ML20148U512

Responds to 780828 Request for Clarification Re ACRS Recommendations Made Between 770101 & 0930.Forwards Clarifying Info

ML20148U512
Person / Time
Site:	Beaver Valley, North Anna
Issue date:	11/28/1978
From:	Case E Office of Nuclear Reactor Regulation
To:	Fraley R Advisory Committee on Reactor Safeguards
References
NUDOCS 7812070094
Download: ML20148U512 (44)

Text

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l o e ss l f  %, UNITED STATES E . T NUCLEAR REGULATORY COMMisslON ,

f. Q$ WASHINGTON, D C. 205% l NOV 2 s 1978 k,, . .... N.s[

MEMORANDUM FOR: Raymond F. Fraley, Executive Director Advisory Committee on Reactor Safeguards FROM: Edson G. Case, Deputy Director l Office of Nuclear Reactor Regulation  !

SUBJECT:

STATUS OF ACRS RECOMMENDATIONS Your memorandum of August 28, 1978, requested additional clarification regarding a number of recommendations made by the Committee during the period January 1,1977 through September 30, 1977. These were initially forwarded by your memorandum of December 1, 1977, to which we responded on April 21, 1978.

The enclosure provides the additional clarification reques by your August 28, 1978, memorandum. 7

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/ Edson G. Case, De'puty Director

/ Office of Nuclear Reactor Regulation

Enclosure:

As Stated cc: H. Denton S. Levine C. Smith R. Minogue J. Davis R. Boyd R. DeYoung R. Mattson V. Stello W. Russell W. Minners D. Bunch

?"llocw 78120700${, A@V PACE O

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J APPENDIX A 201st Meetina, January 6-8, 1977

1. Question: . ,

The Committee looks forward to receiving the Staff evaluation of the-feedwater monitoring program. Feedwater piping vibrations will be discussed at a future meeting of the Fluid Hydraulics Subcommittee.

Response

The staff has not yet completed its review of the feedwater monitoring program at Beaver Valley 1. However, the following is a status of our evaluation to date.

The feedwater system hydraulic transient monitoring program at the Beaver Valley Power Station, Unit 1 was carried out during the period of March l' to August 1,1977. The purpose of this program was to verify the. adequacy of system modifications to prevent certain feed-water line vibrations and to obtain data, in the event of such vibrations, that would facilitate an assessment of piping stresses ,

and allow an analysis of causative factors. The licensee reported the results of its monitoring program by letter dated November R3,1977, and in response to staff questions, it submitted additional informa-tion by. letter. dated March 17, 1978.

The staff has reviewed the information submitted.by the licensee and has not found a recurrence of the same type of vibrations that occurred once in November, once in December 1976 and once in January 1977 during the startup of Beaver Valley, Unit 1. The data showed only one indication of possible vibration that occurred on July 17, 1977, approximately 10 seconds after a load rejection test from 50% power to zero power., The above cited three incidents of significant pipe vibration occurred during operation between 30% to 50% power. The licensee has stated that the indications of vibration on July 17, 1977, were due to the anomalous response of certain instruments to mechanical vibrations caused by closure of the feed-water pump check valves and the turbine stop valves. The staff has not reached a conclusion on this point. The staff intends to request additional information and complete its evaluation in February 1979.

However, since modifying the internals of the feedwater control valves in February 1977, the Beaver Valley, Unit 1 facility has not experi-enced a recurrence of the same type of vibration'that occurred prior

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to the modification. This is a positive indication that the cause of those vibrations has been eliminated.

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s f f ,1 APPENDIX A 201st Meeting, January 6-8, 1977

3. Question:

The resolution is aueptable for North Anna. It is not clear how the North Anna resolution is to be translated into a generic resolution.

Clarification is requested.

Response

The resolution for North Anna on a postulated fuel handling accident inside containment is being pursued by the staff in a generic manner.

Briefly, the staff is requiring that a plant possess the capability for prompt detection of any radioactivity release inside containment and automatic containment isolation using redundant radiation monitors. The staff is revising its Standard Review Plan (Section 15.7.4) to reflect this consideration.

The revised acceptance criteria for consideration of fuel handling accidents inside containment are as follows (extracted from revised Section 15.7.4): i 1

Where an applicant proposes that fuel handling i operations inside containment occur only when contain-ment is isolated, or where the containment is continu-ously vented to the environment via an iodine filter system, this is acceptable. Where fuel handling opera-tions inside containment occur when the containment is open to the environment (i.e., with a containment purge exhaust system) the proposed design is acceptable if it possesses the capability for prompt detection and auto-matic containment isolation by use of redundant radia- l tion monitors. l The revised review procedures (extracted from Section 15.7.4) for consideration of this matter are as follows:

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The proposed systems intended to mitigate the consequences of a fuel handling accident inside contain-ment are reviewed. Where an applicant proposes that fuel handling will occur only when the containment is l isolated, this is acceptable and no radiological con-sequences need be calculated. Where fuel handling

s 1 f a F operations occur only when the containment is exhausted to the environment via an ESF filter system, this is acceptable and the radiological consequences should be calculated giving appropriate credit for this system.

Where the containment will be open during fuel hand]ing operations (as with a containment purge exhaust system), the reviewer should verify that a prompt de-tection and automatic containment isolation capability is provided and that an independent evaluation of the consequences shows that the resulting doses are within the acceptance criteria given in Section II.2. A review should be made of the applicant's analysis and should include examination of the type, location and redundancy of the radiation monitors intended to detect an activity release within containment and verification that de-tection is followed by automatic containment isolation.

The reviewer should assess the time required to iso-late the containment. This should include the instru-ment line sampling time (where appropriate), detector response time and containment purge isolation valve "

actuation and closure time. The containment is consid-ered isolated only when the purge isolation valves are fully closed and seated. The applicant's analysis should be reviewed regarding the travel time of any activity release starting from its release point above the refueling cavity or transfer canal and including travel time in ducts or ventilation systems until it reaches the inner containment purge isolation valve.

Where the applicant claims credit for dilution or mix-ing of a release due to natural or forced convection inside containment prior to release, this is reviewed and assessed. Refs. 3 and 4 may be consulted and used by the reviewer for guidance in estimating dilution and mixing. The time required for the release to reach the inner isolation valve is compared to the time required to isolate containment. If the time required for the release to reach the isolation valve is longer than the time required to isolate containment, then essentially no release outside of containment occurs, and the re-viewer's assessment will reflect this. If the time required for the release to reach the isolation valve is less than that required to isolate containment, and no mixing or dilution credit can be given, the reviewer should assume that the entire activity release escapes from the containment in evaluating the consequences.

Where mixing and dilution within containment isolation, the radiological consequences will be reduced compared  :

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to the entire activity release by the degree of mixing and dilution occurring prior to containment isolation.

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APPENDIX A 202nd Meeting, Februar, 10-12, 1977

1. Question:

The document referenced by Dr. Moeller should have been proposed ANSI Standard N13/42, " Performance Specifications for Reactor Emergency Radiological Monitoring Instrumentation", (Tables 1, 2 and 3).

