ML20039B392

Forwards Request for Addl Info Re Hydrologic Engineering, Matls Engineering,Reactor Sys & Generic Issues,For Licensing Review.Info to Be Submitted by 820115

ML20039B392
Person / Time
Site:	Catawba
Issue date:	12/09/1981
From:	Adensam E Office of Nuclear Reactor Regulation
To:	Parker W DUKE POWER CO.
References
NUDOCS 8112220655
Download: ML20039B392 (31)

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DEC 015 k_/, Lj j[. p-Docket-Nos.t50-413/41aD DISTRIBUTION L k

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DEisenhut RMattson W*SD hh/

EAdensam RHartfield, MPA Docket Nos: 50-413 and 50-414 V.

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!!r. William O. Parker. Jr.

RVo11mer NRC PDR Vice President - Stean Production TMurley NSIC/ TIC Duke Power Company ACRS (16)

P.O. Box 3318S Charlotte. North Carolina 28242

Subject:

Request for Additional Information In the performance of the Catawba station licensing review, the staff has identified the need for additional infornation in the following areas:

1.

Hydrologic Engineering - Safety (Enclosure 1) 2.

Materials Engineering (Enclosure 2) 3.

Reactor Systens - TMI Issues (Enclosure 3) 4.

Generic Issues - Unresolved Safety Issues (Enclosure 4) contains a recent Safety Evaluation Report for Virgil C. Summer-plant, and is transmitted to identify the plant-specific information required by the Generic Issues Branch.

Our review in other areas will be completed in the near future; and we will send you separate requests for additional information related to those areas.

He request that you provide the infornation herein requested no later than January 15, 1982.

If you require any clarification of this request please contact the projact manager. Kahtan Jabbour, et (301) 492-7821.

The reporting and/or recordkeeping requirements contained in this letter affect

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fewer than ten respondents; therefore. 0!!B clearance is not required under P.L.96-511.

Sincerely.

F 8112220655 811209 Elinor G. Adensaa. Chief PDR ADOCK 05000413 Licensing Branch No. 4 A

PDR Division of Licensing

Enclosures:

As stated cc: See next 4 oe OFFKE).uu.bl : d

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N' CATAWBA Mr. William O. Parker Vice President - Steam Production Duke Power Corpany.

P.O. Box 33189 Charlotte, North Carolina 28242 cc: William L. Porter, Esq.

North Carolina Electric Membership Duke Power Company Corp.

3333 North Boulevard P.O.' Box 33189 P.O. Box 27306 Charlotte, North Carolina 28242 Raleigh, North Carolina 27611 J. Michael McGarry, III, Esq.

Debevoise & Liberman Saluda River Electric Cooperative, 1200 Seventeenth Street, N.W.

Inc.

Washington, D. C.

20036 207 Sherwood Drive Laurens, South Carolina 29360 North Carolina MPA-1 P.O. Box 95162 James W. Burch, Director Raleigh, North Carolina 27625 Nuclear Advisory Counsel 2600 Bull Street Mr. R. S. Howard Columbia, South Carolina 29201 Power Systems Division Westinghouse Electric Corp.

Mr. Peter K. VanDoorn P.O. Box 355 Route 2, Box 179N Pittsburgh, Pennsylvania 15230 York, South Carolina 29745 Mr. J. C. Plunkett, J r.

NUS Corporation 2536 Countryside Boulevard Clearwater, Florida 33515 Mr. Jesse L. Riley,. President Carolina Environmental Study Group 854 Henley Place Charlotte, North Carolina 28208 l

Richard P. Wilson, Esq..

Assistant Attorney General t

I 5.C. Attorney General's Office P.O. Box 11549 Columbia,' South Carolina 29211 Walton J. McLeod, Jr., Esq.

General Counsel South Carolina State Board of Health J. Marion Sims Building 2600 Bull Street Columbia, South Carolina 29201 H

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Hydrologic Engineering Questions Catawba fluclear Station Units 1 and 2 240.01 Please provide a description and unreduced supporting drawings for (FSAR)

(2.4.2.3) your site drainage analysis.

Identify drainage areas, conduit sizes, ponding elevation-stprage relationships, peak discharges and discharge points for all sub-basins.

Include the contributions from specific buildings in the reactor complex. Discuss the ability "of the roofs of safety-related structures, particularly the reactor building, to s'afely pond or dispose of PMP runoff assuming that drains are blocked.

Present assumptions and calculation methods in sufficient detail to allow an independent staff evaluation.

.c 240.02 Section 2.4.3.5, Table 2.4.3-8 and figure 2.4.3-6 are inconsistent (FSAR)

(2.4.3.5) with respect to the maximum PMF level.

Please correct accordingly.

240.03 Figure 2.4.1.1 shows an embankment has been constructed behind the l

(FSAR)

LPSW intake structure.

Failure of this. embankment could caure flooding of the power block area. Please provide details of the construction of this emba~nkment and include a plan view and typical cross-section. Also, discuss whether failure of this embankment will l

affect safety-related structures and compone.its and if so how.

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240.04 It is'not clear whether operation of the flood gates on the Wylie (FSAR)

(2.4.3.4)

Dam spillway, during a PMF, is necessary to prevent waves and high water from flooding the power block area or adversely affecting the SNSW dam cr the embankment behind the LPSW intake structure.

Please discuss operator influences on flood levels and what the impacts of faulty gate operation will be.

240.05 One of the design basis. floods that should be concidered is a PMF (FSAR)

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centered critically over the SNSW tributary.

It is not clear that this was considered in your analysis.

Provide the PMF inflow, outflow, and water surface elevation hydrographs for the SNSWP

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considering a PMF on the SNSW tributary coincident with the,[.MF residual flow in Lake Wylie.

If this.results in a water level greater than the one you used for design of the SNWS dam,d'iscuss the wahe forces and static forces on the dam and on the intake and discharge structures.

If these forces are higher than used for' design, compare them with the design limits for the rip-rap and the intake and dis-charge structures. Also,theArea-Volumecurhes, Figure 9.2.5-2 of'theFSAR,extendonlytoelchation571 feet. Please extend these curves.toatleastelehation591 fee'torthePMFlehel,whicheheris greater.

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8e 2f0.06 Please discuss the lateral extent of the rip-rap on the SNSW dam, (FSAR) 1.e'.,doesitextendpasttherightabutmenttothehicinityofthe NSW and SNSW intake structures.

Ifitdoes,towhatelehations does the rip-rap extend.

Ifthereisnorip-rapinthehicinity of these structures, how will erosion affect the st.ructures.

240.07 (FSAR)

(2.4.13.2.4) FSAR Figure 2.4.13-4 is a potentiometric map of the Catawba site for the period 1971-1974. Please prohide a similar map showing the present potentiometric surface, and a map showing the surface once the plant begins operatio,n if it'is expected that water lehels will change between now and then.

,e 240.08 In calculating radius of influence (L) for use in determining the design (FSAR) 2.4.13.5) flow rate for the ground water drainage system, you used an infiltration rate (E) of 0.01 gallons per day per square foot (gpd/sq.ft.). This value was obtained by rounding off the computed infiltration rate of 0.0724 gpd/sq.ft.. HadyounotroundedoffthecomputedYalueof"E"andused 0.0724 instead of 0.10, the calculated value of "L" would hahe been larger. This larger "L" will result in a different flow rate (Q) than wh.at you used for design. Please recompute"Q" using the larger"L" and assure that the system is adequate to maintain groundwater levels at the desired elevation. Also,onpage2.4-28,youstatethatthehalueof l

"E" used in your analysis is 0.01 gpd/sq.ft.

It appears that this is l

a typographical error because the actual value used is 0.1 gpd/sq.ft.

Please correct this. error and assure that 0.1 gpd/sq.ft, was used in all l

your analyses and computations.

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(Q) is frca M. E. Harr, " Groundwater and Seepage". We note however, I

that the formula on page 43.44 of the refererce is not the same as the fomula presented on page 2.4.27 of the FSAR. Pleaseprohidethe appropriate formula and the calculations perfomed to dehelop the projected inflow into the underd' rain system.

Inaddition,pleaseprovidethehalue

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for the plant perimeter used in your calculations.

