Although this report was prepared following the i .vedish Barseback 2 event, the post LOCA ECCS design considerations are applicable to PWRs also.

Musures Token by Member States:

The measures taken by various states illustrate corrective actions that appear to have been implemented in a reactive mode. Clearly blockage is a major concem - however, these actions illustrate dealing with only portions of the assurance of long-term post LOCA cooling capability.

The actions noted for Sweden should be compared to actior.s described in NEA/CSNI/R (95)11 cited above, f

The USA actions discussion should be replaced by the following-The USNRC requires that emergency core cooling systems (ECCS) be able to provide for long-term coaling capability following a LOCA. Unresolved Sate Issue A-43,'Containtnent Emergency Sump Performance" resulted in isseance of NUREG-0897 (Ref.1) and Regulatory Guide 1.82, Rev.1 (Ref. 2) in 1985. The Barseback-2 incident in 1992, and similar blockage occurrences in U.S. BWRs in 1993, resulted in the USNRC issuing GL 96-03 (Ref. 3) which required redesign and installation of much larger ECCS suction strainers. RG 1.82, Rev. 2 (Ref. 4), which provided specific guidance for the design of BWR ECCS suction strainers, was

, issued in with GL 96-03. Installation of larger passive ECCS intake strainers in U.S. BWRs is underway and will be completed in 1991 It has also became apparent that lessons leamed from the BWR re-evaluations should be applied to a re-evaluation of PWR ECCS sump design, which the USNRC has undertaken. In addition to thermal insulation debris, other materials such as aged or degraded containment surface coating materials (e.g. paints) which can result in additional particulate materials are being reviewed to assess potentialimpact on sump debris screens.

References:

1. " Containment Ei.1ergency Sump Pedormance, Technical Findings Related to Unresolved Safety issue A-43," NUREG-0897, Rev.1, USNRC Oct.1985.

2. " Water Sources for Long-Term Recircowtion Cooling Following a Loss-of Coolant Accident," Regulatory Guide 1.82, Rev.1, USNRC Nov.1985.

3. USNRC Bulletin 96-03, " Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling Water Reactors," May 6,1996.

4. " Water Sources for Long Term Recirculation Cooling Following a Loss-of-Coolant Accident," Regulatory Guide 1.82, Rev. 2 USNRC May 1996.

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-Title: Forsign material policy (SM 6)

(Pages 334 335) and Control of temporary installations (SM 7)

(Page 336)

General Comment: These sections were referenced in the ECCS sump issue (see prige 114). USNRC Bulletin 93-02 and Bulletin 95-02, Supplement i hee page 116) are relevant to these subjects and could included.

USNRC Related Action (s):

The USNRC plans to issue Generic Letter 98 XX, Potential for Degradation of Emergency Core Cooling System and the Containment Spray System After a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in the  !

Containment." The USNRC is issuing this letter to alert addresses of potential proolenis associated with the material condition of protective coatings inside the containment and to request information under 10 CFR SC.54(f) to evaluate the addresses' programs for ensuring that protective coatings inside containment do not detach frem their substrate during a DB LOCA and interfere with the operation of ECCS and safety related CSS. In addition, U.S.

licensees are expected to have implemented adequate Foreign Materials Exclusion programs which deal with concems identified in SM 6 ard SM 7. A copy of this generic letter which was sent out for comment is enc:osed for additional information and insights.

Title:

Unreliable Insertion of control rods in PWRs and WWERs (RC 2)

(pages 35 to 38)

General Comment: Significant data has been collected since issue of the bulletin referenced in the original writeup (Bulletin 96-01) which could be included. A Bulletin supplement is being considerod. An excerpt from a letter to the Nuclear Energy Institute in response to NEl's concems follows as potential information for inclusion:

The proposed bulletin supplement does not require tests but states that testing would be one possible way to verify the ability of the control rods to insert in light of recent incomplete control rod insertion events, the invoinplete understanding of the root causes, and the rate at which such problems appear; the ability to insert the control rods fully must be demonstrated at appropriate intervals in order to meet the current licensing basis for each facility. Based on the staff's knowledge of the potential for fuel deformation es a reruit of bum up, limits were identified below which this demonstration clearly would not be needed. The proposed bulletin supplement included these burn up limits to provide licensees with one method to resolve the problem. In other words, should licensees choose to limit bumups to less than those stated, the probicm would be resolved. An evaluation of the safety impact of the options available, such as testing, redesign of the core, or rigorous engineering analysis, is the responsibility of each licensee.

