ML20023B338

Forwards Div of Engineering Proposed Program to Address USGS Clarification Re Seismic Design Earthquakes on Eastern Us Seaboard.Proposed Program Will Complete Ongoing PRA Reviews

ML20023B338
Person / Time
Site:	Millstone
Issue date:	03/02/1983
From:	Vollmer R Office of Nuclear Reactor Regulation
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML20023B340	List: ML20023B338 ML20023B883
References
NUDOCS 8303170135
Download: ML20023B338 (91)

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t UNITED STATES o

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NUCLeiAR REGULATORY COMMISSION g.'

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v MAR 0 21983 MEMORANDUM FOR: Harold R. Denton, Director Office of Nuclear Reactor Regulation FROM:

Richard H. Vollmer, Director Division of Engineering

SUBJECT:

DIVISION OF ENGINEERING GEOSCIENCE PLAN TO ADDRESS USGS CLARIFICATION RELATING TO SEISMIC DESIGN EARTHQUAKES IN THE EASTERN SEAB0ARD OF THE UNITED STATES A plan for our proposed program to address the U. S. Geological Survey's clarification of position relating to seismic design earthquakes in the Eastern Seaboard of the United States is attached (enclosure 1). This plan elaborates on the outline provided as an attachment to a memorandum entitled, " Clarification of U. S. Geological Survey Position Relating to Seismic Design Earthquakes in the Eastern Seaboard of the United States", which was seat from the Executive Director of Operations to the Commissioners on November 19, 1982.

The plan is divided into two parts.

Part one is a short term probabilistic assessment utilizing an extensive new seismic hazard study currently being developed by Lawrence Livermore Natio'nal Laboratory.

Part two is a longer term detenninistic assessc:ent based primarily on long range ORES rcsearch with the possible need for utility sponsored investigations at some locations after an assessment of the long term research results. ' Additionally, we recomtend that an industry sponsored seismic hazard study be solicited.

We estimate that the effort to establish the seismic hazard level for the sites and make appropriate comparisons will take approximately three years to complete, utilizing staff resources of about.2.5-3.0 SY per year, and $300K per year in technical assistance funds. Our preliminary recommendations on which plants, if any, may need further evaluation should be completed in mid-1984 Because of the required res'earch effort, the deterministic element will not be synthesized until 1985.

The proposed program will complement ongoing PRA reviews and the seismic hazard spectra which are developed can also be used for future SEP evaluations. This program, therefore, is basically a continuation, with modification, of our ongoing work. This program does not include resources to complete a reevaluation effort for plants for which design.

spectra may need to be reevaluated. We recommend that this contingency be considered and included in the operating plan for FY 84. This plan also presupposes that our interim position for licensing reviews (enclosure 2) is found to be acceptable by ACRS and ASLB while we implement this pro 5 ram.

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l' There is evidence to support this assumption in the recent Appeal Board i

decisiononSummer(ALAB-710).

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.We have also assessed our ability to implement this plan under the existing regulation, Appendix A to 10 CFR Part 100. We have~ concluded i

that, although Appendix A itself does not explicitly recognize the use

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of probabilistic methods, as a minimum they can be used to assist in j

reaching deterministic judgements (Seabrook Remand, CLIB0-33).

It is j

not clear whether they can be used as the primary tool in setting f

appropriate ground motion levels. Therefore, we recomend that we i

j implement a limited modification or clarification of Appendix A as previously planned in conjunction with ORES as a parallel, yet j

independent effort, along with the Charleston plan. This modification i

has been recommended in SECY-79-300 and endorsed by the Siting Policy j

Task Force in NUREG-0625 and is necessary.to reflect the current state l

of art. This modification will require an additional 1.0 SY per year for -

2 years.

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We recomend that you consider placing this effort equally under three 1

resource areas - Operating Reactor Licensing Actions or Safety j

Technology, Systematic Evaluation Program for older operating plants, I

and Casework for ongoing OL review plants, i

i This plan has been developed as a result of extensive discussion within I

the Geosciences Branch, NRR; and discussions with the Earth Sciences j

Branch, ORES; and the U. S. Geological Survey.

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1 Richard H. Vollmer, Director Division of Engineering 1

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Enclosure:

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D. Eisenhut T. Schmidt l

R. Vollmer W. Russell i

R. Mattson R. Bernero H. Thompson GSB Staff l

J. Knight Z. Rosztoczy 1

R. Jackson P. Williams 1-L. Reiter j

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Enclosure'1 4

i Recommended Plan Eastern U. S. Eart'hiiuakes Introduction i

i On November 18, 1982, the U. S. Geological Survey (USGS) forwarded a i

letter to the Nuclear Regulatory Commission clarifying their past i

i position with respect to the 1886 Charleston earthquake. The USGS j

letter states that:

f "Because the geologic and tectonic features of the Charleston region are similar to those in other regions of the eastern l

seaboard,'we conclude that although there is no recent or i

historical evidence that other regions have experienced strong earthquakes, the historical record is not, of itself, sufficient grounds for ruling out the occurrence in these other regions of l

strong seismic ground motions similar to those experienced near Charleston in 1886. Although the probability of strong ground motion due to an earthquake in any given year at a particular:

j-location in the eastern seaboard may b_e very low, deterministic and probabilistic evaluations of the seismic hazard should be made. for 4

individual sites in the eastern seaboard to establish the' seismic s

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engineering parameters for critical facilities."

j We have evaluated the USGS clarification of position and have concluded j

that it can be addressed predominantly through existing programs at NRC l

with the ' possibility of additional requests for utility - sponsored work. We recommend that a two part program be implemented which will i

address both the deterministic and probabilistic elements mentioned by J

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l Part 1 of the proposed program is a short term probabilistic assessment i

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of plants in the eastern seaboard. This part of the plan is necessary i

because many of the current tectonic working hypotheses are not amenable to investigation by deterministic methods in the short term.

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Part 2 of the proposed program is a longer term deterministic assessment of the causes of large earthquakes, such as the Charleston earthquake, in the eastern seaboard. Specific areas of relatively high se'ismicity 4

and tectonic structures are identified which we recommend be addressed through the ORES long range research plan.

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Based on our evaluation of the research results, some applicants or.

i licensees may be required to investigate tectonic structures which may j

not have been previously identified during the l'icensing procedure.

i Part 1 - Probabilistic Assessment Discussion The November 18, 1982 letter from the USGS represents not so much a new

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understanding but rather a more explicit recognition of existing uncertainties with respect to the causative' structure and mechanism of the 1886 Charleston earthquake. Many hypotheses have been proposed as to the locale in the eastern seaboard of future Charleston-size earthquakes. Some of these could be very restrictive in location while otners would allow this earthquake to recur over very large areas.

Presently, none of these hypotheses are definitive and all contain a strong element of speculation.

Traditional detenninistic' approaches such as that outlined in Section 2.5.2 of the Standard Review Plan are not generally designed to-deal 9

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with this situation.

Probabi11stic metnods which allcw for the consideration of many hypotheses, their associated credibilities, and the explicit incorporation of uncertainty are much better equipped to provide rational frameworks for decision making. The question that needs to be answered is:

Taking uncertainties into account, have licensing decisions for

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plants in the eastern seaboard (i.e., in the region affected by the USGS clarified position on the Charleston Earthyene) resulted in acceptable levels of assumed seismic hazard (exposure to earthquake I

ground motion) at the individual sites?

One means for answering the above question-is a probabilistic assessment of seismic hazard at all nuclear power plant sites east of the Rocky Mountains. Since adequate or acceptable levels of seismic hazard have not been explicitly defined in probabilistic terms, it is assumed that the probability of seismic ground motion exceeding design levels implicitly associated with licensing decisions based upon traditional

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methods in other regions of the U. S. east of the Rocky Mountains is adequate; these other regions include areas such as the Central Stable Region and the Gulf Coastal Plain. The prime tool for carrying out this assessment is an updated version of the Uniform Hazard fieth'odology developed for the Systematic Evaluation Program by Lawrence Livermore National Laboratory (LLNL) and its subcontractor TERA Corporation. This methodology relies upon the incorporation of diverse expert opinion with regard to the input parameters needed to make probabilistic estimates.

4 As'such, it does not' rely up,on single hypotheses which do not account fo' existing uncertainties but rather attempts to incorporate the r

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. s 4-hypotheses'and their uncertainties into the computations.

Identification of plants (if any) in the eastern seaboard at which the probability of exceeding design-level ground motion is signif'icantly greater than has been ass.umed at other locations may result in an integrated seismic evaluation and/or engineering rearalysis to assure the plant's ability to withstand a more severe earthquake. This study

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may also identify selected plants outside of the eastern seaboard whose design levels may be inappropriate, relative'to other plants, with respect to the seismic hazard.

In addition, we are also initiating, through a technical assistance contract, a study to better estimate grcund motion f*om a large earthquake the size of the 1886 Charleston event to gain a better understanding of how this ground motion should be represented..

Major Activities - Probabilistic Assessment The probabilistic assessment portion of the proposed program is divided into the following elements.

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1. January thru April 1983 - Continue development of LLNL study including expert opinion surveys on seismic hazard east of the Rocky Mountains. This study (Seismic Hazard Characterization of the-Eastern U.S.) is presently underway as a joint effort of NRR and t a o

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5-i ORES. No additional resources above those already allocated are needed.

2. May 1983 thru December 1983 - Calculation of seismic hazard spectra by LLNL for all nuclear power plant sites (approximately

75) east of the Rocky Mountains. An estimation of the probability of scismic ground motion exceeding the design level et each site, taking into account specific site conditions, will be completed and provided as a report. An additional 2.0 SY is needed for LLNL and 0.3 SY for NRC effort during this period.

3. September - December 1983 - Comparison of LLNL study with 4

4 existing probabilistic studies such as-Algermissen and others (1982). An additional 0.2 SY is needed for LLNL effort.

4. March 1983 - December 1983 - Sponsorship by the utilities of a probabilistic estimation of seismic hazard for all nuclear power plants east of the Rocky Mountains. This study, while not a requirement, is strongly recommended so as to complement the LLNL study and provide another independent assessment of seismic hazard.

An additional 0.1 SY needed for LLNL and 0.1 SY for hRC effort.

5. December *983-March 1984 - Using LLNL and other studies, the NRC staff will integrate this information and make comparisons of the, probability of seismic ground motion exceeding design levels in the eastern seaboard with probabilities calculated at plants in the rest of the Eastern and Central V. S.

Comparisons will be made in several ways including comparison.by. region alone and by region and plant vintage. Plants _in the eastern seaboard (if any) that are associated with significantly greater hazard than those elsewhere.

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will then be identified. Other comparisons may be needed, but will be decided upon after review of initial results. An additional 0.7 SY is needed for NRC effort.

6. April 1984-September 1984 - Assessment of initial conclusions regarding hazard in light of feedback from expert opinion on original input. A final letter report will be issued with a final

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recommendation on plants which need reevaluation. An additional 0.2 SY needed for LLNL and 0.2 SY for NRC effort.

7. January 1983-December 1983 - Ground motion estimates at different distances and site conditions from a large Charleston type earthquaka. Both theoretical and empirical estimates using data from recent earthquakes will be made.

This study is presently being initiated through a technical assistance contract with LLNL.

No additional resources are required.

Status summary reports of research into probabilistic estimates of

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seismic hazard funded by ORES will be needed by December 1983 so as to incorporate them into task number 5.

Implementation of Probabilistic Assessment Results The implementation of results is outlined above in elements 5 and 6.

NRR Staff and Cpst Re_quirements Probabilistic Assessment The additional effort req' ired for this portion of the program will be u

2.5 SY for LLNL (1.9 in FY 83, 0.6 in FY 84) and 1.3 SY for NRC (0.3 in FY 83, 1.0 in FY 84). This' staff effort can be accommodated with the currently available resources in the Geosciences Branch because this e

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7-program complements ongoing st'af f activities and may replace other staff activities for individual sites.

This program does not include resources to complete the seismic evaluation and/or engineering reanalysis which some plants may require as a result of the I

probabilistic elements.

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Utility-Sponsore'd Study in Conjunction with the Probabilistic Assessment j

A recommended utility-sponsored study is outlined abpve in element 4.

Schedule - Probabilistic Assessment The proposed schedule for implementing this plan appears in Table l.

i Part 2 - Deter:11nistic Assessment 1

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j The deterministic portion of the proposed program is. designed to better 3

understand the causes of large earthquakes, such as the Charleston earthquake, in the eastern seaboard. This effort may require some j '

expansion of immediate and long term ORES programs. Increased j

understanding of the cause of seismicity _in the eastern seaboard will l

allow a reduction in th'e uncertainty in estimating the seismic hazard for nuclear power plants. The primary problem with seismic hazard i-

. characterization in the eastern seaboard.is that no causative mechanism for seismicity has been identified to date and no surface offsets due to l

earthquakes are known. Although there are literally thousands of crustal structures k'nown in the eastern, seaboard, which, if they were i

active, could produce strong earthquakes, none have been demonstrated to have been active during the Quaternary (the last two million years)_ or l_

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proved to be capable. The result is that, to date, there has been no generally accepted association between eastern seismicity and crustal structure.

The overall approach of the deterministic assessment is to study areas of relatively higher seismicity in the eastern seaboard to determine if tectonic features and processes responsible for the seismicity can be identified and correlated. This will be pursued by crustal studies at' hypocentral depths to determine if there is any correlation teiween crustal structures at hypocentral depths and the earthquake hypotenters.

The primary tool for determining crustal structure at hypocentral depths will be the use of multi-channel seis.nic reflection profiles. The i

primary tools for locating the hypocenters will be the cor.tinued monitoring and analysis of earthquakes from the existing microearthquake nets. These e.ets will have to be maintained and upgraded in order to improve depth locations of hypocenters if there is to be an improved ability to correlate between hypocenters and tectonic structures at depths of up to 25 kilometers.

This research will be contracted and monitored by 0RES, and does not represent a radical departure from past programs.

Increased l

ccordination between NRR and ORES will be required, however, to better define the problems that are to be resolved in order to improve our understanding of eastern seismicity in the licensing context. This portion of the program is designed to improve our ability to assess the adequacy of the design of nuclear facilities on the eastern seaboard.

The result, in part, wil1'be summary reports which will represent the current status of research including a review and synthesis of available e

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data. These results will be 'used to modify, if necessary, conclusions drawn from the probabilistic studies and identify individual features, if appropriate, for assessment by utilities.

Major Activities - Deterministic Assessment The deterministic assessment portion of the proposed program is divided into the following elements appropriate to each region listed.

A. Charleston Region Since the causative mechanism of the Charleston earthquake of 1886

. continues to be one of the primary unresolved problems in evaluating seismicity in the eastern seaboard, research in the Charleston area should centinue with the goal of testing the various hypotheses as to the caase of the earthquake.

In particular,' emphasis should be placed on determining if suggested features such as the Ashley River and Woodstock Fault zones constitute the source zores of the Charleston earthquakes.

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May 1983

" Workshop on the 1886 Charleston Earthquake and Its Implications 'for Today" - the U. S. Geological Survey and the scientific community will present a summary and evaluation of the tectonics and seismicity at Charleston.

2.

September 1983 - ORES in consultation with the U. S. Geological Survey and the scientific community should have a program in place to t'est the most likely, tectonic hypothesis for seismicity.

3.

June 1984 - ORES presents the esul'ts of the program e

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w of testing the highest-weighted hypothesis.

4.

January 1985 - ORES presents summary report describing the results of the Charleston work testing the highest-weighted tectonic hypothesis.

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B. Ramapo Fault Zone The Ramapo Fault Zone, a Precambrian fault zone that was intermittently active until the Mesozoic,' is the northwestern boundary of the Newark ~

Triassic Basin. Low level seismicity occurs in the area and may be associated with the fault zone, however, the seismicity in the region

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forms a band 40 kilometers wide. Detailed field work and limited trenching and core drilling suggest that the Ramapo Fault has not been I

recently reactivated. The purpose of studying the fault is to establish whether there is a causal relationship between Mesozoic or older faults i

such as the Ramapo Fault and current seismicity in this area by determining the location and geometry of these faults at hypocentral depths.

1.

April 1983 - ORES initiates a new evaluation of the Ramapo Fault. The study should include culti-channel seismic reflection profiling and other geophysical techniques such as in-situ stress measurements and geodetic measurements to determine the current state of stress at hypocentral depths.

2.

January 1984 - ORES presents preliminary results of the program to date, and plans for the coming year.

3.

January 1985 - ORES presents summary report on this aspect of the Ramapo Fault Study including the identification an'd

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analysis of any seismic source zones.

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C. Central Virginia Seismic Z'ne o

Recent work by earth scientists at Virginia Polytechnic Institute have suggested that there may be a relationship between the seismicity in Central Virginia and the northeast trending thrust faults and deco 11ement of the Piedmont crust of the Appalachian Orogenic Belt. The purpose of this part of the program is to continue evaluation of the

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relationship between the faults and the earthquakes.

j 1.

April 1983 - ORES presents a plan for undertaking the seismic reflection profiling, and applying other geophysical techriiques such as geodetic measurements and_in-situ stress measurements.

2.

January 1984 - ORES presents the preliminary results or progress to date, and plans for the coming year.

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January 1985 - ORES presents a summary report on the -

the Central Virginia Study including the potential identification and analysis of any seismic source zones.

D. Giles County, Virginia The Giles County Seismic Zone is a northeast trending linear zone of seismicity which apparhntly is located beneath the decollement and thrust faults associated with the Valley &' Ridge Province of the Appalachian Orogenic Belt.

It has been suggested that the seismic zone has occurred as a reactivated northeast trending normal fault associated with the opening of the Proto-Atlantic.(called the Iapatus) in the late Proterozoic and early Paleozoic (800-500 million years ago).

' pril 1983 - ORES initiates plannin'g for the proposed research.