240.10 On page 2.4-26 in the third paragraph you state that a pemanent (FSAR) 2.4.13.5 Category I groundwater draina.ge system is inst'alled "... to create and permanently maintain a nomal hroundwater lehel at or near the base of the foundation mat and basement walls, thus eliminating the

,e uplift and hydrostatic forces." 1.ater in the same paragraph you state that lowareasnotreliehedofthisgrcundwaterpressurearedesigned

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to withstand the resultant upiift and hydrostatic loads." These statements implythatintheareaswherethegroundwaterlehelsar.eloweredbythe drainage system, the resultant reduced loads were used as design con-ditions.

Contrary to this, on page 2.4-30, Reh.1, in the third paragraph, you state that the Reactor and Auxiliary Buildings can with-stand the hydrostatic and uplift forces resulting from a groundwater -

7 rebound to yard grade. Please explain and. rectify this apparent con-tradiction.

Compare the design and anticipated operation of your groun~d-water drainage system with Branch Technical Position HGEB-1., Reh. 2 NUREG 0800, Section 2.4.12..

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240.11 Describe the number, location ~and intended disposition of uncapped (FSAR) borings and obserhation wells.

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240.12 Seepage tests were run in the SNSWP dam following construction.

(FSAR) 2.5.6.8)

Flow rates through the-dam were measured at 68 to 76 gallons per minute (gpm). Of this, only 20-28 gpm is stated to be.. seepage and

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the remainder is stated to be groundwater flow. Please. provide the data and analysis used to justify'this breakdown in flow components.

Include a figure showing the flownet developed in your analysis and describe the procedures and references used-to devel'op this flownet.

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7 MATERIALS ENGINEERING BRANCH REQUEST. FOR ADDITI0t!AL.INFORMATION.

CATAWBA NUCLEAR STATION, UNITS 1 AND 2 s

121.1 General Design Criterion 32 requires that the reactor coolant pressure

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boundary shall be designed to permit periodic inspection and testing of important areas and features to assess the'ir structural. and leaktight integrity.

We have recently identified a problem concerriing the effectiveness of ultrasonic examination techniques to examine the primary piping system.

Certain ultrasonic techniques may not be adequate t.o consistently. detect and reliably characterize service-induced flaws during the inservice inspection of thick-wall cast stainless steel components to acceptance standards of Paragraph'IWB-3500.of Section XI.

Discuss the technical basis for determination that your preservice ultra-sonic examination is capable of detecting and characterizing crack-like indications in the reactor coolant boundary piping.

121.2 In Section 3 of your PSI Plan, you-state that mainsteam and feedwater system welds will be examined by radiography.

Standard radiography may not be capable of detecting and characterizing' service-induced degradation because the detection of cracks normally requires the central axis of the p

radiation beam to be within 10' of the predomina"te crack plane. Provide the technical justification, including qualification and special procedure requirements, to. demonstrate that radiography i.s an effective and applicable volumetric examination technique.

121.3 Your PSI Plan does not include the entire Section 9, Relief Requests.

Indicate the anticipated date for submittal of this Section.

To evaluate your compliance with 10 CFR Part 50.55a(g)(2), we will require that all Class 1 and 2 pressure retaining welds that cannot be examined as required by Section XI of the ASME Code be identified with a supporting technical justification.

i A.

Where relief is requested for pressure retaining welds in the reactor vessel and steam generator shell welds, identify the specific welds that did not receive a 100% pre. service ultrasonic examination and estimate the extent of the examination'that was performed.

B.

Where relief is requested for piping system welds (Examination Category B-J, C-F, and C-G), provide a: list of the specific welds that did not receive a complete Section XI preservice examination including a drawing or isometric identification number, system, weld number, and physical configuration, e.g., pipe to nozzle weld, etc.

Estimate the extent of the preservice examination that was performed. When the volumetric examination was performed from one side of the weld, ~ discuss whether the entire weld volume and heat affected zone (HAZ) and base metal on the far side of the weld were examined.

State the primary reason that m.

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7 a specific examination is impractical, e.g., support or component restricts access, fitting prevents adequate ultrasonic coupling on one side, component to component weld prevents ultrasonic examina-tion, etc.

Indicate any alternative or supplemental examinations performed and methods (s) of fabrication examination.

121.4 Thi ASME Code,Section XI,1977 Edition with Addenda through the Summer 1978 Addenda specifies use of Appendix III of Section XI.for ferritic piping welds.. If this requirement is not applicable (for example, for austenitic piping welds), ultrasonic examination is required to be conducted in accord ~ance with the applicable requirements

.of Article 5 of Section V, as amended by IWA-2232.

Provide 'a technical justification for any alternatives used such as Se'ction.XI, Appendix III, Supplement 7 for austenitic piping welds and discuss the following:

All modifications permitted by Supplement 7.

a.

b.

Methods' of assuring adequate examination sensitivity o'ver the required examination volume.

Methods of qualiffing the procedure for examination through the c.

. weld (if complete examination is to be considered for examinations conducted with only one side access).

d.

Any crack-line indication, 20 percent of DAC or greater, discovered during examination of piping welds or adjacent base metal materials should be recorded and investigated by a Level II or Level' III

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examiner to the extent necessary to determine the shape, jdentity, and location of the reflector.

The Owner should evaluate and take corrective action for the disposition e.

of any indication' investigated and found to be other than geometrical or metallurgical in nature.

121.5 The augmented inspection requirement for high energy piping systems is discussed in Standard Review Plan 3.6.1 and Branch Technical Positinn

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ASB 3-1.

High energy lines meeting the " modified break exclusion criteria" need not be subjected to augmented preservice/ inservice l

inspection.

The " modified break exclusion region" criteria may be applied in those special cases in which guard pipes are necessary, and it has been demonstrated that acce'ss to perform an inspection is extremely difficult to achieve.

In such areas the inspection requirements may be l

eliminated provided the guard pipe is designed'for the full dynamic l

effects of a longitudinal or circumferential break of the enclosed l

process pipe including jet impingement, pipe whip impact and environ-mental effects.

Augmented inspection of hihh energy: piping systems should be performed where physical access exists. Discuss the practicality of performing examinations described in SRP 3.6.1 and Position ASB 3-1.

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REACTOR SYSTEFG BRANCH - TMI ISSUES REQUEST FOR ADDITIONAL INFORMATION 440.T.1 16 response ~ to NUREG-0737 item II.B.1, Catawba FSAR Section 1.9 states (1.9) that a description of the reactor vessel head vent system will be pro-vided "later".

We require that this description be provided for our review.

440.T.2 In response to NLREG-0737 item II.K.2.13 (Thermal-Mechanical Report),

(1.9)

Catawba FSAR Section 1.9 states that a report will be provided,. but does not provide a schedule for this submittal.

We require the ap-plicant to either provide a submittal schedule consistent with the requirements of NUREG-0737 or cite an applicable generic report whose submittal schedule complies with NUREG-0737 requirements.

440.T.3 In response to NLREG-0737 item II.K.3.2, Ca(ghba FSAR Section 1.9 (1.9) has referred to a Westinghouse Owners Group report. We require that the applicant either provide a submittal schedule consistent with the requirements of NUREG-0737 or identify the specific reference which ap' plies to Ca.tawba.

440.T.4 In response to NUREG-0737 item II.K.3.3 the applicant has not committed

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(1.9) to report RV and SV challenges annually as required by NUREG-0737. We require the applicant to provide a plan to report these challenges annually.

.c 400.T.5 In response to NUREG-0737 item II.K.3.10 the applicant has proposed to bypass the anticipatory reactor trip on turbine trip at low power levels (below 50%).

We require that he provide analyses to justify the power level at which the trip.is bypassed (P-8).

400.T.6 In response to NLREG-0737 item II.K.3.17.the applicant has committed i

to report ECCS outages, but has not described 'what information would l

be reported.

We require that the applicant c6mnit to include in the report the information specified in NUREG-0737.

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~ GENERIC ISSUES BRANCH

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REQUEST FOR INFORMATION The Atomid' Safety and Licensing Appeal Board in ALAB-444 detennined that the Safety Evaluation Report for each plant should contain an assessment of each significant unresolved generic safety question.

