Text

Evaluation of Fire Protection Exemption Requests from 10CFR50.48 & 10CFR50,App R,Pilgrim Nuclear Power Station, Technical Evaluation Rept
	ML20206Q053
Person / Time
Site:	Pilgrim
Issue date:	08/28/1986
From:	Ahmed N, Bleevan J, Pandey S; CALSPAN CORP.
To:	Stang J; NRC
Shared Package
ML20204J386	List: ML20204J380; ML20204J401; ML20204J417; ML20204J443; ML20204J494; ML20204J531; ML20204J537; ML20204J557; ML20204J579; ML20204J599; ML20204K042; ML20206Q053; ML20207A530; ML20207A535; ML20207A686; ML20207A766;
References
	CON-NRC-03-81-130, CON-NRC-3-81-130, FOIA-88-198 TER-C5506-630, NUDOCS 8609030271
	Download: ML20206Q053 (250)
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TECHNICAL EVALUATION REPORT NRC DOCKET NO. 50-293 FRC PROJECT C5506 NRC LICDiSE NO. DPR-35 FRC ASSIGNMENT 36 N RC CONTRACT NO. NRC-03-81 130 FRC TASK 630 EVALUATION OF FIRE PROTECTION EXEMPTION REQUESTS FROM 10CFR50.48 A!;D APPD; DIX R TO 10CFR50 BOSTON EDISON COMPANY PII4 RIM NUCLEAR POWER STATION TER-C5506-630 Prepared for Nuclear Regulatory Commission FRC Group Leader: N. Ahmed Washington, D.C. 20555 NRC Lead Engineer:

J. Stang August 28, 1996 This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government noe any agency thereof, or any of their employees, makes any warranty, expressed or Implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, appa-j ratus, product or process disclosed in this report, or represents that its use by such third party would not inf ringe privately owned rights.

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TER-C5506-630 CONTENTS Section Title Page 1

INTRODUCTION 1

1.1 Purpose of Review 1

1.2 Generic Background 1

1.3 Plant-Specific Background 3

1.4 Review Criteria.

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EVALUATION 5

2.1 General 5

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2.2 Reactor Building, Elevation (-)l7 Feett Torus Compartment (Fire Zone 1.30A), Control Rod i

Drive Quadrant (Fire Zone 1.6/1.8), and Residual Heat Removal Train A Pump Room (Fire Zone 1.1)

S 2.3 Reactor Building, Torus Compartment, Elevation

(-)17 Feet (Fire Zone 1.30A) 9 2.4 Reactor Building, Steam Tunnel, Elevation 23 Feet (Fire Zone 1.32) 11 3

CONCLUSIONS 13 4

REFERENCES 14 I

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l TER-C5506-630 FOREWORD This Technical Evaluation Report was prepared by Franklin Research Center (FRC) under a contract with the U.S. Nuclear Regulatory Commission (Office of Nuclear Reactor Regulation) for technical assistance in support of NRC operating reactor licensing actions.

The technical evaluation was conducted in accordance with criteria established by the NRC.

Mr. Jacob Klevan contributed to the technical preparation of this report through a subcontract with Rolf Jensen & Associates, Inc.

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INTRODUCTION 1.1 PURPOSE OF REVIEW This technical evaluation report documents an independent review of exemptions er deviations from the fire protection requirements of 10CFR50.48 or Appendix R to 10CTR50 requested for Boston Edison Company's (BECo) Pilgrim Nuclear Power Station.

This evaluation was performed with the following objectives:

o To assess if each exemption request demonstrates an equivalent level of overall protection of plant safo shutdown capability following a disabling fire event To determine the bases for acceptance or denial of each o

exemption request To recommend the minimum level of fire protection the o

Licensee should provide to achieve an equivalent level of fire protection in case a request is d6nied.

1.2 GENERIC BACKGROUND Following a major fire at the Browns Ferry Nuclear Station in March 1975, the NRC established a Special Review Group which initiated an evaluation of the need for improving the fire protection programs at all nuclear power plants.

The group found serious design inadequacies regarding fire protection at Browns Ferry, and its report, "Recommendation Related to Browns Ferry Fire" (NUREG-0050, February 1976), contained over 50 recom-mandations regarding improvements in fire prevention and control in existing facilities.

The report also called for the development of specific guidance for implementing fire protection regulations, and for a comparison of that guidance with the fire protection program at each operating plant.

The NRC developed technical guidance from the technical recommendations in the Special Review Group's report, and issued those guidelines as Branch Technical Position (BTP) APCSB 9.5-1 (1).

This guidance did not apply to plants operating at that time.

Guidance to operating plants was provided later in Appendix A to BTP APCSB 9.5-1 [2), which, to the extent practicable, relies on BTP APCSB 9.5-1.

The guidance in these documents was also published as Regulatory Guide 1.120 (3).

By early 1980, most operating plants had implemented most of the guidelines in Appendix A.

However, the fire protection program has had some significant problems with implementation.

To establish a definitive resolution of these problems in a manner consistent with the general guidelines in Appendix A to the BTP and to assure timely compliance by licensees, the commission issued a proposed fire protection rule and its Appendix R, which was described as setting out minimum fire protection requitcments for the unresolved issues.

The fire protection features addressed TER-C5506-630 included protection of safe shutdown capability, emergency lighting, fire barriers, associated circuits, reactor coolant pump lubrication system, and alternate shutdown systems.

I On February 17, 1981, the final rule 10CTR50.48 (4) and I

Appendix R to 10CTR50 (5) became effective, replacing the proposed rule.

Only three of the 15 items in Appendix R were of such safety I

i significance that they should apply to all plants, including those for which alternative fire protection actions had been approved i

previously by the staff.

These items are protection of safe shutdown capability (including alternate shutdown systems),

emergency lighting, and the reactor coolant pump lubrication system.

Accordingly, the final rule required all reactors licensed to operate before January 1, 1979, to comply with these three items even if the NRC had previously approved alternative fire protection features in these areas.

However, the final rule is more flexible-1 than the proposed rule because Item III.G now provides three i

alternative fire protection features which do not require analysis i

to demonstrate the protection of redundant safe shutdown equipment, l

and reduces the acceptable distance in the physical separation i

alternative from 50 feet to 20 feet.

In addition, the rule now provides an exemption procedure which can be initiated by a i

licensee's assertion that any required fire protection feature will not enhance fire protection safety in the facility or that such modifications may be detrimental to overall safety.

I In summary,Section III.G is related to fire protection j

i features for ensuring that systems and associated circuits used to

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achieve and maintain safe shutdown are free of damage.

Either fire protection configurations must meet the specific requirements of Section III.G or an alternative fire protection configuration must be justified by a fire hazards analysis.

Generally, the staff will accept an alternative fire protection configuration ift i

o The alternative ensures that one train of equipment necessary to achieve hot shutdown from either the control room or emergency control stations is free of fire damage.

4 The alternative ensures that fira damage to at least one o

train of equipment necessary to achieve cold shutdown is limited so that it can be repaired within a reasonable time 1

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(minor repairs using components stored on the site).

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Fire-retardant coatings are not used as fire barriers.

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o Modifications required to meet Section III.G would not enhance fire protection safety levels above those provided by either existing or proposed alternatives.

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Modifications required to meet Section III.G would be o

detrimental to overall facility safety.

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Guidance to the industry and the }TRC's position on certain requirements of Appendix R is covered by various documents, one of them being Generic Letter 83-33 (6), which has recently been superseded by Generic Letter 86-10 (7).

"The Interpretations of Appendix R" and "The Responses to Industry Questions," two sections I

of Reference 7, are written to facilitate industry implementation of Appendix R and represent the NRC's position on all issues covered.

Following the promulgation of the final rule, licensees have requested exemptions and deviations from Appendix R to 10CFR50.

The exemptions and deviations are in the form of a fire hazards I

analysis.

The NRC is to review exemption requests and associated analyses to ensure each alternative to meeting the requirements of the rule provides an equivalent level of overall protection of I

plant safe shutdown capability.

Franklin Research Center (FRC) was to provide technical assistance to the NRC within the context of the following scope of work (8):

l Subtask la Review each exemption request for information deficiencies.

provide Request for Information (RFI) to resolve such deficiencies.

Subtask 2:

Review and evaluate each exemption or deviation request submitted by the licensees and all I

additional information provided for confornance with acceptance criteria, prepare final Technical Evaluation Report (TER) with recommendations, and their basis in support of granting or denying the l

exemption / deviation request.

1.3 PLANT-SPECIFIC BACKGROUND By letter dated November 16, 1983, Boston Edison Company (BEco, the Licensee) requested exemptions from Section III.G of Appendix R.

By letter dated December 27, 1984, the Licensee I

submitted additional information in support of the exemption l

requests.

The exemption requests are the subject of this evaluation.

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The RFI required in satisfaction of Subtask 1 was transmitted to the NRC on February 5, 1986 (9) followed by a site visit on g

April 1, 1986.

The contents of this TER reflect the information contained in the Licensee's submittals identified above, as well as the iniorma-l tion provided or collected during the site visit.

The final TER also incorporates / resolves the NRC's comments received to date.

l 1.4 REVIEW CRITERIA The criteria used in reviewing the Licensee-submitted exemp-tion requests are based on the following documents: I

TER-C5506-630 1.

Fire Protection Program for Operating Nuclear Power Plants, 10CFR50.48 2.

Appendix R to 10CFR50 3.

Standard Review Plan, NUREG-0800, Branch Technical Position (BTP), APCSB 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants" 4.

Appendix A to BTP APCSB 9.5-1 5.

Generic Letter 86-10, "Implementation of Fire Protection Requirements," dated April 24, 1986.

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EVALUATION 2.1 GENERAL This section presents review and evaluation of exemptions or deviations from 10CFR50.48 or Appendix R to 10CFR50 requested by the Licensee (Boston Edison Company, BEco) of Pilgrim Nuclear Power Station.

Evaluation of exemption requests for each fire area / zone r.ingly or collectively follows a format suggested by the NRC and is arranged in the following subsections:

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Exemption requested o

Discussion o

Evaluation o

Conclusion.