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August 1983 - ORES initiates study of the Giles County structure using seismic reflection profiling.

3.

April 1984 - ORES presents preliminary results and' plans for the coming year.

4.

April 1985 - ORES presents sumnary results of this phase of the research including the potential identification and analysis of

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any seismic source zones.

E. New England The research in New England has been underway for several years and will be continued. Increased emphasis should be placed on evaluation of the i

source mechanism for the New Brunswick and_Gaza, N.H. earthquakes, the neotectonics of seismically active areas, and the orientation and magnitude of the stress field in the seismically active areas of the region. An in-situ stress measurement at hypocentral depths will be conducted at Moodus. Depending on the results of the seismic reflection studies described above, additional seismic reflection surveys may be conducted in seismically active areas of New England such as Moodus, Connecticut; New Hampshire; Massena, New York and New Brunswick, Canada.

1.

April 1983 - ORES completes plans for stress measurement at Moodus.

2.

August 1983 - Conduct stress measurements at Moodus.

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3.

April 1984 - ORES presents preliminary results of stress measurements and their relationship to the local seismicity and tectonics.

4.

January 1985

.0RES presents summary results of stress measurements and other studies described above.

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Implementation of Deterministic Assessment Results As the results from the deterministic studies become available, they will be evaluated, and, the effect, if any, on operating plants and plants in the Operating License stage of review will be determined. The need for additional evaluations of particular structures by utilities will be assessed as the information becomes available. Two problems will Le addressed by the deterministic portion of the program; (1) whether"or not the deterministic findings warrant any reassessment of the conclusions drawn from the probabilistic study; and (2) whether there are any particular tectonic structures which are associated with or similar to tectonic structures associated _with seismicity which, because of their proximity to individual sites, should be analyzed by the utilit-ies. The above effort will take about two to three years (early 1985) to complete. The impact of this research on nuclear power plants will be determined by the NRC staff with technical assistance contracts, if necessary.

NRR Staff and Cost Reouirements - Deterministic Assessment This effort will require continuous communication among NRR, ORES and the contractors. As 'research funds are limited and the amount of time is short, careful interaction will be necessary to obtain'the information required to allow a resolution of eastern seismicity.

It is estimated that one staff year per year for three years will be necessary for NRR to implement this detern.inistic part of the overall plan.

The research effort will be funded by,0RES and technical assistance c'ontracts will be funded by NRR.

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deterministic assessment, $200,000 may be required to implement the NRR technical assistance program to determine the impacts of the findings on the nuclear facilities in the eastern U. S.

Utility-Sponsored Studies as P.esult of the Deterministic Asessment During FY 1983 no deterministic work by the utilities is currently recommended, beyond that necessary to pursue their normal efforts to continue to assess any hazards identified by them for their sites.

After the results of the research are available and if any source zones are identified which have particular importance to specifi': sites or have impact on the probabilistic program, some utilities rray be required to investigate structures in the vicinity of their plants.

Schedule - Deterministic Assessment The proposed schedule for implementing this plan follows as Table 1.

Our ability to meet this proposed schedule may be somewhat optimistic and is contingent on implementing the appropriate contracts. We will be better able to assess this schedule when the work has been initiated.

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. REFERENCE I

Algermissen, S. T., D. M. Perkins, P. C. Thenhaus, S. L. Hanson, and B. L. Bender, 1982, Probabilistic Estimates of Maximum Acceleration and Velocity in Rock in the Contiguous United States, United States Department of Interior, Geological Survey. Open-File Report 82-1033, 99 p.

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Calendar Year Schedule for Probabiilstic an_d Deterministic Seismic Hazard Program 1983 1984 1985 Part 1 Short Term Meet with ACRS to Meet with ACRS to discuss Program to discuss Preliminary Recommendations

1. Update LLNL Seismic

+ Complete Methodology Hazard Methodology

2. Calculate Seismic

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+ Report with Spectra e

Hazard Spectra for

. Eastern Sites 3, Ccmpare with other

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+ Report with Comparisons available probability studies

4. Initiation of Industry-

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+ Production of Study Results Sponsored Seismic Hazard Study I

Letter, Report with Preliminary Recommendations 5.. Comparison of Seismic

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+ Final Recommendations Hazard at Sites Initate Feedback Assess Impact on

6. Assessment of Impact of

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+ Previous Results Expert Feedback Initiate Tac with LLNL

7. Charleston Ground Motion

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+ Issue Report Study v

Table 1

Calendar Year Schedule for Probabilistic and Deterministic Seismic Hazard Program 1983 1984 1985 Part 2 Long Tenn Meet with ACRS to Meet with ACRS to discuss Program.

to discuss Preliminary Recommendations Workshop -

Interim Progress Report on Rasults of Synthesis Hypothesis Testing Testing

1. Charleston Research

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Initiate Preliminary Sumary Study Report Report

2. Ramapo Fault Research

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Initiate Preliminary Sumary Study Results Report Report

3. Central Va. Research

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Initiation Preliminary Summary RFP of Study Results -

Report

4. Giles County, Va.

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Research Stress Measurements Conduct Preliminary Sumary Plan Measurements Results Report

5. New England

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Sumarize Review Preliminary Evaluation Summary of of Determinsitic 2

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of Results of RES Source Zones Work

6. Assessment of Impact of

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Deterministic Studies on' Sites Table 1 (cont'd) 4 g

Interim Position on Charleston Earthquake for Licensing Proceeding The NRR Staff position with respect to the Intensity X 1886 Charleston earthquake has been that, in the context of the tectonic province approach used for licensing nuclear power plants, this earthquale should be restricted to the Charleston vicinity. This position was based, in part, on information provided by the United States Geological Survey (USGS) in a letter dated December 30, 1980 from J. E. Devine to R. E.,

Jackson (seeSummerSafetyEvaluationReport).TheUSGShasbeen reassessing its. position and issued a clarification on November 18, 1982 in a letter from J. E. Devine to R. E. Jackson. As a result of.this letter, a preliminary evaluation and outline for NRC action was i

fomarded to the Comission in a memorandum from W. J. Dircks on

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November 19, 1982.

The USGS letter states that:

"Because the geologic and tectonic features of the Charleston region are similar to those in' other regions of the eastern seaboard, we conclude that although there is no recent er historical evidence that other regions have experienced strong earthquakes, the historical record is not, of itself, sufficient grounds for rulinT out the occurrence in these other regions of stron'g, seismic ground motions similar to those experienced near Charleston in 1886. Although the probability of strong ground

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motion due to an-tarthquake in any, given year at a particular location in the eastern seaboard may,be very low, deterministic and probabilistic evalu'ations of the seismic hazard should be made for individual sites in the eastern seaboard to establish the seismic engineering parameters for critical facilities."

~ Yhe USGS clarification represents not so much a new understanding but

, rathe'r a more explicit recognition of exNting uncertainties with respecito the causative structure and mechanism of the 1886 Charleston earthquake. Many hypotheses have been kroposed as to the locale in the

, eastern Taaboard'of future Charleston-size earthquakes. Some of these

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could be very restrictive in location while others would allow this earthquake to recur over very la ge areas.

Presently, none of these hypotheses are definitive and all contain a strong element of a

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speculation.

We are addressing this uncertainty in both longer-tenn deterministic and shorter-term probabilistic programs. The deterministic studies, funded primarily by the Office of Research of the NRC should reduce the uncertainty by better identifying (1) the causal mecnanism of the Charleston earthquake and (2) the potential for the occurrence of large earthquakes throughout the eastern seaboard. The probabilistic studios, primarily that being conducted for NRC by Lawrence Livermore National Laboratory (LLNL) will take into account existing uncertainties. They will have as their aim to determine differences, if any, between the probabilities of seismic ground motion exceeding design levels in the eastern seaboard (i.e. as affected by the USGS clarified position on the Charleston earthquake) and the probabilities of seismic ground motion exceeding design levels elsewhere in the central and eastern U. S.

Any plants where the probabilities of exceeding design level ground motiors are significantly higher than those calculated for other plants h

in the Central and Eastern U. S. will be identified and evaluated for possible further engineering analysis.

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Given the speculative nature of the hypotheses with respect to the recurrence of large Charleston-type earthquakes as a result of our limited scientific knowledge and the generalized low probability associated with such events, we do not see a need for any acticn for 9

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specific sites at this time.

It is our position, as it has been in the l

past, that facilities should be designed to withstand the recurrence of

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.an earthquake the size of the 1886 earthquake in the vicinitf of -

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j Charleston. At the conclusion of the shorter-term probabilistic program and during the longer-term deterministic studies, we will be assessing the need for a modified position with respect to specific sites..

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l REQUEST FOR ADDITIONAL.INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 l

l DOCKET NO. 50-423 I

230.0 GEOSCIENCES BRANCH. SEISM 0 LOGY SECTION 230.3 (SRP 2.5.2) RSP On November 18, 1982 the USGS in a letter from James F.' Devine, USGS to l

Robert Jackson, NRC, clarified its position regarding the localization of the seismicity in the vicinity of Charleston, S.C.

The staff is presently evaluating the significance of the USGS clarification regarding the localization of Charelston seismicity.

Attached are copies of the staff's interim position on the Charleston Earthquake and our recommended plan to address Eastern U. S.

Earthquakes. This position will be included in the Safety Evaluation Report.

230.4 (SRP 2.5.2.2, 2.5.2.3, 2.5.2.4)

In licensing decisions made since approximately 1976, regarding the seismic design basis of nuclear power plants located in New England and the northern Piedmont, the staff has recognized the New England-Piedmont Tectonic Province. On January 9, 1982 a magnitude 5.7 earthquake occurred in south central New Brunswick, Canada, in geologic terrain that is similar to that which characterizes the New England Piedmont Province. As discussed in FSAR sections 2.5.1.1.4 and 2.5.2.2 and shown in figure 2.5.2-10 the northern Appalachian region is sub-divided into a number of tectonic provinces, which is different than the New g

England-Piedmont Province. With respect to the appropriate chcice of tectonic provinces and the effect of the New Brunswick earthquake on the site, two options, either of which would be generally acceptable to the staff, can be chosen to resolve the above difference. We will also review any other approaches that are suggested.

[

Option A:

Assume that the site is located in the New-England Piedmont Tectonic Province and that the m = 5.7 New Brunswick earthquake is the

~

b maximum historical earthquake as defined in Appendix A 10CFR100 for this-province. Calculate a site specific spectra using an m = 5.7 as the s

target magnitude. This can be accomplished by collectiHg a suite of strong motion response spectra recorded on-rock sites, within distances of less than about 25 kilometers, for magnitudes of 5.7 +.50.

Three such collections are currently available, although the target magnitudes are in some cases larger than 5.7 or the set of spectra do not include strong motion recordings from recent earthquakes. Two spectra were-completed by Lawrence Livermore National Laboratory, the most recent of which was completed for the Seabronk site, and the third spectra-was completed by the Tennessee Valley Authority as part of the Sequoyah review. Although one of the above could be used, the staff would reccomend that development of a spectra specifically for an mb = 5.7 using the most recent information that is available.

Option B:

Extensive research is under way regarding the New Brunswick l

earthquake and its relationships to the New England-Piedmont Province.

l A large portion of this effort has been undertaken as a result of I

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l REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATI0'N UNIT 3 DOCKET NO. 50-423 230.0 GEO5CIENCES BRANCH, SEISM 0 LOGY SECTION (CONT'D) reviews of the Seabrook and Maine Yankee sites. We recommend active l

attention and awareness of these studies. Using information provided by l

these studies, update ano provide a complete discussion regarding the current choice of tectonic provinces. Include as a minimum the following information:

1)

A discussion and justification of any association of the Central New Brunswick earthquake sequence with a specific geologic structure or fault within the meaning of Appendix A 10CFR100.

2)

A discussion and justification of any province sub-division with respect to the New England Piedmont Tectonic Province.

3)

An estimate of the ground motion and response spectra resulting from any province sub-division. Both peaks and spectra should be l

compared to that of the SSE. We note that the relationship of Murphy and O'Brien (1977) was used to arrive at a peak acceleration of 0.10g from an Modified Mercalli Intensity VII.

It has been the staff's position to use the " trend of mean" of the relationship in Trifunac and Brady (1975) coupled with a Regulatory Guide 1.60 response spectrum, when intensity is used to describe the SSE.

In addition, in recent OL reviews the staff has requested that the comparison of site specific spectra using the magnitude of the maximum historical earthquake which has not been associated with a fault or structure.

It has been the staff's position that a Modified Mercalli VII corresponds to a m =5.3 (Nuttli and Herrmann 1978).

Rocksitespecificspectraareahailableforuse,however the staff recommends that you develop a spectra specifically for an l

m =5.3 using as much recent information as possible. The existing hrock site specific spectra are discussed in the Wolf Creek (NUREG-0881), Perry (NUREG-0887) and Catawba (NUREG-0954) staff Safety Evaluation Reports.

The staff recommends that a meeting be held to specifically discuss these questions and the discussed options.

230.5 (SRP2.5.2.1) l Update the FSAR to consider all pertinent seismologic information that have been developed in the region since publication of the FSAR. The most recent published seismologic reference in the FSAR bibliography is 1979. Considerable seismological research has been done in the northeastern U. S. since that time.

Evaluate this information and determine whether or not it is significant to the seismological analysis of the site. This update should be completed using Standard Review Plan section 2.5.2 (NUREG-0800 July 1981) and should include as a minimum the following:

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REQUEST FOR ADDITIONAL'INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET N0. 50-423 230.0,GEOSCIENCES BRANCH, SEISM 0 LOGY SECTION (CONT'D) a)

Update Table 2.5-2-3 to include all earthquakes having Modified Mercalli intensity greater than IV or magnitude greater than 3 which have been reported to date in an area within 200 miles from i

the site.

Include the seismic data provided by the Northeastern U. S. Seismic Network.

b)

Update Table 2.5-2-4 to include all recorded and or-felt i

earthquakes, to date, within a 50 mile radius of the site.. Include applicable data provided by the Northeastern U. S. Seismic Network.

i c)

Provide a complete discussion of the 1981 microearthquake swarm sequence near Moodus, Connecticut and assess the significance of these events with respect to the OBE and S5t.

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i REQUEST FOR ADDITIONAL INFORMATION 1

MILLSTONE NUCLEAR POWER PLANT, UNIT 3 DOCKET NO. 50-423 1

PLANT PERSONNEL TRAINING 630.0 LICENSEE QUALIFICATION BRANCH l

l 630.2 Discuss the program which will provide the training to Reactor Operators j

and Senior Reactor Operators in the following areas:

(a) Recognition of emergency conditions.

(b) Classification of observed emergency cor.ditions in accordance with the Emergency Classification System.

(c) Notification of emergency to off-site authorities.

(d) Recorcaendation of protective actions to off-site authorities.

(e)

Direction of station staff to take protective actions.

(Ref. NUREG-0800, Sections 13.2.1.I.B.1 and 13.2.1.II.1.b) 630.3 Provide the outlines of the courses, Fundamentals of Nuclear Training

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and Nuclear Plant Training.

(Ref. NUREG-0800, Section 13.2.1.I.B.1) 630.4 With respect to the simulator training, provide the following infor-mation:

(a) The details of the program in accordance with the guidelines as specified in the Regulatory Guide 1.149.

(b) Discussion of the certification examination provided to demon-strate the candidate's ability.

(Ref. NUREG-0800, Section 13.2.1.1.B.2 and 12.2.1.11.2) 630.5 Discuss the qualification of the training instructors in the training program and the requalification program administered to the instructors in order to have them remain certified as instructors as specified in of H. R. Denton's March 28, 1980 letter to all power reactor applicants and licensees and in Item I. A.2.3 of NUREG-0737.

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 l

DOCKET NO. 50-423 l

630.0 LICENSEE QUALIFICATION BRANCH - continued 630.6 Discuss the certification completed pursuant to Sections 55.10 (a)(6) and 55.33(a)(4) and (5) of 10 CFR Part 55.

Provide the title of the individual who will certify the eligibility of individuals for licensing l

or renewal of license.

(Ref. Enclosure 1 of H. R. Denton's March 20, 1980 letter, Section A.3) l 630.7 Provide a commitment to comply with the following TMI-related require-ments as specified in Item I. A.2.1 of NUREG-0737:

l (a) As an operating license applicant, Millstone 3 is not subiected to the one year experience requirements for cold license SR0 candi-dates. However, after one year of station operation, we will require Millstone 3 to comply with the one year experience require-ment for hot license SR0 applicants.

(b) The requirement for three months on shift experience for control room operators and SR0 candidates as an extra person on shift is not required for cold license candidates and, hence, is not appli-cable to Milistone 3.

thwever, we will require Millstone 3 to comply with this requirement for hot license candidates after three months of station operation.

l 630.8 Provide a detailed training program for mitigating core damage (as described in Item II.B.4 of NUREG-0737) in accordance with the guidance as specified in Enclosure 3 of H. R. Denton's letter dated March 28, 1980. Provide a listing of those individuals and their qualifications who must participate in the training program and provide a schedule for that training as related to the presently scheduled fuel load date.

630.9 Provide a detailed description of the training program for the Shift Technical Advisor in accordance with the guidance as specified in NUREG-0737, Appendix C.

630.10 With regard to fire brigade training, the program for drills should be revised to include all the guidelines as described in NUREG-0800, Section 13.2.2.II.C.6. A.iii(b).

630.11 Describe the training program provided for individuals (non-licensed l

operators) permitted to operate systems or equipment independently that could affect the quality of structures, systems, and components important to safety.

(Ref.NUREG-0800,Section13.2.2.IV.2) l

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1 REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET NO. 50-423 250.0 MATERIALS ENGINEERING BRANCH, Inservice Inspection Section Review of the FSAR and request for additional information regarding preservice (PSI)/ inservice (ISI) inspection program.

250.2 To complete our review, we will require the following (5.2.4) information:

(6.6)

(1)

A preservice inspection plan.

