ML19340B369

Forwards Request for Addl Info Re Radiological Controls. Requests Amend to FSAR to Reflect Responses to Encl Requests

ML19340B369
Person / Time
Site:	Grand Gulf
Issue date:	09/22/1980
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Mcgaughy J MISSISSIPPI POWER & LIGHT CO.
References
NUDOCS 8010220235
Download: ML19340B369 (25)

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UNITED STATES j

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NUCLEAR REGULATORY COMMISSION l

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-C WASHINGTON. D. C. 20S55

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ShP 4 L bau Docket Nos.: 50-416/417 Mr. J. P. McGauchy Assistant Vice President - Nuclear Production l

Mississippi Power and Light Company l

P. O. Box 1640 l

Jackson, Mississippi 31205

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION - GRAND GULF NUCLEAR STATION, UNITS 1 AND 2 As a result of our review of the information contained in the Final Safety Analysis Report for the Grand Gulf Nuclear Station, Units 1 and 2, we have developed the enclosed request for additional information.

Included are questions from the Radiological Assessment Branch.

lie request that you amend your Final Safety Analysis Report to reflect your responses to the enclosed requests as soon as possible and to inform the Project Manager, Joseph A. Martore, of the date by which you intend to respond.

l Sincerely,,

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Robert. Tedesco Assistant Director for Licensing i

Division of Licensing

Enclosure:

As stated cc: See next page 24 g

p 801622

O cc: Robert 8. McGehee, Esq.

Wise, Carter, Child, Steen & Caraway P. O. Box 651 Jackson, Mississippi 39205 Troy B. Conner, Jr., Esq.

Conner, Moore & Corber 1747 Pennsylvanit Avenue, N. W.

Washington, D. C.

20006 i

.Mr. Adrian Zaccaria, Project Engineer Grand Gulf Nuclear Station Bechtel Power Corporation d

Gaithersburg, Maryland 20760 Mr. Alan G. Wagner, Resident Inspector P. O. Box 469 Port Gibson, Mississippi 39150 e

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4 331.0 RADIOLOGICAL ASSESSMENT BRANCH 331.17 Provide a descrsption of turbine building airborne radio-(12.3.4) activity monitors in Table 12.3-5.

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331.18 Your'respense to item 331.13 covers the shielding of the spent (12. 3. 2. 2. 2) fuel transfer tube. Discuss your administrative procedures for positive access control and radiation monitoring to the areas where access to the spent fuel transfer tube may be occasionally required.

It is our position that all accessible portions of the spent fuel transfer tube shall be clearly marked with a sign stating that potentially lethal radiation fields are possible during fuel transfer.

If removable shielding is used for the fuel transfer tubes, it must also be explicitly marked as above.

i If other than permanent shielding is used, local audible and visible alanning radiation monitors must be installed to alert i

personnel if temporary fuel transfer tube shielding is removed during fuel transfer operations. Please provide a description q

of your methods for compliance with these positions.

4 331.19 Health Phy' sics Organization, Unit 1. Figure 12.5-1, shows staff (13.1.2.1) of 8 Health Physics. Technicians, while GGNS Plant Organization, Unit l' and 2 Firgure 13.1-2, show staff of 9 Health Physics Technicians for two units.

It appears that this is an inadequate j

number of-Health Physics Technicians for both one and two unit operations.

Section 13.1.2.1, Plant Organi:ation and Figure 13.1-3 and 13.1-4 4

specifying shift crew ccmposition.coes not indicate that a healtn pnysics technician will be onsite at all times. NUREG-0654, "Critaria for T

Preparation and Evacuation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants", requires tnat a i

radiation protection technician be onsite at all times. Secticn 13.1.2.1 or Figures 12.1-5,13.1-2,13.1-3, and 13.1-4 snould be revised to snow a health physics technician on each shift and to increase the number'of technicians

.necessary for a one and two unit plant operation. Otherwise, you should justify the proposed staffing level based on proposed w ek assigreents that considers work done in-house by technicians, and that contractec to service organizations (such as whole body counting, cosi :etry etc.).