It is the staff's view that the generic issues identified as " Unresolved Safety Issues" (NUREG-0606) are the substantive safety issues referred to by the Appeal Board. Accordingly, we are requesting that you provide us with a sumary description of your relevant investigative programs and the interim measures you have devised.for dealing with these issues pending the completion of the investigation, and what alternative courses of action might be available should the program not produce the envisaged result.

There are currently a total of 26 Unresolved Safety Issues discussed in NUREG-0606. We do not require infonnation from you at this time for a number of the issues since a number of the issues do not' apply to your type of reactor, or because a generic resolution has been issued.

Issues which have been resolved have been or are being incorporated into the NRC licensing guidance and are addressed as a part of the normal review process. However, we do request the information noted above for each of the issues listed below:

1.

Waterhammer (A-1)

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2.

Steam Generator Tube Integrity (A-3) 3.

'ATWS (A-9) 4.

Reactor Yessel Materials Toughness (A-11) 5.

Steam Generator and Reactor Coolant Pump Support (A-12) 6.

Systems Interaction (A-17) 7.

Seismic Design Criteria (A-40) 8.

Containment Emergency Sump Performance (A-43) 9.

Station Blackout (A-44)

Shutdown Decay H'at Removal Requirements (A-45) 10.

e 11.

Seismic Qualification of Equipment in Operating Plants (A-46) 12.

Safety Implications of Control Systems (A-47) 13.

Hydrogen Control. Measures and Effects of Hydrogen Burns on Safety Equipment (A-48)

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GENERIC ISSUES BRANCH

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SAFETY EVALUATION REPORT b

VIRGIL C. SUMMER APPENDIX C NUCLEAR REGULATORY COMMISSION (NRC)

UNRESOLVED SAFETY ISSUES C.1 Unresolved Safety Issues

. The NRC staff continuously evaluates the safety recuirements used in its reviews acainst new information as it becomes available.

Information related to the safety of nuclear power plants comes from a variety of sources includino experience from operating reactors; research results; NRC staff and Adytsory Committee on Reactor Safeguards (ACRS) safety reviews; and vendor, architect / engineer and utility design reviews.

Each time a new concern or safety issue is identified from one or more of these sources, the need for immediate action to assure safe operation is assessed. This assessment includes consideration of the generic imolications of the issue.

In some cases, innediate action is taken to assure safety, e.g., the derating of boiltng water reactors as a result of the channel box wear problems in 1975.

In other cases, interim measures, such as modifications to operating procedures,' may be suff.icient to allow further study of the issue criar to making licensing decisions. In most cases, however, the initial assessment indicates that immediate licensing actions or changes in licenstng criteria are not necessary. In any event, further stcEv may be deemed acerceriate to make judgments as to whether existing N'C R

staff recuirements should be modified to address the issue for new clants 'or if backfitting is accrocriate for the.long tenn coeration of clants alreadv under c:nstruction or in cceration.

These issues are sometimes called "ceneric safety issues" because they are eslated to a particular class or type of nuclear facility rather

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than a soecific olant. These issues have also been referred to as

" unresolved safety issues." However, as discussed above, such issues are considered on.a generic basis cnly after the staff has made an initial determination that the safety significance of the issue does not l

orchibit continued operatitn or require licensing actions while the loncer-term generic review ts underway.

C.2 ALAB 444 Recuti ements These lo'nger-term geieric studies were the subject of a Decision by the Atomic Safety and Lictising Acceal Board of the Nuclear Reculatory ) in Commission. The Decision was tssued on November 23,1977 (ALAB-444 _

connection with the Acpeal Board's consideration of the Gulf States Utility Company anplica. tion for the River Send Station, Unit Nos. I and.

2.

In the view of the Appeal Scard, (po. 25-29)

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health and safety schere are not confined 'to the considera' tion and C-1 G

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disposition of those issues which may have been presented to it by a party or an " Interested State" with the required degree of specificity.

To the contrary, irrespective of what matters may or may not have been properly placed in controversy, prior to authorizing the

-issuance of a construction permit the board must make the finding, inter alia, that there is '" reasonable assurance" that."the proposed Ya'BTity can be constructed and operated at the proposed location without undue risk to the health and safety of the oublic." 0f necessity, this 10 CFR 50.35(a) determinatten will entail an inouiry into whether the staff review satisfactorily has cbme to grips with ~

any unresolved generic safety problems which might have an imcact upon operation of the nuclear factlity under consideration."

"The SER is, of course, the princioal document before the licensing board which reflects the content and outcome of the staff's safety review. The board should therefore be able to icok to that document to ascertain the extent to which generic unresolved safety problems which have been previously identified in an FSAR item, a Task Action Plan, an ACRS report or elsewhere have been factored into the staff's analysis for the carticular reactor--and with what result. To this end, in our view, each SER should contain a sununary description of those generic prcblems under continuing study which have both relevance to facilities of the type under review and potentially significant public safety imolications."

"This sumary description should include information of the kind now contained in most Task Action plans. More soecifically,'there should be an indication of the investicative orogram which has been or will be undertaken with regard to the problem, the program's anticioated time scan, whether (and if so, what) interim measures have been devised for dealing with the crablem cending the comoletion.

of the investigation, and wnat alternative courses of action mient" be available snould the program not produce the envisaged result."

"In short, the board (and the public as well) should be in a casition

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to ascertain from the SER itself--without the need to resort to s

extr.insic cocuments--the staff's cerceotion of the nature and extent of the relationship between each'significant unresolved generic safety ouestion and the eventual ' operation of the reactor under scrutiny. Once again, this assessment might well have a direct. bearing upon the ability of the licensing board to make the safety findings recuired of it on the construction pemit level even though the ceneric answer to the cuestion remains in the offing. Among other thinos, the furnished information would likely shed light on such alternatively important considerations as whether:

.(1) the problem has already been resolved for the reactor under study; (2) there is a reasonable basis for concluding that a satisfactory solution will be obtained before the reactor is out.in oceration; or (3) the problem would have no safety imolications until after several years of reactor operation-and, should it not be resolved by then, alternative means will be available to insure that continued oceration (if cemitted at all) would not pose an undue risk *a the i

pub 1ic."

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This appendix is specifically included to respond to the. decision of the Atomic Safety and Licensing Appeal Board as enunciated in ALAB A44, and as applied to an operating license proceeding Vircinia Electric and power Comoany (North Anna Nuclear Power Station,-Unit Nos 1 anc TT7 ADT-491, NRC 245 (1978).

C.3 " Unresolved Safety Issues" In a related matter, as a result of Congressional action on the Nuclear Regulatory Commission budget for Fiscal Year 1978, the Energy Reorganization Act of 1974 was amended (PL 95-209) on December 13, 1977 to include, among other things, a new Section 210 as follows:

" UNRESOLVED SAFETY ISSUES PLAN" "SEC. 210. The Commiss' ion shall. develop a plan providing for specif.ication and analysis of unresolved :sfety issues relating to nuclear reactors and shall take such action cs may be necessary to implement corrective measures with respect to such issues. Such clan shall be submitted to the Congress.on or before January 1, 1978 and progress reports shall be included in the annual report of the Commission thereafter."

The Joint Explanatory Statement of the House-Senate Conference Comnittee for the Fiscal Year 1978 Aopropriations Bill (Bill 5.1131) provided the followinc additional information regarding the Cennittee's deliberations on this portion of the bili:

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"SECTION 3 - UNRESCLVED SAFETY ISSUES" "The House amendment required develcpment of a plan to resolve generic safety issues. The conferees agreed to a requirement that the plan be submitted to the Congress on or before January 1,1978.

The conferees also expressed the intent that this clan should identify and describe those safety issues, relating to nuclear power reactors, which are unresolved on the date of enactment.

It should set forth:

(1) Commission actions taken directly or indirectly to develoo and imolement corre:tive measures; (2) further actions planned concerning such measures; and (3) timetables and cost-estimates of such actions. The Commission should indicate the oriority it has assigned to each issue, and the basis on which priorities have been assigned."

In response to the reporting requirements of the new Section,210, the NRC staff submitted to Congress on January 1,1978, a report, NUREG-0410, entitled "NRC Program for the Resolution of Generic Issues Related.

to Nuclear Power Plants," describing the NRC generic issues program.