(2)

All requests for relief with a supporting technical justification.

(3)

An inservice inspection plan submitted six months after licensing.

250.3 Plans for preservice and inservice examinations of the reactor (5.2.4) pressure vessel welds should address the degree of compliance witn Regulatory Guide 1.150, Rev. 1, as required by Generic Letter 83-15, dated March 23, 1983.

250.4 Section 6.6.3 of the Millstone 3 Nuclear Power Station FSAR (6.6) references ASME Code Section XI, Winter 1975 Addenda, and Paragraph IWA-2200 of the Summer 1976 Addenda which has not been referenced in 10 CFR 50.55a(b).

Note that 10 CFR 50.55a(b) states that when applying the 1974 ASME Section XI Code Edition, only the Addenda through Summer 1975 may be used.

If Appendix III is used it must be used in conjunction with Summer 1978 Addenda or later Addenda as referenced by 10 CFR 50.55a(b).

When using Appendix III of Section XI for preservice or inservice examination of either ferritic or austenitic piping welds the following should be incorporated:

A.

Any crack-like indication, regardless of amplitude, l

discovered during examination of piping welds or adjacent base metal materials should be recorded and investigated by a Level II or Level III examiner to the extent necessary l

to determine the shape, identity, and location of the reflector.

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_ REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET NO. 50-423 4

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250.0 MATERIALS ENGINEERING BRANCH, Inservi_ce Inspection Section - continued Confirm that high energy fluid system piping between containment l

isolation valves will receive an augmented examination as follows:

i.

1 A.

Protective measures, structures, and guard pipes should not l

prevent the access required to conduct the inservice examina-l tion specified in the ASME Code,Section XI.

s-1 B.

For those portions of high energy fluid system piping between J

containment isolation valves, the extent of inservice examina-

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tion completed during each inspection interval (ASME Code

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Section XI) should provide 100% volumetric examination of j-circumferential and longitudinal pipe welds within the i

boundary of these portions of piping.

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C.

For those portions of high energy fluid system piping enclosed in guard pipes, inspection ports should be provided l

in the guard pipes to permit the required examination of circumferential pipe welds.

Inspection ports should not 4

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be located in that portion of the guard pipe passing through the annulus of dual barrier containment structures.

j D.

For those items requiring ISI, a baseline or preservice examination for establishing the integrity of the original condition is also required by the ASME Code.

Confirm that the augmented examination for high energy system piping is maintained throughout the entire piping system up to f

the outboard restraint.

If the re'straint is located at the isolation valve, a classification change at the valve interface is acceptable, j

Confirm that welds between outboard containment isolation valves and piping restraints are included in the PSI and ISI program j

plan as required.

j 250.5 FSAR Sections 5.2.4 and 6.6 state that the "preservice/ inservice 1

(5.2.4) inspection program has been developed using the criteria of the 1

(6.6)

ASME Code,Section XI, 1974 Edition, Summer 1975 Addenda along.

i with existing construction drawings as they are issued.

A l

PSI /ISI program will be finalized and submitted to the NRC pursuant to 10 CFR Part 50, at which time relief requests will be identified as necessary."

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Indicate.the anticipated date for submittal of.this information i

.and all requests for relief from impractical examinations.

The j

Preservice Inspection Program should include the following i

j information:

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REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET N0. 50-423 250.0 MATERIALS ENGINEERING BRANCH, Inservice Inspection Section - continued B.

The Owner should evaluate and take corrective action for the disposition of any indication investigated and found to be other than geometrical or metallurgical in nature.

In later editions of the ASME Code, Appendix III of Section XI, is specified for ferritic piping welds.

If this requirement is not applicable (for example, for austenitic piping welds), ultra sonic examination is required to be conducted in accordance with the applicable requirements of Article 5 of Section V, as amended by IWA-2232.

Discuss the criteria for applying Article 5 of Section V, as amended by IWA-2232.

Provide a technical justifi-cation for any alternatives used such asSection XI, Appendix III, Supplement 7, for austenitic piping welds and discuss the following:

A.

All modifications permttted by Supplement 7.

B.

Methods of ensuring adequate examination sensitivity over the required examination volure.

C.

Methods of qualifying the procedures for examination through the weld (if complete examination is to be considered for examinations conducted with only one side access).

250.6 Clarify the statemnt in the Millstone FSAR Section 6.5, Augmented (6.6)

Inservice Inspection to Protect Against Postulated Piping Failures, which states " Welds in certain portions of high energy fluid system piping will receive supplemental examination."

High energy lines within the " break exclusion" of the containment penetration area, whether encased in guard pipes or not, must receive augmented preservice/ inservice examination regardless of the requirements of Section XI of the ASME Code as discussed in SRP 3.6.1 and 3.6.2.

However, high energy lines meeting the

" modified break exclusion region" criteria need not be subjected to augmented preservice/ inservice examination.

The " modified break exclusion region" criteria may be applied in those special cases in which guard pipes are necessary, and it has been demonstrated to the satisfaction of the NRC that access to perform an examination is extremely difficult to achieve.

In such areas the examination requirements may be eliminated provided the guard pipe is designed for the full dynamic effects of a longitudinal or circumferential break of the enclosed process pipe including jet impingement, pipe whip impact and environmental effects.

If the high energy fluid system piping does not meet the

" modified break exclusion region" criteria, submit the required augmented preservice/ inservice examination program for this piping.

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REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR PnWER STATION, UNIT 3 DOCKET NO. 50-423 250.0 MATERIALS ENGINEERING BRANCH, Inservice Inspection Section - continued A.

For ASME Code Class 1 and 2 components, provide a table similar to IWB-2600 and IWC-2600 confirming that either the entire Section XI preservice examination was performed on the component or relief is requested with a technical justification supporting the request.

B.

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

C.

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 nt,t receive a complete-Section XI preservice examination including 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 volunetric examination was performed from one side of the weld, discuss whether the entire weld volume and the heat affected zone (HAZ) and base metal on the far side of the weld we e examined.

State the primary reason that a specific examination is impractical; e.g., support or component restricts a:: cess, fitting prevents adequate ultrasonic coupling on ofe side, component-to-component weld prevents ultrasonic examination, etc.

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

Detailed guidelines for the preparation and content of relief requests are attached as Appendix A to these questions.

250.7 Confirm that you will comply with NUREG-0452, Rev. 4, (5.4.2.2)

Standard Technical Specification, which states that the PSI will be performed on 100% of the length of all tubes.

This supersedes the guidance in R.G. 1.83 Rev. 1.

APPENDIX A f

GUIDANCE FOR PREPARING REQUESTS FOR RELIEF FROM CERTAIN CODE REQUIREMENTS PURSUANT TO 10 CFR 50.55a(g)

A.

Description of Requests for Relief The guidance in this enclosure is intended to illustrate the type and extent of information that is necessary for " request for relief" of items that cannot be fully examined to the requirements of Section XI of the ASME Code.

The preservice/ inservice inspection program should identify the examination and pressure testing requirements of the applicable portion of Section XI that are deemed impractical because of the limitation of design, geometry, radiation considerations or materials of construction of the components.

The request for relief should provide the information requested in the following section of this appendix for the examinations and pressure tests identified above.

B.

Request for Relief from Certain Examination and Testing Itequirements Many requests for relief froa. examination and/or testing requirements submitted by che Applicants or Licensees have not been supported by adequate descriptive and detailed technical information.

This detailed information is necessary to:

(1) document the impracticality of the ASME Code requirements within tM limitations of design, geometry, and materials of construction of components; and (2) determine wiether the ne of alternatives will provide an arwptable level of quality and safety.

Relief request (s) submitted with a justification such as " impractical",

" inaccessible", or any other categorical basis, require additional information to perrait an evaluation of that relief request.

The objective of the guidance provided in this section is to illustrate the extent of the information that is required to make a proper evalua-tion and to adequately document the basis for granting the relief in the Safety Evaluation Report.

Subsequent requests for additional information and delays in completing the review can be considerably reduced if this information is provided in the initial relief request submittal.

For each relief submitted, the following information should be included:

1.

An identification of the component (s) and/or the examination requirement for which relief is requested.

2.

The number of items associated with the request relief.

3.

The ASME Code class.

4.

An identification of the specific ASME Code requirement that has been determined to be impactical.

5.

The information to support the determination that the requirement is impractical; i.e., state and explain the basis for requesting relief.

. 6.

An identification of the alternative examinations that are proposed:

(a) in lieu of the requirements of Section XI; or (b) to supplement examinations performed partially in compliance with the requirements of Section XI.

7.

A description and justification of any changes expected in the overall level of plant safety by performing the proposed alternative examinations in lieu of the examination required by Section XI.

If it is possible to perform alternate examinations, discuss the impact on the overall level of plant quality and safety.

For inservice inspection, provide the following additional information regarding the inspection frequency:

1.

State when the request for relief would apply curing the inspection period or interval (i.e., whether the request is to defer an examination).

2.

State when the proposed alternative examinations will be iaplemented and performed.

3.

State the time period for which the request for relief is needed.

Technical justification or data must be submitted to support the relief request.

Opinions without substantiation that a change will not affvt the quality level are unsatisfactory.

If the relief is requested for inaccessibility, a detailed description or drasing which depicts the 1

inaccessibility must accompany the request.

A relief request is not required for examinations and/or tests prescribed in Section XI that not apply to your facility.

A statemc.it of "N/A" (not applicable) i or "None" will suffice.

C.

Request for Relief for Radiation Considerations l

j Exposures of personnel to radiation to accomplish the examinations prescribed in Section XI of the ASME Code can be an important factor in determining whether, or under what conditions, an examination must be performed.

A request for relief must be submitted in the manner described above for inaccessibility and must be subsequently approved by the NRC Staff.

Some of the radiation considerations will only be known at the time of the examination or test.

However, from experience at operating facilities, the Applicant or Licensee generally is aware of those areas where relief will be necessary and should submit as a minimum, the following informa-tion with the request for relief:

1.

The total estimated man-rem exposure involved in the examination or test.

. l 2.

The radiation levels at the examination or test area.

L 3.

Flushing or shielding capabilities which might reduce radiation l

levels.

t i

j 4.

A proposal for. alternate examination techniques.

1 5.

A discussion of the considerations involved in remote examinations.

6.

Similar welds in redundant systems or similar welds in the same systens which can be examined.

7.

The results of preservice examinution and any inservice results for l

the welds for which the relief is being requested.

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8.

A discussion of the failure consequences of the weld which was not examined.

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REOUEST FOR ADDITIONAL INFORMATION' MILLSTONE Nt; CLEAR POWER STATION, UNIT NO.~ 3 Docket No. 50-423 451.0 METB-Meteorology Section 451.1 Provide a meteorological data tape of onsite measurements in accordance (SRP 2.3.3) with the attached meteorological data tape format. The data should in-clude at least two consecutive annual cycles including the most recent one-year period as described in Regulatory Gaide 1.70, Section 2.3.3.

451.2 Describe any proposed supplemental meteorological monitoring, on or (SRP 2.3.3) near the site, to aid in characterizing effluent transport during the occurrence of sea / land breeze circulations.

460.0 MEiB-Effuent Treatment Systems Section 460.5 Table 6,5.1-1 of SR? 6.5.1, providing guidance on minimum instrumenta-(SRP 6.5,1) tion of ESF atmospheric cleanup systems, and Position C.2.g of Regulatory Guide 1.52, call for recorded ind,ication of flow rates in the control room. Sections 6.5.1.5,' 9.4,' 1:8, and 1.9 of the FSAR indicate that ne flow rate instrumentation is included aQd that flow can be estimated by fan curves verified every 18 months. The staff does not consider these fan curves as an acceptable alternative for instrumentation. Provide the necessary flow rate instruments for indication and recording on the four ESF systems (Section 6.5.1),

including filtered and unfiltered flow operation.

460.6 Acceptance Criterion II.A.6 and Position C.2.c of Regulatory (SRP 11.3)

Guide 1.140 call for remote recorded indication of flow rates in

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET N0. 50-423 460.0

,METB, EFFLUENT TREATMENT SYSTEMS SECTION (CONT'D) 460.6 nonnal ventilation systems.

Sections 11.3, 9.4, and 1.8 of the FSAR (SRP 11.3) excludes flow rate instrumentation. In order to assure representative nionitoring/ sampling by the off-line monitors on the nonnal ventilation systems, the staff reconmiends that flow rate instrumentation be pro-vided in lieu of fan curves.

Provide your justification for not including this instrumentation in the design.

460.7 Acceptance Criteria, Requirement II.a calls for meeting the Positions (SRP 11.1) in Regulatory Guide 1.110.

Section 11A of the FSAR does not provide the cost of borrowed money. Provide the cost of borrued money ex-pressed in percent that was used in your cost-benefit analysis fer Appendix I to 10 CFR 50.

460.8 Acceptance Criteria, Requirement II.b calls for meeting the Positions (SRP 11.1) in Regulatory Guide 1.112.

Section 11.3.3 of the FSAR states that the release points are indicated on Figur'e h2-1.

Clarify if this should be Figure 1.2-2.

Compare Figure 1.2-2 with Figug 2.1-5 of the ER.

Provide the infonnation requested in Appendix B to Regulatory Guide 1.112, item 6d, for the height and location relative to adjacent structures. Include the turbine building, warehouse, steam relief vents, and outside tank vents, for example. Provide some detail on how the release lines lead to the stack and the reactor plant vent.

460.9 In cur review, Subsection 1.1 of SRP 11.3, the staff has located (SRP 11.3) several P&ID differences frot. the description provided in the FSAR.

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET N0. 50-423 460.0 METB, EFFLUENT TREATMENT SYSTEMS SECTION (CONT'D) 460.9 Provide clarification on the following:

(SRP 11.3) a)

The flow direction of the reactor plant gaseous waste and the condenser air removal lines to the radioactive waste line on Figure 9.3-5 (Sheet 1) at K-9 disagrees with Figure 10.4-2 (Sheet 2) at K-2.

b)

The waste tank inlet and outlet are connected on Figure 9.3-6 (Sheet 1) at K-4 or Figure 11.2-1.

c)

The BRS inlet line EM-109A at E-10 is not shown on Ficurc 9.3-9 (Sheet 1) at B-2.

d)

Figure 11.5-1 indicates monitor 3HVR-REID. Should this be 3HVR-RE10A and for IOC, as given on Figure 3.8-62 (Sheet 4)?

e)

Figure 9.4-3 (Sheet 2) at I-1 indicates ventilation release to Unit No. 2.

How is this lir.e monttored prior to gas erlaase at Unit No. 3?

460.10 Acceptance Criterion II.B.6 requires special provisions for radioactive (SRP 11.3) gaseous wastes that have the potential for a hydrogen explosion.

Sections 11.3.2.2, 9.3.5.2 and 15.7.2, describing the use of the GWS degassifier on input to the BRS, does not clearly state that hydro-genated reactor coolant can not be collected in the BRS holdup tanks.

Your description of the BRS tanks does not indicate provisions, such as diaphragms on the holdup and test tanks, to assure that hydrogen gas can not mix with the air above the liquid with an open vent to the atmosphere.

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET N0. 50-423 460.0 METB, EFFLUENT TREATMENT SYSTEMS SECTION (CONT'D) 460.10 a)

Provide all possible flow directions fer the four 3-way valves on (SRP 11.3) the input to the BRS'shown on Figure 9.3-4 (Sheat 1).

b)

Provide an analysis of 3-way valve and control failure for the valving in a) above.

c)

Describe how you plan to mitigate explosive gas mixtures of hydrogen and air in the BRS components and vents.

460.11 Branch Technical Position ETSB 11.5 attached to SRP 11.3 provides (SRP 11.3, BTP-ETSB the staff position for the analysis of the consequ>nces of a failure 11.5) in the waste gas system.

Section 15.7.1 provides an analysis of the waste gas system based on lir.e rupture upstream of the adsorbers.

Table 1.9-2 states that an amendment is proposed. Clarify your position or provide a date for the additional information.

460.12 Regulatory Guide 1.52 and 1.140 recommend leak testing of dampers (SRP 6.5.1, 11.3) used in ESF Jnd non-ESF air filtrat' ion syst: ems.

In Table 1.8-1, pages 21 and 55, you have taken exccption to testing evggy damper, and propose to test every type of damper.

Since leakage is a func-tion of valve size, we recomnend that you determine Class B leakage rates for at least one damper of each size and type used in the ESF and non-ESF atmospheric cleanup systems, as an acceptable alternative.

460.13 SRP 11.4 calls for a description and design bases for solid radio-(SRP 11.4) active waste handling systems. Table 1.8-1, page 17, of the FSAR, l

states that the charcoal adsorber in filtration trains will be I

t REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET N0. 50-423 460.0 METB, EFFLUENT TREATMENT SYSTEMS SECTION (CONT'D) 460.13 replaced using an external vacuum system.

Provide the description, (SRP 11.4)

P& ids, bases, and details relative to this equipment and describe the provisions _ for handling the charcoal adsorber media from the trains to the SWS.

460.14 SRP 11.4 calls for a description and design bases for solid radio-(SRP 11.4) active waste handling systecs.

Although Table 11.4-1 includes spent 4

resins from the cor.densate demineralizers, we find no provisions fcr f

handling these spent resins (Section 10.4.6.2).

Provide the informa-tion, design bases and the appropriate P&ID figure (s).

463.15 Acceptance Criteria II.6 of SRP 11.3 calls for meeting the Positions (SRP 11.31 of Regulatory Guides 1.140 and 1.143 for gaseous radwaste systems including normal ventilation subsystems that operate during normal operation, including anticipated operational occurrences.

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Sections 9.4.8.2.1 and 11.3 do not address the design bases for the condensate demineralizer regeneration and drain systeisiin the Warehouse No. 5.

The evaporator feed tanks and the neutralization f

sumps, for example, vent to the room air and, therefore, do not meet GDC 64.