331.20' Reviewing Qualifications of Plant Personnel, Section 13.1.3, it appears (13.1.3) that the Radiation Protection Sucervisor (RPM), Resume No. 20 is not fully qualified in accordance with Regulatory Guide 1.3, " Personnel Selection and Training".

o Resume No. 20 should be upcated, and the Radiation Protection Supervisor's 4

applied radiation protection experience from.1975 to date, if available, should be described;~ the presently cascribed

. radiation protection experience does not appear to be sufficient.

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. 331.21 Describe the extent and functions of the corporate support available to (13.1.2.2.8) the Radiation Protection Supervisor and to the radiation protection staff.

331.22-Based on infomation contained in the draft cocument " Criteria for Utility (13.1.2.2.3)

Management and Technical Competence", it is our position that the organization chain contain a qualified health physicist to provide back-up in the event of the absence of the Supervisor of Radiation Protection. The December 1979 revision of ANSI 3.1 specifies that individuals temporarily filling the RPii position shculd have a 3.S. degree in science or engineering, 2 years experience in raciation protecticn, 1 year of which should be nuclear power plant exoerience, 6 months of wtiich should be on-site, It is our position that such experience' be^ professional experience.

Indicate how you intend to provide back-up in the event of. the absence of the Supervisor of Radiation Protection.

331.23 Specify tne numcer of a'nddescr1ce tne quatirications of health physicists (12.5.1.1) wno are listed in Figure 12.5-1 and in Section 12.5.1.1, ano submit their resumes.

331.24 Our letters dated September 27, and November 9,1979, (reference 3 and 5 of Appendix A), specified new requirements for applicants for an operat-ing license.

Your should revise your application to incorporate these requirements i

which are summarized in NUREG-0594, "Tril-Related Requirements for New f

Operating Licenses," Item II.B.2 (for vital area access), Item II.F.1 (centainment monitor), and III.D.3 (iodine sampling and analysis).

A more complete description of these requirements and :neir documentaticn

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is enclosed in Appendix A to this letter.

331.25 Section 12.5.3.9, page 12.5-17, should be revised to require that all (12. 5. 3. 9 )

licensed radiation sources, not just those with activities in excess of Appencix C, of 10 CFR 20, or Schedule B of 10 CFR 30, be subject to l

material controls for radiological protection.

331.25 Table 12.5-1, Portable Health-Physics Instrument show s the numcer of (12.5-1) radiation detection instruments which will be availcble for both units.

It is our position that the number of portable radiation survey instru-l ments, (especially those which are most frequently used by radiation protection personnel 0-5000 mR/hr) be increasec to reflect following l

considerations (a) both units can be shut down for recairs at the i

same time (maximum usage of survey instruments),(b) a numoer of l

instruments cut of service (in need of calibration or recairs), and.

(c)anumberofspare,operaticr.al.instrumentsshouldbe.always available for use in unusual occurrences.

Justify the proposed gumber.of portable radiation detection instruments, which will e available on ne site, ror both units.

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'Specify in Table 12.5-1, how many Air Purifying Respirators, Atmosphere.

Supplying Respirators, Self-Cantained Breathir.g Apparatuses and how many disposable Atmosphere Supplying Respirators will be available.

State if quantitative respirator fit test will be available for both testing and for training of personnel'in the usage of respirators.

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'331.27-

- In Section 12.5.1.1 you state that you would implement a bioassa'y program l

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in accordance with Regulatory Guide 8.9, " Acceptable Concepts, Models Equations and ~ Assumptions for a. Bioassay Program." Elements of Regulatory L

Guide 8.9 were extracted in the development of ANSI N343-1978,. "American

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National Standard for Internal Dosimetry for Mixed Fission and Activation Products," which is specific to reactor bioassay programs.

Planned and u'nplanned outages or unusual occurrentes at power may require a large number of whole' body counts or prompt unplanned counts. You have not'specified that a.whole body counter will'be onsite.