The NRC program was already in place when PL 95-209 was enacted and is

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of considerably broader scope than the " Unresolved Safety Issues Plan" recuired by Section 210. In the letter trt.nsmitting NUREG-0410 to the -

Congress on December 30, 1977, the Comission. indicated that "the progress reports, which are required by Section 210 to be included in future NRC annual recorts, may be more useful to Congress.if they focus on-the spe.cific Section 210 safety items."

It is the NRC's view that the intent of Section 210 was to assure that plans were developed and implemented on issues with potentially significant public safety implicatio,ns. In'1978,.the NRC undertook a review of over 130 generic issues addressed in the NRC program to deternipe which issues fit this description and cualify as " Unresolved Safety Issues" for reporting to the Congress. The NRC review included the development of proposals by the NRC Staff and review and final approval by the NRC Comissioners.

This review is described in a report NUREG-0510, " Identification of 01 resolved Safety Issues Relating to Nuclear Power Plants - A Repor.t to Congress," dated January 1979. The report provides the following definition-of an " Unresolved Safety Issue:"

"An Unresolved Safety Issue is a matter affecting a number of nuclear power plants that poses important questions concerning.the adecuacy of existing safety recuirements for which a final resolution has not yet been developed and that involves conditions not 14kely to be accepable over the lifetime of the plants it affects."-

Further the recort indicates that in acolying this definition, matters that pose imcortant cuestions concerning the adecuacy of existing I

safety recuirements" were judced to be those for whien resolution is neces'sary to (1) comoensate for a nossible major reduction in the degree of protection of the oublic health and safety, or (2) provide a potentially significant decrease in the risk to the oublic health and safety. Ouite simoly, an " Unresolved Safety Issue" is potentially significant from a public safety standooint and its resolution is likely to result in NRC action on the affected plants.

All of the issues addressed in the NRC program were systematically evaluated against this definition as described in NUREG-0510. As a result, seventeen " Unresolved Safety Issues" addressed by twenty-two tasks in the NRC program were identified. The issues are listed below.

Progress on these issues was first discussed in the 1978 NRC Annual Report. The number (s) of the generic task (s) (e.g., A-1) in the NRC program addressing each issue is indicated in parentheses following the titl e.

" UNRESOLVED SAFETY ISSUES" (APPLICABLE TASK NOS.)

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1.

Waterhamer - (A-1) 2.

Asymetric Blowdown Loads on the Reactor Coolant System - (A-2)

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3.

Pre'ssurized Wate.r Reactor Steam Generatg Tube Integrity - (A-3, A-l 4,A-5) l 4

SWR Mark I and Mark II Pressure Suppression Containments - (A-6, A-7,A-8,A-39)

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An'ticipated Transients Without Scram - (A-9) 6.

.3WR -Nozzle. Cracking - (A-10) 7.

Reactor Vessel Materials Toughness - (A-11) 8.

Fracture Toughness'of Steam Generator' and Reactor Coolant Pump Supports - (A-12) 9.

Systems Interaction in Nuclear Fower Plants - (A-17)

10. Environmental Qualification of Safety-Related Electrical Ei;uipment -

(A-24)

11. Reactor Vessel Pressure Transient Protection - (A-7.5)

12. Residual Heat Removal Requirements - (A-31)

13. Control of Heavy Loads Near Spent Fuel - (A-36) 14 Seismic Design Criteria - (A-40)

15. Pi'pe Cracks at Boiling Water Reactors - (A 42)

16. Contair. ment Emergency Sump Reliability - (A-43)

17. Station Blackout - (A-44)

In the view of the staff, the " Unresolved Safety Issues" listed above are the substantive safety issues referred to by the Appeal Board in

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ALAB-444 when it spoke of "... those generic problems under continuing study which have.... potentially significant public safety implications."

Eight of the 22 tasks identified with the " Unresolved Safety Issues" are not applicable to Virgil C. Sumer Nuclear Station, Unit 1 and six of these eight. tasks (A-6, A-7, A-8, A-39, A-10 and A-42) are peculiar to boiling water reactors. Tasks A-4 and A-5 address steam generator tube oroblems in Combustion Engineering and Sabcock and Wilcox plants. With regard to the remaining la tasks that are applicable to this facili.tiy, the NRC staff has issued NUREG reports providing its proposed resolution of five of these issues. Each of these have been addressed in this Safety Evaluation Report or will be addressed in a future supolement.

The taole below lists those issues and the section of this Safety Evaluation Report in which they aie discussed.

Safety Evaluation Task Number NUREG Recort and Title Recort Section A-2 NUREG-0609, " Asymmetric 3.9.3 Blowdown loads on PWR Primary Systems" A-24 NUREG-0588, " Interim Staff 7.7.2 Position on Environmental Qualification of Safety-Related Electrical. Equipment" A-26 NUREG-0224, " Reactor Vessel S.4.2 Pressure Transient Protection for Pressurized Water Reactors" and RSB STP 5-2

.A-31 Regulatory Guide 1.139, Will be addressed Guidance for Residual Heat in a future Removal" and RSB BTP S-1 supplement.

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. Safety Evaluation Task Number NUREG Reccrt and Title Recort Section

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A-36 NUREG-0612, " Control of 9.2.4 Heavy Loads at Nuclear Power Plants" The remaining issues applicable to this facility are listed in the~

following table:

GENERIC TASKS ' ADDRESSING UNRESOLVED SAFETY ISSUES THAT ARE APPLICABLE TO THE VIRGIL C. SUMMER NUCLEAR STATION, UNIT 1 1.

A-1 Waterhammer 2.

A-3 Westinghouse Steam Generator Tube Integrity 3.

A-9. Anticipated Transients Without Scram 4

A-11 Reactor Vessel Materials Toughness 5.

.A-12 Potential for low Fracture Toughness and lamellar Tearing -

on PER Steam Generator and Reactor Coolant Pump Supoorts 6.

A-17 Systems Interactions in Nuclear Power Plants 7.

A 40 Seismic Design Criter.ia 8.

A-43 Containment Energency Sump Reliability 9.

A 44 Station Blackout With -the exception of Tasks A-9, A-43, and A 44 Task Action Plans for the generic tasks above are included in NUREG-0649, " Task Action Plans for Unresolved Safety Issues.Related to Nuclear Power Plants." A technical resolution for Task A-9 has been prooosed by the NRC staff in Volume 4 of NUREG-0460, issued for comment. This served as a basis for the staff's crocosal for rulemaking on this issue. The Task Action Plan for.

Task A J3 was issued in January 1981, and the Task Action Plan for. A 44 was issued in July 1980. Draft NUREG-0577 which represents staff resolution of USI A-12 was issued for com ent in Novemoer 1979. The Draft NUREG contained the Task Action Plan for A-12.

The information provided in NUREG-0649 meets most of the informational requirements of ALA8 444 Each Task Action Plan provides a.descr4ption of the arablem; the staff's approaches to its resolution; a general discussion of the bases upon which continued plant licensing or coeration can proceed pending completion of the task; the technical organizations involved in the task and estimates of the manoower recuired; a description cf the interactions with other NRC offices, the Advisory Committee on Reactor Safeguarts and outside organizations; estimates of funding required for contractor supplied technical assistance; prospective dates for comoleting the task; and a description of potential problems that could alter the planned approach on schedule.

In addition to the Task Action Plans, the staff issues the " Office of Nuclear Reactor Regulation Unresolved Safety Issues Summary, Aqua Book" (NUREG-0606) on a quarterly basis which provides current schedule information for each of the " Unresolved Safety Issues." It also includes information relative to the imolementation status of each " Unresolved Safety Issue" for which technical resolution is complete.

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i' We have reviewed the nine " Unresolved Safety Issues" listed above as they relate to Virgil C. Summer-Nuclear Station, Unit 1.

Discussion of each of these issues including references to related discussions in the Safety Evaluation Report are provided below in Section C.5.

Based on our review of these items, we have concluded, for the reasons set forth in Section C.5, that there is reasonable assurance that this facility can be operated prior to the ultimate resolution of these generic issues without endangering the health and safety of the public, C.4 New " Unresolved' Safety Issues" An in-depth and systematic review of generic safety concerds identified since January 1979 has been performed by the staff to determine if any of thec a issues should be designated as new " Unresolved Safety Issues."