What provisions are included to prevent GWS waste from being drawn into the distillate tank. Provide the information, design bases and consider changes to the P&lDs.

460.16 SRP 11.5 requires additional review of monitoring / sampling instru-(SRP 11.5) mentation at the FSAR-OL stage, as indicated by Subsection 1.2.

i

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET N0. 50-423 460.0 METB, EFFLUENT TREATMENT SYSTEMS SECTION (CONT'D)

~

460.16 In order for the staff to complete a review for Millstone, Unit (SRP 11.5)

No. 3, we need the following infomation for the instrumentation described in Sections 11.5, Tables 11,5-1 and 11.5-2, and the figures.

a)

Describe the monitors by type (e.g., in-line, off-line, etc.).

b)

Describe the particulate monitor detector (11.5.2.2.9).

c)

Specify redundancy, where applicable. Describe range switching.

d)

Clarify tag numbers vs. plant name.

Clarify use of letters A end B for different monitor detectors and different ranges or redundant monitors.

e)

Locate the monitors by P&lD and building layout figure.

f) locate the sampling points and sampling stations.

g)

Describe the actions performed (manual or automatic) by signal from the monitor (s).

h)

Describe the calibration laboratory and the sample analysis laboratory (i.e., location, purp'ose', facilities) for calibrating, repairing and testing monitors or providing radiochemical analysis.

1)

Provide a composite figure of the liquid process and effluent monitors listed in Table 11.5-2.

j)

Specify the sample locations for iodine adsorber sampling devices per Table 1 of SRP 11.5.

460.17 The Subsection 11.4.2.4 of the FSAR states that a Process Control (SRP 11.4)

Prcgram (PCP) will be implemented. Provide a commitment to submit

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET NO. 50-423 460.0 METB, EFFLUENT TREATMENT SYSTEMS SECTION (CONT'D) d60.17 the PCP for solidification and packaging of wet solid waste at least (SRP 11.4) six months prior to fuel loading, for review of the Technical Specification 3.11.3.

Provide an acceptable reference, such as a Topical Report, for the moble solidification system shared with Unit Nos.1 and 2, and the dewatering equipment for spent resins.

We need this infonnation to confirm compliance with the BTP ETSB 11-3, attached in SRP 11.4, i

~

2nd to confinn your design bases (11,4.l. item 4) in the FSAR.

460.18 Describe the hydecgen and oxygen monitor shown on Figure 11.3-2 (SRP 11.3)

(Sheet 2) at X-5.

Provide the instrament and readout location, range, i

set point, sensor checks, and calibration.

Will the instruments be nonsparking and capaMc of withstanding a hydrogen explosion as required by Acceptance Ceiterion 6 of SRP 11.3?

460.19 Section 11.5.2.5 refers to the use of glass sampling bulbs. Experi-(SRP 11.5) ence at many nuclear power plants indicates that it is not prudent to use glass containers for collecting and transferring radioactive

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samples due to breakage of the container. Justify your position for using glass sampling bulbs.

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Density (1600 BPI - 800 will be accepted)

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mgnet ic tapes with unformatted or spanned records.

At the beginning of each tape, use the first five (5) records (whicn Include is the equivalent of ten cards) to give a tape description.

plant name, and location (latitude, longitude) dates of data, in-formation explaining data contained in the *other' fields if they are used, height of measurements, and any additional informaticn Make sure all five records per tinent to idatification of the tape.

are included, even if some are blank. Format for the first five records will be 160A1. Meteorological data format is (I6, I2, I3, 14, 25FS.1, FS.2, 3FS.1). Decimal points 'should not be included when copying data onto the tape.

All data should be given to a tenth of a unit e}c'ept solar radiation which should be given to a hundreth of a unit.

mis does not necessarily indicate the accuracy of the data.

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All nines in any field indicates a lost record (99999). All sevens in a wind direction field indicates calm (77777).

If If only two levels of data, use the upper t. lower levels.

only one level of data, use the up;cr level.

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FS.1,

Wind Speed (meter /sec)

.F5.1 Sigma 2 eta (degrees)

FS.1 Ambient Temperature (*C)

F5.1 Hoisture:

F5.1 Other: _

w F5.1 Intermediate Measurements: Irvel = meters F5.1 _

Wind Direction (degrees)

F5.1 Wind Speed (meters /sec)

FS.1 Sigma meta (degrees)

F5.1 Ambient wmperature (oC)

_ FS.1 misture:

F5.1 Other:

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FS.1 Wind Speed (meters /sec)

F5.1 Sigma 'Iheta (degrees)

F5.1 Ambient ' Ibm,mrature ("C)

F5.1 misture:

F5.1,

Other:

FS.1 Terp Diff (Upper-Lower) ("C/100 meters)

FS.1 Wr.p Diff (Upper-Intermediate) ( C/100 mters)

F5.1 Te.Tp Diff (Intermediate-7.nwer) ( C/100 meters)

FS.1 Precipitation (m)

FS.2 Solar Radiation (cal /cm2/ min)

F5.1 Visibility (km)

F5.1 Other:

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ij REQUEST FOR ADDITIONAL INFORMATION s _

MILLSTONE NUCLEAR POWER STATION UNIT 3 s

DOCKET N0. 50-423 i

l 430.0 POWER SYSTEMS BRANCH, ELECTRICAL 1

i 430.3 Criterion '50 of Appendix A to 10 CFR 50, IEEE Standard 485, Regulatory (SRP8.1)

Guide 1.63 and branch technical positions ICSB 4, PSB-1 and PSB-2 have not been identified in Table 8.1-2 of the FSAR; thus, a positive

.]

statement as to compliance with these criteria and staff guidelines i

has not been provided in the FSAR.

Provide a statement of compliance and justify areas of noncompliance.

430.4 There are four transmis.sion lines between Millstone and Hunts Brook 4

(SRP 8.2)

Junction that follow a common right-of-way.

It is the staff position

~

j i

that no other transmission lines cross over-these four lines and that l

the four lines be physically separate and independent so that no single event such as a tower falling or line breaking will be sble to simultaneously affect all circuits in such a way that none of the four l

9 circuits can be returned to service in. time to prevent fuel design 3

l limits or design conditions of the reactor coolant pressure boundary from being exceeded. Line cross overs.and physical separation of I

these four transmission lines has not been describad in the FSAR.

Provide the description and justify areas of nonconpliance with the above staff position.

430.5 The Millstone design provides two immediate v.ces offsite circuits

{

(SRP S 2) between the switchyard and the 4.16 kv Class 1E busses.

It is the staff position that these two circuits be physically separate and independent such that no single event can simultaneously affect both circuits in such a way that neither can be returned to service i

in time to prevent fuel design limits or design conditions,of.the i

reactor coolant pressure boundary from being exceeded. The physical H

separation and independence of these two circuits has not been described or analyzed in the-FSAR.

Provide the~ description and analysis and justify areas of noncompliance with the'above staff position. The analysis should include separation and independence of control and p*otective relaying circuits as well as the power

]

circuits.

430.6 The Millstone design arrangement provides tw'o immediate access offsite I

(SRP8.2) circuits. One of.these circuits utilizes a generator circuit breaker to isolate the turbine generator from the. main and normal station service transformers. Other facilities that utiliie generator.

~

circuit breakers have been required to parform verification testing.

Provide a verification test program with-results to demonstrate the breaker's ability to perform its inte~nded function during steady-state

, operation, power system transients, and major faults. (Seeadditional guidelines for this question in Attachment 1.)

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~~' REQUEST FOR ADDITIONAL INFORMATION

//

MILLSTONE' NUCLEAR POWER STATION UNIT 3 DOCKET N0. 50-423

~

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,430.0 POWERSYSTEMSPRANCH, ELECTRICAL"{CMD_),

N n

a.

It is the staff @sition that the Hillstone grid stability

',430.7 (SRP 8.2) analysis must show that loss of the large'st single supply to

^~

the grid does not result in the, complete loss of preferred s

power. The analysis should c'onsider the. loss, through a single event, jf,the largest caracity being supplied to the grid, removal of the largest load from the grid, or loss of the most critical tehnsmission line. The combined capacity m of Millstone Units 1, 2, and T is'to, be supplied to the grid 1

through the komman Millstone switchyard. The combined capacity of the threeliits appears to be the largest capacity being

' supplied to the grid and should-be considered in the Millstone grid stability analysis.

Provide the results of the grid stability analysis when simultaneous loss of the combined capacity of Units 1, 2,and 3-foconsidered and justify areas

~

of noncoiapliance with the.above staff position.

'b.

There are four transmission circuits that connect the Millstone switchyard to the grid system. "The four circuits are routed on two tower lines - two circuits per tow'er line. Section 8.1.3 of the FSAR indicates that a simultaneous failure of eitner of the two tower' lines with only one circuit in service on the other' ther line, may result in instability of Millstone generation. The applicant, in order to prevent instability, has installed a protection scheme to automatically

' ~

redu'ce generato*-output at Millstone Unit 3.

Describe the

,. f protection scheme.

~/

~

430.8 Each of the 4.l6'kV Class-1E busses at Millstone.is supplied power (SRP 8.2 from preferred offsite and standby onsite circuits.

It is the

&8.3.1) staff position that these circuits should not have common failure modes.

Physical separation and independence of these circuits has not been described or analysed,in the FSAR. ~ Provide a description and analysis in accordance with*Section 5.2.l(5) of IEEE Standard 308-1974.

430.9 Section 8.3.1.1.4 of the FSAR iridicates that a; degraded voltage (SRP 8.3.1, scheme with two-out-of-four logic is provided on each of the 4.16 kV Class lE buses.

Provide reference tb electric schematic Appendix.

8A) drawings that describe the degraded voltage scheme and provide a-description, with voltage and time sit'oints, to indicate how p

the Hillstone design complies with the guidelines of-position 1

, of branch technical position PSB-1 (NUREG-0800 Appendix 8A) and

' provide justificat,lon for.any deviations.

~

C 430.10 As stat'ed in' Section 8,3hl.l.3 of'tbe FSAR, th'e emergency generator (SRP8:3.1, load sequences (EGLS)'has the capability to automatically reset Appendix du' ring-a sustained low v51tage condition on the essential bus.

8A)

It is the staff. concern that this capability may unnecessarily

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REQUEST FOR ADDITIONAL INFORMATION MILLSTONE ~ NUCLEAR POWER STATION UNIT 3 DOCKET NO. 50-423 430.0 POWER SYSTEMS BRANCH, ELECTRICAL (CONT'D) i i

delay the connection of the required mitigating loads within the l

times allowed by the accident analysis. Address the staff concern, j

describe the design of the EGLS for automatic reset during sustaine.d low voltage conditions, and describe how the design-meets position 2 of branch technical position PSB-1 (NUREG-0800, 4

Appendix 8A) and justify areas of noncompliance with position 2.

430.11 The voltage levels at the safety-related loads should be optimized (SRP 8,3.1, for the maximum and minimum load conditions that are expected j

Appendix throughout the anticipated range of voltage variations of the

}

8A) offsite power sources. Perform a voltage analysis and verification by actual measurement in accordance ::nh the guidelines of positions 3 and 4 of branch technical position PSB-1 (NUREG-0800, Appendix 8A).

Provide the voltage at the terminals of each Class lE load as determined by analysis and by actual measurement for all modes of plant operation. Verify that all Class 1E loads will operate at or within design voltage limits under all condition of operation.

Where tenninal voltage determined by analysis is not adequate to meet the design voltage rating of the equipment, provide justification, 430.12 Provide the results of a reliability analysis for the solid state (SRP 8.3.1) load sequencer that demonstrates that overall reliability or capability of the onsite power system to supply power to safety loads on demand has not been significantly reduced by the use of solid state load sequencers.

430.13 Section 8.3.1.1.3 of the FSAR indicates that diesel generator

' (SRP 8.3.1, protective relaying is bypassed under accident condition in Appendix accordance with branch techr,1 cal position ICSB 17. Provide 8A) drawing reference numbers that describe the design of the bypass circuitry, the 2-cut-of-3 logic circuitry, and relaying that is not bypassed under accident conditions.

430.14 Section 8.1.7 of the FSAR indicates that the diesel generator (SRP 8.3.1) voltage (prior to connection of the first load block) may drop below the 75 percent minimum level permitted by position 4 of Regulatory Guide 1.9 (Revision 2).. Provide' justification for this exception to Regulatory Guide 1.9 and correct inconsistency between statements of compliance found in Sections 1.8 and 8.3.1.2.6 of the FSAR.

430.15 Section 1.8 of the FSAR indicates that the Millstone design does (SRP 8.3.1) not comply with position C11 of Regulatory Guide 1.9 Revision 2.

Describe and justify the areas of noncompliance.

430.16 Section 1.8 of the FSAR indicates that the Millstone desian does (SRP 8.3.1) -

not comply with position C2(a)4 of Regulatory Guide 1.108:' The applicant has implied that the diesel generator load shedding test will be conducted using the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating for rejection of the single largest load and the continuous rating for complete loss of load. Justify use of continuous versus 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating for

+

complete loss of load.

x

.. - - - - -. =..

I, REQUEST FOR ADDITIONAL INFORMATION-MILLSTONE NUCLEAR POWER STATION UNIT.3 DOCKET N0. 50-423 I

430.0 POWER SYSTEMS BRANCH, ELECTRICAL (CONT'D) 430.17 Section 1.8 of the FSAR indicates that the Millstone design does (SRP 8.3.1) not comply with positions C2(a)2 of Regulatory Guide 1.108. The capability of the diesel generator to accept the design accident j

load sequence is to be demonstrated under conditions as close to design as 'possible. Provide clarification and justification.

j 430.18 Section 1.8 of the FSAR indicates that the Millstone design does 1

(SRP 8.3.1) not comply with position C2(a)3 of Reguitory Guide 1.108.

It

]

appears that the full load carrying capability of the diesel i

generator may not be tested for the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating. Justify l

not testing at the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating and define the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />

[

rating of the diesel generator at Millstone 3.

i j

430.19 Section 6.4.2 of IEEE Standard 387-1977 requires,in part, that j

(SRP 8.3.1) the load acceptance test consider the potential effects on load i

acceptance after prolonged no load or light load operation of the diesel generator.

Provide the results of load acceptance tests or analysis that demonstrates the capability of the diesel i

generator to accept the design accident load sequence after i

prolonged no load operation. This capability should be demonstrated over the full range of ambient air temperatures that may exist at the diesel engine air intake.

If this capability cannot be derconstrated for minimum ambient air temperature, conditions, describe design provision that will assure an acceptable engine air intake temperature during no load operation.

2 l

430.20 In accordance with Section 5.6.2.2(1) of IEEE -Standard 387-1977, SRP 8.3.1)

Section 5 of IEEE Standard 338-1977, and position C2a(8) of i

Regulatory Guide 1.108, it is the staff position that the diesel generator, when in the test mode and parallel with the offsite power system, be capable of responding to an accident signal.

l Describe how the Millstone design meets the staff position and justify areas of noncompliance.

430.21 The FSAR does not provide a complete description of how design 4

l (SRP 8.3.1) criteria, testing. and analysis is being implemented for the onsite power system diesel generator at Millstone.

Provide the description that as a minimum addresses each section~ o'f IEEE Standard 387-1977 as supplemented by Regulatory Guide 1.9 and i

each section of Regulatory Guide 1.108 I

430.22 Section.8.3.1.1.3 of the FSAR describes the surveillance I

(SRP8.3.1 instrumentation provided to monitor the status of the diesel i

Appendix generator. Expand the FSAR to describe how the Millstone. design 8A) complies with the guidelines of branch technical position PSB-2 i

,(NUREG-0800 Appendix 8A) and provide justification for any 1

deviations.

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REQUEST FOR ADDITIONAL INFORMATION ftILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKEY N0. 50-423 430.0 POWER SYSTEf1S BRANCH, ELECTRICAL (CONT'D) 430.23 The FSAR does not provide a complete description of how physical (SRP 8.3.1, separation of Class lE systems will be accomplished at Millstone 8.3.2)

Unit 3.

Provide a description that as a minimum addresses each section of IEEE Standard 384-1974 as augmented by Regulatory Guide 1.75.

The description should include but not be limited to the following items.

1.

Separation between redundant Class lE and between Class 1E and non-Class lE cables located in cabinets and control switchboards.

2.

Separation of actuated Class lE equipment.

3.

Separation of sensors and sensor to process connections.

4.

Separation of cable entrances to control switchboards.

5.

Identification of cables inside cabinets.

6.

Separation between Class lE conduit and non-Class lE cable trays.

7.

Separtion between redundant and between Class 1E and non-Class 1E terminations 8.

Compatibility with mechanical systems 430.24 You imply (by taking exception to position 05 of Regulatory Guide (SRP 8.3.1, 1.75) that the Millstone design for cable separation does not 8.3.2) meet the minimum separation distances spe:ified in IEEE Standard 384-1974.

Identify each circuit and the location where it does not meet the minimum separation distance.

Provide an analysis for each location identified that demonstrates the adequacy of the lesser separation.

430.25 In Section 1.8:of the FSAR, you identified the following exception (SRP8.3.1, or clarification to position C4 of Regulatory Guide 1.75:

8.3.2)

" Associated circuits are identified by the same color code as the Class lE circuit with which they are associated. This color code exists up to and including an isolation device."

Position C4 of Regulatory Guide 1.75 requires that associated circuits (up to and including an isolation device) be subject to all requirements placed on Class 1E circuits unless it can be demonstrated that the absence of such requirements cannot significantly reduce the availability of Class lE circuits. The applicants clarification or exception inplies that associated circuits meet only the color code requirement versus all requirements of Class lE circuits. Provide justification for the implied exception to position C4.

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REQUEST FOR ADDITIONAL INFORMATION l

MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET NO. 50-423 430.0 POWER SYSTEMS BRANCH, ELECTRICAL (CONT'D) 430.26 In Section 1.8 of the FSAR, you identified an exception to (SRP 8.3.1, position C10 of Regulatory Guide 1.75.