Response

The staff has noted this item. The reference appears to be to BNWL-1635, dated May 1972 which is incorporated as Reference 3 in Reg. Guide 1.97, Rev. 1.

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APPENDIX A 202nd Meetino, February 10-12, 1977

3. Question:

It is not clear from the response what final action the NRC Staff has taken. Clarification is requested. ,

Response

The staff erred in its previous response. We should have noted that although the regulations exclude consideration of nuclear weapons detonation from consideration, electromagnetic interference (EMI) that might arise from other sources is being investigated.

NUREG-0153 discusses the effect of EMP from a nuclear weapon and two ORNL technical reports on the subject. Other external sources of .

electromagnetic interference, such as-lightning, would produce ampli- l tudes that would be only a small . fraction of that from a nuclear  ;

j weapon. Thus the following quote from NUREG-0153 would apply to these other possible sources.

"In all nuclear power plants, the reactor and sona of the protection system circuitry are located within the containment building, which is either built of steel plate or is a concrete structure lined with steel plate.

In both cases, the shielding from EMP provided by the steel plate is excellent and there should be no adverse effects within the containment structure. However, a substantial part of the protection system circuitry is outside the containment, in the control room, the cable spreading room, and in portions of the auxiliary build-ing, where essential auxiliary systems are located.

The control room and auxiliary buildings are normally constructed of reinforced concrete of heavy construc-tion since they are built to withstand tornado missiles, differential pressures and seismic events. The mul-tiple courses of reinforcing ba: s in the walls and ceil-

.ings of these structures should provide substantial attenuation of EMP. It appears that up to 30 to 40 db of attenuation are available from this sort of heavily reinforced concrete construction. Further shielding is

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. , . , .i provided by steel cabinets, cable raceways, and electrical conduits for wire and cable runs inside these structures.

The ORNL reports find that the most serious effects would be on digital logic circuits. They find that analog-type control circuits are more resistant to pulse damage. There is also a strong effect from.large pulsas on solid state circuitry, because the solid

state elements (diodes, transistors, etc.) are typically unable to accept large temporary overloads as are .

vacuum tube elements. Digital computers with solid state components are probably the most vulnerable kind of equipment to EMP exposures."

Effects of inplant EMI phenomena from sources other than nuclear weapons, especially on safety-related digital equipment, are being addressed on recently submitted standard design applications; e.g.,

BSAR-205 (RPS-II), CESSAR (CPC) and RESAR-414 (IPS).

Typical of our approach on the standard designs is the recently completed review of ANO-2. The staff expressed concern with the sus-ceptibility of the AN0-2 core protection calculator (CPC) system to EMI. A position was developed and a test program was set up to verify that the proper operation of the CPC system will not be compromised by radiated or conducted noise signals that can be expected during nuclear power plant operation. The test procedure and test results are addressed in the AN0-2 SER. The susceptibility tests for EMI radiation and conduction were~ run in accordance with MIL require- ,

ments.* Also as a guide, the staff utilizes in its review RDT Standard CI-IT, " Instrumentation and Control Equipment Grounding and Shielding Practices", as a methodology to minimize the effects of EMI phe.omena. A common practice within industry is to provide a shield ,

around a twisted pair of wires and ground one end of the shield.  !

This minimizes the capacitive coupling from the external voltage sources to the pair of wires inside the shield. In conjunction with j shielding, in-line filters are used to suppress the undesirable  !

• 1. MIL-STD-416A; Military Standard Electromagnetic Interference Requirements for Equipment.

2. MIL-STD-464; Military Standard, Electromagnetic Interference 4 Characteristics, Measurement of.

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l frequencies. In order to evaluate the effectiveness of the shielding and filtering, it is necessary to measure the actual levels of fre-quencies of EMI in-situ and evaluate their impact on equipment /

system susceptibility. At the present time, such tests are being conducted at the ANO-2 plant. Similar rev' s are being conducted on the PDA applications listed above.

We believe that the ongoing review of th 'f EMI phenomena on safety-related digital equipment will ide .1eed and priority for any further study on this subject for rious safety-related features of nuclear power plani.s.

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APPENDIX A 204th Meeting, April 7-9, 1977 la. Question:

The response relative to pump fly.< heels is accept ible for the present, but there is no indication of the schedule for ultimate resolution of this matter. The Connittee recommends that the Staff make quarterly reports until a technological solution to this problem is identified.

Response

The staff now considers this issue to be resolved. The requirement for the maximum acceptable flaw size in pump flywheel material in Regulatory Guide 1.14 is primarily based on the capabilities of current manufacturing processes and inspection methods, since the calculated critical flaw size that could result in failure is larger by a significant margin. The specified maximum acceptable flaw size is well above the detectable limits of current inspection methods.

The specified flaw size is also within the capabilities of current manufacturing processes and available from commercial sources. The staff believes that sufficient technical basis exists to support our current requirements. However, the staff remains open to receive and review any new information that might support a change to our position.

APPENDIX A 204th Meeting, April 7-9, 1977 lb. Question:

1 The Committee would like to be kept informed regarding the development '

and application of this probabilistic methodology to this subject.

Response

Subsequent to the July 1, 1977 staff report to Libarkin from Denton, NRR prepared a research request on probabilistic flood assessments.

Enclosed is a copy of the October 26, 1977 memo to the Office of Nulcear Regulatory Research suggesting activities in this area. The intent of the proposed research is to assess and potentially improve the acceptability of methodology associated with estimating proba-bilities of severe floods. No formal program has yet been received from research on this subject.

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OCT 2 6 377

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1 MEx0RAn;un FOR: 3aul Lavina, ctrector -

Office of liuclear Rc;ulatcry Researc.y FRON: Edson G. Case, Acting Diro: tor Offica.of Ecclect Reset:r Regulation .

SUBJECT:

RESEARCH REQUIREME':TS FOR EVALUATION OF MARGIftS l I

/NAILABLE IM FLOOD PROTECTIOM OF UUOLEAR PGUER PU.NTS (RR-P.RR77-16)

NRR requests RES to initiate confimatory research related to evaluating a the r.argins inherent in flood protaction of nuclect power plants. Both -

WASH-1400 and ifcensing experience indicate that identification of such cargins is important to either confirming that present practica is -

adequate, or for modifying future practice. The centacts for this work are W. 5. Bivins and L. G. Hulman, both at 492-7233.

BACKGROUtiD .

Our currant cathods and criteria for snalysis of flood potential and for "

flood protection are ss=narized in the following documents:

a. R.G.1.59, Design Basis Ficods for Nuclear Power Plants;

b. American Rational Standard 5-170, Standards 'for Determining Design l Easis Flooding at Power Reactor Sites. '

c. R.G.1.70, Section 2.4, Standard Format.and C:ntant of SARs for 4 Muclear Fower Plants; .

d. R.G.1.102 Flood Protection for Nuclear Power Plants. .'

i The as:umptien and underlying practice in this subject area is that a nuclear power plant hardened against the =est severe flooding conditions reasonably probable is adequate to protect the pubite health and safety.