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Title:

Inconnel-600 cracking (Cl 5)

(pages 61 to 64)

General Comment: The NRC issued GL 97-01 (accession number 9703260336) on this issue a%r the industry dragged their feet on resolving the issue on their own. An e :cerpt from a letter to the Nuclear Energy Institute in response to NEl's concems follows as possible information for inclusion:

Concern 1: NRC issued the generic ietter (GL) despite staff conclusions that cracking in control rod drive mechanisms and other vessel head penetrations were "not of ~

immediate safety significance" and were likely to " result in detectable leakage" that would provide ample opportunity for corrective action before a penetratica would fail.

Response: As stated in GL 97-01, "the NRC staff has conciuded that vessel head penetration cracking does not pose an immediate or near term safety concern, in the '

long-term, however, degradation of the CRDM [ control rod drive mechanism) and other VHPs [ vessel head penetration) is an important safety consideration that warrknts further evaluation."[ emphasis added] Further, as stated in the staffs response, of December 26,1996, to NEl's, public comments on the draft generic letter, " relying solely upon detection of leakage, when the Standard Technical Specifications (STS) and most facility's TSs state that no pressure boundary leakage is acceptable, is not consistent with maintaining defense-in-depth. Further, as was reported in IN [Information Notice)86-108, Supplement 1 " rapid and severe corrosion caused by boric acid leakage can occur before it is noticed." Therefore, the staff concluded that the issuance of a gener.Y letter et this time was an appropriate action because of the long-term safety significant,,.

of the issue.

Concem 2: 'We believe the proposed generic letter esscntially requires licensees to define and commit to an augrnented inspection program beyond the GL 88-05 visual inspections presently performed by licensees. The. Be_ quested Information. by implication and inference, place inordint.',. ,A essur.: on a licenseo to establish an augmented inspection plan. To not do so places the licensee in a position of justifying not responding in the preferred manner."

Response: The staff responded to the public comments as follows: The proposed generic letter is requesting information from the licensees. The generic letter does n01 require a commitment to an integrated, long-term inspection program, but rather asks what, if any, periodic inspections licensees are performing. It also asks the licensee's baser for concluding acceptability of these plans to (not) perform inspections, it should be noted that the staff interacted extensively with NEl and the industry on this issue since 1993, and the NRC suggested on numerous occasions that the industry develop an integrated program for nondestructive examinations of CRDM penetrations.

The staff also stated that it did not believe every plant would need to perform 4

V,. * *. "

Inspections but that an integrated approach to inspection planning based on relative susceptibilities of plants should be acceptable. The industry committed, through NEl, to develop such a program and indicated that necessary models had been developed and that licensees were implementing an integrated approach to inspection. However, the industry would not share the specifics of the industry models and the bases for ,

inspection planning with the NRC. Without this information, the NRC was left in a situation where it could not assess the acceptability of the industry program. Therefore, a generic letter was issued to obtain the needed information.

Title:

Diversion of recirculation water (holdups in containment)(SS 5)

(pages 124 to 125)

General Comment: The recent Architect Engineer inspection at DC Cook raised concerns about inadequate passageways for water to drain back to the containment sump. A paraphrased excerpt from a director's highlight on 'he issue follows as possible information for inclusion:

The NRR/AEOD/RES Events Assessment Panel, on November 4,1997, classified as a Significant Event for the Performance Indicator Program the multiple deficiencies found at the D.C. Cook Nuclear Power Plant, Units 1 and 2 affecting the emergency core cooling and containment heat removal systems. As a result, the safety margin of these systems to perform their intended recirculation and containment heat removal safety functions following a loss-of-coolant accident (LOCA) was significantly diminished.

During the recirculation phase following a LOCA, the lower containment spray nozzles deliver water from the containment spray system to an annulus area beneath the ice condenser. The plant construction did not provide a flow path from thie annulus area to tF e ntainment sump. Also, the plant design requires a manual switch over of the emergency core cooling system (ECCS) pump suctions from the refueling water storago tank (RWST) to the containment sump on low RWST level. As a result of flow bias

• errors and instrument uncertainties associated with the RWST level instrument, switch over could be performed before the assumed amount of water was available in the sump to support pump operation. Only by adding the volume of ice melt water to the sump inventory could vortexing be avoided. This deficiency could have created a common cause failure of the recirculation system to circulate water due to insufficient water in the sump.

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