Class 1E cables are to be 8.3.2) marked at intervals not exceeding 15 feet versus 5 feet as required by the Regulatory Guide.

In justification of the exception you documented, that the 5 foot requirement is e typographical error in Regulatory Guide 1.75 which has been confirmed by the NRC. The staff does not consider the 5 foot regirement to be a typographical error. A 5 foot maximum marking distance is considered necessary i

to facilitate easy visual verification that the cable installation is in confonnance with separation criteria. Justify noncompliance with the 5 foot marking requirement for cables located in raceways as well as inside panels or cabinets.

430.27 In section 1.8 of the FSAR, you imply taking exception to Section i

(SRP8.3.1, 5.1.4 of IEEE Standard 384-1974 Where plant arrangements 8.3.2) preclude maintaining the minimum separation distance, you state that redundant circuits will be routed in non-solid raceways.

Solid versus non-solid raceways are required by Section 5.1.4 of IEEE 384.

Provide clarification and justification for noncompliance.

430.28 In Secticn 1.8 of the FSAR, you imply taking exception to position (SRP 8.3.1, C12 of Regulatory Guide 1.75.

Position Cl2 indicates that 8.3.2)

(1) power supply feeders to instrument and control room distribution panels installed in enclosed raceways should not be considered acceptable, (2) traversing power circuits separated from other circuits in the cable spreading area by a minimum distance of l

3 feet and barriers should not be considered acceptable, and j

(3) traversing power circuits routed in imbedded conduit which in l

effect removes them from the cable spreading area should be considered acceptable.

Power circuits that traverse the cable spreading area at Millstone are installed in enclosed raceways (rigid steel conduit).

In accordance with position C12 of 'Regulatiry Guide 1.75, the routing should not be considered acceptable. Justify the adequacy of the proposed routing in steel conduit.

430.29 In Section 1.8 of the FSAR, you imply taking exception to position J

(SRP 8.3.1, C8 of Regulatory Guide 1.75.

Redundant Class 1E cables or Class 1E 8.3.2) and non-Class lE cables are to be routed separately within the same confined space such as the cable tunnel at Millstone that is effectively unventilated.

It is the staff concern that routing of cables along opposite sides of a confined space may.not provide adequate separation.

Identify all confined spaces at Millstone j

that are effectively unventilated and have Class lE cables routed in the space.

Describe the separation and justify the adt.quacy of the separation between redundent Class 1E circuits and between Class 1E and non-Class lE circuits.

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REQUEST FOR ADDITIONAL INFORMATION i

MILLSTONE NUCLEAR P0'ER STATION UNIT 3 W

D0CKET N0. 50-423 i

430.0 POWER SYSTEMS BRANCH, ELECTRICAL (CONT'D) l 430.30 In Section 1.8 of the FSAR, you take exception to position C16 of (SRP8.3.1, Regulatory Guide 1.75 and Section 5.6.2 of IEEE Standard 384-1974.

8.3.2)

Minimum separation between redundant Class 1E cables or between i

Class lE and non-Class 1E cables is identified to be 1 inch versus 6 inches inside control switchboards and instrument cabinets.

l Provide the analysis that demonstrates the adequacy of 1 inch minimum separation.

i 430.31 A third or spare charging pump may be connected to either Class lE (SRP8.3.1, bus 34C or 340.

Describe the interlocks that preclude two charging 8.3.2) pumps from being powered from the same Class lE bus and preclude redundant buses from being tied together.

Provide a similar description for the third or spare reactor plant component cooling pump.

j 430.32 You state in Section 9.5.4.3 of the FSAR, in part, that one fuel oil 1

(SRP 8.3.1, transfer pump on each fuel oil storage tank is arranged to allow 8.3.2) transfer from the A electrical bus to the B electrical bus, or visa versa, by means of a 480-volt, seismically qualified Class lE transfer switch manually operated under administrative control.

It appears that the Millstone design includes provision for manually transferring loads between redundant Class 1E divisions other than those described in Chapter 8 of the FSAR.

It is the staff position that the designs of each interconnection should prevent a single failure or inadvertent closure of one interconnecting device from compromising division independence.

~

An acceptable design includes a minimum of two series connected disconnect devices that are physically separated, interlocked, administrative 1y kept normally open, and annunciated in the control room upon closure.

Identify all interconnections between redundant distribution systems; describe how each interconnection meets the above staff position; and justify areas of noncompliance.

430.33 The electrical 4.16 kV one line diagrams (No. 12179-10-1K, IL, IM, (SRP 8.3.1, and IN) included in Section 1.7 of the FSAR indicate that there 8.3.2) is an interconnection between Millstone Unit 2 and Unit 3 and between redundant buses 34C cnd 34D.

Figure 8.1-3 and Section 8.0 of the FSAR indicate no interconnection between Units 2 and 3 or between buses 34C and 34D.

Provide clarification and justify the interconnection.

430.34 Section 8.3.1.1.4(8) of the FSAR indicates that piping subject to (SRP8.3.1, freezing or boron precipitati6n are electrically heat traced. Two i

8.3.2) heat tracing circuits are provided for each pipe one heat trace i

circuit is connected to Class 1E division A while the other circuit l

is connected through an isolation transformer to division B.

Provide a description and justification of the physical and electrical independence between the two heat tracing circuits and between the redundant Class 1E divisions.

1

l REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET NO. 50-423 4

l 430.0 POWER SYSTEMS BRANCH,' ELECTRICAL (CONT'D) i 430.35 In Section 8.3.2.1.1 of the FSAR you state that battery charger 5 (SRP8.3.1, is powered from a Class 1E emergency bus, furnishes de power to 4

8.3.2) nonsafety loads, and meets all the requirements of an isolation device.

Provide test results and/or analysis that demonstrates that any failure or combination of failure or malfunction in the nonsafety circuits will not cause unacceptabls influence on Class 1E circuits.

In addition, define the requirements for this isolation device.

I 430.36 In Section 1.8 of the FSAR, you take exception to position Cl of (SRP 8.3.1, Regulatory Guide 1.75.

Interrupting devices, actuated only by 8.3.2) fault current, are used as isolation devices.

It is the staff position that non-essential circuits (powered from Class lE buses) be either disconnected by an accident signal or connected to the Class 1E bus through two series connected and coordinated interrupting devices actuated by fault current.

Identify and describe each non-Class 1E or nonessential circuit that is to be isolated from Class 1E circuite by an interrupting device actuated only by fault current and that is in nencompliance with the above staff position. In order to justify noncompliance with the staff positfon, provide the test or analysis that demonstrates that each non-Class 1E j

circuit identified will not cause unacceptable influence on Class lE circuits.

4 l

430.37 In Section 8.3.2.1.1 of the FSAR you state that nonsafety 480 volt (SRP 8.3.1, stub bus 32-3T (that supplies power to a number of nonsafety de loads 8.3.2) located in a nonseismic building) is powered from a Class lE bus and is automatically shed upon loss of offsite power.

It is the staff position that this stub bus should also be automatically shed upon an accident signal.

Provide justification for noncompliance with this staff position.

430.38 Non-Class lE NSSS loads are connected to the Class 1E 120V vital ac (SRP8.3.1, buses through transformers.

You have stated that these transformers 8.3.2) are qualified as isolation devices. Provide test results and/or analysis that demonstrates that any failure or combination of failures (including hot short) in the nonsafety circuits will not cause unacceptable influence on any Class 1E circuits.

In addition, provide a description of the non-Class 1E load with respect to its size and the capacity and capability of the Class 1E system to supply the non-Class 1E load.

430.39 In Section 8. 3.2.1. 2.1 of the FSAR you state that the battery charger (SRP8.3.1, is sized based on normal bus loads. Position C.l.b of Regulatory 8.3.2)

Guide 1.32 on the other hand, requires that the battery charger be sized based on the largest combined dmands of the various steady-state loads. Provide clarification and description as to how the 4

Millstone design meets position Clb of Regulatory Guide 1.32.

The clarification and description should address but not be limited to items: size of each load identified in Table 8.3-4, the following(normal, continuous, transients or momentary),

type of load i

_ _. _., _,. ~ _ _ _ _ _ _

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REQUEST FOR ADDTTIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET NO. 50-423 i

430.0 P,0WER SYSTEMS BRANCH,-ELECTRICAL (CONT'D) identification of non-Class 1E loads, and the inconsistencies between Tables 8.3-4 and 8.3-5 and between Table 8.3-4 and the 125V dc one line diagrams for battery buses 301A-1 and 301B-1 presented in Section 1.7 of the FSAR.

430.40 Recent operating experience has shown that an incompatibility (SRP 8.3.1, between the battery rack and the battery may cause cracking J' 8.3.2) the battery case. The cracking may be caused'in part by the improper support at the battery stress points (the plate support bridge). Describe the relationship between the plant support bridge and the battery rack supports and how the seismic 1

qualification test program encompasses the subject stress-related i

aging of the battery case.

430.41 Loads connected to the de bus may be subject to voltage variations (SRP8.3.2) from 90 to 143 volts due to battery discharge and equalizing charge.

It is the staff position that de loads be designed and qualified to to operate when subject to these voltage variations. Describe the extent of compliance of the Millstone design to this position and justify any areas of noncomplisnce.

430.42 Describe the extent to which the recommendations of IEEE Standards (SRP8.3.2) 338, 450, 484 and 485 and Regulatory Guides 1.118,1.128, and 1.129 have been followed in regard to testing, maintenance, installation, i

and sizing of Class 1E batteries and de systems.

430.43 The specific requirements for de power system monitoring derive from (SRP8.3.2) the generic requirements in Section 5.3.2(4), 5.3.3(5), and 5.3.4(5) of IEEE 308-1974, and in RG 1.47, " Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems." In summary, these general requirements state that the de system (batteries, distribution systems, and chargers) shall be monitored to the extent that it is shown to be ready to perform its intended function.

It is the staff position that the following indications and alarms of the Class lE direct current power system status shall be provided in the control room:

i

-battery float charge (ammeter)

-battery circuit output current (ammeter)

-battery charger output current (ammeter)

-dc bus voltage (voltmeter)

-battery discharge alarm

-dc bus overvoltage alarm

-dc system ground alarm

-battery disconnect open alarm

-battery charger disconnect open alarm

-battery charger failure alarm (one alarm for a number of abnormal conditions which are usually indicated locally)

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REQUEST FOR ADDITIONAL INFORMATION i

MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET NO. 50-423 '

P_0WER SYSTEMS BRANCH, ELECTRICAL (CONT'D) 430.0 0

)

The staff has concluded that the above-cited monitoring, augmented by the periodic test and surveillance requirements that are included in the Technical Specifications, provide reasonable assurance that the Class lE de power. system is ready to perform its intended safety functisn.

Describe the extent to which the above staff position is followed and justify areas of noncompliance.

l 430.44 In Section 8.3.2.1 of the FSAR you state that power will be i

(SRP 8.3.2) available to de system loads for at least two hours in the event of loss of all ac power. After 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> you have assumed that ac power is either restored or that the emergency generators 1

i are available to energize tne battery chargers. Based on the staff's review of recent applications, this period for restoration of ac power appears to be too short.

Provide the s

4 basis end operational experience data for the assumption that l

ac power can be restored within two hours.

Emergency procedures and training requirements for station i

blackout events are described in generic letter 81-04.

Provide j

a statement of compliance with these generic requirements.

1 i

430.45 Provide a description as to how the onsite Class lE power system (SRP8.3.1, meets the guidelines of IEEE Standard 338 and Regulatory Guide 8.3.2) 1.118.

Identify and justify deviations.

j 430.46 In Section 8.3.1.1.4 (items 2 and 4) of the FSAR you indicate that (SRP8.3.1, primary and backup containment electrical penetration protection j

8.3.2) is provided only where the available fault-current exceeds the current-carrying capabilities of penetration conductors for loads connnected to safety related buses that are not qualified to the containment accident environment. This design for containment electrical penetration protection does not meet the guidelines of position 1 of Regulatory Guide 1.63.

Position 1 requires:

a) primary and backup protection where maximum available fault-current exceeds the current-carrying capability of the penetration i

versus capability of the conductors and b) all conductors, that pass through containment electric penetrations, to have primary 4

and backup protection versus only those that are connected to safety related buses and loads that are not qualified to the containment accident environment.

Provide justification for noncompliance with the guidelines a.

of position 1 of Regulatory Guide 1.63.

b.

Describe how the Millstone design complies with each of the guidelines of IEEE Standard 317-1976 as augmented by Regulatory Guide 1,63 and provide justification for any deviations.

Provide coordinated fault-current versus time curves for each c.

representative type cable that penetrates primary containment.

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET N0. 50-423 430.0 POWER SYSTEMS BRANCH, ELECTRICAL (CONT'D)

For each cable, the curves must show the relationship of the fault carrying capability between the electric penetrations, the primary overcurrent protective device, and the backup overcurrent protective device, d.

Provide the test report with results that substantiates the capability of the electrical penetration to withstand the total range of time versus fault current withcut seal failure for worst case environmental conditions.

430.47 In Section 1.8 of the FSAR you provided clarifications as to how the (SRP 8.3.1, guidelines of Regulatory Guide 1.63 are to be implemented in 8.3.2) the Millstone design for protection of containment electrical penetrations. The clarifications state that overcurrent protective devices are not required to comply with criteria listed in IEEE 279 (except Section 4.2) and need not be Class lE or seismically qualified.

Position 1 of Regulatory Guide 1.63, on the otherhand, states that overcurrent protective devices should conform to the criteria of IEEE ?79. The proposed Millstone design does not meet the guidelines of position 1 of Regulatory Guide 1.63.

Provide justification for noncompliance.

430.48 Describe how the Millstone design complies with the guidelines of (SRP 8.3.1, NUREG-0737 items II.E.3.1 and II.G.1 and justify areas of 8.3.2) noncompliance.

430.49 In Section 8.3.1.4.1 of the FSAR, you define design criteria for (SRP8.3.1, independence and availability of Class lE systems. The 8.3.2) definition includes the statement that " separation of equipment is maintained to prevent loss of redundant features for single events and accidents." Similarly, in Section 8.3.1.1.2 of the FSAR, you state that redundant Class lE buses are physically and electrically separated so that any credible event which might affect one bus will not jeopardize proper operation of the other bus.

The above statements imply that, with sufficient separation, only one of the redundant Class lE divisions need be protected from the effects of any single event or accident. Such a design does not meet the protection requirements of GDC 2 and 4, the single failure requirement of GDC 17, or the guidelines of IEEE Standard 308-1974. Define all credible events, accidents or design basis events and describe how each Class lE power system component is designed and qualified to withstand (or is protected from) the effects of each defined credible event.

Defined credible events should include but not be limited to: Design basis events listed in Table 1 of IEEE Standard 308-1974 and failures of non-Class lE l

or nonseismic Category I structures, systems, or components.

Where separation is used to prevent loss of redundant features from any single event or accidents, justify noncompliance with the requirements of GDC 2, 4 and 17.

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 j

DOCKET NO. 50-423 430.0 POWER SYSTEMS BRANCH, ELECTRICAL (CONT'D)

]

430.50 In item 5 of Section 6.3.1.4.2 of the FSAR you state, in part, (SRP 8.3.1, that Class lE cables of only one train will be installed in 3

8.3.2) potential missile-producitig areas or adequate missile protection will be provided when Class lE cables of redundant trains are installed 'in missile-producing areas. Based on this. statement, 1

it appears that Class lE equipment and cables are not protected from the effects of accident generated missiles.

Identify each Class lE equipment and cable not protected from the potential

}

effects of missiles.

For each cable or equipment identified, provide the results of an analysis that demonstrates that the a

number of circuits and equipment remaining is sufficient so

{

that the protective functions required can be accomplished assuming a single failure.

i 430.51 Identify all electrical equipment, both safety and nonsafety, (SRP8.3.1, that may become submerged as a result of a LOCA.

For all such J

8.3.2) equipment that is not designed and qualified for service in i

such an environment provide analysis to determine the j

following:

1 The safety (significance of the failure of this electrical 1.

equipment e.g. spurious actuation or loss of actuation l

function) as a result of flooding.

2.

The effects on Class lE power sources serving this f

equipment as a result of such submergence; and 3

3.

Any proposed design changes resulting from this analysis.

)

430.52 Provide additional information regarding the power sources 1

(SRP 8.3.1, supplied to the RHR isolation valves. The staff's position is f

8.3.2) that a single failure of a power supply should not prevent isolation of the RHR when RCS pressure exceeds the design

]

pressure of the RHR system. Additionally, loss of a single power supply should not result in the inability to initiate j

at least one 100 percent RHR train.

y I

1 430.53 In Section 8.3.1.1.2(1) of the FSAR, you state that the Class lE (SRP8.3.1, switchgear rooms contain automatic fire protection systems.

l 8.3.2)

Provide indication in Section 8.0 of the FSAR that Class IE equipment in all areas of the plant are either protected from t

automatic fire protection effluent or designed and qualified to operate in the environment that may be caused by the effluent.

t 430.54 In Sections 8.3.1.1.2 and 8.3.1.1.3 of the FSAR, you state that l

(SRP8.3.1, controls for the diesel generator and Class lE circuit breakers 8.3.2) are located in the control room and at remote locations. Describe the electrical independence between these two controls.

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REQUEST'FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET NO. 50-423 430.0 POWER SYSTEMS BRANCH, ELECTRICAL (CONT'D) 430.55 Section 8.3.1.1.4 of the FSAR implies that there may be safety (SRP8.3.1, and nonsafety related electrical equipment connected to Class 8.3.2) 1E power supplies and located inside of containment that are i

not design,ed and qualified to the containment accident environment.

For all such equipment provide analysis to determine the following:

1.