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The fire area / fire zone numbering used in this section l

corresponds to that used in the Licensee's submittal.

i 2.2 REACTOR BUILDING, ELEVATION (-)l7 FEET:

TORUS COMPARTMENT (FIRE ZONE 1.30A), CONTROL RCD DRIVE QUADRANT (FIRE ZONE 1.6/1.8), AND RESIDUAL HEAT REMOVAL TRAIN A PUMP ROOM (FIRE ZONE 1.1) i 2.2.1 Exemption Requested Exemptions were requested from the techaical requirements of Section III.G.2.a to the extent that it requires separation of redundant trains of residual heat removal (RMR) and core spray syhtsms, automatic depressurization system (ADS), and emergency dicy.el generator fuel oil transfer pump cables located in Fire Zones 1.1, 1.6/1.8, and 1.30A, respectively, by 3-hour fire rated

%arriers.

2.2.2 Discussion 2.2.2.1 Elevation (-)17 Jeet The Licensee has identified the following conditions which do not meet the technical requirements of section III.G.2.at redundant trains of the RHR and core spray systems, ADS, and emergency diesel generator fuel oil transfer pump cables are not separated from each other by 3-hour rated fire barriers at the boundary between Fire Zones 1.6/1.8 and 1.30A, as well as et the boundary of Fire Zones 1.30A and 1.1.

Each of the subject fire zones is located in the reactor building.

The reactor building is divided by concrete floor elabs into six elevations:

(-)li feet, 6 inchest 2 feet, 9 inches; 23 feet 51 feet; 74 feet, 3 inches; 91 feet, 3 inches; and 117 feet.

I It is divided into fire areas and several fire zones.

The firo 1

TER-C5506-630 sepurated from each other by fire rated barriers with area. uv protected openings.

The fire zones are separated from each other by fire rated barriers penetrated by unprotected openings.

This exemption request involves fire zones located on or adjacant to elevation (-)l7 feet.

The reactor building, elevation (-)l7 feet, is divided into five fire zones.

Fire Zone 1.30A comprises the majority of this elevation.

It is bounded by Fire Zone 1.1 in the southeast, Fire Zone 1.2 in the northuest, Fire Zone 1.6/1.8 in the northeast, and Fire Zon0 1.5/1.7 in the southwest quadrants of this elevation.

It is separated from the other zones (quadrants) by 36-inch hick concrete walls.

Penetrations in each wall consist of an unprotected doorway and a small amount of nonrated mechanical and electrical penetrations.

Elovation (-)l7 feet is connected to elevation 23 fast by open stairwayr located in Fire Zones 1.6/1.8, 1.1, 1.2 and 1.5/1.7.

Fire Zones 1.6/1.8 and 1.1 are open to Fire Zone c.9 on elevation 23 feet.

Fire Zones 1.2 and 1.5/1.7 are open to Fire Zone 1.10 on elevation 23 feet.

The combustible contents of Fire Zone 1.1 consist of cable insulation and lube oil.

12,200 Btu per square foot, which produces an equivalent fireThe combustible l severity of 9 minutes on the ASTM E-119 time-temperature curve.

Fire protection in this zone consists of smoke detectors and a manual hose station.

The combustible contents of Fire Zone 1.2 consist of cable insulation and lube oil.

12,200 Btu per square foot, which produces an equivalent fireThe combustible severity of 9 minutes on the ASTM E-119 time-temperature curve.

Fire protection in this zone consists of portable fire extin-guishers and a manual hose station.

insulation and lube oil.The combustible contents of Fire Zone 1.5/1.7 c 14,400 Btu per square foot, which produces an equivtlent fireThe combustible l severity of 11 minutes on the ASTM E-119 time-temperature curve.

The fire protection in this zone consists of smoke detectors, portable fire extinguishers, and a manual hose station.

insulation and lube oil.The combustible contents of Fire Zone 1.6/1.8 c 4,800 Btu per square foot, which produces an equivalent fireThe combustible l coverity of a cture curve. pproximately 4 minutes on the ASTM E-119 time-temper-fire extinguisher and a manual hose station. Fire protection in this zone consists The combustible contents of Fire Zone 1.9 consist of cable insulatir.1.

The combustible loading is approximately 39,200 Btu per square foot, which produces an equivalent fire severity of opproximately 30 minutes on the ASTM E-119 time-temperature curve. I i

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Fire protection in this zone consists of portable extinguishers and manual hose stations.

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The combustible contents of Fire Zone 1.10 consist primarily of cable insulation.

The combustible loading is approximately 30,400 Btu per square foot, which produces an equivalent fire severity of 23 minutes on the ASTM E-119 time-temperature curve.

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Fire protection in this zone consists of portable fire extinguishers and manual hose stations, i

The combustible contents of Fire Zone 1.30A consist of 24 pounds of cable insulation in one cable tray, which is approxi-mately 125 feet long and approximately 8,'200 pounds of fire-retardant painted. wood. staging t (scaf folding) encircling the torus.

The combustible loading is approximatelys5,900 Stuiper square foot,.

which produces an equivalent fire severity of approximately 4 minutes on the ASTM E-119 time-temperature curve.

No fire protec tion systems or equipment are installed in this fire zone.

However, there is one hose reel and one portable extinguisher in each adjacent quadrant that will reach and cover this fire zone.

2.2.2.2 Fire Zone 1.30A Fire Zone 1.30A contains cables associated with the following i

safe shutdown systems:

o RHR train B o

Core spray train B i

Emergency diesel generator trains A and B fuel oil transfer o

pumps Reactor core isolation cooling (RCIC) o High pressure coolant injection (HPCI) o o

Torus water temperature.

The Licensee has committed to reroute power cables out of Fire Zone 1.30A that feed MCC B18 components required for train B of the RHR and core spray systems.

The cables will be routed out of the reactor building through the west wall, around the exterior, and in through the east wall.

The Licensee has also committed to reroute i

cables associated with torus water temperature and both trains of fuel oil transfer pumps out of this fire zone as described in the Licensee's letters dated June 25, 1982 and May 17, 1983.

Upon completion of the modifications, no train A or B safe chutdown components or cables will be physically present in this fire zone.

However, this fire zone is not separated by complete 3-hour fire rated boundaries from Fire Zones 1.2 and 1.10, which contain train B components required for safe shutdown.

This fire i

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zone is also open to Fire Zone 1.6/1.8 via an open stairway.

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Z'ne 1.6/1.8 does not contain any safe shutdown components, but it o

is open to Fire Zone 1.9 by an open stairwell which contains train A components required for safe shutdown.

Therefore, Fire Zone 1.30A provides a path between train A components located in Fire Zone 1.9 and train B components located in Fire Zones 1.2 and 1.10.

The minimum distance between train A and B components along this path is at least 100 feet horizontally (between the openings in the fire barriers separating Fire Zone 1.30A from Fire Zones 1.2 and 1.6/1.8) and 40 feet vertically between elevations (-)17 and 23 i

feet.

2.2.2.3 Fire Zones 1.2, 1.5/1.7, and 1.10 Fire Zone 1.2 contains cables and equipment associated with train B of the RHR and core spray systems.

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Fire Zone 1.5/1.7 contains cables and equipment associated' with the RCIC system only.

t Fire Zone 1.10 contains cables and equipment associated with train B of the RHR and core spray systems, ADS, and emergency diesel generator fuel oil transfer pump, as well as the HPCI and RCIC systems.

Fire Zones 1.2, 1.5/1.7, and 1.10 are separated from each other and from Fire Zone 1.30A as described in Section 2.2.2.2.

In addition, Fire Zon6L 1.10 and 1.9 are r"Jarated from each other by a sprinkler water curtain on elevatiot feet.

The separation distance between the train B componenth

,r cables in these zones and the closest train A-designated zone is at least 100 feet horizontally.

2.2.2.4 Fire Zonea 1.1, 1.6/1.8, and 1.9 Fire Zone 1.1 contains cables and equipment associated with train A of the RHR and core spray systems.

The closest redundant train B components are located in Fire Zone 1.2, approximatsly 150 feet from Fire Zone 1.1, which contains ccunterpart train A.

j fire Zone 1. 6/h4*e49nteine-ce6 saf a shutdetsv9thkee ow.e@ip-ment.

However, ~3't is open to71re Zone'T. 9dn' e'levati'on '23 feet, which contains cables associated with trains A and B of the RHR and core spray systems, ADS, and emergency diesel generator fuel oil transfer pumps.

The Licensee has committed to relocate cables associated with train B of the above-named systems out of Fire Zone 1.9.

The closest train B components are in Fire Zone 1.2, which is located,4 Bone 136/b 4, which contains train J.

approximately 100 feet horizontally across Fire Zone 1.30A l

from Pit 2.2.3 Evaluation The fire protection in Fire Zones 1.30A, 1.6/1.8, and 1.1 does not comply with the technical requirements of Section III.G.2.a of Appendix R because redundant trains of the RHR and core spray i,

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TER-C9506-630 systems, ADS, and emergency diesel generator fuel oil transfer pump cables are not separated by fire barriers having 3-hour ratings.

The concern was that the lack of 3-hour fire rated barriers between the redundant trains may result it. a loss o.! redundant safe shutdown capability.

However, the equivalent fire severity in any of these fire zones is less than 30 ainutes.

Therefore, a,iire of significant magnitude or severity is not expeci,ed to occur.

Also, the burning rate of the combustibles is expected to be limited because most combustibles are enclosed (oil and lubricants in pumps) or treated to reduce combustibility (fire-retardant painted wood and fire retardant coated cable).

If a fire should occur in Fire Zone 1.30A, it is expected that it would be detected by fire detectors in Fire Zones 1.5/1.7 or 1.1.

The detectors annunciate in the control room to alert the control room operators.

They, in turn, would alert the fire bri-gade to respond to the reactor building and extinguish the fire.

If the fire was not detected promptly, it is expected that it would not result in a loss of safe shutdown capability for the following reasons:

The separation distance is 100 feet or more between redundant safe shutdown systems in Fire Zones 1.2 and 1.1 or 1.6/1.8 and the combustible loading in Fire Zone 1.30A is limited, which would not allow fire gas temperatures to exceed the limit at which damage to cables of equipment is expected to occur.

In addition, the openings between elevations (-)l7 and 23 feet would further limit maximum fire eas tempratures because of the mixing with erler air.

If a fire occurred in one of the quadrants at elevation b)l7 feet, similar results are expected.

With the installed fire protection features in conjunction with the committed modifications, reasonable assurance exists that a fire originating in the above-described sections of Fire Zones 1.30A, 1.2, 1.1, 1.6/1.8, or 1.5/1.7 would not prevent the plant from safely shutting down.

2.2.4 Conclusion 1

3 Based on the above evaluation, it is concluded that the l

existing fire protection features combined with the proposed modifications provide an acceptable level of protection for redundant trains of ths RHR and core spray systems, ADS, and l

emergency diesel generator fuel oil transfer pumps on elevation

(-)l7 feet.