Potential flooding conditfor.s are anelyzed detarainistically using techniques and procedures evolved from practica by other Federal agencies  :

(primarily the Corps of Engineers, liOAA, FPC, and Bureau of Reclemation).  :

Furtherscre, these techniques and procedures consider the range of causative mechanisms, including tropical stor s, large and small scale extra tropics 1 precipitation and wind storms, se: seismic activity and dam failures. %o assessment is made of the probability of the "1 cod conditions postu-lated. Futhermore, no evaluation is mde of the likelihood of failure of flood protection, the consequences of failure, the residual risks

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fnherent in in:dequate need cor dition/ flood phbtectien criterie, er the degree of conservatis: associated with the present methodology. _

We have follcwed the evolution of probabilistic techniques with consider-able interest, particularly those associated with UAti-1400. and have

, attempted to utilize their application in flooding assessments on several

- occasions. Our latest application is summari:ad in a :sso to ACP.S fr =

H. R. Denton, dated July 1,1977 (copy enclosed) which indicates , cur concerns related to probabilistic techniques applied to esti=ating the 11ko11 hood of severe floods. Our primary concerns include the following:

a. A single =casura of an ovant cutco=e..such as water level or dis-chargo, is generally used :s an indic: tor of event =agnitude. No differentiation is made as to the cause of the event, however, and

• experience indicates that a flood record contains events caused by at least two cc=pletely different phenomena (e.g. , tropical and extra tropical storms). A typical flood record may not contain a -

large enough sa=ple of floods caused by each type of event to be representativn. Furthermore, even if a flood record is not considered composed of =f xed events, the representativeness of a relativsly shcrt-tarm recced for predicticn of very icw likelihood events say be questionable. ,

b. The selection of confidence limits that (1) miniat:a the residual error in estimates cf event =agnitudes, and correspondin;1y, (0) mini =ize the range of event likelihood,

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c. If likelihood estinstes are $a~de using dependent and indopendent cesponents of event na;nitude (e.g. , rainfall cagnitude, areal distribution of rainfall, ground wetness, etc.), hew are individual ce=ponant confidence limits reconciled to minimize the residual errer in estimates of the outccma ::agnitude and outcome likelihood?

Flocd protection requirements vary considerably frc= site to site. For exacple, if all safety-related facilities are located above design basis flood levels, no flood protection provisions are required. Many sites fall in this category; others do not, prior to the issuance of Reg.

• Guide 1.102, flood protection provisions at those sites susceptible to flooding often included many provisions requiring emergoney action to  ;

provide external water barriers. With the advent of Reg. Guide 1.102, hardened protection has been the staff goal such that water barriers are ,

permanently in place. E4 sed upon this history, designs and costs of 1 providing f1ced protection vary considerably from site to site. l

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To assess the overall risk, we have consistently concluded that a plan

• acccamodating a design basis flood ccndition (which could be caused by a l i

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severe precipitation, da 1 failure, hurricane, wave actien, or seis=tcally induced event) is adequate. No detailad assessment has been made of the overall risk of a severs flood for which either ficod protection is inadequate, or for the likalthood and consequences of a fa'ilure of ,

, design flood protection. Both of these situations should be assessed to 3 assure that (1) flood protection requiracents are adequate, and (2) '

residual risks are appropriata1y minimized. ,

INFORMATION HEEDS Our infor:ation needs arc divided into two categories: assessments of methodology uncertainties for applying probabilistic =ethods to pre- i dicting severe flooding events, and the residual risk associated with t

present flood protection requiracents. We are aware of no programs in j any of the .Mational Laboratories that are compatible with the work pro-  !

posed herein. There are, however, several researchers that have evaluated *).

cxtreme natural phenomena probabilistically, and residual risk assess- t' ments in the area of earthquakes have been undertaken. Futhermore, a j' numerical evaluation of accident risk is under study with PNL, and may ',

provide a basis for the work requested herein, j

5pecifically, the following caterial should be provided: i I

a. assess the long tars representativeness of stream, lake and coastal j flood records with particular emphasis on causative mechanisms  ;

(including hurricanes, large scale extra tropical storms and thunder- ]

showers),

b. identify acceptable methodology (or methodologies) for selecting 1 confidence ifmits that (1) minimize residual risks 1 '

cagnitude evaluation at design levels of 10' to10~9cxttcmeevent per year, and (2) minimize the uncartainty in probability estimates at the same design levels.- ,

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c. if individual ecmponents of floed events are used to assess event j likelihood, instead of a single outeeme, identify an acceptable y methodology (or methodologies) that also satisfies b. above.  !

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d. assess the likalthood of flood protection not performing its  !

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required function and the resulting potential consecuences. ,'

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We recognize, hawaver, that a conclusion frod this research cay be that extreme ficed events cannot be predictad with an acceptable level of '

confidene . '

i The desired time frame for completing this activity is the first cuarter of FY 79 to allow for developzont of any indicated changes in staff revie.v methodology and changes in Standard Review Plans. -

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t . < Ii LICENSING IMPACT '

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This program may provide a basis for a probabilistic assessment of the . '

flood potential at' nuclear power plants. Overall, identification of the safety margin available in flood potential and flood protection will provide a basis' for, considerations of revisions to standard review plans -

and safety guides presently employing deterministic approaches. '

t RESEARCH EFFORT Mo assessment of the level of effort required has been made. We suggest proposals be sought and we will be glad to participate in their review.

This recuest has been discussed informally with Ian Wall and Jerry Harbour ,

of your office.

VALUE/ IMPACT ASSESSPENT ,

No quantification of flood likelihood is available for use to judge the' utility of regulatory requirements. A well considered probabilistic analysis will provide the basis for (1) maintaining the present level of flood evaluation and protection requirements, (2) recuiring less protection,-

or (3) requiring more protection, and changing present evaluation -

methodologies. <

Three alternatives to this proposal were considered as follows:

a., continue the present methodology; -

b. arbitrarily increase or decrease the level required for flood protection by simply adding or subtracting an increment of eleva-tion; or

c. requiring flood protection redundancy.

The latter two alternatives are considered purely arbitrary without the results of the requesteG research. The first alternative, business as usual, will be continued until we can evaluate the results of the recuested a research based upon not only our own experience that no historical flood 1 e

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event has produced conditions worse than postulated, but similar enerience

, of ether Federal agencies. -

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E. c. 0.:. f Edson G. Case, Acting Director.

Office of Nuclear Reactor Regulation

Enclosures:

l As Stated  !

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cc: w/ enclosure H. Denton D. Muller -

L. Rubenstein a ,

F. Miraglia '

O. Wiggin'an .

DSE ads DSE BCs  :

J. Harbour, RES .

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L. Beratan, OSD

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. J. Knight '

R. Tedesco .

I. Sthweil .

V. Benaroya . -

.N W. Bivins ~

HES Personnel .