The safety significance of the failure of this equipment (e.g. spurious actuation or loss of activation function).

2.

The' effects on Class lE power sources serving this equipment.

3.

Any proposed design changes resulting from this analysis.

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ATTACHMENT 1 ADDITIONAL GUIDELINES FOR QUESTION 430.6 Generator Circuit Breakers / Load Break Switches A.

Background

Generator circuit breakers have been used in recent nuclear ger.erating stction designs (McGuire, Catawba) as a neans of providinc immediate access of the onsite ac power systems to the offsite circuits by isolating the unit generator from the main step-up and unit auxiliary transformers and allowing backfeeding of power through these circuits to the onsite ac power system.

Generator load break switches can be used as a means of providing access to the offsite circuits as described above, but only on a delayed basis.

Since this is a new design feature, the staff made the use of generator c"cuit breakers and load break switches a generic item no. B-53.

In the case of McGuire and Catawba, an expert consultant was retained to evaluate the generator circuit breaker verification testing program and its results.

These reovirements are formalization of the results of that extensive work.

Also requirements for the load break switches are incorporated, as these devices have some common functional requirements as generator breakers as described above.

The staff has made a determination that only devices which have the capability of interrupting the systen maximum available fault current i.e., circuit breakers will be approved as a means of isolating the unit generators from the offsite power systen in order to provide immediate access in accordance with GDC 17. This is necessary because a non fault-current interrupting device i.e.,

load break switch must delay its trip for electrical faults until the switchyard circuit breakers have interrupted the current.

Following opening of the load break switch, the switchyard circuit breakers must then be reclosed to establish offsite power to the -

unit, A generator circuit breaker, however, could interrupt the fault current and isolate the unit generator at the same time, maintaining continuous pcwer to the onsite ac power system.

B.

Specific Guidelines 1.-

Only devices which have maximum fault current interruptino ccoability i.e. circuit breakers can be used to isolate the unit generator from the offsite and onsite ac power systems in order to provide immediate access for the onsite ac power system to the offsite source.

Generator load break switches can only be used for isolating the unit generator for the g

purpose of provi, ding a delayed access offsite source.

2.

Generator circuit breakers should be designed to perform their intended function during steady-state operation, power systen transients and major faults; tests should be perforned on the l

circuit breaker to verify these capabilities.

As a minimun, the folicwina performance. tests and capabilities should be demons trated:

J

h /

'a.

Dielectric Tests The circuit breaker should be given dielectric strength tests in accordance with the requirements and ratings contained in the applicable ANSI C37 series standards.

b.

Load Current Switching For applications which use only one generator circuit breaker, the cir:uit breaker should be cycled through 40 load interruption operations (a lesser number requires suitable justification) at a current equal to the normal full load continuous current rating of the circuit breaker.

For applications which utilize two generator circuit breakers in a parallel circuit, the circuit breaker should be given 40 load interruption operations (a lesser number requires suitable justification) at a current equal to twice the normal full load continuous current rating of the circuit breakers.

The procedures and acceptance cr:teria utilized for this test should be based upon those given in ANSI C37.06 and C37.09.

c.

Fault Current Interrupting Capability The circuit breaker should have, as a minimum, the capability of interrupt.ing the maximun assynetrical and symmetrical fault current available at the instant of primary arcing contact separation. This current should be calculated by assuming a bolted three phase fault at a point on the system which causes the maximum amount of fault current flowing through the generator circuit breaker.

The fault current interrupting capability (short circuit current rating) of the circuit breaker should be demonstrated by performing a series of tests similar to those called for in ANSI C37.04 and C37.09.

The tests should include close/open (CO) operations and should be perforned at the circuit breaker minima rated air pressure and control voltage and with a rate of rise of recovery voltage not less than the following rated

value, d.

Maximum Rate of Rise of Recovery Voltaae The rated maximum rate of rise of recovery voltage (RRP.V) of the circuit breaker should not be less than the maximum RRRV imposed on the breaker in the circuit in which it is used.

e.

Short-Time Current Carrying Capability The circuit breaker should have the capability of carryinq a fault current for the length of time that the fault exists assuming feilure of the primary protective

e o

device to clear it.

The fault current chosen should be that due to a fault on the system at a point which causes 2

the largest I t heating of the circuit breaker. The short-time current carrying cepability should be demonstrated with a current carrying test.

f.

Monentary Current Carryino Capability The circuit breaker should have the capability of carrying the maximun crest value of current calculated for the worst case bolted three phase fault on the system. This capability should be demonstrated by test.

g.

Transformer Magnetizing Current Interruption The circuit breaker interruption of an unloaded station main and/or auxiliary transformer magnetizing current should not generate excessively high surge voltages which could damage the connected bus and transfomer insulation.

This should be verified by test.

h.

Themal Capability The themal capability of the circuit breaker should he

. demonstrated by a test at its continuous current rating.

The test should be in accordance with the requirenents and ratings contained in, ANSI C37.04 and C37.09.

For applications which use two generator circuit breakers in parallel, a test should be conducted to determine the time to reach the maxinum pemissable temperature on the most limiting compnnent of the breaker when going from the rated continuous current to twice rated continuous l

current.

i.

Mechanical Operation Test l

A sufficient nunber of no-load mechanical operations should be perforned by the circuit breaker to provide a reasonable indication of its mechanical reliability and life.

The demonstrated life should be adequate for the l

plant life expectancy.

l 3.

The availability of offsite power to the onsite loads for designs utilizing generator circuit breakers should be no less than comparable designs which utilize separate offsite power transfomers to supply offsite power to the station loads.

In this regard the trip selectivity between the generator circuit brea'Kers and the' switchyard high voltage generator circuit breakers should insure against unnecessary tripping of the switchyard generator circuit breakers during abnormal events

)

in order to maintain offsite power to the station loads.

4.

Load break switches should be designed to perform their i

q

4 intended function during steady-state operation, power systen transients and major faults.

Except for iten 2.C. the switches should have the same capabilities as defined in position 2 for generator circuit breakers.

In addition the symmetrical interrupting capability of the load break switch should be at least equal to the maximum identified peak loading capability of the station generator.

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a REFERENCES 1.

Safety Evaluation Report related to operation of McGuire Nuclear Station, Units 1 and 2 NUREG-0442 dated March 1978.

2.

FSAR McGuire Nuclear Station Docket 50-369/370 3.

FSAR Catawba Nuclear Station Docket 50-413/414 4

ANSI Standard C37.04, Rating Structure for AC High-Voltaae Circuit Breakers Rated on a Symmetrical Current Basis 5.

ANSI Standard C37.06, Preferred Ratings and Related Required-4 Capabilities for AC High-Voltage Circuit Breakers Rated on a Synnetrical Current Basis 6.

ANSI Standard C37.09, Test Procedure for AC Hiah-Voltage Circuit Breakers Rated on a Synnetrical Current Basis i

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REQUEST FOR ADDITIONAL INFORMATION 4

MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET NO. 50-423 640.0 PROCEDURES AND SYSTEMS REVIEW BRANCH 640.1 Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants (Regulatory Guide 1.70) specifies that Section 14.2.7 of an FSAR include a list of Regulatory Guides that are applicable to the facility's initial test program and include justification for any excep-tions to Regulatory Positions stated in those Guides.

The enclosed list of applicable Regulatory Guides shows the specific revisions which should be addressed in Section 14.2.7 of the FSAR.

Section 14.2.7 should be amended as necessary.

1.18 Rev. I 1.20 Rev. 2 1.37 Rev. O 1.41 Rev. 0 1.52 Rev. 2 i

1.68 Rev. 2 1.68.2 Rev. 1 1.68.3 Rev. 0 1.72 Rev. 1 1.79 Rev. 1 1.95 Rev. 1 1.108 Rev. I 1.116 Rev. 0-R 1.140 Rev. 1 e

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REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET N0. 50-423 t

640.0 PROCEDURES AND SYSTEMS REVIEW BRANCH - continued 640.1 Regulatory Guides which need not be addressed in 14.2.7:

(cont.)

1.30 Listed in Standard Review Plan, but has been withdrawn.

1.80 Listed in Standard Review Plan, but has been withdrawn effective April 20, 1982, in favor of Regulatory Guide 1.68.3.

1.128 Listed in Standard Review Plan, but none of the plants are required to. conform since their construction

, permit docket dates were all before the implementation date of December 1, 1977.

1.139 Listed in Standard Review Plan, but none of the plants are required to conform since their construction permit docket dates were all before the implementation date of January 1, 1978.

1.9 Refe'renced in Regulatory Guide 1.68, but diesel testing is cov.ered in 1.108 and this guide does not directly bear on the initial plant test program.

1.104 Referenced in Regulatory Guide 1.68, but withdrawn August 16, 1979.

NOTE: All other Regulatory Guides referenced in Regulatory Guide 1.68 or listed in the Standard Review Plan are contained in the Applicable Regulatory Guide List.

' REQUEST FOR ADDITIONAL INFORMATION N

MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET N0. 50-423 260.0 QUALITY ASSURANCE BRANCH 260.1 Describe the criteria for determining the size of the QA organization (s)

(IAS)*

including the inspection staff (s).

260.2 Figure 1.6 of NU-QA-1 identifies onsite and offsite organizational (IAS &

elements. The figure (per letter of 8/17/82) appears to reflect 17.2.1)** more the construction phase than the operations phase. Update the-figure to include functions such as inspection, test, instrumentation and control, operations, maintenance for the operations phase of Millstone 3.

The figure shows both Operations QA and Station Super-intendent Millstone as being offsite.

Clarify.

Identify which positions on the figure are in the NUSCO organization and which positions are in the NUP0C organization.

260.3 Section 1.3 of NU-QA-1 describes responsibilities for the Millstone (181) 3 QA program.

It states that overall responsibility is assigned to the_ Director-Nuclear Engineering and Operations Services. This seems to conflict with the Policy Statement which says that " corporate authority is delegated to the NUSCO Manager, Quality Assu~rance for the preparation and :.dministration of the Quality Assurance Program."

Cla rify. Assigning the overall responsibility for the QA program to the Director-Nuclear Engineering and Operations Services is acceptable if this position has no other duties or responsibilities unrelated to QA that would prevent his full attention to QA matters and has respon-sibility for approval of the QA Manual (s). Clarify in the FSAR whether this is indeed the case.

260.4 Clearly differentiate between the authority, duties, and responsibil-(Gen-ities of Operations QA organization under the Manager Quality ' Assurance eral) and the Quality Assurance Supervisor Millstone under the Station Super-intendent Millstone.

• Alphanumeric designations in parentheses refer to the acceptance criteria in Section 17.1 of the Standard Review Plan (NUREG-0800, July 1981)

• • Numerical designations in parentheses refer to acceptance criteria in Section 17.2 of the Standard Review Plan (NUREG-0800, July 1981).

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET N0. 50-423 260.0 QUALITY ASSURANCE BRANCH (CONT'D) 260.5 Identify (by organization) the individuals responsible for the verifi-(1B2) cation of conformance to requirements and discuss their independence from the " doers," their training, and their qualification require-ments.

260.6 The last paragraph of 1.4.3.2c of NU-QA-1 enumerates 4 responsibil-(1B3&4 ities of the Millstone Quality Assurance Supervisor and his staff.

identify who (by position title) has the same responsibilities in the Manager, Quality Assurance's organization. Describe k these actions are accomplished.

260.7 The last sentence of the introduction of NU-QA-1 states the NUP0C and (1B5)

NUSCO interrelationship shown in Figure 1.6 " indicates how conflicts are resolved." This is not clear.

Describe provisions for the reso-lution of disputes involving quality arising from a difference of opinion between QA personnel and cther department (engineering, pro-curement, manaufacturing, operating, etc.) personnel.

260.8 Describe those provisions which assure that designated QA individuals' (186) are involved in day-10-day plant activities important to safety (i.e.,

the QA organization routinely attends and participates in daily plant work schedule and status meetings to assure they are kept abreast of day-to-day work assignments throughout the plant and that there is adequate QA coverage relative to procedural and inspection controls, acceptance criteria, and QA staffing and qualification of personnel to carry out QA assignments).

260.9 Describe previsions which assure that implementation of the QA program, (IC1) procedures, and instructions is mandatory as stated in section 2.2.5 of NU-QA-1.

260.10 Appendix B of NU-QA-1 lists the qualification requirements for the (1C2)

NUSCO Manager, Quality Assurance.

The staff is interested in the qualification requirements for the position that retains overall authcrity and responsibility for the QA program (Ref. question 260.3).

The qualifications for this position should be established in a position description which includes the following prerequisites:

a.

Management experience through assignments to responsible positions.

b.

Knowledge of QA regulations, policies, practices, and standards.

c.

Experience working in QA or related activity in reactor design, construction, or operation or in a similar high technological industry.

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REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET N0. 50-423 260.0 QUALITY ASSURANCE BRANCH (CONT'D)

The qualifications for this position should be at least equivalent to those described in Section 4.4.5 of ANSI /ANS-3.1-1978, " Selection and Training of Nuclear Power Plant Personnel," as endorsed by the regu-latory positions in Regulatory Guide 1.8.

Provide such a commitment or propose an alternative for staff evaluation.

l 260.11 Identify by position title the person responsible for the onsite QA (103) program.

Clarify that this position is free from non-QA duties and l

can thus give full attention to assuring that the QA program at the plant site is being effectively implemented and that it has appro-priate organizational position, responsibilities, and authority to exercise proper control over the QA program.

l 760.12 Describe how the QA program will be applied to the development, (2Alc) control, and use of computer programs.

260.13 Appendices D and F address commitments to regulations, regulatory (283) guides, and ANSI standards.

For Millstone 3, add:

a.

Appendix A to 10 CFR Part 50, General Draign Criteria 1 b.

10 CFR Part 50, Section 50.54, Condition of Licenses j

c.

10 CFR Part 50, Section 50.55a, Codes and Standards d.

Regulatory Guide 1.26, Rev. 3 (2/76) e.

Regulatory Guide 1.29, Rev. 3 (9/78) f.

BTP ASB 9.5-1 (Fire Protection) and delete:

a.

Regulatory Guide 1.70 b.

Draft Regulatory Guide concerning preoperational and startup test programs.

i For systems, components, and structures covered by the ASME Code l

Section III (Classes 1, 2 and 3) describe measures which assure that the QA code requirements are supplemented by the specific guidance j

addressed in the regulatory positions of the applicable Regulatory Guides.

Clarify that the QA organization and the necessary technical organiz~

ations determine and identify the extent QA controls are to be applied to specific structures, systems, and components. This effort involves applying a defined graded approach to certain structures, systems, and components in accordance with their importance to safety and affects such disciplines as design, procurement, document control, inspection tests, special processes, records, audits, and others described in 10 CFR Part 50, Appendix B.

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3-DOCKET NO. 50-423 260.0 QUALITY ASSURANCE BRANCH (CONT'D) 260.14 Sections 1.2.5.1 and 2.2.6 of NU-QA-1 describe an annual Management QA (2C1)

Review under the sponsorship of the Senior VP - Nuclear Engineering and Operations. The involvement of the Senior VP - Nuclear Engineering and Operations should also include frequent contact with the QA program through reports, meetings, and/or audits. Provide such a commitment-or describe an alternative for staff evaluation.

260,15 Describe provisions which assure that:

(20) a.

Documentation of formal training and qualification programs includes the objective, content of the program, attendees, and date of attendance.

b.

Proficiency tests are given tG those personnel performing and verifying activities affecting quality, and acceptance criteria are developed to determine if individuals are properly trained and qualified.

c.

Certificates of qualification clearly delineate the specific functions personnel are qualifed to perform and the criteria used to qualify personnel in each function.

260.16 Describe provisions for assuring the QA program for operations is (17.2.2-2) implemented at least 90 days prior to fuel loading.

260.11 Provide confirmation to commit to continued implementation of the (17.2.2-3) PSAR QA program for the remaining design and construction activities and the preoperational test program or provide an acceptable altern-ative.

260.18 Describe internal and external design interface controls, procedures, (3D) and lines of communication among participating design organizations and across technical disciplines for the review, approval, release, distribution, and revision of documents involving design interfaces to assure structures, systems, and components are compatible geo-metrically, functionally, and with processes and environment.

260.19 Describe provisions which assure a documented check to verify the (3E1) dimensional accuracy and completeness of design drawings and specifications.

260.20 Describe provisions which assure that design drawings and specifi-(3E2) cations are reviewed by the QA organization to verify that the documents are prepared, reviewed, and approved in accordance with company procedures and that the documents contain the necessary QA-requirements such as inspection and test requirements, acceptance requirements, and the extent of documenting inspection and test

results, d

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REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET N0. 50-423 260.0 QUALITY ASSURANCE BRANCH (CONT'D) 260.21 Describe provisions which assure that design verification, if other (3E4b) than by qualification testing of a prototype or lead production unit, is completed prior to release for procurement, manufacturing, con-struction or to another organization for use in other design activities.

In those cases where this timing cannot be met, the design verification my be deferred, providing that the justification for this action is documented and the unverified portion of the design output document and all design output documents, based on the unverified data, are appropriately identified and controlled.

Site activities associated 4

with a design or design change should not proceed without verification past the point where the installation would become irreversible (i.e.,

require extensive demolition and rework).

In all cases, the design verification should be complete prior to relying upon the component, system, or structure to perform its function.

260.22 Describe provisions which assure that procedural control is established (3E4c) for design documents that reflect the commitments of the SAR and that this control differentiates between documents that receive formal de-l sign verification by interdisciplinary or multi-organizational teams and those which can be reviewed by a single individual (a signature and date is acceptable documentation for personnel certification).

Design documents subject to procedural control include, but are not limited to, specifications, calculations, computer programs, system l

descriptions, SAR when used as a design document, and drawings in-cluding flow diagrams, electrical single line diagrams, structural systems for major facilities, site arrangements, and equipment locations.