Therefore, the exemptions can be granted.

I 2.3 REACTOR BUILDI!iG, T'mlS COMPARTMEllT, ELEVATIoli (-)17 FEET (FIRE 20!!E 1.30A) 2.3.1 Exemption Requested An exemption was requested from the technical requirements of Section III.G.2.a to the extent that it requires structural steel forming a part of or supporting the fire barrier between redundant l

TER-C5506-630 trains of safe shutdown components in Fire Zone 1.30A and Fire Zones 1.9 and 1.10 to be protected to provide fire resistance equivalent to that required of the barrier.

2.3.2 Discussion The concrete floor slab which separates Fire Zone 1.30A from Fire Zones 1.9 and 1.10 above is supported by unprotected otructural steel beams.

The combustible materials in Fire Zone 1.30A are primarily located 18 feet below the structural steel in the form of fire-retardant painted wood staging.

The other significant combustible material, including cable insulation, is located 2 feet below the steel in a 12-inch-wide cable tray.

The Licensee has committed to modifications described in Section 2.2.2.

Upon completion of these modifications, Fire Zone 1.30A will contain only RCIC and HPCI components, which provide citernate shutdown capability for the RHR and ADS components contained in Fire Zones 1.9 and 1.10 located above Fire Zone 1.30A.

The Licensee has performed an analysis of the effect of instantaneous complete combustion of the entire combustible load of this fire zone on the structural steel.

The analysis assumed that all of the heat from the fire was immediately absorbed by only the otegl, and that the steel would fail if it reached a temperature of 650 F.

The Licensee's analysis indicated that based on the total combustible loading, the sgeel would not experience more ghan an average temperature of 326 F, which is well below the 650 F assumed to fail the structural steel.

2.3.3 jdalestion The fire protection in Fire Zone 1.30A does not comply with the technical requirements of Section III.G.2.a of Appendix R because structural steel forming a part of or supporting the fire barrier between redundant safe shutdown systems in Fire Zone 1.30A and Fire Zones 1.9 and 1.10 is not protected to provide fire resistance equivalent to that required of the barrier supported.

The Licensee's analysis indicates that the structural steel would not fail even if it instantaneously absorbed the entire heat of combustion of the combustible materials in Fire Zone 1.30A.

However, the Licensee's analysis does not take inte account the offect on the steel of a fire plume impinging directly on a otructural steel member.

Because the cable tray is located spproximately 2 feet below the structural stgel, a fire in the tray might create air temperatures as high as 650 F at the lower flange of the steel.

Therefore, there is not reasonable assurance that a fire in this zone would not jeopardize the structural steel, creating a situation which will impair the safe shutdown capability.. - -

TER-C5506-630 2.3.4 Conclusion Based on the above evaluation, it is concluded that the existing fire protection features do not provide an acceptable level of protection for redundant trains of cables and equipment located in Fire Zones 1.9 and 1.10.

should not be granted.

Therefore, the exemption 2.4 REACTOR BUILDING, STEAM TUNNEL, ELEVATION 23 FEET (FIRE ZONE 1.32) 2.4.1 Exemption Requested An exemption was requested from the technical requirements of Section III.G.2.a to the extent that it requires structural steel forming a part of or supporting the fire barrier between Fire Zone 1.32 and Fire Zones 1.11 and 1.12 to be protected to provide fire resistance equivalent to that required of the barrier.

2.4.2 Discussion The Licensee has identified the following condition which does ~

not meet the technical requirements of Section III.G.2.a:

The structural steel beam supporting the floor slab separating Fire Zone 1.32 from Fire Zones 1.11 and 1.12 is not protected to fire resistance equivalent to that required of the barrier. provide Fire Zone 1.32 is located on elevation 23 feet.

It adjoins the containment to the north, Fire Zone 1.9 to the east, and Fire Zone 1.10 to the west.

It is located below Fire Zones 1.11 and 1.12 on elevation 51 feet of the reactor and turbine buildings.

Fire Zone 1.32 is separated from Fire Zones 1.11 and 1.12 by a concrete floor slab supported by one structural steel beam.

Fire Zones 1.11 and 1.12 contain redundant safe shutdown systems.

The combustible contents of Fire Zone 1.32 consist of a few exposed electrical cables.

The majority of the cables in this fire zone are routed in conduits.

There are no other combustible materials in the fire zone.

Fire protection consists of a portable fire extinguisher and a manual hose station in an adjacent area.

The steam tunnel (Fire Zone 1.32) conthins the RCIC and HPCI systems. The Licensee has stated that the loss of these systems does not prevent safe shutdown.

Fire Zones 1.11 and 1.12 contain safety-related core spray and RHR valves and safety-related cable trays and panels.

l The Licensee performed an analysis to determine the quantity of combustible material which wguld be required to raise the temperature of the steel to 650 F, above which it would fail to oupport the floor.

The analysis indicated that a combustible loading of 21,500 Btu per square foot would be required.

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i TER-C5506-630 Licensee concluded that, since the actual combustible loading in this fire zone is negligible, the steel would not experience high temperature to fail.

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2.4.3 Evaluation The fire protection in Fire Zone 1.32 does not comply with the technical requirements of Section III.G.2.a of Appendix R because i

structural steel forming a part of or supporting the fire barrier between Fire Zone 1.32 and Fire Zones 1.11 and 1.12 is not protected to provide fire resistance equivalent to that required of the barrier supported.

The Licensae's analysis indicates that the structural steel would not fail even if it instantaneously absorbed the entire heat of combustion of the combustible materials present in Fire Zone 1.32.

Although the Licensee's analysis does not take into account the effect of a fire plume impinging directly on a structural member, it is not expected that such an exposure fire would be significant because of the negligible combustible loading.

Therefore, reasonable assurance exists that a fire originating in this fire zone will not prevent the plant from safely shutting down.

2.4.4 conclusion Based on the above evaluation, it is concluded that the existing fire protection features for the structural steel in Fire Zone 1.32, which support the floor of Fire Zones 1.11 and 1.12, provide an acceptable level of protection for the redundant trains of cables and equipment located in Fire Zones 1.11 and 1.12.

Therefore, the exemption can be granted.

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CONCLUSIONS This section is provided to consolidate the results of the cvaluation contained in Section 2 concerning the exemptions requested by the Licensee from the requirements of Section III.G of Appendix R to 10CFR50 for Pilgrim Nuclear Power Station.

It is not meant as a substitute for the specific conclusions reached in the various subsections of Section 2 for which the reader is referred to specific subsections.

Based on the evaluation, the level of fire nafety in the fire zones listed below is equivalent to that achieved by compliance with the technical requirements of Section III.G of Appendix R and, therefore, the Licensee's requests for exemptions in the following creas can be granted.

1.

Fire Zones 1.30A, 1.1, and 1.6/1.8.

Lack of fire barriers separating redundant trains of cables and equpiment.

Refer to Section 2.2 for details.

2.

Fire Zone 1.32.

Lack of fireproofing of structural steel supporting or forming a part of the fire barrier between Fire Zone 1.32 and Fire Zones 1.11 and 1.12.

Refer to Section 2.4 for details.

Based on the evaluation, it is found that the level of fire safety in the zone listed below has not been shown to be in compliance with the technical requirements of Section III.G of Appendix R and, therefore, the Licensee's request for exemption should not be granted:

1.

Fire Zone 1.30A.

Lack of fireproofing of structural steel supporting or forming a part of the fire barrier between Fire Zone 1.30A and Fire Zones 1.9 and 1.10.

Refer to Section 2.3 for details.

TER-C5506-630 4.

REFERENCES 1.

BTP APCSB 9.5-1 "Fire Protection Program," July 1981 (Standard Review Plan, NUREG-0800) 2.

Appendix A to BTP APCSB 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976,"

August 23, 1976 3.

Regulatory Guide 1.120, "Fire Protection Guidelines for Nuclear Power Plants," November 1977 4.

10CFR50, "Fire Protection Program for Operating Nuclear Power Plants," November 19, 1980 5.

Appendix R to 10CFR50, "Fire Protection Program for Nuclear Power Facilities operating Prior to January 1, 1979," November, 19, 1980 6.

Generic Letter 83-33, "NRC Position on Certain Requirements of Appendix R to 10CFR50," October 19, 1983 7.

Generic Letter 86-10, "Implementation of Fire Protection Requirements," April 24, 1986 8.

Final Work Assignment No. 36, transmitted by M. Carrington (NRC) to Dr.

S.

Pandey (FRC) on July 17, 1985 9.

Request for Information (RFI), transmitted by N. Ahmed (FRC) to J. Stang (NRC) on February 5, 1986.

I Dated August 28, 1986 l

I BWR CONTAINMENT PERFORMANCE BWROG ACTIVITIES 0

MEETING WITH R. BERNER0 -- SEPTEMBER 11, 1986 o

PRESENTATION TO ACRS SUBCOMMITTEE ON CONTAINMENT PERFORMANCE -- SEPTEMBER 23, 1986 O

ONG0ING/ ANTICIPATED WORK ON SPECIFIC PROPOSALS (SHORT TERM)

APPROPRIATE VENT SIZING DRYWELL SPRAY DESIGN BASIS FEASIBLE MODIFICATIONS OF EXISTING DRYWELL SPRAY CONFIGURATIONS EVALUATION OF DEINERTED OPERATION o

STAFF REVIEW 0F EMERGENCY PROCEDURE GUIDELINES (REV. 4)

UNDERWAY BWROG/NRC SEMINAR IN BETHESDA -- OCTOBER 7-9, 1986 0

INTERACTION WITH NUMARC WORKING GROUP o

MEETING WITH R. BERNER0 PRIOR TO CRGR REVIEW 0F PROPOSED GENERIC LETTER (MID-NOVEMBER 1986)

SN

STATUS OF EFFORTS RELATED TO SPECIFIC PROPOSED REQUIREMENTS 1.

HYDR 0 GEN A.

MARK I/II:

EVALUATION LIMITATIONS ON TIME DEINERTED B.

MARK 111:

COMPLETION OF HC0G PROGRAM 11.

SPRAYS A.

REEVALUATION OF CURRENT DESIGN BASIS REQUIREMENTS B.

EVALUATION OF HOW BEST TO MODIFY CURRENT SPRAY CONFIGURATIONS C.

FURTHER REVIEW 0F POTENTIAL BENEFITS / RISKS III.

PRESSURE A.

PRELIMINARY ANALYSIS (SIZE VENT FOR DHR, NOT ATWS)

B.