C. Jupiter RES ..

DISTRIBUTION - '

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APPENDIX A 205th Meeting, May 5-6, 1977

1. Question:

The Staff reply is not responsive to the inquiry that addressed the procedures for identifying drawings and descriptive material that should be withheld from public disclosure. Apparently, the decision to withhold has been turned over to the Commission. For the present, information is being withheld on a proprietary basis. The Committee is interested in the ground rules for establishing what information should be withheld assuming that the proprietary alternative can be implemented. The memorandum from Goller suggests that the licensee will make judgments concerning information to be safeguarded. It is not clear whether the NRC Staff has a basis for testing the licensees' judgment. Clarification is requested.

Response

Generally, the staff will withhold from public disclosure any drawing or descriptive material which details, displays, identifies or ampli-fies a licensee's or applicant's site specific method for safeguarding licensed special nuclear material or security measures taken for the physical protection of a licensed facility or plant in which licensed special nuclear material is processed or used. This position is be-lieved to be fully justified pursuant to the provisions of 10 CFR 2.790(d)(1). It is recognized that in some cases, however, sit may be necessary or prudent for the staff to disclose information which would not significantly or adversely affect a licensee's or applicant's physical security system. Such disclosure, though, would normally only concern generic physical security requirements or matters of common knowledge.

In other cases, the staff can and has challenged the validity of a licensee's or applicant's request to withhold information from public disclosure. Should the staff challenge such a request, however, they must determine:

1. Whether the information has been held in confidence by its owner.

2. Whether the information is of a type customarily held in confidence by its owner and whether there is a rational basis therefore.

3. Whether the information was transmitted to and received by the Commission in confidence.

4. Whether the information is available in public sources.

5. Whether public disclosure of the information sought to be withheld is likely to cause substantial harm to the competitive position of the owner of the information.

Should a request for withholding pursuant to the above be denied, the Commission notifies the licensee and provides the licensee with a statement of reasons for the denial.

In any instance, a balancing of the interests of the person or agency urging nondisclosure and the public interest in disclosure is made by the staff.

APPENDIX A 205th Meeting, May 5-6, 1977 ,

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2. Question:

l Formation of a damage control team should receive consideration. A fairly modest investment might lead to a considerable improvement in response time and capability. An ACRS Subcommittee will follow up this response with a meeting to discuss the subject.

Response

It appears at this time that a specific " damage control team" is not needed to provide an effective response to acts of sabotage. This position is based on the assumption that the Safeguards Contingency Plans, which the licensees are required to submit to the Commission f in accordance with 10 CFR 50.34(d) and Appendix C to 10 CFR Part 73, will provide licensee response forces with predetermined measures or actions to be initiated should a sabotage event be attempted.

The main purpose of the contingency plans is to identify credible

} events capable of disrupting plant operations, e.g., attempted sabc-l tage. The plans require statements of. the objective (s) to be achieved for each event and the actions to be accomplished by the response force.

In addition to the above, the licensee is required to provide the l Commission with an Emergency Plan in accordance with the provisions of ,

Appendix E to 10 CFR Part 50. The plan must provide reasonable l assurance that appropriate measures can and will be taken in the event of an emergency to protect the public health and safety.

In view of the above, it is currently believed that the requirements detailed in each plan, including the duties defined and assigned to i specific personnel, provide the same degree of contcol as that of a specially designated " damage control team".

ApPF.NDIX A 206th Meeting, June 9-10, 1977

2. Question:

See response to 205th Meeting, item 2.

Response

We assume this statement implies a Committee interest in the establishment of " damage control teams" to handle fires. Such a team is, in fact, in existence at the Zion Station.

The staff requires that ifcensees establish " fire brigades" for immediate response to fire threats. A general description of the staff requirements as regards minimum manning levels for the fire brigades is attached as Enclosure 1. For the Zion Station, these man-power requirements are established in the Technical Specifications.

Copies of applicable pages of these Technical Specifications are in-cluded as Enclosure 2.

The staff has evaluated the overall fire protection program for the Zion Station and has reported the results of its review in a Safety Evaluation. This Safety Evaluation is incorporated as an enclosure to a letter of March 10, 1978, to the licensee, which issued amend-l ments regarding fire protection to the Zion Station units. Copies of .

this letter with the Safety Evaluation were previously forwarded to the Committee.

The staff considers that its evaluation of the Zion Station for fire

, protection, as reported in the Safety Evaluation, includes the matters l raised by R. Pollard in his testimony on North Anna before the ASLB.

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2 0 AD?F' ~STRATIVE CONTROLS .

. 1 Organization, Review, Investigation and E. Retraining and replaccrent training of Stad Audit tion personnel shall be in accordance with-'

Ai!S1 M18.1," Selection and Training of Mu- .

A. The Station Superintendent shall have clear Pouer Plant Personnel", dated " arch S, overall full-time responsibility for 1971. A training program for the Fire Brig-safe operation of the facility. ade shall be naintained under the direction During periods when the Station of the Station Fire Marshall and shall Superintendent is unavaiable, he meet or exceed the requirements of Section shall designate'this responsibility 27 of the MEPA Code - 1975,. except that Fire to an established alternate who satifies Brigade training will be conducted quarterly.

the AHS1 N18.1 experience requirements for plant manager. P. Retraining shall be conducted at intervals not exceeding two years.

B. The corporate management uhich relates to the operation of this station is G. The Review and Investigat.vei Function and thc shown in figure 6.1.1. Audit Function of activities affecting qua-lity during f acility operations shall be con-C. The normal functional organization stituted~and have the responsibilitics and

< for operation of the station shall authoritics outlined below:

', be as shown in Figure 6.1.2. The shift manning for the station shall be as The Supervisor of the Offsite Review anc.

shown in Figure 6.1.3. A Fire Investigative Function shall- be appoin-Ericade of at leact 5 members shall ed by the Vice President of Construction he r.2intained on-site at all tines. Production, Licensing and Environental The fire brigade shall not include Affairs. The Audit Function shall be th the ninimum shift creu necessary for responsibility of the Manager of Quality safe shutdown of the plant (4 menbers) Assurance and shall be independent of or ar; personnel required for other operations.

escential functions during a fire cnergancy. a. Offsite Review and Investicative Function D. Cuclifications of the station management and cp; rating staff shall n.ect minimum The Supervisor of the Offsite Revich accc,rtable levels as described in AUSl and Investigative function shall:

"3clection and Training of nuclear Power (1) provide directicas fer the revie Plent rcrsonnel", dated " arch 8, 1971 with end investigative fu:ction cad appc-the c::ccptien of the Radiological Chemical int a senior participant to provide Surarvi nar who shall meet or c.:ceed the appropriate direction, (11) select cual:fications of Regulatory Guide 1.3 cach participant for this function, Scpt mber, 1975. The individual filling (111) select a complement of more the resi tion of Adminictrative Assistant than one participant who collectivel shall act the minimum accaptable level for possess backround and cualifications "Tecnnical Manager" as described in 11.2 4 of in the subject natter under review to provide I ann e ,

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MANPOWER REOUIREMENTS FOR OPERATING REACTORS The NRC has established requirements for persennel at operating reactors for purposes of plant operation, industrial security, and fire fighting. The following discussion considers the extent to which plant personnel assigned to either plant operatien or security may also be temporarily allowed to man a fire brigade in the event of a fire fer a single unit facility and sets f:rth an acceptable sharing :: heme for operating reactors.