Specialized reviews should be used when uniqueness or special design considerations warrant.

260.23 Describe provisions which identify the responsibilities of design (3E4d) verifiers, the areas and features to be verified, the pertirent a

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considerations to be verified, and the extent of documentation.

260.24 When design verification is by test, describe procedures which assure (3E5) the following:

a.

Procedures provide criteria that specify when verification should be by test.

b.

Prototype, component, or feature testing is performed as early as possible prior to installation of plant equipment, or prior to the point when the installation would become irreversible.

c.

Verification by test is performed under conditions that simulate the most adverse design conditions as determined by analysis.

260.25 Clarify that procedures are established to assure that verified -

(3E6) computer codes are certified for use and that their use is speci-fied.

REQUEST FOR DDITIONAL1INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET.NO. 50-423 j

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260.0 QUALITY ASSURANCE BRANCH (CONT'D) 260.26 Describe measures which assure that responsible plant personnel, on (17.2.3-2) a timely basis, are made aware of design changes / modifications which may affect the performance of their duties.

260.27 Describe organizational responsibilities for bid evaluation.

Include (4B1) the involvement of the QA organization (s).

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260.28 Expand the list of documents in section.6.2.1 of NU-QA-1 to include (6A1) documents related to computer codes and as-built documentation. -

260.29 Describe provisions which assure that the QA organization, or an 2

(6A2) individual other than the person who generated the document but qualified in quality assurance, reviews and concurs with the docu-ments listed in section 6.2.1 of'NU-QA-1 with regards to QA-related aspects.

260.30 Describe provisions which assure that obsolete or superseded docu-(681) ments are removed and replaced by applicable revisions in work areas i

in a timely manner.

260.31 Describe provisions which assure the preparation of as-built drawings (6C1) and related documentation in a timely manner to accurately reflect the actual plant.

260.32 Describe provisions which assure that maintenance, modification,-

(17.2.6-2) and inspection procedures are reviewed by qualified personnel knowledgeable in QA disciplines (nonnally the QA organization) to determine:

a.

The need for inspection, identification of inspection per-sonnel, and documentation of inspection results.

b.

That the necessary inspection requirements, methods,Jand i

acceptance criteria have been identified.

I Identify the organizational element (s) responsible for this review, 260.33 The_ fifth paragraph of section 4.2.1 of NU-QA-1 eddresses the l

(7A4) procurement of spare and replacement parts. Describe provisions which assure that such procurements are subject to the pertinent I-provisions of the latest QA program controls.

260.34 Describe measures which assure that, for commercial "off-the-shelf" (7B4).

items where specific. quality assurance controls appropriate for

'l nuclear applications cannot be imposed in a practicable manner, special quality verification requirements shall be established and

.i described to provide the necessary assurance by NUSC0/NUPOC of an acceptable item.

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o REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET N0. 50-423 260.0 QUALITY ASSURANCE BRANCH (CONT'D) 260.35 The last paragraph of section 7.2.1 of NU-QA-1 addresses the independ-(785) ent verification of Certificates of Conformance.

Clarify that these verifications are documented.

260.36 Describe provisions which assure that procedures provide criteria (10A) for determining the accuracy requirements of inspection equipment and for deternining when inspections are required.

Clarify NUSC0/

NUP0C QA's involvement in the preparation of insoection plans and schedules as discussed in section 10.2.2 of NU-QA-1.

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260.37 Clarify in section 10.2.3 of NU-QA-1 whether inspections are l

(10B1&

performed by huSC0/NUPOC QA personnel or by personnel outside the 17.2.10.2) QA organization.

If inspections are made by personnel outside l

the QA organization, describe provisions which assure (in addi-tion to commitments in section 10.2.3) that these personnel do not report directly to the immediate supervisors who are responsible for the activity being inspected.

If the individuals performing inspections are not part of the QA organization, the inspection procedures, personnel qualification criteria, and independence from undue pressure such as (.ost and schedule should be reviewed and found acceptable by the QA organization prior to the initiation of the activity.

If inspections associated with normal operations of the plant (such as routine maintenance, surveillance, and tests) are performed by individuals other than those who performed or directly supervised the work, but are within the same group, de-scribe measures which assure the following controls are also met:

a.

The quality of the work can be demonstrated through a functional test when the activity involves breaching a pressure retaining barrier.

b.

The qualification criteria for inspection personnel are reviewed and found acceptable by the QA organization prior to initiating the inspection.

260.38 Clarify in section 10.2.4 of NU-QA-1 that inspection procedures (10C1g) and checkliFts specify the necessary measuring and test equip-ment and the required accuracy of this equipment.

l 260.39 Describe provisions which assure that inspection results are l

(10C3) evaluated for acceptability and indicate what organization is responsible for this activity.

260.40 Describe provisions which assure that test procedures provide (11A1) criteria for determining the accuracy requirements of. test equipment and for determining when a _ test is required.

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REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET NO. 50-423 260.0 QUALITY ASSURANCE BRANCH (CONT'D) 260.41 Describe QA and other organizations' responsibilities for establishing, (12.2) implementing, and assuring effectiveness of the calibration program.

260.42 Identify the organization (s) responsible to establish, review, and (12.3) concur with procedures for calibration, maintenance, and control of M&TE. Clarify that NDE equipment and instrumentation used during the operations phase of Millstone 3 will be controlled in accordance with section 12.0 of NU-QA-1.

206.43 Identify the management authorized to accept calibration of M&TE (12.6) against standards that have an accuracy of less than four times the accuracy of the equipment being calibrated.

260.44 Identify the management authorized to allow the use of calibrating (12.7) standards with the same accuracy as standards being calibrated.

260.45 Describe provisions which assure that reference and transfer standards (12.8) are traceable to nationally re. cognized standards and that', where national standards do not exist, provisiens are established to document the basis for calibration.

260.46 Describe provisions for the storage of chemicals, reagents (including (17.2.13.2) control of shelf life), lubricants, ar.; other consumable materials.

260.47 Describe provisions to control altering the sequence of required (14.3) tests, inspections, and other operations important to safety.

Such actions should be subject to the same controls as the original review and approval.

260.48 Identify the organization responsible for documenting the status (14.4) of nonconforming, inoperative, or malfunctioning structures, systems, and components.

260.49 Expand section 15 of NU-QA-1 to include nonconforming services (15.1)

(including computer codes).

260.50 Describe provisions to assure that nonconformances are corrected (15.3) or resolved prior to the initiation of the preoperational testing of the item.

260.51 Describe provisions which assure that nonconformance reports are (15.5) periodically analyzed by the QA organization to show quality trends, and the significant results are reported to upper manage-ment for review and assessment.

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""o-REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET NO. 50-423 s

i 260.0 QUALITY ASSURAICE BRANCH (CONT'D)

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260.52 Describe provisions which assure that'HUSC0 QA reviews and documen'.ts

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(16.1) concurrence with corrective action procedures.

260.53 The first paragraph of section 16.2.1 of NO QA-1 indicates that (16.2) conditions adverse to quality are evaluated to determine the need for corrective action.

Clarify that the procedures which require

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such evaluation also require documentation of this evaluation. Al so',._

describe provisions which assure QA concurrence of the adequacy of -

the corrective action.

260.54 Describe provisions which assure that followup action is taken by the (16.3)

QA organization to verify proper implementation of corr'octive action and to close out the corrective action in a timely manner.

260.55 Describe the facilities for the storage of Millstone 3 records.

(17.4) 260.56 Expand the list of areas audited in the second paragraph of section L

(18A4) 18.2.1 of NU-QA-1 to include:

a.

Corrective action, calibration, and nonconformance control systems.

s, b.

SAR commitments.

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Activities associated with computer codes.

260.57 Section 3.4 and Appendix A of the Millstone 3 " Fire Protection Eval-i (FP) uation Report" address quality assurance.

It is not clear that the pertinent provisions of NU-QA-1 will be applied to the fire protection j

program (and related hardware) during the cperation phase of Millstone 4

3.

Clari fy.

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REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET N0. 50-423

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471.0 RAD _IOLOGICAL ASSESSMENT GRANCH, RADIATION PROTECTION SECTION (CONT'D) 471.10 Section 12.1.II,0f the Standard Review Plan (SRP), NUREG-0800, Rev. 2, lists Reg'alato[y GuidF 8.8 as an acceptable means of meeting the requirements of,10 CFR 20.~1(c).

In accordance with R.G. 8.8 Section C.I.d(2), describe the radiation protection aspects of decommissioning

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which you have included in your plant, design to ensure that exposures towo[kers,duringdecommissionin[,willbeALARA.

471.11 In accordance with the acceptancecriteria of Section 12.2 of the SRP, NUREG-0800, your FSAR indicates that borated silicon shields are employedintheannularrehionforneutronstreaming. Provide a description of this shield that includes shield thickness, boron loading and the source strength that,yTaur design is~ based on.

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471.12 Section 121.2 of your FSAR-states that af te(a point by point comparison

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of your plant design with provisions in R.G. 8.8 "in nearly all cases, the plant designed was determined to be ALARA".

Provide a listing of whic6 specific cr.iteria are not in accordance with R.G. 8.8, Also, since

. R.G. 8.8 is referenced in Section 12.1 Acceptance Criteria as a means of demonstrating to the staff that occupational radiation exposures will be ALARA, provide the basis for con'cluding that those aspects of plant design not in conformance with R.G. 8.8 are acceptable.

471.13 Acceptance criteria for Section 12.3 of the SRP, NUREG-0800, include meeting the criteria of Sections II.B.2 of NUREG-0737, Clarification of TMI Action Plan:Requii ements.

In accordance with II.B.2 provide the analysis that

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demonstrates vital system operation and occupancy of the TSC.

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REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET NO. 50-423 4

l 470.0 RADIOLOGICAL ASSESSMENT BRANCH, RADIATION PROTECTION SECTION (CONT'D) 471.14 In accordance with Section 12.5.1.B.3 of the SRP, NUREG-0800, provide a description of the radiation protection facilities. This description should be complete enough to demonstrate how such facilities and services will allow both male and female workers to receive the necessary protection.

Include a description of such features as separate locker rooms, shower rooms, decontamination area and dress out areas.

471.15 In accordance with Section 12.3-12.4.I.1 of the SRP, NUREG-0800 and Section 12.3.1 of R.G.1.70 " Standard Format and Content of Safety Analysis Report for Nuclear Plants, Rev. 3, it is our position that the FSAR l

should include plant layouts showing shield wall thickness around each major radiation source. The shield wall thickness of major radioactive equipment can be provided in a separate table. Section 12.3.2 of your FSAR shoul(I be revised accordingly to provide this information.

471.16 Provide the information specified in Section 12.3.11.1 " Facility Design Features" of the SRP, NUREG-0800, as it refers to the stringent access control around the spent fuel transfer tube.

Include in your discussion i

any access routes to the spent fuel transfer tube, type of marking on 9

accessible poritions of transfer tube, and a description of the alarming radiation monitors employed if other than permanent shielding around'the transfer tube is used.

4 471.17 In accordance with Section 12.2.I.1 of the SRP provide the radioactive source geometry parameters missing from tables 12.2-2 and 12.2-4.

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET N0. 50-423 470.0 RADIOLOGICAL ASSESSMENT BRANCH. RADIATION PROTECTION SECTION (CONT'D) 471.18 10 CFR 20.103(b)(1) states that the licensee shall use process or engineering controls, to the extent practicable, to limit the concen-trations of radioactive materials in the air, to below 25% of the concen-trations given in Column 1, Table I of 10 CFR 20 Appendix B (as specified in 20.203(d)(1)(ii). Section 12.3.3.1.1 of your FSAR indicates that your ventilation system is designed only to maintain concentrations given in Column 1, Table I of Appendix B.

Discuss steps taken to up grade your ventilation system so as to meet the requirements of 10 CFR 20.103(b)(1). Also, Section 12.3.3.1 of your FSAR states that " airflow...

is normally from areas of lower to higher potential airborne contamination."

Describe the exceptions to this good ventilation practice (which is also required in 12.3.11.3 of the SRP).

471.19 In accordance with the criteria of 12.3.II.4.b of the SRP, NUREG-0800, indicate how your airborne radioactivity monitoring system will detect ten NPC-hours of particulate and radio-fodines from any compartment, normally occupied, which may contain this airborne contamination. Also, in accordance with 12.3.I.4.c, describe your procedures for locating suspected high activity areas.

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REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET NO. 50-423 470.0 RADIOLOGICAL ASSESSMENT BRANCH, RADIATION' PROTECTION SECTION (CONT'D) 471.20 In accordance with the criteria in 12.3.4 of the SRP, you should I

describe the details of your fi;:ed area radiation and continuous air-borne radioactivity monitoring instrumentation, as enumerated in 12.3.4 of R.G. 1.70. This should include a description of auxilliary l

or emergency power supplies; sensitivity, accuracy, precision and alann setpoints of the instruments; calibration methods and frequencies and also the criteria used for selecting the location for these instru-l ments (include selection criteria for locating portable CAMS).

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471.21 Section 12.3.II.4 of the SRP states that an acceptable area radiation l

monitoring system must meet the provisions in R.G. 1.97, Rev. 2.

Table 1 of R.G. 1.97, Rev. 2 specifies two High Range Containment Radiation monitors with a range of from 1 R/hr to 107 R/hr. Describe your means of implementing this provision. The placement of these monitors in containment, the accuracy and sensitivity of the monitoring employed, should all be discussed.

471.22 In accordance with 12.3-12.4.5.b of the SRP, describe any additional dose reducing measures, if any, taken as a result of the dose assessment provided in Chapter 12.4 of your FSAR.

471.23 Based on the criteria in C.l.e of R.G. 8.10, the RPM should have l

sufficient authority to enforce safe plant operation.

It is the Staff's l

position that to ensure the RPM's ability to communicate promptly with an appropriate level of management about haulting an operation he deems unsafe, he should report directly to, the Station Superintendent. Your organization description in the FSAR'should be revised to reflect this position.

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET N0. 50-423 470.0 RADIOLOGICAL ASSESSMENT BRANCH, RADIATION PROTECTION SECTION (CONT'D) 471.24 Indicate (and provide the resume's for) the individuals named as RPM and his backup.

In accordance with 12.5.II.A of the SRP, the RPM should meet the qualification specifications of R.G. 1.8.

Also it is the staff's position that the individual who will act as RPM in the RPM's absence (e.g., while en vacation), should have at least a B.S. degree in science or engineering, 2 years experience in radiation protection, 1 year of which should be nuclear power plant experience, 6 months of which should be onsite.

In addition your FSAR should be changed to address the qualification for health physics technicians as specified in ANSI N18..l.

471.25 Section 12.5.2 of your FSAR states that " areas in the RCA above 10 mrem per hour will have a locked or guarded barrier." The regulations require a locked or guarded barrier at 100 mrem per hour.

If this is a typo-graphical error, your FSAR should be changed accordingly.

471.26 Section 12.5.3 of your FSAR ctates that "special control techniques will be used to minimize airborne exposure arising from special work projects."

In accordance with the requirements of Section 12.5.II.C of the SRP, describe what these special control techniques are. Describe your technique for obtaining breathing zone air samples. What conditions require special air sampling and what are your reporting practices for airborne contamination surveys?

In addition, although Section 12.3.3.1 states that design and expected airborne concentrations are given in Section 12.4, there is no reference to airborne concentrations in that section.

In accordance with Section 12.2.I.2 of the SRP describe the j

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radioactive material sources used for design of personnel protective l

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REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET N0o 50-423 470'.0' '

• RADIOLOGICAL ASSESSMENT BRANCH, RADIATION PROTECTION SECTION (CONT'D) measures and for dose assessment.

This description should include a tabulation of the calculated concentrations of radioactive material, by nuclide, expected during normal operation, anticipated operational occurrences and accident conditions for equipment cubicals, corridors, and operating areas of the plant.

Include the models and parameters used as a basis for these concentrations.

471.27 Section 12.5.II of the SRP lists R.G. 8.13, R.G. 8.27, R.G. 8.29 and NUREG-0731, as acceptable guidance for establishing a radiation protection training program. Your FSAR should be changed to describe your compliance with these references.

471.28 In accordance with the acceptable criteria of Section 12.5 of the SRP, NUREG-0800, describe your plants capability of meeting the iodine sampl-ing criteria listed as item III.D.3.3 of NUREG-0737, clarification of TMI Action Plan Requirements.

471.29 Section 12.5 IIC of the SRP notes that compliance with R.G.1.33,

" Quality Assurance Program Requirements" is an acceptable means to show compliance with the requirements of 10 CFR 50, Appendix B.

In Section 12.5.3 of your FSAR, you state that quality assurance inspections will be performed on the radiation protection procedures identified in R.G. 1.33 Rey. 2.

Since the listing of procedures in R.G. 1.33, Rev. 2 is not intended to be an all inclusive listing, you should change your FSAR to include all radiation protection procedures (including those such as emergency procedures, and instrument storage, calibration and maintenance procedures, not specifically listed,in R.G.1.33, Rev. 2).

REQUEST FOR INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET NUMBER 50-423 220.0 STRUCTURAL AND GE0 TECHNICAL ENGINEERING BRANCH, STRUCTURAL ENGINEERING SECTION 220.09 FSAR Section 3.3.2.3 states that the metal siding and (SRP 3.3.2) roofing of the service, turbine, waste disposal, containment enclosure buildings, and portions of the fuel building are assumed to blow off under tornado wind load, but the structural steel framing of these structures is designed to withstand tornado wind loads.

Was it assumed that the metal siding and roofing remained intact to the structural steel framing in calculating the tornado wind loads acted on the framing?

220.10 Is there any concrete barrier whose thickness is less (SRP 3.5.3) than that shown in Table 1 of SRP Section 3.5.3? If yes, please identify and justify them.