FURTHER WORK (PROBABLY) TO DETERMINE APPROPRIATE SIZING 0F VENT FOR DHR C.

PROVIDE GUIDANCE TO UTILITIES REGARDING VENTING PATHWAYS /AVAILABLE OPTIONS

r STATUS OF EFFORTS RELATED TO SPECIFIC PROPOSED REQUIREMENTS (CONTINUED)

IV.

CORE DEBRIS A.

BENEFITS LOW, VARIOUS NEGATIVE IMPACTS B.

N0 FURTHER STUDY WARRANTED OR PLANNED V.

TRAINING AND PROCEDURES A.

COMMITMENT T0 IMPLEMENT EPGs REV, 4 B.

PROPOSAL TO PERFORM REVIEW 0F REV, 4 T0 "FINE TUNE" FOR SEVERE ACCIDENT INSIGHTS

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October 29, 1986 j

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Mr. Edward A. Thomas FEMA - Room 462 J. W. McCormack Post Office and Court House Boston, Massachusetts 02109 1

RE:

Petition for Show Cause Concerning Pilgrim I Nuclear Power Station Dear Mr. Thomas-As me had previously discussed, enclosed is Boston Edison Company's response to the "Emergency Response Plan" allegatiors set forth in the "Petition for Show Cause Concerning Pilgrim I Nuclear Power Station" that was submitted to the NRC July 15, 1986.

Should you have any questions, clease feel free to call either myself or Roger Silva.

Very truly yours, I

cc: Mr. Robert Boulay i

Mr. Peter Agnes j

/Imc Enclosure i

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10/29/86 j

Boston Edison Company Response to Emergency Response Plan Allegations Contained in Petition for Show Cause Concerning Pilgrim I Nuclear Power Station INTRODUCTION On July 18, 1986 a document entitled "Petition for Show Cause Concerning Pilgrim I Nuclear Power Station" (the "Show Cause Petition") was submitted to the United States Nuclear Regulatory Commission by William 8. Golden et al.

The Show Cause Petition is based upon allegations relating to licensee l

management, the radiological emergency rasponse plan and the facility's 1

containment structure.

Subsequently, the NRC has referred the portions of the Show Cause Petition (i.e., Sections 14-20) pertaining to emergency response plans to the Federal Emergency Management Agency for review.

This document contains Boston Edison Company's response to the emergency response plan allegations contained in the Show Cause Petition.

This response consists of a general overview of the allegations pointing out several common deficiencies, l

and is followed by an item by item response to each major allegation, including documentation wherever appropriate.

GENERAL RESPONSE Sections 14-20 of the Show Cause Petition entitled "Emergency Response Plan" are in essence a re-hash of a 1983 "study" by MASSPIRG entitled "Blueprint for Chaos !!" (the "Chaos II Report").

The Show Cause Petition liberally cites or is based upon various "findings" of the Chaos !! Report; however, the Show Cause Petition falls completely to note that the Chaos II Report was previously the basis of a 10 CFR 2.206 petition to the hRC (the "W SSPIR3 Petition"), which was previously referred to FEMA for analysis and

response, and t;hich tas subsequently rejected by both FEMA ("Analysis of Emergency Preparedness Issues at Pilgrim Nuclear Power Station Raised by the Massachusetts Public Interest Group (MASSPIRG)", November 3, 1983; "Analysis Report on Issues Related to the Pilgrim Evacuation Time Estimate Pilgrim Nuclear Power Station Plymouth, Massachusetts", May 1,1984) a',10 the NRC (Director's Oe:ision 00-84-5, February 27, 1984; Director's 0icision 00-84-15, July 3, 1984).

The reasons for rejection of the Chaos II Report remain valid.

The current Show Cause Petition contains essentially no new information and in addition falls to acknowledge significant improvements in the emergency planning area since that time.

)

Apart from almost verbatim repetition of allegations originally contained in the 1983 Chaos II Report, essentially the only other source of information relied upon in the Show Cause Petition is a June 18, 1986 legislative hearing j

before a committee of the Massachusetts State Senate.

Although Boston Edison representatives attended this hearing, presented testimony and responded to many of the issues raised at the hearing and repeated in the Show Cause Petition, there 15 no reference to any of that testimony or those responses.

In addition it is our belief that many of the statements attributed to various individuals are either inaccurate, incomplete or taken out of context.

Subsequent to receipt of the Show Cause Petition, Boston Edison attempted to obtain a transcript of this hearing in order to verify some of those statements and in order to place those statements in proper context.

Those efforts were not successful and there is substantial doubt whether there is in fact any transcript of this hearing.

While this response will address each of J

the issues assertedly arising from this hearing, it should be observed that 2_

such a hearing ci%h no writ %en record represents a very insubstantial and unreliable basis for the numerous allegations that have been made.

A third significant shortcoming of the Show Cause Petition is its failure to address developments that have occurred since the time that much of its information base was developed.

By its very nature emergency planning is a dynamic process involving continually changing situations and personnel.

Key elements such as training, drills, tests and dissemination of pubile information are updated frequently.

In the event problems are observed, it is frequently possible to undertake appropriate remedial actions.

In many cases i

the Show Cause Petition cites a deficiency that may have existed at one particular time but then overlooks actions that were taken to remedy identified shortcomings.

In other cases actions are either planned or underway to address such issues.

In no case has an issue been identified that has not been resolved or that is not susceptible to reasonable and timely resolution.

In summary, the allegations contained in the Emergency Response Plan section of the Show Cause Petition are without merit and do not form a basis for action by the NRC under 10 CFR 2.206 or by FEMA under 44 CFR 350.13.

Recent actions, such as the September 29, 1986 stren test, and upcoming actions, such as the emergency exercise currently scheduled for December, 1986, when coupled with ongoing NRC and FEMA inspection and oversight provide a far more substantial and significant basis for evaluation of emergency l

response capability, l

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fTEM BV ITEM RFSPONSE Section 14: "Deficiencies in the Radiological Emergency Response Pla_n (RERP)"

Although Boston Edison has substantial disagreement with the caption and the broadly stated and unsupported conclusion of tnis Section, this Section does contain a generally accurate description of the substantive regu'latory provisions relating to emergency planning.

It should be observed, however, that the paragraph refers throughout to 10 CFR 50.47, which deals with i

applications for licenses, rather than to 10 CFR 50.54(s) and (u), which applies to those plarts which already have operating licenses. One further point that should be noted in discussing emergency planning regulations is that under both NRC (10 CFR 50.54(s)(2)(ii)) and FEMA regulations (44 CFR 350.13(a)), the appropriate action in the event of a significant emergency plan deficiency is to provide formal notification and thereaf ter allow a period of time for corrective actions or for a plan for corrective actions.

In contrast a petition such as this Show Cause Petition under 10 CFR 2.206 in effect by. passes the regulatory process designed to deal with emergency plan deficiencies. Allegations such as those contained in the Show Cause Petition should be reviewed in light of the 10 CFR 50.54 and 44 CFR 350.13 process for remediation of deficiencies and such process should be allowed to function to the maximum extent oossible.

Simply stated a petition under 10 CFR 2.206 is an inappropriate vehicle for review of the issues raised in the Show Cause Petition.

1 4

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Seg% ion 15: "Deficiencies in Adeance Information" a) Subsection (a) repeats allegations contained in the 1983 Chaos !! Report, specifically referring to a 1983 telephone survey which purports to show that area residents have received inadequate advance information concerning the Pilgrim emergency plan.

Boston Edison would seriously question the scientific and statistical basis for this survey, particularly with reference to its current applicability to conditions in 1986.

In any event, these allegations were rejected by FEMA in its November 3, 1983 Report on the MASSPIRG Petition and the reasons for such rejection remain valid.

t Subsection (a) also makes the claim that there have been no improvements in advance information procedures since 1983.

This claim is completely false.

First there has been substantial review of and subsequent revisions to the "Emergency Pubile Information" pamphlet.

(Attachment I

1).

Additional mailings of the pamphlet took place in September, 1985 and August, 1986.

The August, 1986 mailing was accompanied oy a press release (Attachment 2) and newspaper advertisements appearing in local media (Attachment).

Since 1983 procedures have been implemented to assure annual malling to all residences and businesses in the ten mile Emergency Planning Zone.

(Attachment 4).

In addition, as discussed in greater detail below in response to Section 20, there have been substantial j

leprovements in Boston Edison's facilities, staffing and organization as i

related to emergency planning in order to assure continued progress in a

this area.

I j :

1 1

Finally the claim is made % hat the "only" me% hod being used for adtrance public educa%1on is the distribution of pamphlets by mall.

While %his is certainly a method, if not the principal method, this allegation rather simplistically overlooks the tremendous amount of public information and education concerning the emergency plan transmitted through the media, public meetings and other modes of public communication.

It also overlooks the existence of signs and posters in various public locations (see response to Section 15(c)).

In addition next year it is planned to include ads in the yellow pages.

The recent extensive public and media debate over emergency planning at Seabrook, over management and safety issues at Pilgrim and ' esulting from the recent political compaigns focussing upon nuclear issues has also undoubtedly led to additional public awareness.

The same could be said of the media attention (including news articles (Attachments 5-7) and Boston Edison advertisements in the local newspapers (Attachment 8) surrounding the September 29, 1986 siren test as well as of the expected media attention that will accompany the upcoming emergency exercise currently scheduled for later this Fall.

In addition Boston Edison provided an informational mailing to all residents concerning the recent stren test.

Considering the "Emergency Public Information" pamphlet in conjunction with all of the other Avenues of public information. Boston Edison would conclude that there is in fact a great deal of public awareness of the existence of an emergency plan and of the key element of that plan - i.e.,

when alerted, go inside and tune to a local radio or television station and await further instructions.

b) Subsection (b) alleges that the current public information pamphle%s contain no information on public transportation.

Again this is a repetition of an allegation contained in the Chaos II Report which was dealt with in FEMA's November, 1983 Report. More fundamentally the allegation is both incorrect and fails to recognize the proper role of the advance public information pamphlet as opposed to the more detailed Town or State emergency plan or the specific information which would be communicated over the television or the radio.

First it is inaccurate to say the pamphlet contains "no" information on public transportation since there are several directions (on p. 7, for example) relatir.g to those who will need transportation assistance.

In addition there is a card to fill out and return to the Massachusetts Civi; Defense Agency for those who will need such assistance.

Second the principal focus of the pamphlet is upon the vast majority of the populace who may be expected to utilize private transportation - either their own or with friends or neighbors.