The candidate issues originated from concerns' identified in NUREG-0660, "HRC Action Plan as a Result, of the TMI-2 Accident;" ACRS recommendations; abnormal occurrence reports and other coerating experience. The staff's proposed lis't was reviewed and commented on by the ACRS, the Office of Analysis and Evaluation.of Operational Data (AE00).and the Office of Policy Evaluation. The ACRS and AE00 also proposed that several addi-tional

" Unresolved Safety Issues" be considered by the Commission. The Commission considered the above information and approved the following four new " Unresolved' Safety Issues:"-

A 45 Shutdown Decay Heat Removal Reqcirements

, s-A-46 Seismic Qualification of Ecuipment in Operating Plants.

A 47 Safety Implications of Control Systems A 48 Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment A description of the above process together with a list of the issues considered is present in NUREG-0705, " Identification 'of New Unresolved Safety Issues Relating to Nuclear Power Plants, Special Report to Congress," dated March 1981. An expanded discussion of each of the new

" Unresolved Safety Issues" is also contained in NUREG-0705.

The apolicability and bases for licensing prior to ultimate resolution of the four new USIs for Virgil C. Summer, Unit 1 are discussed in Section C.S.

C.5 Discussion of Tasks as they Relate to Virgil C. Summer Nuclear Station. Unit 1 A-1 Waterhammer Waterhammer events are intense pressure pulses in fluid systems caused by any one of a number of mechanisms and system conditions.

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Since 1971 there have been ever 100 incidents involving waterhammer in pressurized water reactors and boiling water reactors. The water -

hamers have involved steam generator feedrings and piping, decay heat removal systems, emergency core cooling systems, containment spray lin'es, service water lines, feedwater lines and steam lines. However, the systems most frequently affected by waterhammer effects are the feedwater systems. The most serious waterhamer events have occurred in the steam generator feedrings of pressurized water. reactors. These types of waterhamer events are addressed in Section 10.4.3 of this Safety Evalua-tion Recort.

With regard to protection against other potential waterhaniner events currently provided in plants, piping design cedes require consideration of impact loads. Approaches used at the design stage include: (1) increasing valve closure times, (2) piping layout to preclude water slugs-in-steam lines and vapor. for1 nation in water lines, (3) use of snubbers and pipe hangers, and (4) use of. vents and drains. In addition, as described in Section 3.9.2 of this Safety Evaluation Report, we require that the applicant conduct a preoperational vibration dynamic effects test program in accordance with Section III of the ASME Code for all ASME Class 1 and Class 2 piping systems and piping restraints during startup and litial operation. These tests will provide adecuate ' assurance

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that the piping and piping restraints have been designed to withstand dynamic effects due to valve closures, cumo trics and other ocerating modes associated with the design operational transients.

-c Nonetheless,'in the unlikely event that a large pipe break did result from a severe waterhamer event, core. cooling is assured by the emercency core cooling systems described in Section 6.3 of this Safety Evaluation Recort and crotection acainst the. dynamic effects of such pipe breaks inside and outside of centainment is provided as described in Secticn 3.5 of this Safety Evaluation Report.

Task A-1 may identify some potentially significant waterhamer scenarios that have not exolicitly been accounted for in the design and operation of nuclear gewer plants. The task has not as. yet identified the need for requiring any additional measures beyond those already required in the short tem.

Based on the foregoing, we have concluded that the. faci,lity can be operated prior to ultimate resolution of this generic issue without undue risk to 'the health and safety of the public.

A-3 'destinchouse~ Steam Generator Tube Intecrity The primary concern is the capability of steam generator tubes to maintain theit integrity during nomal cperation and postulated accident conditions. In addition, the requirements for increased steam generator tube inspections and repairs have resulted in signifi-cant increases in occupartional exposures to workers. Corrosion resulting in steam generator tube wall thinning (wastage) has been observed in C-a

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several Vestinghouse plants for a number of years. Plants operating exclusively with an all volatile secondary water treatment process have not experienced this fctm of degradation to date. Another major corrosion-related phenomenon has also been observed in a number of plants in recent years, resulting from a buildup of support plate corrosion products in the annulus between the tubes and the support plates. This' buildup

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eventually causes a diametral reduction of the tubes, called " denting,"

and defomation of the tube support plates. This phenorcenen has led to other problems, including stress corrosion cracking, leaks at the tube / support plate intersections, and U-bend section cracking of tubes which were highly stressed because *of support plate deformation.

Specific measures such as steam generator design features and a secondary water chemistry control and monitoring program, that the applicant has employed to minimize the onset of steam generator-tube problems are described in Section gf this Safety Evaluation Report. In addition, Section of this Safety Evaluation Report discusses the inservice inspection. requirements. *As described in Section

, the applicant has met all current recuirements regarding steam generator tube integrity.

The Technical Specification wil.1 include requirements for actions to be taken in the event that steam generator tube leakage occurs during clant operation.

Task A-3 is expected to result in imorovements in our current requirements for inservice inspection of steam generator tubes. These improvemtnts will include a better statistical basis for inservice inspection program recuirements and consideration of the cost / benefit of increased inspection.

pending comoletion of Task A-3, the measures taken at this facility snould minimi:e the s, team generator tube oroblems encountered. Further

-he - iervice inscecticn and Technical Soecification recuirements will.

assure that the applicant and the NRC staff are alertad to tube degradation should it occur. Acprocriate actions such as tuce plugging, increased and more frequent inspections and pet,er derating could be taken if necessary. Since the improvements that will result'from Task A-3 will be procedural, i.e., an improved inservice inscection crogram, they can be implemented by the aoplicant after operation of this facility begins, if necessary.

Sased.on the foregoing, we have concluded that this facility can be operated prior to ultimate resolution of this ganeric issue without undue risk to the health and safety of the public.

1-9 Anticicated Transients Without Scram Nuclear plants have safety and control systems to limit the consecuences of temporary abnomal operating conditions or " anticipated transients."

Some deviations from nomal operating conditions may be minor; others, occurring less frequently, may 1 moose significant demands on plant equipment.

In some anticipated transients, rapidly shutting down the nuclear reaction (initiating a " scram".), and thus ranidly reducing the generation of heat in the reactor core, is an important safety measure.

If there were a cotentially severe " anticipated transient" and the C-9 1

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reactor shutdown system did not " scram" as desired, then an " anticipated transient without scram," or ATWS, would have occurred.

The anticipated transien't without scram issue and the requirements that must be met by the applicant prior to operation of the facility are discussed in Section 15.3.5 of this Safety Evaluation Report.

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The ATWS issue is currently scheduled for rulemaking in mid-summer 1981.

~ The applicant will be required to comply with.any further requirements on ATWS which may be imposed as a result of the rulemaking.

Based on our review, we have concluded that there is reasonable ' assurance that this facility can be operated prior to ultimate resolution of this generic issue without endangering the health and safety of the public.

A-11 Reactor Vessel Materials Touchness

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f Re'sistance to brittle fracture, a rapidly propagating catastrophic failure mode for a component containing flaws, is described quanti -

tatively by a material property generally deroted as " fracture toughness."

Fracture toughness has different values and charactersitics depending upon the material being considered. For steels used in a nuclear reactor pressure vessel, three considerations are important. First, fracture toughness increases with increasing temperature; second, fracture toughness decreases with increasing load rates; and third; fracture toughness decreases with neutron irradiation.

4 In recognition of these considerations, power reactors are operated withirt restrictions imposed by the Technical Specificaticns on the cressure during heatuo and cooldown operations. These restrictions assure that the reactor vessel wi}l not be subjected to a comoination cf pressure and temoerature that could cause brittle fracture of the vessel if there were significant flaws in the vessel materials. The effect~of neutron radiation on 'the fracture toughness of the vtssel material is accounted for in developing and revising these Technical Soecification limitations.

For the service times and operating conditioris typical of current operating plants, reactor vessel fracture toughness for most plants provides adequate margins of safety against vessel failure under operating, testing, maintenance, and anticipated. transient conditions, and accident conditions over the life of the plant. However, results from'a re' actor vessel surveillance program and analyses performed for up to 20 older coerating pressurized water reactors and those for some more recent vintage plants will have marginal toughness, relative to required margins at nomal full power after comparatively short periods of operation.