220.11 FSAR Section 3.5.3.1 states that the barriers are (SRP 3.5.3) designed so that the calculated ductility ratio of the barriers for any load combination is less than the i

maximum allowable ductility ratio. However, the term

" ductility ratio" was never defined.

Please provide a definition for it and use a numerical example of concrete barriers to show how "the calculated ductility ratio" was computed.

220.12 The last sentence of FSAR Section 3.5.3.1 states that (SRP 3.5.3) if a concrete barrier is not required to carry other

]

loads during and af ter impact, the maximum allowable ductility is limited to correspond to a rebar elongation of 5 percent.

Is "the maximum allowable ductility" the same as "the maximum allowable ductility ratio" mentioned in the last question? If not, provide a new definition for it. Ordinarily. the ductility ratio of a reinforced concrete section is defined as the ratio of the calculated curvature at failure to the calculated curvature at yield. A curvature calculation involves both the steel reinforcing strain and concrete compressive strain.

Therefore, pleast explain why ductility is measured by rebar strain alone and not concrete compressive strain in the FSAR. Should that be the case, can the text be rephrased as "the maximum allowable rebar elongation is limited to 5 percent" to not involve the undefined word

" ductility"? There are several places in Section 3.5.3.1 where the words " ductility" and " ductility ratio" have been misplaced as interchangeable. Please correct them.

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER-STATION UNIT 3 DOCKET MUMBER 50-423' 220.0 STRUCTURAL AND GE0 TECHNICAL ENGINEERING BRANCH, STRUCTURAL ENGINEERING SECTION (CONI'D) 220.13 FSAR Section 3.5.3.1 states that for beam-column (SRP 3.5.3) members, where the compressive load is equal to or less than one-third of that which could produce balanced conditions, the allowable ductility is 10. Should the word " ductility" be replaced with " ductility ratio"?

If not, please explain the meaning of "the allowable ductility is 10".

We presume i.nat your " balanced conditions" mean the same as " balanced strain conditions" as defined in Section 10.3.2 of ACI 318-77 Code. Notice that the balanced strain conditions of the ACI Code exist at a cross section, but not structural members as implied in your description.

Therefore, please use an example to illustrate how you actually applied this requirement to beam-column members.

220.14 FSAR Section 3.5.3.1 states that for beam-column (SRP 3.5.3)-

members, where the design is controlled by compression, the allowable ductility is 1.3.

Please define "the allowable ductility" mathematically, and provide technical reasons to justify the number 1.3.

220.15 The first paragraph that describes the overall barrier (SRP 3.5.3) response of the FSAR Section 3.5.3.3 is difficult to read.

Please rephrase it.

Complete the sentence that reads "For beams, walls, and slabs where flexural controls design, the permissible ductility ratio is based on".

Also, use an example to illustrate how the i

flexural strength is determined from an ultimate strength theory with the limitations on ductility, as stated.

l 220.16 FSAR Section 3.78.1.1 states that Regulatory Guide 1.60 (SRP 3.7.1) spectra are not used. Does this mean that " site specific spectra" have been developed for this plant?

If so, have these site specific spectra been reviewed and approved by the Geosciences Branch of NRC?

220.17 FSAR Section 3.78.1.3 states that the values of the (SRP 3.7.1) percentage of critical damping used in the analysis of Seismic Category I structures, systems, and components depends on the seismic input motion used in the analysis.

Please explain this in detail, 220.18 The format and some of the percent of critical damping (SRP-3.7.1) values of Table 3.7B-1 are different from that of Regulatory Guide 1.61, " Damping Values for Seismic Design of Nuclear Power Plants." Either revise the contents in Table 3.7B-1 to comply with Regulatory Guide 1.61 or provide justifications for the deviations, as stated in SRP Section 3.7.1 11 2.

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET NUMBER 50-423 220.0

~ STRUCTURAL AND GE0 TECHNICAL ENGINEERING BRANCH, STRUCTURAL ENGINEERING SECTION (CONT'D) l 22.0.19 SRP Section 3.7.2 II 4 " Soil-Structure Interaction" (SRP3.7.2) requires that-two different ways of modeling the supporting soil media of a soil-structure interaction system be considered: half-space and finite boundary, and then envelop response results in structures for the l

use of designing Seismic Category I structures, systems, and components. Please revise your FSAR to comply with these requirements.

220.20 FSAR Section 3.7B 2.9 states that floor response (SRP3.7.2) spectra for the cracked and uncracked cases were enveloped. Provide the criteria that was used to determine the cracked or uncracked cases.

220.21 SRP Section 3.7.2 II 11 requires that an additional (SRP3.7.2) seismicity effect based on a consideration of f

5% of the maximum building dimension at the level under consideration shall be assumed to l

account for accidental torsion. Since FSAR Section l

3.7B.2.11 has not addressed this requirement, please l

revise the FSAR to comply with this requirement.

l 220.22 Provide response spectra that correspond to the time (SRP3.7.2) histories used at the foundation level of the contain-ment structure shown in Figure 3.7B-9.

Also, provide the values of horizontal, vertical, rocking and torsional subgrade springs and subgrade damping values (both translational and rotational).

Explain why there and 1, M, and M5 are far-coupling situations among i 2

between M and M in Figure 3.7B-9 1

3 220.23 Provide values of M's, K's, and damping of the (SRP 3.7.2) structure end values of rock springs and damping in Figure 3.7B-10.

220.24 Provide response spectra that correspond to the time (SRP 3.7.2) histories used at the bedrock (EL. - 14'-0") in Figures 3.78-11 and 3.78-12.

220.25 FSAR Section 3.7B.2.2.1 mentioned "the amplified (SRP 3.7.2) response spectra (ARS)".

Please define the new term l

and explain how do they differ form " design response spectra" and " floor response spectra"?

REQUEST FOR ADDITIONAL-INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET NUMBER 50-423 220.0 STRUCTURAL AND GE0 TECHNICAL ENGINEERING BRANCH, STRUCTURAL ENGINEERING ~SECTION (CONT'D) 220.26-FSAR Section 3.8.1.3.1 states that the allowable (SRP 3.8.1) compressive stress in concrete is 0.45fc'. However, the ASME Section III, Division II Code allowable for primary membrance is only 0.30fc'.

Please justify the deviation.

220.27 FSAR Section 3.8.1.3.1 states that structural members (SRP3.8.1) subjected to test prerssure, temperature, or wind, when combined with other forces, are designed for the allowable stresses increased by 33 percent.

Notice that SRP Section 3.8.1.11 Sa allows the increase only for temperature, not others, and you should revise the FSAR to comply with it.

220.28 The contanment load combinations and factors in FSAR (SRP 3.8.1)

Section 3.8.1.3.1 are not identical to that of SRP Section 3.8.1.II 3.

You should either revise the FSAR to comply with SRP requirements or list the deviations and justify them.

220.29 The purposes of the descriptions in FSAR Section (SRP 3.8.1)

(SRP 3.8.1) 3.8.1.3.2 are not clear. You should rephrase them so that the purposes can be understood.

Specifically, Tables 3.8-1 and 3.8-2 are not in i

compliance with the requirements of SRP Section 3.8.1 II 41. You should either revise the FSAR to comply with SRP requirements or list the deviations and justify them.

220.30 Does the method used for tangential shear design in (SRP3.8.1)

FSAR Section-3.8.1.4.1 satisfy the requirements of SRP Section 3.8.1 11 f " Tangential Shear"?. You should rephrase the FSAR to indicate that if it does, or justify the deviations if it does not.

220. 31 Provide temperature profiles that were used for (SRP3.8.1) containment (SRP 3.8.1) thermal analysis for both operating and accident conditions.

220.32 FSAR Section 3.8.1.5.1 states that design of the (SRP 3.8.1) containment equals or exceeds ACI 318-71 requirements for serviceability. Since the ACI 318 Code is for conventional building structures, not for containment structures, we fail to see the connection between containment serviceability and conventional building serviceability. Please list the requirements for serviceability and explain the connection.

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET NUMBER 50-423-220.0 STRUCTURAL AND GE0 TECHNICAL ENGINEERING BRANCH,-

STRUCTURAL ENGINElRING SECTION (CONT'D) 220.33 SRP Section 3.8.1 II 4 requires that an analysis should (SRP 3.8.1) be performed to determine the ultimate capacity of the containment. Please revise the FSAR to comply with this requirement.

220.34 Have the requirements of ACI 318-71 Section 11.6 (SRP 3.8.3)

"Special Provisions for Walls" been adopted as acceptance criteria in FSAR Section 3.8.3.5?

If not, provide reasons to justify it.

220.35 FSAR Section 3.8.4.1 states that non-safety related (SRP 3.8.4) partitions are solid concrete block. Were these concrete block partitions designed to withstand seismic loads so that they will not collapse?

220.36 The loads and loading combinations in FSAR Section (SRP3.8.4) 3.8.4.4 are not identical to that of SRP Section 3.8.4 II 3.

You should either revise the FSAR to comply with l

SRP requirements or list the deviations and justify them.

220.37 Identify the allowable stresses for steel structures (SRP3.8.4) in'FSAR 3.8.4.4 which deviate from that of SRP Section 3.8.4 II 5 and justify them.

220.38 Are any safety-related masonry walls in the plant? If l

(SRP 3.8.4) yes, revise the FSAR to comply with the requirements in Appendix A to SRP Section 3.8.4.

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REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR."0WER STATION UNIT 3 DOCKET NUMBER 50-423~

~

241.0 STRUCTURAL AND GE0 TECHNICAL ENGINEERING BRANCH,

_GE0 TECHNICAL ENGINEERING SECTION Q241.1 Subsurface Profiles 2.5.4.2.4 and You have stated in Section 2.5.4 of the FSAR that some sections of the 2.5.4.5.1 circulating water discharge tunnel and service water intake lines are (SRP 2.5.4) founded on soil. Please identify those sections on a site plan and provide cross-sections along soil. supported sections of discharge tunnel and intake lines. On those cross-sections, subsurface profiles disclosed by exploratory borings and foundation excavation should be presented.

If compacted fill was used to raise the foundation levels, the extent of the fill should be identified and the results of the field moisture and density tests should be provided.

Q241.2 Rock Failures 2.5.4.5.1 (SRP 2.5.4)

Rock failures resulting from blasting during excavation have been reported in the FSAR.

Please provide additional infonnation to identify the locations and extent of those failures.

Cross-sections showing the high angle jointing should be provided.

Q241.3 Structural Backfill 2.5.4.5.2 (SRP 2.5.4).

You have stated in the FSAR that the structural backfill was used to support the control building and diesel generator enclosure building.

Provide plan views and cross-sections showing the extent of the fill placed and the subsurface conditions.

Identify the source of the backfill material used for these buildings and provide the backfill compaction test results obtained during construction.

Q241.4 Soil-Structure Interaction 2.5.4.7 3.7.1.4 You have stated that soil-structure interaction analyses were performed and for control building and emergency generator enclosure.

Identify the 3.7.2.4 subsurface layers and their material properties used in your analyses.

(SRP 2.5.4)

Provide and discuss the results of your analyses.

Q241.5 Concrete Encasement 2.5.4.7 (SRP 2.5.4)

You have reported that the service water intake pipes between the.

circulating and service water. pumphouse and main plant area are embedded j

in a rectangular concrete encasement. Provide the sectional profiles l

and details about the concrete encasement. Also, provide the results of compaction tests performed along the intake pipes.

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REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 0

DOCKET NUMBER 50-423 241.0 STRUCTURAL AND GE0 TECHNICAL ENGINEERING BRANCH, GE0 TECHNICAL ENGINEERING SECTION (CONT'D)

Q241.6 Sliding Stability 2.5.4.7 (SRP 2.5.4)

You state that the service water encasement has been analyzed for sliding stability due to seismic loading.

Provide the details of the j

analysis and identify the cross-section used in your analysis.

Q241.7 Dynamic Response Analysis of Beach Sand 2.5.4.8.3 4

(SRP 2.5.4)

Based on FSAR Figure 2.5.4-35, the thickness of beach sands varies from a few feet to about 50 feet and the thickness of basal till is I

also variable.

In view of this variation of profile, justify that the one dimensional computer program SHAKE is suitable for analyzing j

the dynamic response of the shorefront sand deposits.

Identify the location of the idealized profile used in the analyses. Substantiate i

the assumption that the groundwater is 10 feet below the ground surface using field monitoring results.

2.5.4.8.3 Provide the bases for not using artificial time history conforming 2.5.4.8.4 to R.G. 1.60 design response spectr.a in your dynamic response and analyses of beach sands.

2.5.4.9 (SRP 2.5.4)

Q241.8 Liquefaction Analysis of Beach Sand 4

2.5.4.8.3 (SRP 2.5.4)

Discuss the assumptions used in your liquefaction analysis of beach sand about the possible pore pressure build-up and the potential 4

strength reduction under seismic shaking. Also, discuss the after-earthquake effects, slope instability and lateral movements,on beach 1

sand.

Q241.9 Equivalent Stress Cycles 4

2.5.4.8.3 (SRP 2.5.4)

Provide the bases for assuming that the irregular shear stress time history of the SSE can be represented by five uniform cycles of loading.

Q241.10 Liquefaction Potential 2.5.4.8.3 (SRP 2.5.4)

You have stated that the standard penetration resistance data have been i

used in assessing liquefaction potential. Since the standard penetration resistance depends on many factors, i.e., drill rig type, hammer type, the fall height, the number of turns around the cathead and operator characteristics, provide those information and other relevant information associated with the data cited in the FSAR. Discuss the effects of these parameters on your SPT test data and their influence on liquefaction potential evaluation results.

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET NUMBER 50-423 241.0 STRUCTURAL AND GE0 TECHNICAL ENGINEERING BRANCH,

_GE0 TECHNICAL ENGINEERING SECTION (CONT'D)

Q241.ll Bedrock P_r_ofile 2.5.4.8.4 (SRP 2.5.4)

In the'FSAR Section 2.5.4.8, you state that in the vicinity of the ventilation stack north of Millstone 1, bedrock drops sharply to a trough.

Identify the location of this bedrock trough on a plot plan and provide the subsurface profiles of the trough and the overlying soils.

Information pertinent to the disclosing of the bedrock trough, such as exploratory boring and/or trenching, should be identified and discussed.

Q241.12 Dynamic Response Analysis of Ablation Till 2.5.4.8.4 (SRP 2.5.4)

Identify the location where the idealized profile was obtained for the dynamic response analyses of Ablation Till and justify the groundwater level assumption.

Q241.13 Bearing Capacity 2.5.4.10 (SRP 2.5.4)

Provide the results of your ultimate bearing capacity calculations shown in Table 2.5.4-14 of the FSAR and demonstrate that the minimum safety factor for the allowable bearing capacity exceeds 3.

Also provide bearing capacity evaluation for the discharge tunnel and pumphouse.

Q241.14 Rock Bearing Capacity, 2.5.4.10 (SRP 2.5.4)

Provide the bases for allowing the bedrock bearing load as high as 200 KSF.

Q241.15 Settlement Records 2.5.4.10 and Provide settlement monitoring records for the control, fuel, waste 2.5.4.13 disposal, and emergency diesel generator enclosure buildings.

(SRP 2.5.4)

Q241.16 Lateral Earth Pressure 2.5.4.10.3 (SRP 2.5.4)

Provide the design values of the lateral earth pressures used in the design of rigid, unyielding, foundation walls.

Q241.17 Ring Beam 2.5.4.12 2.5.4.13 You have stated that a reinforced concrete ring beam was placed in the and annular space between the excavation face and the containment exterior 2.5.5.1 wall to stabilize the wedges.

Provide the criteria and bases used for (SRP 2.5.4) the ring beam design.

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET NUMBER 50-423 241.0 JTRUCTURAL 'AND GE0 TECHNICAL-ENGINEERING BRANCH, GE0 TECHNICAL ENGINEERING SECTION (CONT'D)

Q241.18 Dynamic Slope Stability 2.5.5.1 and You state that the computer program LEASE II was used to analyze the 2.5.5.2 stability of the shoreline slope and an undrained shear strength of

'(SRP 2.5.5) 0 equals 20 degrees and C equals 1200 psf was used for beach sands.

Provide the bases for justifying the use of this shear strength for beach sands. Provide a summary of all dynamic soil properties used in the dynamic analyses. Justify the groundwater conditions used in the analyses.

Discuss the possibility of strength loss of beach sands resulting from earthquake shaking and post-earthquake failure potential.

Q241.19 Final Moisture Contents Appendix F Section 3.2 Provide final moisture contents of the soil samples used for cyclic (SRP 2.5.5) triaxial, and resonant column tests (Table 2 & 3).

Q241.20 Grain Size Tests Appendix F Section 3.3 You have presented Lee & Fitton and Kishida's liquefaction envelopes on (SRP 2.5.5)

Figure 2.5.4-30 of the FSAR. Compare the results of your grain size test data on beach sands with these envelopes.

Provide the results of your comparison on a figure and discuss these results as they affect the liquefaction potential evaluation of beach sands.

ENCLOSURE 2 LIST OF QUESTIONS IN ENCLOSURE 1 BRANCH QUESTION NOS.

TOTAL ASB 410.7-410.31 25 CMEB - Fire Protection 280.2-280.27 26 CMEB - Chemical Engineering 281.3-281.13 11 CPB - Physics 491.1 1

GIB 730.1 1

GSB - Seismology 230.3-230.5 3

GSB - Geology 231.1-231.3 3

EHEB - Hydrology 240.1-240.9 9

LQB 630.2-630.11 10 METB - Meteorology 451.1-451.2 2

NETB - Effluent Treatment 460.5-460.19 15 MTEB - ISI 250.2-250.7 6

PSB - Electrical 430.3-430.55 53 PTRB 640.1 1

QAB 260.1-260.57 57 RAB 471.10-471.29 20 SGEB - Structural 220.9-220.38 30 SGEB - Geotechnical 241.1-241.20 20