Thirdly the pamphlet is not designed to contain the entire emergency plan l

- only to contain certain categories of information (i.e., that specified in NUREG-0654, Evaluation Criteria G.I. G.2).

As a final note Boston Edison has no objection to reasonable enhancements to the public information partphlet in this or in other respects and will, in conjunction with appropriate Town, MCDA and FEMA officials, consider revisions in this area.

j c) Subsection (c) alleges that there is inadecuate advance information for tourists and transients and that no public information signs have been posted. Again this is a repetition from the Chaos II Report which was previously addressed by FEMA.

Again the allegation is inaccurate.

In --

fact there are posters (for a sample, see Exhibl% No. 4 to A22achment 2) at numerous locations throughout the ten mile EP2. Approximately 30 such Posters have been placed at fire and police stations, parks, town halls, public libraries, etc. throughout the area.

In addition copies of the Emergency Public aformation pamphlet have been distributed, and will be updated annually, to the various motels and other places of public accommodation in the area. is the Boston Edison Company procedure providing for such distribution.

Section 16: "Deficiencies in Notification During an Accident" This Section includes a number of allegations regarding the stren system.

Again these allegations are without merit and are based entirely (except for the limited references to the June 18, 1986 hearing) upon the Chaos II Report.

The false alarm problem caused by electrical storms is essentially ancient history - there is at the present time no disconnection of sirens during storms.

The audibility concern is also an issue from the past.

In any event, the results from the September 29, 1986 stren test, accompanied by any correcth2 actions that may be found appropriate, will bJ a far more accurate basis for evaluating the adequacy of the siren system than are the allegations in this Section of the Show Cause Petition, Preliminary reports of the September 29 test show that 90 of 94 strens worked properly and that 88.2% of the population surveyed heard the test.

In addition to the full scale test Boston Edison also provides FEMA and MCDA with monthly $1ren test results.

(See Attachment 9).

One further point that should be made with regard to emergency notification is that the siren system is not the only means of providing such

-a.

public notification.

In addition to the s'rens there eili be notification by means of radio and television (the "EBS"), by means of tone-alert radios and by means of mobile teams.

Beyond such formal means there will also be informal communications between neighbors, family members and friends.

The i

combination of such means, spurred perhaps initially by the sound of very loud

{

l sirens which will be heard by most if not all persons in the vicinity, gives reasonable assurance that all will be notified in a prompt manner of the existence of an emergency situation.

l Section 16 also raises a concern over notification of the hearing l

impaired. Again this is an issue that was raised initially in the Chaos II Report and which was responded to at that time.

The vague reference to testimony of unnamed persons at the June 18, 1986 meeting adds nothing additional in support of the allegation.

In fact, Boston Edison has engaged in efforts to identify the hearing impaired population (see, e.g., letter to the Massachusetts Commission for Deaf and Hard of Hearing dated July 23, 1986 which is Attachment 10) in order to investigate installation of teletypewriter (TTY) systems where appropriate, and Boston Edison will continue those efforts as necessary.

Apart frc,m the use of a TTY system, it should also be observed l

that many of the hearing-tmpaired, as well as other persons with special l

needs, will be a part of a household that includes persons who are not hearing impaired or will have friends, relatives or neighbors nearby who can provide assistance.

Whlie sole rellance should not be placed on such Informal

.ietworks of cccounication and assistance, as a matter of fact it should be observed that such networks exist and function throughout society.

l The final issue raised in this Section concerns a purported failure to deliver to FEHA certain technical data on the $1ren system.

Whatever the

.g l

i prior ealidity of such an issue, the technical data has noa been delleered (see Attachment 11) and the full scale siren test has occurred.

Insofar as this issue is raised to demonstrate alleged deficiencies in cooperation and coordination, please refer to Boston Edison's response to Section 20.

Section 17: "Deficiencies in Evacuation Plans" a)

Subsection (a) repeats allegations concerning the inaccuracy of evacuation time estimates which were initially presented in the Chaos II Report and which were reviewed and rejected by FEMA at that time.

In this connection, it should be observed that the principal purpose of evacuation time estimates is to enable decision makers to decide on the appropriate protective action in a given situation rather than to mandate evacuation of all persons in all situations.

It is also the case that the methodology underlying the performance of evacuation time estimates has evolved greatly over the past decade.

Initial evacuation time estimates were submitted for Pilgrim in January, 1980 in response to a Novembs: 29, 1979 request for information from the Emergency Preparedness Task Group of the NRC.

NRC comments on these initial time estimates as reported in NUREG CR-1856 published in May, 1981 rank the time estimate studies for Pilgrim as representing the state of the art and being among the best such studies for any nuclear plant in the country.

Existing evacuation time estimates are based upon a September, 1980 study by HMM Associates, Inc.

In addition to the basic study there was also a supplemental study by HMH Associatesaddressing the Sagamore / Buzzards Bay area immediately to the south of the EPZ as well as a supplemental study including projections of future population growth eith time estimates for later years reflecting such increases.

See Massachusetts Radiological Emergency Response Plan, Appendix 3, Section C-1, pp. C-83 to C-86.

All of these reports were thoroughly reviewed by the NRC and by FEMA in denying the 1983 MASSPIRG Petition.

Although we believe these existing evacuation time estimates fully comply with existing regulatory guidance for such estimates, Boston Edison has recently contractually undertaken a new evacuation time estimate study in order to reflect.known population, tr. ?portation and other changes and in l

order to utilize updated computer codes and other state-of-the-art evacuation time estimation methodology.

See Attachment 12.

Such revised study will reflect input from local, MCDA and FEMA officials and is expected to be available in 1987.

This updated study will address many, if not all, of the specific issues or allegations raised in the Show Cause Petition as can be seen through a review of Attachment 12.

Subsection (a) also repeats Chaos !! Report allegations regarding the failure of the evacuation time estimates to deal with panic, traffic disorder, shadow evacuation outside of the ten-mile EPZ and adverse weather.

Several of these issues represent generic concerns related to evacuation time estimates around all nuclear plants, or in all disaster situations.

Boston Edison understands that such generic concerns are being addressed by FEMA expert contractors in terms of evaluating the significance of such issues in affecting existing time estimates or in establishing methodologies to assess such impacts.

Certainly there is nothing in the Show Cause Petition suggesting site specific concerns i 1

relating to a greater propensity to panic in the Pilgrim EPZ than could be the case in any evacuation situation.

Insofar as the issue relates to traffic obstacles outside of the EPZ, (i.e., the Sagamore rotary and shadow evacuation from the Cape) this is precisely the issue which has already received enhanced review in connection with the 1983 MASSPIRG Petition and which was subsequently rejected in FEMA's May 1, 1984 "Analysis Report".

The final concern regarding adversJ weather is an issue which Boston Edison will address in greater detail in the new l

evacuation time estimates being prepared.

Insofar as subsection (a) includes allegations relating to plans for communities outside of the ten-mile EPZ, it is not clear exactly what type of plans are being referred to in the Show Cause Petition as opposed to what type of plans may have been discussed at the June 18, 1986 hearing.

By definition the ten-mile EPZ 15 the area for which detailed planning is needed.

NUREG-0654, Section 1.0.2.

The type of plans that would be i

appropriate for host communities outside of the ten-mile EPZ are orders of magnitude less involved than the detailed planning base required within the EPZ.

This is explicity recognized in NUREG-0654 which provides that detailed planning within 10 miles would provide a substantial base for expansion of response efforts if this proved necessary.

NNEG-0654, Section !.0.2.d.

Also plans for outside of the EPZ do not implicate the same level of safety issues involved in evacuating the persons from within the EPZ.

Nevertheless, should additional or updated plans be sought for areas outside of the EPZ Boston Edison is committed to assisting MCOA and the local coneunities in this effort to the extent possible.

b) Subsection (b) repeats an allegation made in the Chaos 17 Report regarding inadequate planning for evacuation of groups with special needs.

This continues to be an issue where the principal obstacle is identification of the individua1r or groups who will need such assistance.

In'the case of the handicapped or those who will need transportation assistance, such efforts have been discussed in response to Section 15(c).

Provisions for identified groups such as school ;hildren are included in existing plans and involve either returni."; children home to families when there is sufficient time, or evacuation by buses.

In addition to evacuation of any of these identiffed special grovos, it is also important to recall that there are a nu:nber ~'

r protective measures, most significantly that c/ sheltering,

>8 ;. ma.., anpropriate ir. a given situation. Again this area remains one t-w:

1 ~.iton Edison is committed to providing all reasonable assistence te 1ccel and MCM of ficials, c) Subseccion (c) raises the pri.acipal alltgation relating to a lack of contractual agreements w th but conpanies or other transportation providers.

The very short answer to this is that there is no requirement that there be any such "contracts". On the other hand, there must be reasonable plans and arrarjement?..

The State Plan for providing such transportation wou'<; I wol"e utilizing buses and drivers of the various existing transportatici p.' ovide s in Castern Hassachusetts, including most significantly the HBIA.

The Governor of the Commonwealth has the esisting statutory responsibiliti and autha-ity to take all necessary actions in the event of an emergency, including taking possession of necessary vehicles and utilation of other state resources and personnel in order !

to protec% 9he lives and proper %y of citizens.

See generally ALM Spec. L.

c. 31, 65; Acts of 1950, c. 639. MCOA officials have previously acknowledged the existence of this authority and that they would utilize such authority in the event of a radiological emergency.

(See Attachment 13).

As a practical matter, Boston Edison submits that it may be assumed that State officials will comply with their legal responsibilities in those situations.

See, e.g., In the Matter of Long 151&nd Lighting Co.

(Shoreham Nuclear Power Station, Unit 1) CLI 86-13, NRC

, July 24, 1986.

Again insofar as more specific arrangements, including provision for training and other logistical arrangements are concerned.

Boston Edison rematns willing to assist the MCOA in all reasonable respects.

Subsection (c) also repeats allegations regarding the purported inadequacy of evacuation routes.

The short answer is that these "inadequacies" have already been factored into existing evacuation time estimates.

The issue of Cape Cod traffic, including plans relating to the possible closirg of the Sagamore Bridge as well as to the possible closing of Route 3 southbound at Route 128, has already been extensively addressed in FEMA's May 1, 1984 Analysis Report.

The Show Cause Petition presents no new information on either of these issues.

Section 18: "Deficiencies in Medical Facilities" a) Subsection (a) is almost entirely a repeat of the 1983 MASSPIRG Petition insofar as allegations are made of inadequate provisions for treatment of contaminated injured individuals.