In addition, results from analyses performed by pressurized water reactor manufacturers indicate that the integrity of some reactor vessels may not be maintained in the event that a main steam line break of a loss-of-coolant accident occurs after approximately 20 years of operation. The principal objective of Task A-11 is to develop an improved engineering method and safety

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• are available during norml operation and transients in older reactor vessiils with marginal fracture toughness and of the safety margins available during accident conditions for all plants.

Based on our evaluation of this facility's reactor vessel materials toughness, we have concluded that this unit will have adequate safety margins against brittle failure during operating, testing, maintenance and anticipated transient conditions over the life of the units. Since Task A-11 is projected to be completed well in advance of this facility's reactor vessel reaching a fluence level which would noticably reduce fracture resistance, acceptable vessel inteefity for the postulated accident conditions will be assured at least entil the reactor vessel is reevaluated for long-tem acceptability.

In addition, the surveillance program required by 10 CFR 50, Appendix H-will afford an opportunity to reevaluate the fracture toughness periodically j

during the first half of design life.

Therefore, based upon the foregoing, we have concluded that this facility can be operated prior to resolution of this generic issue without undue risk to the. health and safety of the public.

i A-12 Fracture Touchness of Steam Generator and Reactor Coolant pumo Sucoorts I

Curing the course of the licensing action for North Anna Power Statfon i

Unit No. I and 2 a number of questions were raised as to the potential for lamellar tearing and low fracture toughness of the steant generator and rea'ctor coolant pump support materials for those facilities. Two different steel scecifications (ASTM A36-70a and ASTM A572-7Ca) covered most of the material used for these succort.t. Toughness tests, not criqinally soecified and not in the relevant ASTM specifications, were made on those heats for which excess material was available. The toughness of the A36 steel was found to be adecuate, but the toughness of the A572 steel was o

relatively peor at an operating temcerature of 80 F.

Since similar materials and designs have been used on other nuclear plants, the concerns regarding the suoports for the North Anna facilities are applicable to other PWR plants.

It was therefore necessary to reassess the fracture toughness of the steam generator and reactor coolant pumo succort materials for all coerating PWR plants and those in CP and OL review.

NUREG-0577, " Potential for Low Fracture Toughness and Lamellar Tearing on PWR Steam Generator and Reactor Coolant Pump Supports," was issued I

for comment in November 1979. This report sumarizes work perfomed by the NRC staff and its contractor, Sandia Laboratories, in the resolution l

of this generic activity. The report describes the technical issue,

the technical studies perfomed by Sandia Laboratories, the NRC staff's -

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technical positions based on these studies, and the.NRC staff's plan for implementing its technical positions' As a part of initiating the implementation of the findings in this report,. letters were sent to all applicants and licensees on May 19 and 20, 1980. In these letters a revised proposed implementation plan was presented and specific criteria j

for material qualifications were defined.

C-11

Many coments on both the draft of NUREG-0577 and the letters of May 19 and 20 have been received by the NRC staff and detailed consideration is presently being given to these comments. After completing our review and analysis of the coments provided, we will, issue the final revision of NUREG-0577 which will include a full discussion and resolution of the coments and a final plan for implementation.

We estimate that our implementation review will require approximately two years. Since many factors (initiating event, low feacture toughness in a critical support member in tension, low operating temperature, large flaw) must be simultaneously present for failure of the support system we have determined that licensing for pressurized water reactors should continue during the implementation phase. Our conclusions regarding licensing and sucsequent operation are not sensitive to the estimated length of time required for this work.

i A-17 Systems Interaction in Nuclear power plants The. licensing requirements and procedures used in our safety review.

address many different types of systems interaction. Current licensing requirements are founded on the principle of defense-in-depth. Adh'erence i

to this principle results in requirements such as physical separation

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and independence of redundant safety systems, and protection against events such as high energy line ruptures, missiles, high winds, flooding, seismic events, fires, operator errors, and sabotage. These design provisions supplemented by the current review procedures of the Standard Review plan (NUREG-75/08'/) which require interdisciplinary reviews-and which account, to a large extent, for review of potential systems interactions, orovide for an adequately safe situation with respect to such ~ interactions.

The cuality assurance program which is followed during the design, construction, and operational phases for each plant is expected to provide added assurance against the potential for adverse systems. inter-actions.

In November 1974, the Advisory Comittee on Reactor Safeguards recuested that the NRC staff give attention to the evaluation of safety systems from a multidisciplinary point of view, in order to identify potentially i

l undesirable interactions between plant systems. The concern arises because the design and analysis of systems is frequently assigned to teams with functional engineering specialties--such as civil, electrical, mechanical, or nuclear. The cuestion is whether the work of these functional specialists.is sufficiently integrated in their design and analysis activities to enable them to identify adverse interactions between and among systems. Such advery events might occur, for example, because designers did not assure that redundancy and independence of safety systems were provided under all conditions of. operation required, which might happen if the functional teams were 'not adequately coordinated.

In mid-1977, Task A-17'was in tiated to confirm that present review procedures and safety criterla provide an acceptable level of redundancy and independence for systems required for safety by evaluating the potential for undesirable interactions between and among systems.

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The NRC staff's curren't review procedures assign primary responsibility for review of various technical areas and safety' systems to soecific

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e organizat.ional units and assign secondary responsibility to other units wher(there is a functional or interdisciplinary relationship,.,0esigners follow somewhat similar procedures and provide'for interdisciplinary reviews and analyses of systems. Task A-17 provided an independent -

study of methods that could identify important systems interactions adversely impacting safety; and which are not considered by cur' rent 1

review procedures. The first phase of this study began in May 1978 and was completed in February 1980 by Sandia Laboratories under contract to the NRC staff.

The Phase I investigatiert was structured to identify areas where inter-actions are possible between and among systems and have the. potential of negating or seriously degrading the performance of safety functions.

The study concentrated on corrrnon cause on linking failures among systems that could violate a safety function. The investigation then identified where NRC review procedures may not have' properly accounted for these interactions.

The Sandia Study used fault-tree methods to identify component failure comoinations (cut-sets) that could result in loss of a safety function.

The cut-sets were reduced to minimal combinations by incorporating six comon or linking systems failures into the analysis. The results of the Phase I effort indicate that, within the scope of the study only a few areas of review procedures need improvement regarding systems interaction.

However, the level of detail needed to identify all examples of potential' system interaction candidates observed in some operating plants was'iiot within the Phase I scope of the Sandia Study.

It is expected that the development of systematic ways'to identify and evaluate systems interactions will reduce the likelihood of co=en cause failures resulting in the loss of plant safety functions. Hcwever, tne -

studies to date indicate that current review procedures and criteria supplemented by the aoplication of post-TMI findings and risk studies

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provide reasonable assurance that the effects of potential systems interaction on plant safety will be within the effects on plant safety previously evaluated.

Therefore, we concluded that there is reasonable assurance that Virgil C. Sumer, Unit I can be operated prior to the final resolution of this generic issue without endangering the health a'nd safety of the public.

A-10 Seismic Desian Criteria - Short-Term Procram NRC regulations require that nuclear power structures, systems and components important to safety be designed to withstand the effects of natural phenomena such as earthquakes. Detailed requirements and guidance regarding the seismic design of nuclear plants are provided in the NRC regulations and in Regulatory Guides issued by the Comission. However, there'are a number of plants with construction permits and coerating licenses issued before the NRC's currer;t regulations and regulatory

. guidance were in place. For this rea' san, rereviews of the seismic

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design of various plants are being undertaken to assure that these plants do not present an undue risk to the public. Task A-40 is, in

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effect, a compendium of short-term efforts to support _suca reevaluation.

efforts of the NRC staff, expecially those related to older, operating, plants. In addition, some revisions to the Standard Review >lan sections %.,

and Regulatory Guides to bring them more in line with the state-of-tho' ~N art v il result.