Summary of Mancewer Recuirements

1. Fire Brigade: The staff has cencluded that the minimum si:e of the fire brigade shift should be five persons unless a specific site evaluation has been completed and some other number justified. The five-man team would consist of one leader and f ur fire fighters and would be excected to previde defense against the fire for an initial 30-minute peried. See Attachment A for the basis for the need for a five-man fire brigade.

2. Plant Operation: Standard Review plan Section 13.1.2 requires that for a statien having one licensed unit, each shift crew should have at least three persens at all times, plus two additional persens when the unit is operating. For ease of reference, Attachment B contains a copy of this SRP.

3. Plant Security: The requirements for a guard force are Outlined in 10 CFR Part 73.55. In the c:urse of the staff's review of proposed security plans, a recuired minimum security rescense force will be established for each specific site. In addition to the respense team, two additicnal memcers of the security force will be recuired to centinueusly man the Central Alar n Statien (CAS) and Secondary Alar n Station (SAS). It is expected that many facilities will have a security organi:atica with greater numbers of personnel than the minimum numcer assumed for purposes of discussion in this paper.

The NRC staff has given c:nsideratien to the apcropriateness of per-mitting a limited degree of sharing to satisfy the recuirements of plant cperati:n, security and fire protection and has c:ncluded that, (1) subject to certain site and plant specific ::nditions, the fire brigace staffing c uld generally be provided through Oceratiens and security perscnnel, and (2) the recuirements fer Opera: Ors and One security force should remain unc:mer:mised. Until a site scecific review is ccmcleted, the follcwing indicates the interim distributien and justifica:icn for these dual assignments, and therefere our interim minimum recuirements for a typical presently operating c:mercial single unit facility. The staff believes that mancewer for the fire brigade f:r multi-unit facilities is not new a ;reciem because Of the larger numcers of :eocle generally : resent at the sites. Situatiens whien de pose problems will be reviewed :n a :ase-cy-case basis.

1. Plant Oceration: The st:.ff has concluded that for most events at a s1ngle unit nuclear facility, a minimum of three ocerators should be available to place the reacter in a safe c:ndition.

The two additional operators required to be available at the nuclear facility are generally required to be present to per#enn routine jobs which can be interrupted to accomodate unusual situations that may arise. That is, there is the potential for the remaining two members of the coerating crew to assume other short-term duties such as fire fighting. In light of .hs original rationale for providing extra plant operators to cope with off-normal conditions, it accears justified to rely on these personnel fer this function. The staff rece. mends that one of the two operators assicned to the fire brigade should be designated as leader of the fire brigade in view of his background in plant operations and overall familiarity with the plant. In this regard, the shift super /isor should not be the fire brigade leader .

because his presence is necessary elsewhere if fires occur in certain critical areas of the plant.

2. plant Security: In the event of a fire, a contingency plan and prececures w111 be used in deploying tne security organi:atien to assure that an acpropriate level of chysical ;rotection is maintained during the event. The staff has deternined that it is :essible in the planning for site response to a fire, to assign a maximum of three members of the security organi:ation to serte en the fire brigade and still provide an ac:ectable level of pnysical protecticn. 'While certain security posts must be manned c:ntinuously (e.g. , CAS, SAS), the persennel in other assignments, including the rescense force, c:uld be temcorarily (i.e., 30 minutes) assigned to the fire brigade. In judging the merits of this allowance the underlying cuestien is whether the minimum security force strength must be maintained c:ntinuously in the event of a clant emergency such as a fire. Further examinatien of this issue leads to two potential rationales for reacning an affirmative decisien. First, c:uld there be a causal connection between a fire and the security threat? Secend, are there c:mpelling policy reasens to postulate a simultaneous threat and fire?

The first potential rationale would cnly be credible if, (1) the insider (pesed as part of the threat definitien) was an active

articicant in an assault and startec a fire :cincident witn the attack en the plant Or, (2) a diversienary fire was started oy an attack f:rce somewhere external to the slant itself wnere no ecuiement recuired for safe snuttewn is located. The role of the insider will be discussed first. 'While 73.55 assicns an active status to the in:ider, tne rule also recuires that measures be implemented :: c:ntain nis activities and therecy recuce his

effectiveness. At present, these measures include background checks on plant employees, limited access to vital plant areas, badging systems and the two-man rule. Here, limited access means that only designated employees are allowed in vital areas and that their entry is centrolled by either conventional locks or card-key systems. Also, if separate trains of safety ecuip-ment are involved, then either compartmentalization or the two-man rule is required. These measures to contain the insider are presently being implemented and will provide assurance that people of questionable reliability would not be able to gain employee status at a nuclear plant and shculd they bec:me an employee with unescorted access, significant restraints would be inter-pesed on the ability of such a person to carry cut extensive damage to plant vital areas. Reccgni:ing that additional safeguards may still be appropriate, the staff has rec: mended to the Comission that plant personnel also be required to ootain an NRC security clearance. The staff believes that the attendant background investigation asscciated with a clearance, in con-juncticn with the other 73.55 measures, will provide a high cegree of assurance that plant personnel will not attempt to take an active sabotage role. If the clearance rule is adopted the staff believes seme of the measures, such as tne two-man rule, designed to contain the insider can be relaxed. Thus, i

there dces not new appear to be a reascnably credible causative relatienship between a fire intentionally set by an insider i and the postulated external security threat. For the case of diversionary fires set external to the plant itself, adecuate security ferees can still be maintained by allcwing enly part of the fire brigade to rescend wnile both fire fighters and security force ar ned responders maintain a high degree of alertness for a cossible real attack semewhere else en the plant. Thus, the effective numcer of armed respenders required by 73.55 can be maintained for external diversionary fires.

The second potential rationale concerns whether a sericus, scontanecus fire should be :estulated coincident with an external security threat as a design basis. In evaluating such a recuire-ment it is useful to censider the likelihood of eccurrence of this ccmcination of events. While it is difficult to quantity the er:cability of the 73.55 threat, it is generally accectec that it is small, ccmcarable er bably to other design basis type events. The pr:bability of a fire wnich is scontaneous and located in or in c1cse prcximity to a vital area of the :lant and is sericus enougn : :ese a significant safety c:ncern is aise small. It wculd aceear, therefore, hat the randem coincicence of these two unlikely events wculd be sufficiently smail to not

.4.

require protection against their simultanecus occurrence. In addition, it should oe noted that the short time period (30 minutes) for which several memoers of the security force would be dedicated I tc the fire brigade wculd further reduce the likelihood of coincidence.