The allegations were rejected then and the reasons chich supported rejection then (see FEMA Repor9, November 3, 1983, pp. 9-10) remain valid today.

In addition the Show Cause Petiticn is flatly incorrect in suggesting that there have been no material changes or improvements in this area since 1983.

First there have been additional hospit:15 which have been identified and with which arrangements have been made or updated concerning treatment of contaminated individuals or radiation victims.

See, e.g., Attachment 14, Letter Agreement with Affiliated Hospitals, Inc.; Attachment 15, Letter Agreement with Shriners Hospitals Burns Institute; Attachment 16 Letter Agreement with Jordan Hospital, Inc.

Secondly, there h6s been considerable additional clarification of NRC policy and requirements 19 this area, particularly in light of GUARD v. NRC, 753 F.2d 1144 (D.C. Cir. 1985).

See 51 Fed. Reg. 32904, September 17, 1986. Current arrangements at Pilgrim regarding emergency medical facilities go considerably beyond the mere list of facilities criticized in the GUARD case and are in full compliance with this policy.

b)

Subsection (b), dealing with the failure to provide for potassium todide distributton, repeats another Chaos II Report allegation and, as with other allegations, appears to show no recognition or awareness of the fact that such allegation has already been addressed and rejected.

See FEMA Report, November 3, 1982, p. 10.

The Petition contains no new information to suggest that the existing Com mnwealth of Massachusetts determination in this regard is incorrect or in violation of any NRC or FEMA requirement.

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Section 19: "The Emergency Planning Zone Is Too Small" I

i a) Subsection (a) contains a generally accurate statement regarding EPA's protective action guidelines.

Such guidelines were a principal impact in the NRC's determination regarcing the apprcpriate size of the EPZ.

b) tubsection (b) includes several generally correct statements concerning the regulatory basis for establishing the size of the pluce exposure EPZ.

In fact the Pilgrim EPZ size is based upon precisely this regulatory guidance.

Previous allegations regarding the purported inadequacy of the size of the EPZ have been reviewed and rejected in FEMA's response to the 1983 MASSPIRG Petition.

SUBJECT:

HISTORY OF REGULATORY PERFOPF.ANCE AT PILGRIM NUCLEAR POWER STATION In res;cnsa !c Ri:hard Ucell's re:ues; cf w'une 5,1986, this.temerancum for4 arcs a hist:ry of regula :ry pr:elems relacive to the Pilgrim Nuclear Fewer Station.

milestones and enforcement acticns, such as:This hist:ry includes a tabulation of

-: ficant milestones, civil penalties / orders, significant management meetings,shutdcwn enforcement 1,0nferences, and Innediate/Confir natory Acticn letters.

als: a kr: led;e his su::lemer.:ai re:ves; f:r five s:ecific itsms; h: wever, sia.:e it wi:' re:uire ;cre :ime :: develes, we have ceciced ne: :: ce ay this

tMa:fcn 2.1c we will provide the ::her 1:ams in ateut a week.

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P!LGR!M STATION Occke: No. 50-309 REGULATORY ptRFopyANCE HISTORY A tabulatien of sienifican: :nfies :nes and enfer:2.7.e n : actiens s

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June 1972 Issued o;erating license.

Dectmber 1973 Shutdown Order issued to inspect for and re-pair fuel channel box damage.

December 1974 Fuel failure: Hydriding and pellet-clad interaction failures resulted in high gasecus ac'tivity.

Operation with the fuel cladding perforations resulted in high dese rates in locations requiring access for cceration and maintenance.

During 1975, 75 and 77, power was limited between 60-80:

to maintain offgas activity within regulat:ry requi remen:s.

The las: cf the defective fuel bunc'.es was replacec curing :. e 1977 refueling cu age.

May 1975 A civil penalty (512,000) was assessed for violations c:ncerning Inservice Inspecticn activities identified during an, inspection conducted Cecember 1973 - Fe:r.:ary 1375.

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o'uly li'5 F.anagemen: meeting to discuss c:ncems rela:ed to the managemen: and implementaticn cf the Health Physics Program.

Oc::ber 1976 Panagemen meeting to discuss cencarns related t: me'agement and im

ne Healtn 7:/ sics Fregram. plementa-icn c'-

Ncvem:e# 1977

. Manage en: meeting :: revies licensee e#f:r:s to streng:nen Radia:icn Protecticn Fr: gram.

March 1978 A civil penalty (516,000) was assessed for violations identified in inspecticn re;or 50-293/77-31.

The violations were:

ever-exposure of one individual; failure to in-struct personnel in acecreance with IC CF:t 19; failure to perform recuired air samcling; and failure to follow precedures.

September 1973 Panagemen; mee:ing t: discuss cencerns en recent inspecticn findings (all areas).

Sectember 1979 Panagement meeting to dis:Uss viola:1:n of primary con:sincent in*agri;y.

Oc:cher 1973 A civil cer.alty (Si,CCC) was assessed fer a 5

vicia:icn icentified in inspection recert 50-E93/79-15 involving a failure to fellcw the fecuri:y Flan.

- - - - - - - - " '~

February ISEC A civil penal:y (!!,CCO) was assessed fer shipoing radicac:ive materials with externai rtdiatien levels in excess of regulatory l imi ts.-

larch 1981 (SALP)

Management meeting to discuss the results of the SALP for the period January 1,1980 to December 31, 1980.

April 1981 A civil penalty (513,000) was assessed for events surrounding movement of irradiated fuel without secondary containment as identified in inspection reper:

50-293/EC-C9.

July 1951 A managemen; meeting was held in July 1,96.*

cis:uss c:ncerns for TML Ac:icn Plan 1: ems involving pos ac:icent sampling precedures and equipment and an imediate Action Letter was issued regarding implementation of these items. Meeting was prempted by a' June 1981-radiation pr:tec:icn inspection

(!C-233/51-14) fcunc :ne licensae failed t:

c:nf:rm witn NRC cri:eria in c:nnecti:n with 4 of the 5 NtJREC-05'8 Categcry A f tems insp;:ted.

June - September 1981 Inspections 50-293/81-18 and 81-22 identified six pr:blems; incperable c:mcustible gas c:ntrol syste:n; failure to perf:rm an adecuate 50.59 review failure to pr: vide a::- :ria a ; :cecures; and drawings; failun to make a report recuirec by Tecnnical Specifica:icns; failure to provide ac: urate inf:r ation to NRC; and failure te satisf an Limiting Cendition fer Opera:len (LCO)y regarding primary centainment isolatten valves. These ins;ections were subsequently the subject of enfor:ement actient taken in January 1982.

July - August 1921 A Pe #cr:ance Aperaisal Ins:ecticn (50-293/

81-20) fcund 6 of 3 areas examired bel:w average.

These were:

quality assurance audits; maintenance;c:n:nittee a c:rnc-ive ac-icn systems; licsnsed anc nea-licensed training; Ind pr:curemen:.

Plan:

eperatiens and design changes and mcdifica-

{

tiens were fcund to be avertge; heaever,

(

significant weaknesses were identified in both areas.