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Specify that a.whole body counter will be onsite or describe L

arrangements you have made, to assure that counting capacility will ce l'

- available when necessary.

331.28 Provide maximum routron and gamma exposure levels, in routinely visited l

areas in'the containment, in the vicinity of major drywell shield-L penetrations. Areas ofinterest are i.e., Reactor Water Cleanup and Standby Liquid Control System (drywell purge penetrations), TIP Station (personnel and equipment lock drywell. penetrations, fl. el.120'-10")

l etc. Describe location, dimensions and shielding for the drywell shield penetrations. Frovide average neutron and gamma exposure levels at tne CRD hydraulic centrol' units, at the CRD master control and containment t

- personnel lock area and provice an estimate of average daily-personnel

' exposure time.in these areas.

331.29 Provide a breakdown of time (hours) personnel will spend in the

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containment, during plant operation performing specific tasks (i.e.,

inspecting,. maintaining equipment etc.-).

i Provide an estimate of personnel exposure (similar to Table 12.4-8),

resulting from actuation of safety relief valves, based on the follow-

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.ing considerations:

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use design basis' radiation sources:

a)' Noble gas concentrations corresponding to an off gas release rate of 0.1 Ci/sec after 30 minutes decay; j

b) Halogen concentration in reactor water FSAR, Table 11.1-2; l

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Operator working at TIP drive floor el. 120'-10", at a location closest to.the low-set safety relief valve' discharge; i

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3.. Assume that all safety reifef valves open; low set relief valves remain open following closure of others (TYPE 2 occurrence);

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Operator exposure '4 minutes; 5.

Use average radiohalogen (reactor water to steam) carryover cf 1.21 by weight;

' 6.~ ENomal ' ventilation in containment (do not assume homogeneous.

mixing of airborne contaminants.in the entire containment volume-

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300,000-cubic feet within :ne ffrst fcur minutes);

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Cose reduction. factor can be applied.if a clear-air shower is provided in the vicinity of the containment personnel lock;

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f S. -Containment airborne concentrations should not be corrected for plate cut on walls, as it will be negligible in the first few minutes.

Specify the noble' gas and halogen pool retention factars and other significant, dose effecting factors employed in the calculations.

331.30 Section 12.3.4.2 describes Airborne Radioactivity Monitoring Instrument-(12.3.4.2) ation and refers to Table 12.3-5, Airborne Radioactivity Mcnitors.

Explain now monitors detecting noble gases can be also used to detect airborne iodine and particulates simultaneously.

331.31 Section 12.3.4.2.5, paragraph (b), (page 12.3-33, and 12.3-33a) states (12.3.4.2.5) that air monitors can detect 10 MPC-hours of airborne radicactivity.

(12.3-5)

Explain how building exhaust menitors can detect 10 MPC-hours in a i

' particular area of tne building. For example, exclain how exhaust monitor can detect 10 MPC-hours in the cask wash dcwn area, and how the auxiliary building exhaust monitor can detect 10 MPC-hours in the area of emergency decontamination station (fl. el. 93'-0", coord. P-P.4 and 6.2-7.5, Drwg. M-1014). Explain why, in Table 12.3'5, Airborne Radioactivity Monitors, the warning alarms fer tne containment and t'

Auxiliary building exhaust conitors areset for 2.5,15,-and 5 times the 10 (MPC)-hours limit. Describe how one would icentify the specific enclosure having hign airborne radioactivity after an alarm.

331.32 (12.3.2.2.2) _

Describe shielding during spent fuel transfer and provide resulting radiation levels in the drywell at approximately 6 feet (height of 1

person) above floor el.161'-0, below the fuel transfer path.

Assume the folicwing conditions:

a) normal fuel transfer spent fuel assembly being transferred from the reactor vessel, passing over reactor vessel flange, and the too of'drywell annulus, through the fuel pool gate into the containment fuel pool.