As neither of the two potential rationales appear to preclude the use of members of the security force in the event of a fire the staff has concluded that the short assignment of security personnel from the arned resocnse force or other available security personnel to the fire brigade under these conditions would be_ acceotable.

To ensure a timely and effective respense to a fire, while still ,

preserving a flexible security response, the staff celieves that the fire brigade shculd operate in the following manner. In the event of an internal fire, all five memoers of the fire brigade should be dispatched to the scene of the fire to assess the nature and seriousness of the fire. Simultaneously, the plant security force should be actively evaluating the possibility of any security threat to the plant and taking any actions which are necessary to counter that threat. For external fires, a lesser number than the five-man brigade should respond for assessment and fire fignting.

As the overall plant situation becomes apparent it would be expected that the most effective distribution of manpower between plant coerations, security and fire protection would be made, allowing a balanced utilization of manpower resources until offsite assistance becomes available. The manpcwer poci provided by the plant operations personnel and security force are adequate to respend to the occurrence of a design basis fire or a security threat equivalent to the 73.55 per%rmance requirements. It is also recogni:ed that other, more likely comoinatiens of postulated fires and security threats of a lesser magnitude than the design basis, could be considered. While tne procabilities of these higher likeliheed events may be sufficient to warrant protecting against them in ccmcination, the manoewer requirements required to coce with each event would be similarly reduced thereby allcwing adecuate coverage by plant persennel.

Conclusien The staff believes that it would be reasonable to allcw a limited amount of sharing of plant ;ersennel in satisfying the recuirements of plant operation, security, and fire protectien. An acceptable snaring scheme would entail reliance on two plant ocerators and nree memoers of the security organizatien to constitute the fire brigade. Since availability of the full fire brigace wculd only

.s-be recuired for fires with potential for serious damage, actual distribution of plant personnel during a plant emergency would be governed by the exigencies of the situation. Of course, all personnel assigned to the fire brigade would have to fulfill all applicable training requirements. It should also be recogni:ed that the diversion of personnel to the fire brigade would be of short duration and that substantial additional offsite assistance would be forthcoming in accordance with the emergency and contingency plan developed for each facility. In evaluating licensee procesals for manpower sharing due consideratien will also have to be made of unique facility characteristics, such as terrain and plant lay. cut, as well as the overall strengths of the licensee's fire and security pl an s . Minimum protection levels in either area could preclude the sharing cf manpcwer.

~

Attachment A Staff position Minimum Fire Urigade Shift Size T HTO.CC_UCTION Nuclear power plants depend en the response of an ensite fire brigade for dr.fense against the effects of fire on plant safe shutdewn cacebilitics. In scme areas, actions by the fire brigade are the In other areas, that are cretected

,only means of fire suopression.

by correctly designed cutematic dettction and suppression systems, manual firo fignting offerts are used to extinguish: (1) fires tco small to actuate the automatic system; (2) well develcoed fires if the autenctic system fails to function; and (2) fires that are not ecmoletely controlled by the automatic system. Tnus, an adecuate fire brigade is j essential to fulfill the defense in dectn requirements wnicn protect ~

~

safe chutdewn systems from the effects of fires and their related ccmeur.tien by-procucts.

DISCUSSICH There are a numeer of' factor that should be considered in establishing the minimum fire bristce shif t si:e. They include:

1) plant geometry and si:e;

2) cuantity and ' quality of cotection and suceressien systems; 31 fire ficating strategics for postulated fires;

4) fire brigace training; 51 fire brigado ecuipment; and

6) fire brigade su;plements by plant personnel and local fire deparment(s).

In all plants, the majority of postulated fires are in enciesed windew-less structures.

In. suen areas, the working envir:nment of the brigace created by the heat and smcke tuildup within the enclosure, will recuire the use of seif-centainec brectning apoaratus, smcke ventilation ecui; ment, and a personnel replacement capacility.

Cartain functions must be perfonned for all fires, i.e. , comand brigade actions, inform plant management, fire suppression, ventilation control, provide extra equipment, and account for pessible injuries. Until a site specific review can be completed, an interim minimum fire brigade si:e of five persens has been established. This brigade si:e shculd provide a minimum working number of personnel to deal with those postulated fires in a typical presently operating ccmercial nuclear pcwer station .

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If the brigac2 is composed of a smaller number of personnel, the fire Attack may be Sr. pped whenever new equipment is needed or a person is injured or fatigued. h'e note that in tfie career fire service, the minimum engine ecmpany manning considered to be effective for an initial attack on.a fire is also five, including one officer and four team members.

It is assumed for the purposes of this position that brigade training and equipment i.s adequate and that a backup capability of trained individuals exist whether through plant personnel call back or from the local fire department.

POSITI0l! .

1. The minic":m fire brigade shift si:o should be justified by an analysis of the plant specific f actors state,d above f.or the plant, after modifications are comp 1ste.

2. In the interim, the minimum fire brigade shift size 'shall be five pertons. These persens shall be fully cualified to perform their assigned resocnsibility, and shall include:

O_ne_Sujj,crvisor : This individual must have fire tactics training.

He wil l aswme all command res:ensibilities for fightine the fire. ~

During plant emergencies, the origade supervisor snould not have other mscensibilities tnat would detract from his full attention being devotec to the fire. This sucervisor sh uld not be activel'v engaged in the fighting of the fire. His total function should be to survey the fire area, ccm:Pand the brigade, and keec the upper levels of plant management informed. .

Two Ho:e Men - A 1.5 inch fire hose being handled within a window-less enclosurn would require tao trained incividuals. The tuo team memoers are renuired to chysically hancle the active hose line and to protect cach other while in the adverse environment of the fire.

. Two additienal Team Meners - One of these individuals would be recuir:c :o su:piy f11100 air cylinders to the fire fignting

. memaces of tne brigade and the sec nd to' estaclish.smcke ventilation and aid in filling tne air cylinder. These two individuals would also act as the firs- backup to the engaged team.

o - . . .

s, ATTACHMENT B 4 a. Assigruments of personnel wetino AM51 N18.1-1971 cualifications. Section 4.3.1 or Section 4.5.1. should be made to casite shif t coeratine enws wmoers not less than the following: ,

e For a station naving ofte licensed unit, each snift : Pew should have at least three persens at all tims, plus tw additional persons unen tne unit is operating.

For a eteltf aunf t staticn. eace snsf t crew snould nave at least three persons per licensed unit at all times, plus one additional person ser,0cerating unit.

b. 0 ertter license cualifications of :ersons assigned to acerating snif t crews snould be as follows:

(1) A licensed senior cotest:r anc is also a memser of the station sacervisory staff should se onsite at all times wnen at least One unit is teaced with fuel.

(2) For any station attn more than :na reactor c:ntaining fuel, (t) the numcer of licensec senice crerat:rs :nsite at all times snouic act se less inan the number of control reces from snien the fueled units art monit:rtd. and (2) tne numcer of Itcensed senior : erators sncule act :e tess than t3e numcer of reactors :erating.