. - - - - - - - - - ~

- - - - ~

~~~

Cc:cber 1951

\\

Enforcement conference to discuss management c:n:rois of safety related activities in-cluding the violations identified during inspections 50-293/81-1E and 81-22, the Performance Appraisal Inspection results, and an interim SALP review (period September 1, 1980 - August 31,1981).

January 1982 Civil penalty (3350,000) assessed for failure to comply with requirements of 10 CFR 50.44; submittal of false information to NRC and subsequent delay of notification to NRC of known inaccurate information; and failure to comply with LCO for RCIC c:ntainmen-isolation valves.

(?!?)

Order modifying license required licensee to submit a comprehensive plan of action that would yield an independent appraisal of.sita and corporate management, rec rn:endatiens*

for improvaments in management c:n:rols'.and ovarsight, and a reviev ef ;revitus

m-pliance with NRC requiremen...

Management meeting to discuss i=plementing requirements of the NOV/ proposed civil penalty and order modifying licensa regarding the independen: appraisal of Bost:n Edisen Ccepany (!!Co) management practices.

January *5E2 Ins:er,:icn re:cr: !C-252/31'2f icen-ified a severity level III violati:n for transporta-tien of radioactive materials with liquid in the containers.

This violatien was based en an inspecticn in Aucust 1921 by the State of Scuth Carolina whien resuited in issuance of a civil penalty (51,000).

March 1982 Best:n Edisen Company (SECo) submi::ed the Per#:rmance Im;revement Pr: gram (.:!P) required by the January !!E2 Order.

NRC Managemen; meetings

review status of the Performance Imorevement Pr: gram were held a:or:x mately every six weeks until d

fe;;em:er 1964

June 1982 A special inspection (50-293/82-20) c:nducted of licensee actions after radicactive spen:

i j

resin was found on roof tops anc. pavement

\\

within the protected area.

No iiolations i

identified.

Confirmatory Acti<,n Letter issued c ncerning actions to ia taken j

regarding the spent resin.

i i

July 1S52 Enforcement Conference to disuuss aceeding an LCO associated with the Rea-*Le Protection System water level instrumentation.

August 1922 Enforcement Conference to dis:uu exceeding an LCD asscciated wi:n the '/acuum Breaker 1

Alarn System.

September 1982 (SAL?)

Management meeting te discuss the results of the SAL? for the peried September 1,.1981 to June 30, 1982.

August 1983 A shutdcwn order was issued re;uiring the,

licensee to shu:d:wn in Cecam:er 1553 and inspect the recirculatien sys.am piping fer Intergranular Stress Cerrosien Cracking.

1:

required them to remain in c:Id shutdown until authorized to restart by the Director of NRR.

The licensee replaced the recirculation system piping and was autscri:ec to res ar in Decem:er '!!4 Rec tm:er.'352 (IAL?)

.vanagemen mes:ing t:

is:us's the results ' ef the SAL? fer the peried July 1,1922 to June 30, 1983.

Novi.-ter 1953 Management mee:ing to discuss refueling /pi;e replacement preparations.

January 1984 Cenfirma: cry Actice Letter issued regarding licensee actions relative to health ;nysics practices following the discovery of small, highly radicactive scur:es in the c:ntrol red drive repair r:cm.

February 1984 Enforcement conference regarding the une:n-trolled handling of small, hignly radicac-ive scur:es in the c:ntrol roc drive repair r:cm.

April 1534 A civil pendity (540,0C0) was assessed fer pr:blems in c:nnection with the unc7ntrolled handling of small, highly radioactive sources in the control red drive repair recm between Januar9'14 and 18,1984 The violation involved identified problems with the labeling of c:ntainers, the use of extremity desimetry, and the adequacy of instructions given to individuals working in tt.e repair recm.

September 1984 Management meeting to discuss a second instanc of the uncontrolled presence of i

small, hi c:ntr:1 r:ghly radicactive scurces in the d drive recair r:cm.

Oc::cer 1954 Enfor: ment c:nference en the unplanned ex-tremity expcsure (within regulater ' limits) connected with the small, highly r',aMaetive scurces in the control red drive repair -

(Fo11cw-up to September l'9EA rcem.

management meeting en same subje d I

Ccnfir at:ry Action Letter issu:d in j

c:nnecticn with recurring radiatien protection program weakn0sses.

Tne letter-outlined licensee plans fo* evaluating and correcting these weaknesses

.ioverter 1;EA An order mcdifying the license was issued in c:nnecticn with recurring weaknesses it, raciati:n pr tecti:n pr: grad.

the quired the licensee to c:molete 3nThe Order re-independent centrac;cr assessment cf the radiological c:ntrols pr: gram and te submit to NRC rev Improvemen:iew and a;:reval a Radic1cgical ' '

Pian (RI?) for upg radiological controls program.rading the Follewup inspecticns cencucted in May, August, and November 1955 and April 1986.

A Severi:y level III vielatien (ne civil penalty) was issued for failure to perform radiation surve workers in ac::ys; failure to instruct rdance with 10 CFR 19; and failure t: pr:cerly imCitment a pr:cedure in connecticn wi:n tne unplanned exposure noted above.

(

Enforcement conference to discuss weakn in the c:ntrol and meni:: ring of neutren in-s rwren: :i:n curing refueling c;erstens.

January 1955 (3AL?)

Management meeting to discuss the rtsults of the SAL? for tne peried July 1,1923 to September 30, 1984 Enforcement conference to discuss an unplanned cccupational radiation exposure within regulatory limits asscciated with sludge-lancing operations on a waste tank as identified in inspection 50-293/24-44 August 1985 Enforcement conference to discuss licensee's sction on abnormal surveillance tast results and a degraded vital area barrier.

October 1985 A civil penalty (!50,0C0) was assessed f:r

ne degracation of a vital area barrier.

Novemcer 1985 A safety system functienal team inspection (50-293/85-30) was conducted by the Office of Inspection and Enforcement to. assess' the operational readiness and fune:1cn of selected safe;y systems.

The ins:ecticn icentified na: :ne licensee..ac net effectively mitigated a wa r hammer pr:blem ass:ciated with the HFCI turoine exhaust line which had been occurring sjnce the '

)*? cning_o' plant _g;teration.

4 Weaxnesses were also icentified wit design ~cha'n'e process; c:h tha licensee's g

ntrol of plan:

ins;rumentatien; hancling of vender inferma-

)

tien; pr: gram f:r a::r:ving and va emerSency c:erating :r:cecures; ca:lidating

c:ncuct a atility c:ntrol r:cm; plant shu:dewn from outside the and maintenance pr: gram for met:r operated valves.

February 1986 Inspection re; ort 50-293/Se-Ca identified a severity level !!I violation for failure to meet packaging requirements for low specific activity radicactive materials.

This viola-tien was based en an inspecticn in January 1985 by the 5:ste of South Car: lina which

($1,0C0).resulted in issuance cf a civil penalty s

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e

..-__.-_,,.,.----,a--

Mar:h 1986 (SAL?)

Management meeting to discuss the results of the sat? for the period October 1,1984 -

October 31, 1985.

February - March 1.986 A special diagnostic teau inspection (50-293/

86-06) was c nducted to deter nine the under-lying reasons for the licensee's poor pericts.ane.e described in the most recent SALP and to ascertain whether they could have an adverse impact en the safety of plant operations.

April 1386 An Augmented Inspection Team (AIT) conducted an ins:ec:1cn of recen ecerstienal events wnien incluced 1) :ne s:urious gr:ue one pri.7.ary containmen; isola: fens (and asscciated reactor scrams) that occurred on April 4 and 12, 1986, 2) the failure of the main steam isolation valves to prcmptly,

3) pen after the centainment isolatiens, and,

reo the re. curring pressuri:sticns' of the

~

resicual heat removal systam.

The A!T feutd the licensee's evaluatiens folicwing the -

I second event to be carefully structured and thor: ugh.

A Confir:natory Action Letter t

concerning the events was issued which required the licensee to provide a written report prior to rest?rt centaining the results of the evalua9 fen and c:rrec-ive acticns.

The CAL also recuired Regi:nal Ad.?inistra :r au;ncri:atien f:r restar..

Inspection (!O-293/86-10) reviewed implementa:icn of the RIP.

The inspection found t:a licensce adequately addressed 13 nf the 34 items reviewed.

May 1986 Management meeting to discuss evaluatier:s and c:rrective actions c:ncerning the operat.icnal events of April 4 and 12,1986.

June 1986 The first in a planned series of management meetings scheduled to review SEco managemen:

oversight of the implementation of the

)

licensee improvement pr:grsms in pr:gress.

t.

Descriccien c# Imcrevement P : cram I.

Per er ance Imerovement Peceram (DIP) d a)

Required by Order in January 1962 b)

Areas for Improvement i

1 1)

Independent Review and Evaluation (MAC) i 2)

Organization Review /Revis.cn l

3)

Management Centrol System Review / Revision 4)

Training'cn Changes l

c) 126 milestenes established

- e.tamcles - Precedure Update P egram I

(660 precedures)

- Update Design Occ::ments Prcgram (450 drawings) d)

Status

,C:mp'.e:a Licensee CA verificatien of final cem.it ent perfcrme'd Octrber 1985 II.

Radiolceical Imerovement Precram a)

Required by Order in November 1984 b)

Areas fer Im;reve.aent 1)

Indecencent Assessmen: cf Fregram 2)

Radiciegical Organizaticn Review /Revisien 3)

Radiolegical Centrols Review / Revision 4)

Managemen: Oversight and Corrective Actiens 5)

Training on Changes c) 209 Milestenes Established As cf Cecember 1955 cre item remains c;en (recenfigure access centrc1)

C:ntinueus ! :-:ve e : Precram a)

Initiated by BEco in June 1985 b)

A;tions 1)

Visited plants with gecd SALP evaluattens 2)

Ccnducted internal survey to icentify prcblems/cause 3)

Issued re;cr: of fincings in Decemcer ISEE

c)

Pr:blem Areas Identified 1)

Attitude 2)

Accountability 3)

Weak Rcot Cause Analysis 4)

Ccmunica tion 5)

Effectiveness Assessment d)

Sta tus Implementation of sixteen of eighteen recomendations in progress

.'T'.

(

J

The f ollowing material was p ovided by Mr. Joseph Bernero, Chief of the Boiling Water Reactor Division of the NRC.

"' severe accidents' is the term most commonly used to describe accidents in which the reactor core is severely damaged.

As happened at Three Mile Island, prolonged loss of core cooling can allow the heat of radioactive decay on the core to build up to the point that the fuel begins to disintegrate, the zirconium metal cladding melts or reacts with residual steam to form combustible hydrogen, and even the ceracic uranium oxide fuel pellets can melt.

A great deal of attention is being given to understanding the behavior of reactors and their containments in severe accidents, especially since the Tnree Mile Island accident.

The objectives are to ensure that the likel! hee.' ef core calt acci?cnts is very Inw and thst, shnuld th.rv is s@stantial assurance that the cent.i:n w.t ene cccur, will mitigate its consequences.

I The severe accident behavior of a BWR with a Mark I

]

containrent, the Peach Ecttom plant, was assessed in the reacter Sefety StuJy

('..a.il'- 1 J C 0 c r NUE EG ~ 5. C l J ) wh i c h wa s r#!::"' : r.

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Ste.'.

i t..f i e s t f ' i r 4 t : s. l e,1 crerall risk for the EWk, principally Jue to its ability te prevent core melt.