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v As discussed in Section 3.7 of this Safety Evaluation Report'the seismic

' m, design basis and seismic design of the facility have been evaluated at the operating license stage and have been found accept:ble. We do not

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exoect the results of Task A-40 to affect these conclusions because t' a techniques under consideration are essentially those utilized in the

' review of this facilityt Should the resolution of Task A-40 indicate a change is needed in licensing requirements, all operating' reactors, we have~g Sunner will be reevaluated on a case-by-case basis. Accordingly, includin concluded that this facility can be operated prior to. the ultimte resolution of this generic issue without endangering the health ands-safety of the public.

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'A A 43 Containment Emercency Sumo Reliability Following a postulated loss-of-coolant accident, i.e., a break'in M reactor coolant system piping, the water flowing from the break woulhbe 1

collected in the emergency sump at the low point in the containment.\\.

This water would be recirculated through the reactor system bycthe.

emergency core cooling pumas to maintain core cooling. This water would also be circulated through the containment spray system tc remove. heat; s and fission products from the containment. Loss of the ability to ' draw '

water from the emergency sump could disable the emergency core cooling and containment spray systems.

One costulated means of losing the ability to draw water frem the ' emergency

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sumo could be blockage by debris. A principal source of such debris-could be the themal. insulation on the reactor coolant system piping'.

In the event of a piping break, the subsecuent violent release to the.

high pressure water in the reactor coolant system could rip off the insulation in the area of the break. This debris could then be sweet into the sumo, potentially causing blockage.,

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Currently, regulatory positions regarding sump -design are presented in Regulatory Guide 1.32, " Sumps for Emergency Core Cooling and Conta.inment Spray Systems," which address. debris (insulation).. Regulatory Guide 1.82 recommends, in addition to providing redundant sep'arated sumps, that two protective screens be provided. A low approach velocity in the vicinity of the sump is required to allow insulation to settle out before reaching the sump screening; and it is required that the sump remain functional assuming that one-half of the screen surface area is blocked.

A second postulated means of losing the ability to draw water from the emergency sump coulc be abnamal conditions in the sump or at the pump inlet such as air entrai.nment, vertices, or excessive pressure drops.

These conditions could result in pump cavitation, reduced flow and.

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Currently, regulatory positions regarding sump testing are containedin Regulator'y Guide 1.79, "Preoperational Testing of Emergency Core Cooling

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the-rectreulation function. Both in-plant and scale model tests have s

been perfcmed by applicants to demonstrate that circulation through the s

'. sumo can be reliably accomplished.

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• As 'Eldicated in Section 6.3.3 of this Safety Evaluation Report, the iN L

"upplicant'wfli.cerform out-of-plant scale model tests of. the containment v'f.

-sumo design. Tha applicant will be required to demonstrate that there

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fs renscoable cssurance that the sump design will perform as expected f' followins a less-of-coolant accident.

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-The near tenn implementation of Task A-43 for this facility is expected l*

N to be procedural in nature and assure adequate housekeeping and emergency

? % q, procedures tesupplement the sumo tests discussed above. Accordingly, '

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s resolution of this generic issue without endangering the health and safety of the public.

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fi ElectricaF deser for safety systems at' nuclear power plants must'be supplied;by at least two redun' dant and independent divisions. The

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x systems us8d to re. ove decay heat to cool the reactor core folicwing a r,

reactor shutdgwn are included among the safety systems that must megt e

these requirements. Each electrical division for safety systems includer an offsite~:lternating current power connection, a standby emergency diesel generator alternating current power supply and direct current m-

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sources.

Tcsk A M jnvol/es a study of whether or not nuclear power plants should be designed to accomodate a complete loss of all alternating current o

power, i.e., loss of both the offsite and the emergency diesel generator t

alternating current power suoplies. This issue arose bec"Jse of operating s

excerience regarding the reliability of alternating cbrrent power suoplies.

I A number of operating plants have experienced a total loss of offsite electrical power, and more-occurrences are expected in the future.

During each of these loss of offsite power events, the onsite emergency alternating current power supplies were available to supply the power needed by vital safety equipment. However, in some instances, one of r

I the redundant emergency power supplies has been unavilable. In addition, -

there have been numerous reports of emergency diesel generators failing to start and run in operating plants during periodic surveillance

tests, i

A loss of all alternating current power was not a design basis event for-the Sumer facility. Nonetheless, a combination of design, operation and testing requirements that have been imposed on the applicant will assure that these units will have substantial resistance to a loss of all alternating current and that, even.if a loss of all alternating current should occur, there is reasonable assurance that the core will be cooled. These are discussed below.

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A loss of offsite alternating. current power involves a loss of both the preferred and backup sources of offsite power. Our review and basis for C-15 O

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acceptance of the design, inspection, and testing provisions for the offsite power system are described in Section 8.2 of this Safety Evaluation Report.

If offsite power is' lost, two diesel generators and their associated distribution systems will' deliver emergency power to safety-related equipment. Our review of the design, testing, surveilla. ice, and maintenance provisions for the onsite emergency diesels is described in Section 8.3 of the SER. Our requirements include preoperational testing to assure the reliability of the installed diesel generators in accordance with our requirements discussed in the SER.

In addition, the applicant has been requested to implement a. program for enhancement of diesel generator reliability to better assure the long-tem reliability of the diesel generators. This program resulted from recommendations of NUREG/CR-0660, " Enhancement of Onsite Emergency Generator Reliability.",

Event if both offsite and onsite alternating current power are lost, cooling water can still be provided to the steam generators by the

. auxiliary feedwater system by employing a steam turbine driven pump that does not rely on ' alternating current power for operation. Our-review of the auxiliary feedwater system design and operation is described in Section of the Safety Evaluation Report.

The issue of station blackout was also considered by the Atomic Safety

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and Licensing Appeal Board (ALAB-603) for the St. Lucie Unit No. 2 -

facility. In addition, in view of the completion schedule for Task A-44 (October 1982), the Appeal Board recommended that the Commission t#te -

i expeditious action to ensure that other plants and their operators'are ecuicped to accennodate a station blackout event. The Commission has-reviewed this recommendation and determined that some interim measures should be taken at all facilities including Sumer while Task A la is being conducted. Consequently, in'teria emergency procedures and coerato'r training for safe operation of the facility and restoration of alternating current power will be ' required. The staff notified the applicant of these requirements in a letter from D. Eisenhut, NRC, to the applicant dated February 25, 1981. We will condition the coerating license for Sumer that their procedures and training te completed by fuel load date.

Based an the above, we have concluded that there is reasonable assurance l

that Sumer can be operated prior to the ultimate resolution of this generic issue without endangering the health and safety of the oublic.

A 45 Shutdown Decay Heat Removal Recuirements Under normal operating conditions, power generated 'within a reactor is removed as steam-to produce electricity via a turbine generator.

Following a reactor shutdown, a reactor. produ~ es insufficient power c

to operate the-turbine; however, the radioactive decay of fission products continues to produce heat (so-called " decay heat"). Therefore, when reactor shutdown occurs, other measures must be available to remove r'_

decay heat from the reactor to ensure that high temperatures and pressures do not develop which could jeopardize'the reactor and the reactor coolant

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- system. It is evident, therefore, that all light water reactors (LWRs) share two comon decay heat removal functional reouirements: (1) to C-16

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inside the reactor vessel to ensure adequate cooling of the. reactor fuel. The reliability of a particular power plant to perform these functions depends on the frecuency of initiating events that require or jeopardize decay heat removal operations and the~ probability that required-systems will respond to remove the decay heat..

This Unresolved Safety Issue will evaluate the benefit Of providing alternate means of decay heat removal which could substantially increase the plants' capability to handle a broader spectrum of transients and accidents. The study will consist of a generic system evaluation and will result in recomendations regarding the desirability of and possible design requirements for improvements in existing systems or an alternative decay heat removal method if the improvements or alternative can significant reduce the overall risk to the public.

The primary method for removal of decay heat from pressurized water reactors is-via the steam generators to the secondary. system. This energy is transferred on the secondary side to either the main feedwater-or auxiliary feedwater systems,-and it is rejected to either the turbine condenser or the atmosphere via the steamline safety / relief ~ valves.

Following the TMI-2 accident, the importance of the auxiliary feedwater system was highlighted and a number ~of steps were taken to improve the -

reliability of the auxiliary feedwater system. The staff's review of these items is contained in Section of this Safety Evaluation" Report.