(3) For esca react:r cntaining fuel, there snould te at least One licensed scerat:r in Me : ente:1 -:c:r at all times. Shift c e= ::me:sitions shoule se specified sucs inat tnis cancitten car :e satis #4ed inde:endently of licensed senice oce at:rs assignec to shift crews :: net tne criterta of (1) and (2) acove.

(4) For eacn control room from nien one or more reacters are in coeration. an additiona Ocerator should se ensite and ava114 ele to serve as relief operator for tnat ::ntrol reem. Shift crew :omectitices snculd :e s:ecified such nat inis ::ndition :an se satisfie ince:encently of (1). (2). and (3) and for eacn suen ::nte:1 room.

. Raciation :retectier ::.ali't:ati:ns of at least era :ersen en eacn :cerating snift saculd te as 'ollcws:

'he mar. age-e'it of eaca stati:n aaving ene :r :re unit s :: staining 'wei saculd eitner (1) avalify are :esi;nate at least one **:er :( eacn sm't ::erating crew to im;1emnt raciati:n :retection crecedures, including reuttre Or sc4Cial raciat10n surveys usi99 per*a:Ie radia**3n Oetec*:rs. Jsa :# P0tec*

tive barriers and signs, use yf :rere:tive :ictning anc :reatning a::aratus.

DerfcManct of n'aminatten surveys. che?hs On radiaticr. *cet**rs. and limits of exCosure Patas and ac:Ut"u l a t ec :: s e . O r ( 2 ) a s s i gn a r e a ; *.a. :ayst:s tec*r1:ian to eacn snift, sucn assignment t: :e in addition :: inose assi;*ec *: shift Cerating Orews in ac:gr0ance d1*n (a) anc (b) accve.

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Selecti:n ar: w asis :f vars:us asge e s :f ino sreas ::verec :y tris *ev'ew : tar .ilt :e made my the *evitt.er On esca :ase. *he hd;" eat :n t?e areas : 2e jiven attan*':r fue'*;

13.1.2*3

• 11/2:/75

APPENDIX A l 208th Meeting, August 11-13, 1977

2. Question:

The intent of this request was to be sure that open items identified for the PDA are addressed when the PDA is used in a construction per- .

mit application. The response is not clear on this matter; clarification is requested.

Response

The ACRS memorandum of August 28, 1978 clarified the intent of the original ACRS request made during the 208th meeting (August 11-13, 1977). The original request was misinterpreted in the staff response of April 28, 1978. The request is now understood to be a concern "that open items identified for the PDA are addressed when the PDA is used in a construction permit application".

Open items identified during a PDA review are addressed in the review of a construction permit application referencing a PDA or PDA application.

In most cases, the open items in a PDA review are resolved prior to the issuance of a PDA; for such items the resolution of the issue is applied to any and all construction permit application (s) (or other PDA application (s)) under review which reference the PDA application under review.

In some cases, a PDA may be issued subject to the resolution of '

certain issues, i.e., the resolution of an issue might not be com-pleted at the time of issuance of the PDA. An applicant referencing a PDA will be required to resolve any outstanding issue, within the applicant's scope of responsibility, as stated in the NRC Safety Evaluation Report (SER) pertaining to the original issue of the PDA referenced by the applicant. If one or more supplements to the SER have been issued, the requirements, if any, to be addressed by a referencing applicant would be identified in +.he supplement (or sup-plements) pertaining to the issuance of any and all amendments to the original PDA, during the effective period of the PDA.

For example, the application for a construction permit for the Phipps Bend nuclear facility (Docket Nos. 050-553,-554) addressed open issues then remaining.on the GESSAR 238 (Nuclear Island) PDA (No. 1).

- _ _________________m_ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

Similarly, applicants referencing the original BSAR-205 PDA (No. 12) will be required to address the matter of reactor cooldown using only safety-grade systems, as required by paragraph (6) of the PDA,

/

=

/

/

e 5

APPENDIX A 208th Meeting, August 11-13, 1977

3. Question:

According to our records we have not received a Staff response on this item.

Response

A response to this item was inadvertently omitted from our original memorandum. The original Committee request is as follows:

3. Dr. Bush requested that the NRC staff discuss its requirements for snubbers, the potential consequences from snubber failures, and methods for assuring that they will in fact work when needed.

The staff in its reports on snubbers at the August 11-13, 1977 ACRS meeting addressed all of the ACRS concerns. Further staff actions are presented in Task Action Plan A-13, " Snubbers".

The staff believes that the completion of this task action plan will resolve all outstanding concerns of the ACRS.

APPENDIX A l

209th Meeting, September 8-10, 1977

3. Question:

It is not clear whether the premises described by Portland G.E. with respect to. failure to isolate' containment is representative of the case in question. The intent of the inquiry was to obtain an assess-ment concerning the habitability of a control room with degraded containment capability subsequent to an accident where radiation re-leases are a variable. For example, the intent was to determine whether a large number of fuel cladding failures coincident with a LOCA and partially ineffective containment closure could influence the habitability of control rooms. One approach might be to consider the effects of 1% fuel clad perforation, 5% fuel clad perforation, and 50% fuel clad perforation as possible conditions coincident with a LOCA and incomplete containment as a way of assessing control room habitability contingencies. The Committee would appreciate a re-sponse on this matter.

Response

Control room habitability is reviewed by the staff for the case of a postulated LOCA using the source term of Regulatory Guides 1.3 or 1.4, coupled with the operation of engineered safety features designed to l mitigate the consequences of the event and assuming the containment is leaking at the design leak rate. The single failure criterion is invoked-in evaluating the performance of engineered safety features designed to mitigate the consequences of this event. Thus, where iodine removal sprays are employed for example, the evaluation assumes ,

1 out of the 2 spray trains fails to function. The radiological con-sequences are required to be within the criteria given in GDC 19 of Appendix A to 10 CFR Part 50. In comparing the hypothesized source term to that which is representative for a fuel failure of 10%, for example, the staff estimates that the hypothesized source term is approximately 250 to 500 times greater then that represented by failure of 10% of the fuel rods. On this basis, the staff concludes that realistic fuel failure rates would be within the GDC 19 limits even for containment leak rates significantly higher than the design leak rate.

i APPENDIX B 1

i 201st Meeting, January 6-8, 1977 l 2b. Question:

l The Committee recommends that the Staff provide guidance and a  ;

schedule for implementation of Reg. Guide 1.97.

Response

The staff currently is in he process of revising its approach towerd implementation of Regulatory Guide 1.97, Revision 1. A description of the present status and the proposed future course of action is pro-vided in the attached memorandum from R. H. Vollmer, dated October 12, 1978. As noted in the draft schedule, included with Mr. Vollmer's memorandum, we plan to discuss this matter further with the Committee.

It now appears that we could be prepared to meet with the Committee during its January 1979 meeting.

OCT121978 MEMORAtlDUM FOR: Domenic B. Vassallo, Assistant Director for Light Water Reactors, DPH ..