The containment was estimated to provide very little mitigation of core melt consequences because the buildup of pressure under accident conditions would be a direct cause of c'atainment failure unless adequate cooling was preserved.

Consistent with operating procedures in place in 1075, the Study assured little effort by the reactor operators which might effectively preserve the containment's integrity.

1 The situation, more than ten years later, is different and still changing for the better.

It is recognized today that melten core material melting i nto the ground through the thick containment base is not the principal threat; rather, it is an atmospheric release of radioactive material which is the principal threat.

The principal factors which can cause containment tallure with atmospheric release are hydrogen ignition, gas overpressure buildup to rupture, and direct attack of the drywell by core melt debris.

The general situation for each of these is sumn.3ri:ed as follows:

livd ro g e n Ignition Recogni:ing that combustible hydrogen can be generated and relasse' in s<vera acci8'nts, all Mark I centainnents now are purged ani failed w th inert nitrogen gas during eperatien 5>

that even if hyJrogen gas is formed it has insufficient cxygen available to suppert combustion.

Remaining questions in this area relate to how long the containment may be without this inert atresrhere in order to permit inspectier.s, and how air might lesk in or hy3rogen leak out to nearby roers under a c c i. ":, *

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Overpressure failure Careful analysis indicates that a typical Mark I containment can withstand pressures o) more'than twice the design pressure without rupture.

Nevertheless, severe accidents in the extreme can generate such pressures and cause containment rupture.

Overpressure damage control procedures have been developed for pressure s'uppression containments and are already in place for operator use.

With these procedures the containment remains closed for most accident conditions; but, if overpressure failure threatens, large vent valves above the suppression pool chamber are opened so that the excess pressure is released gradually by bubbling the releases through the pool, forming a filterad vent containment system.

With this path assured, virtually nothing but the noble gases are rei,..id.

The radioactive ncble gases pose a modest exposure threat offsite only in the area very close to the plant.

A number of questions are being pursued in this area.

All the plants have suitably large vent valves anc cc:O tut tr.ey vary cr.e to another in the ability to open tnese valves under accident ccnditions.

The valves are t'esigned for highly reliable closure, not opening.

Consideration is b2ing given to modifying valve contro's.

In addition, the vent ductwork downstream of the valves may warrant modification.

In most plants it is fairly light gauge ductwork and cignt be breached in a:cicent venting.

If so, consiceration is being given to the e'fe:ts cf se:orda y re'. ease of rat!:5:tive gas, hydrogen, a-d n-nacs steam into the reactor building.

Direct Attack The core melt debris, since it has melted through the reactor vessel into the 5

drywell may, by direct radiation of heat, cause failure of rennections in the drp ell shell; or the debris, if sufficiently fluid, may flow out to the wall and melt through the steel.

The Park I containments have one or more spray systems in the drywell which are at'e to spray water along the walls and onto the floor of the drywell inhibiting direct attack.

Concerns in this area are in three general areas:

core debris m:deling, shell and concrete attack modeling, and spray reliability.

In the first area, it is recognized that a noiten reactor core i

to melt through the bottom of a EWR, must dissolve a very i

i

.=. - - _.

large amount of inert metal in the lower reactor vessel, probably diluting the are melt.

The key que...on is whether the melt would come out moving

. sluggishly like Hawaiian volcano lava or as a hot free flowing liquid.

The latter is the more threatening condition, i

If core melt debris reaches the conr. rete floor and steel shell of the wall, it is important to understand that the path to the outside that might be opened bypasses the beneficial scrubbing of radioactive material passing through the pool.

As noted earlier all these p1 ave dr>well spray systems, but they are designed as a s2condary mode ation for a reactor safety system.

Strong i

consideration is being given bling hookup of these systems to fire protection systems so that s,

ability is almost always available.

l SLtsta.tially different emergen:y c;erating procedures and tr<.ining were put in pla:e at all eactors af ter the Three Mile Island accident; further deprovements in these procedures are still being made.

For the Mark I t

ontainments both industry and NR studies are being used to identify the best

.cebined strategy for pro:edures a.d perhaps scee changes in equipment su:h as alternate vent paths, e-ir:re.'ed valse c e-atility.

The Mark I studies are j

being given higtest priority tj ine hE: staff and the industry.

Tne '

ex;ectation is that, with m: dest ic;rovements of this type, one can achieve substantial assurance of core melt consequences mitigation by a Mark I containment."

t i

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8 8 9's, p

UNITED ST ATes g,

i NUCLE AR REGUL ATORY COV. MISSION I'

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r W ASHINGTCN, D. C. 20555 ll October 7, 19EE L CEt.!EES:

EbE OW' EF!' Gi%

FACILITIES:

A.L SWF. PL A!.T!

SLE'ECT:

SLM A:Y 0: SEFTEFEti. 3:,19Ef MEET!!$ Of GEf.Er.:(

EECLIEEMEti:5 F0; EkF C0tiiAlteEt.it A r+etir.( teimter ttt CWF Cortrs GrcL: !EWFCC; ar.c the tiFC was heic is Cett.e sda, l'a ryie rc, cr. Se;ter.t.er 11, IFE(.

Tr.e pLrpest cf the tr.eetirg war fcr it.t AFC tc presert tc tt.e EWRCG a st; cf strawfr.ar. gereric reepirer+r.t! tse r.;

cer11ctrec cf tre i,FC f(r tr.t PL rrc st cf er.r.a r.cint, BWF cor.tainter.t pe rf orrar ce ir. sesere at:icer.: cencititr1.

int EWRCO was te present the resbits cf its ptrallel stvey usirg ticrtrert itate! Fewr-Cer;ary's Mcrticelic plar; at re fe rer ce Ei;T wi;r t.t i.1 cca:a irr e *..

itt cf rettir.g atter.c'ets it r

er.closec (Er,:les.rt 1,'.

Tr.c f.T* ; tsr';(;

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ccrsictra.'tr! c' sLitat'e ge e i; rt:Lirerer:s.

rt Er FCc gese 6 :resertati:r c' its it!';i:r en its ;t ralie: s:Lc (Er.c'.cstri 2..

Tre,. rest :: L:e:

er: se':;. : r::v:' re SLicelires, res's'c 4

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PC.cNer, tre is!Lt! c' Fj : ';t' cer,

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CF.GF scr e:. lee f t f.:ver:e r ;f, ;9ti.

jh.4b Fchar. C. Tracari, Frc;e:t Fa a;t e Gerer1: ReqLirerer::

for EWK Cor,tainrerts E r *,I C 1 L r t s ;

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Enclosure (1 NRC-BWROG MEETING ON C0hTAlhMEhT REQUIREPENTS SEFTEMBER 11, 19e 7 NAME AFHLIATION R. Bernero NRC/ DBL H. Denten NRC/NRR C. Reed Cerr.onwealth Edison Johr Raulstor.

Tennessee Valley Authcrity Vincent Boyer Philadelphia Electric Cor.pany Richard Diederich Philadelphia Electric Company James C. Carter IT Corp /IDCOR Edward Howard Boston Edison R. E. Skavdahl General Electric Jnhr.. F. Ful ter.

Boston Edison Cor. pan; T. H. Landers New York Power Authority J. A. Gray, Jr.

New York Power Autherity S. O. Flcy c Carolina Power & Light C upary A. B. Cutter Carolina Power & Light Ccepany H. k te 'se.

Ferrsylvar.it Pcwer & Light Cce;ary Briar PcCa"rty Lc'5 Island Ligrting Ccrpany Charles Castric Lorg Island Lighting Cerrpen)

L. T. Guc*a Gecrgia Power Corpany S. H. Chesrut Georgia Power Cor.pany f

J. R. Largley Gulf States Utilities /HCOG Terry Fickers herthern States Power Kevir. Felt: claw Gere al Electric Cer;ar.y Dear EcLiter he.C/ ACES Staf' Fa c' l Elta..la hER/CSR0/RIE l

Jir teccers G:. States Lt'11 ties Cer;ar.

Cernis E. hackir5 Enercon Services /KCG Wayr.6 hecSes hRC/ DBL /RSE L. G. Hulrar NRR/ DBL /PSB A r.st ic F.a ri r e s hRR/ DEL S. W. Wilczen, Jr.

hiagara Mohawk Eruce A. Prester Power Supply & Ergineerir.g & Cor.st.

Pichard F. Fw "ay Power Supr.ly & Engineering & Cctst.

Fichael J. Ma.,

Tennessee Valley Authcrity - BFN Woody Streupe Enercon Services Fat Pete11 Washington Public Power Supply Syster Lynn Conner Loc Search Associa'es Larry Gifferd General Electric John Stang NRR/EWRd/ Lacrosse Prcject Manager lhoras S. Retella hRR/08L/BWO4 Big Rock Point PP Raean Pichur.ari NER/ DBL / FOB Donald R. Hof fraa-Cleveland Electric 111crinating Ver1non L. Rooney bcR/ DBL /BWD2 John Lartirs 0:M/LZ Kazir.ieras Carpe hER/0EL/PSE berald E. Gears hRR/ DEL /FDP2 KAME AFFILI ATION Jack Kudrick NRR/ DEL /PSB Jay Thayer Yankee Atomic R. J. Lodwick Vermont Yankee Stephen P. Schult:

Yankee Atomic Electric David H. Wagner NRR/ DBL /BWD3 Dorenic B. Yassalle NRR/ DBL / FOB Daniel R. Muller NRR/ DBL /PD2 John A. Zwolirski NRR/ DBL /BWD1 Raj Auluck NRR/ DBL /BWD1 Jack Denchew NRR/ DBL /BWC1 E. G. Adensan NRR/ DBL /BWD3 Earl PaSe Detroit Edison Phillip L. Paull Vermont Public Service Dept.

G. Tarrar.t VT Dept. Pub. Service N. W. Edwards NUTECH C. L. Reic' Bechtel Oaer M. Scet:

Southern Co. Services G. t. Lair.as NRR/CBL Eve ic:cpecies SE:.CH Licensirg, 5.echtei Akira Drric The Tckyc Electric Pcwer Te:st Ir TEF: 0 Sans Ec

).'

hER/ DBL /PSB Mer,ar. C. Tnadari hRR/ DBL /BWC2 f

Kathleer F. Sr.et hewrar, & Holtzinger David W11scr Icas Electric Lignt ar.c Foner Stept.er. Ma ler t.

Deve r.rt e Ccr Lenis P3ssissippi Fewe. & Ligrt FaL1 Leeca NE:/BW:.

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l DISCUSS 10ti 0F A 4

GENERIC LE~TER ON EWR CONTA!NMENT PERF0 MAN;E t

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ROBERT M. BERNER0, USNRC f

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I GENERIC LETTER ON Ek'R CONT AINMENT PEP:0RMANCE i

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P.At;T EVA' UAT10NS U.CER THE SEVERE ACCIDENT PC.lCY

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THE SU5JEC~5:

37 EWRS W:TH FRESSURE T'JPP:ESS10N C0',TAIN.I'i S THE METHOD:

A GENERIC LETTER OF REQUIREMENTS TO IMPLEMENT CHAfGES BASED ON GENERIC EVALUATION i

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NRC SEVERE ACCIDTNT PDLICY STATEMEN',

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e THE MOST COST-EFFECTIVE OFT 10NS FOR REDUCING THIS VULNERA31LITY SHALL BE IDENilFIED AND A DECISION SHALL BE REACHED CONSISTENT WITH THE COST-EFFECTIVENESS CRITERIA 0F THE COMt;!SS10N'S EACKFIT POLICY AS TO WHICH OPTION OR SET OF OPTIONS (lF ANY) AF.E JUSTIFlABLE AND REQUIRED TO BE IMPLEMENTEE.

e IN THOSE 1NSTAN ES WHERE THE TECH:slCAL ISSUE GOES 3EYCD CURRE'iT RE3L" ' TORY RE:'.'!REMENTS, GENER!C RULEMAKIN: VILL EE THE FREFERRE:. S.U !CN.

IN CTHER CASES, THE ISSL'E S 'OL".2 EE DISPOSED OF THROUGH THE CCNVENTIONAL FRACTICE 0? ISSUlh3 BULLETINS A!;D ORDERS OR GENERIC LETTE:S WHERE f.3DIF] CAT 10NS ARE JUST!FIED THR^ UGH BACKFlT POLICY, OR THROUGH PLANT-SPECIF]C DECISION M' KING ALONG THE LINES OF THE INTEGRATED SAFETY ASSESSMENT PROGRAM (ISAP) C0i;CEPL ON.

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GDC 16:

CRITER10N 16 - CONTAINMENT DESIGN.

--AN ESSENTIALLY i

LEAK-TIGHT EAF,P,lER AGAlliST THE UNCONTROLL ED RELEASE OF RADIDACTIVITY TO THE ENVIRONMENT AND TO ACSURE THAT THE CONTAINMENT LESIGN CONDITIONS IMPORTANT TO SAFETY ARE NOT EX:EEDED FDE AS LON: AS PDSTULATED ACCIDENT C0iCIT10NS REC'JIRE "

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