It was also stipulated that plants must be capable of providing the required AFW flow for at least two hours from one auxiliary feedwater pump train, independent of any alternating current power source (that is, if both off-site and on-site alternating current cower sources are lost).

pressurized water reactors also have alternate means of removing decay heat if an extended loss of feedwater is postulated. This ciethod is known as " feed and bleed" and uses the high pressure injection system to add water coolant (feed) at high pressure to the primary system. The decay heat increases the system pressure and energy is removed through the power-operated relief ' valves and/or the safety valves (bleed), if necessary.

At low primary system pressure (below about 200 psi), the long-tenn decay heat is removed by the residual heat removal system to achieve cold shu'tdown conditions.

Based on the foregoing, we have concluded that Virgil C. Summer, Unit 1 can be operat(d prior to ultimate resolution of this generic issue

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without endangering the health and safety of the public.

A-46 Seismic Oualification of Ecuioment in Coerating plants The design criteria and methods for the seismic qualification of mechanical and electrical equipment in nuclear power plant

.hange during the course of the comercial nucl,s have undergone.significant ear power program.

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Consequently, the margins of safety provided in existing equipment to resist seismically induced loads and perform the intended safety functions may vary considerably. The seismic qualification of the equipment in operating plants must, therefore, be reassessed to ensure the ability to bring the plant to a safe shutdown condition when subject to a seismic event. The objective of this Unresolved Safety. Issue is to ' establish an

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explicit set of guidelines that could be used to judge the adequacy of the seismic qualification of mechanical and electrical equipment.at all operating plants in lieu of attemoting to backfit current desi.gn criteria for new plants. This guidance will concern equipment required to safely shut down the plant, as well as equipment whose function is not required for safe shutdown, but whose failure could result in adverse conditions which might impair shutcown functions.

Virgil C. Summer Unit I was designed using current seismic criteria and the design has been reviewed and approved by the Commission-staff 'in accordance with current design criteria and methods for seismic qualifica-tion. Therefore, we conclude that Virgil C. Summer Unit 1 can be operated prior to resolution of this generic issue without undue risk to the health and safety of the public.

A 47 Safety Imolications of Control Sfstems This issue concerns the potential for transients.or accidents being made more severe as a result of control system failures or malfunctions.

These failures or malfunctions may occur independently or as a resukt of the accident or transient under consideration. One concern is the-potential for a single failure such as a loss of a power supply, short circuit, open circuit, or sensor failure to cause simultanecus malfunction of several control features. Such an occurrence could cenceivably result in a transient more severe tnan those transients analyzed as anticipated operational occurrences. A second concern. is for a postulated accident to cause control system failures which could make the accident more severe than analy:ed. Accidents could conceivably cause control system failures by creating a harsh environment in the area of the control ecuipment or by physically damaging the control equipment.

It is generally believed by the staff that such control system failures would not lead to serious events or result in conditions that safety systems cannot safely handle. Systematic evaluations have not been rigourously performed to verify this belief. The potential for an accident that could affect a p~ articular control system..and effects of the control system failures, may differ from plant to plant. Therefore, it is not possible to develop generic answers t'o these concerns, but -

rather plant-specific evaluations are required. The purpose of this Unresolved Safety Issue is to define generic criteria that will' be used for plant-specific evaluations.

The Summer control and safety systems have been designed with the goal of ensuring that control system failures will not prevent automatic or manual initiation and operation of any safety system equipment required to trip the plant or to maintain the ' plant in a safe shutdown condition C-18

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foll,owing any " anticipated operational occurrence" or " accident." This has been accomplished by either providing independence between safety and non-safety systems-or providing isolating devices between safety and non-safety systems. These devices preclude the propagation of non-safety?

system equipment faults to the protection system. This ensures that coeration of the safety system equipment is not imoaired.

A systematic evaluation of the control system design, as contemplated for this Unresolved Safety Issue, has not been performeid to determine whether postulated accidents could cause significant control system

' failures which would make the accident consecuences more severe than presently analyzed. However, a wide range of bounding transients and accidents is presently analyzed to assure that the postulated events such as steam generator overfill and overcooling events would be adequately safety systems have been_

In addition, systematic reviews of mitigated by the safety systems.

control system failures (perforned with the goal of ensuring thatsingle or multiple) will action.

Based on the above, we 'have concluded that there is reasonable assurance i

that the Sunner Unit can be operated prior to the ultimate resolution of this generic issue without endangering the health and safety of'the public.

A 48 Hydrocen Control Measures and' Effects of Hydrocen Burns on Safety _,

_tculoment Following a loss-of-coolant accident in a light water reactor plant, combus;ible gases, J rincipally hydrogen, may accumula_te inside '

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the primary reactor containment as a result of:

(1) metal-water reaction involving the fuel element cladding; (2) the radiolytic decomoosition of.

the water in the reactor core and the containment sume; (3) the corrosion of certa'in construction materials by the spray solution; and (a) any synergistic chemical, thermal and radiolytic effects of pos,t-accident

. environmental conditions on containment protective coating systems and electric cable insulation.

Because of the potential for significant hydrogen generation as the i

result of an accident,10 CR Section 50.44, " Standards for Combustible Gas Control-System in Light Water Cooled pcwer Reactors" and the General l

Design Criteria 41, " Containment At:nosphere Cleanup" in Appendix. A to 10 CR part 50 require that systems be provided to control hydrogen l

concentrations in the containment atmosphere following a postulated accident to ensure that containment integrity is maintained.

10 CR Section 50.44 requires that the cor:bustible gas control system provided be capable of handling the hydrogen generated as a result of l

degradation of the emergency core cooling. system such that the hydrogen release is five times the amount calculated in demonstratir.g compliance with 10 CR Section 50.46 or the amount. corresponding to reaction of the cladding to a depth,of 0.00023 f.nch, whichever amount is greater.

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The accident at TMI-2 on March RS,1979 resulted in hydrogen generation well.in excess of the amounts specified in 10 CFR 50.44.'

As a result of this knowledge it became apparent to NRC that specific design measures are needed for handling larger hydrog'en releases, particularly for..

smaller low pressure containments. As a result, the Comission detennined

. that 'a rulemaking proceeding should be undertaken to define-the manner and extent to which hydrogen evolution'and other effects of a degraded core need to be taken into account in plant design. An advance notice of this rulemaking proceeding on degraded core issues was published in the Federal Register on October 2,1980.

Recognizing that a numbei of years may be required to compl.ete this r0lemaking proceeding, a set of short-tem or interim actions relative to hydrogen control requirements were developed and implemented. These interim measures were described in a second October 2,1980 Federal Register notice. For plants with large dry containmente such as Virgil C.

Sumer, Unit 1, no near-tern mitigation neasures are requ' ired by the interim rule.

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The Virgil C. Sumer plant has about two million cubic feet of net i'ree volume. Assuming 30 to 50% metal-water reaction in the core, the resulting uniformly mixed concentration of hydrogen in the containment will range

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from 6.to 10%. This is well below the concentrations for detonation and even below the limits for combustion if there were more than 50% steam in the' containment atmosphere.

Design pressure of the Virgil C. Sumer plant is 57 psig. Analyses performed on the Ifon and indian point plants show that the failure pressures are greater than twice the design pressures.

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If the substantial amount of metal-water reaction were to occur shortly.

following onset of a large LCCA and while the containment is still near its peak pressure, the pressure increase. caused by the noncondensible hydrogen gas and its associated exothermic formation energy will be substantially less than the failure pressure. If the metal-water reaction were to occur well after onset of the large LOCA, then the centainment heat removal system would have condensed much.of the steam in the containment and reduced the containment pressure. This would provide a substantial margin for accommodating the hydrogen generated by the metal-water reaction.

In addition, the "Short Tenn Lessons Learned" frem the iNI-2 accident have been implemented on the Virgil C. Sumer plant. This action will reduce the likelihood of accidents that could lead to substantial 1

amounts of metal-water reaction.

Accordingly, pending resolution of this Unresolv' d Safety Issue and the-e rulemaking proceeding on hydrogen generation, the Virgi'. C. Sumer plant can be operated without undue risk to the health and safety of the public.

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