Robert L. Tedesco. Assistant Director for Plant Systems, DSS Brian K. Grimes, Assistant Director for Engineering and Projects, DDR Darrell G. Eisenhut, Assistant Director for Systems and Projects, 00R Frank Schroeder, Acting Assistant Director for Reactor Safety, DSS FROM: Richard H. Vollmer, Assistant Director for Site Analysis, DSE

SUBJECT:

IMPLEMENTATION OF R. G. 1.97 As you know we have been in the process of implementing R.G. 1.97, Rev.

1 for some time within the program defined by TAP A-34. This program initially envisioned the use of the lead plant concept to work out the details of the sometimes complex requirements of the regulatory guide. Our experience with this approach to date has not been fruitful although in the process, guidance has been developed upon which further efforts in implementat.on can be based. '

Because of the diffianties encountered in utilizing the lead plant concept, we are abandoning this approach and propose to proceed in a more straight-forward conventional manner as discussed in the attached outline. Please review this proposed approach and the enclosed draft schedule for implementation and provide your conraents to me by fMvember 1,..

1978.

/b (1/ f/Wh hRichardH.Vollmer,AssistantDirector for Site Analysis Division of Site Safety and .

Environmental Analysis -

Enclosures:

Distribution As stated Central File '"

AAB Reading ,

cc: R. DeYoung R. Mattson AAB File (TAP-A-34)

G. Chipman V. Stello e .n , . R. Vollmer M,,Ag.

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Proposed Course of Action for Implementation of R.G. 1.97 10/12/78 TaskActionPlanA-34was_developedtoprohideasystematicapproachto implementation of R.G. 1.97. It called for the use of the lead plant concept in working out the detailed requirements in two operating phases; -

one phase for position C-3 of the guide (instrumentation for beyond design basis events) and the other for implementation of position C-1 (instrumen-tation for design basis events). Our work to date in dealing with the selected lead plants has been unsatisfactory for developing guidance. As a result, we believe that implementation should proceed without further reliance on the lead plant concept. TAP A-34 is being revised to reflect a different approach as described below.

All plants will be required to implement position C-3 (except for C-3.d) on a reasonable time schedule. C-3.d is excluded because no current instruments are available which will fulfill the requirements needed to monitor the

large and variable releases for identifiable release points (C-3.d). A

, contract to determine the feasibility of and overall performance require-ments for such instrumentation will be let. The results of this contract will be utilized to provide appropriate criteria for implementation and backfit of position C-3.d.

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Beginning with the review of the3Haven application, we anticipate requiring applicants to provide the analysis required in position C-1. Our evaluation of these analyses will determine the specific instrumentation needs related to' design basis events. As there instrumentation needs are identified on

' current and future licensing reviews, a determination will be made concerning backfit of such instrumentation on operating plants.  ;

m- _

-A , . .,_.. .. . . . . , _ . . _ . s Draft Schedule for Implementation of R. G.-l.'97 Revise TAP .tcr reflect proposed approach, 11/78 l

Present approach.to ACRS. 12/78 Approherevised. TAP. 12/78 ImplementpositionC-1rehiewonHahen '

application. 12/78 Letter to applicants and licensees on all LWR plants requiring backfit of position C-3 (except C-3.d). 1/15/79 Let contract for feasibility study of instruments required.by position C-2.d and develop design criteria. 1/15/79 Required response _date by applicants - committment .

I and schedule for SAR submittal and installation. 3/15/79 Completedehelopmentofdesigncriteriafor position C-3.d instruments. 6/15/79 Letter to applicants and licensees on all LWRs requiring implementation of position C-3.d in accordance witn enclosed guidance. 8/1/79 ,

Completion date for position C-3.d implemented on all plants - to be developed.  ? -

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, *q UNITED STATES y

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NUCLEAR REGULATORY COMMISslON WASHINGTON, D. C. 20555

%*****/ OCT 2 51978 MEMORANDUM FOR: Domenic B. Vassallo, Assistant Director for Light Water Reactors, DPM Robert L. Tedesco, Assistant Director for Plant Systems, DSS Brian K. Grimes, Assistant Director for Engineering and Projects, D0R Darrell G. Eisenhut, Assistant Director for Systems anc Projects, D0R Frank Schroeder, Acting Assistant Director for Site Analysis, DSS FROM: Richard H. Vollmer, Assistant Directro for Site Analysis SUBJECT IMPLEMENTATION OF R. G. 1.97 My October 12, 1978 memo included an enclosure that described a proposed course of action for implementation of Regulatory Guide 1.97, Revision 1.

The proposal incorrectly recommended that we require applicants to provide the analysis required in Position C.1 beginning with the review of the Haven application. In fact, we propose that the analysis be required beginning with the New Haven application. Please make your comments based on this correction by November 1, 1978.

gb<u & dle d Richard H. Vollmer, Assistant Director for Site Analysis Division of Site Safety and Environmental Analysis cc: R. DeYoung R. Mattson V. Stello S. Varga K . Crocker

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APPENDIX B l

l 201st Meeting, January 6-8, 1977 I I

2d. Question: l The Committee wishes to be kept informed regarding the results of the Licensee's reliability study, the Staff's evaluation of it, the final fix required and its generic implications, if any.

Response

No further information is available at this time. The license requires submittal of an analysis and, if required, installation of the final fix at first refueling which is scheduled for February 1980.

When the information is available we will inform the committee as requested.

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APPENDIX B 201st Meeting, January 6-8, 1977 3a. Question:

The Committee desires information regarding stress levels for various structures and components required for safe shutdown and long-term cooling presented in such a manner that the margin against an in-crease in seismic stress can be determined.

Response

The staff, in a memorandum from E. Case to S. Lawroski dated June 14.

1978, stated that a report on the available seismic margin in the systems for safe shutdown and continued shutdown heat removal at North Anna Power Station Units No.1 and 2 would be prepared. This report is now scheduled to be sent to the ACRS during December 1978.

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d. .

APPENDIX B 203rd Meeting, March 10-12, 1977

2. Question: 1 The approach suggested by the Staff concerning auxiliary system reliability might be adquate, but there is insufficient descriptive information to provide a basis for judgment. It would be useful for the Staff to provide an illustrative example with ficitious data if no meaningful statistics are available as a way of displaying their approach to answering the question.

Response

An evaluatien of the reliability of auxiliary feedwater systems was also requested in a letter from R. F. Fraley to L. V. Gossick, dated July ll, 1978. As stated in the memo from H. R. Denton (September 26, 1978) in response to this request "the Division of Systems Safety is now in the process of initiating a technical assistance contract to evaluate the reliability of various auxiliary systems, such as the component cooling water system, the auxiliary feedwater system for PWR's and the steamline isolation valve leakage control system for BWR's. We expect to complete the preparation of the proposed work scope for this contract early in Fiscal Year 1979. At that time DSS will arrange to brief the Committee on the Program, and our expected schedule for completion".

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