.b) accident condition spent fuel assembly has been dropped, one ene laying on reactor vessel flange and the other end of the fuel assembly on the floor of the fuel pool gate.

331.33 Explain or correct the differences of N-16 sources given in

. Table 12.2-7; Table 12.2-14; subsection 12.4.1.1 (p.12.4-2; and in srbsection 12.4.2-2 (p.12.4-5).

- 331.34 In the GGNS Plant Organization, Figure 13.1-2, the Radiation (13.1-2)

Protection Supervisor (equal to the Radiation Protection Manager.

-(RPM) in Regulatory Guide 8.8) reports to Support Services Super-intendent, who reports to Plant Manager.

The RPM should have direct recourse to responsible management personnel and be independent of station divisions as specified in Regulatory Guide 8.8,'Section C.l.b(3) and in " Criteria for Utility Management and Technical Competence". -Thus, the RPM should report directly to the plant 11anager. Revise Figure 13.1-2 accordingly (see subsection 12.5.1.1, 3rd paragraoh (p. 12.5.1.1) and Figure 12.5-1),

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APPENott A DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL QUALIFICATION OF EOUIPMENT FOR SPACES / SYSTEMS WHICH MAY BE USED IN POSTACCIDENT OPERATIONS (II.B.2)

POSITION With the assumption of a postaccident release of radioactivity equivalent to that described in Regulatory Guides 1.3 and 1.4 (i.e., the equivalent of 50%

of the core radiciodine,100% of the core noble gas inventory, and 1% of the core solids, are contained in the primary coolant), each licensee shall per-form a radiation and snielding design review of the spaces around systems that may, as a result of an accident, contain highly radioactive materials. The design review should identify the location of vital areas and equipment, such as the control rocm, radwaste control stations, emergency power supplies, motor control centers, and instrument areas, in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during postaccident operations of these systems.

E,ach licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shield-ing, or postaccident procedural controls. The design review shall determine which types of corrective actions are needed for vital areas throughout the

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CLARIFICATION (Applicable to vital area access, not equipment qualification)

Any area which will or may require occupaacy to permit an operator to aid in the mitigation of or recovery from an accident is designed as a vital area.

The control room, Technical Support Center (TSC), sa=pling station and sample analysis area must be included among those areas where access is considered vital after an accident. Your evaluation to determine the necessary vital areas should include but not be limited to, consideration of the control room, Technical Support Center (the letter dated April 25, 1980, from Darrell G.

Eisennut (NRC) to All Pcwer Reactor Licensees allows substitution of in onsite TSC with an' offsite TSC), post-LOCA hydrogen centrol system, containment isola-tion reset control area, sampling and sample analysis areas, manual ECCS alignment area (if any), motor control centers, instrument panels, emergency power supplies, security center and radwaste control panels.

As a minimum, necessary modifications must be sufficient to provide for vital system operation and for occupancy of the control room, TSC, sampling station, and sample analysis area.

In order to assure that personnel can perform necessary postaccident operations in the vital areas, the following guidance is to be used by licensees to evaluate the adequacy of radiation protection to the operators:

1.

Source Term The minimum radioactive source tern should be equivalent to the source terms recommended in Regulatory Guides 1.3, 1.4, 1.7 and Standard Review

i Plan 15.6.5 with appropriate decay times based on plant design (i.e.,

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you may assume the radioactive decay that occurs before fission products can be transported to various systems).

a.

Liquid Containing Systems: 100% of the core equilibrium noble gas inventory, 50% of the core equilibrium ha'iogen inventory and 1% of all otners are assumed to be mixed in the reactor coolant and liquids i

j injected by HPCI and LPCI (or the equivalent of these systems).

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determining the source term for recirculated, depressurized cooling water, you may assume that the water contains no noble gases.

b.

Gas Containing Systems:

100% of the core equilibrium ncble gas inventory and 25% of the core equilibrium halogen activity are assumed to be mixed in the containment atmosphere. For vapor con-taining lines connected to the primary system (e.g., SWR steam lines) the concentration of radioactivity shall be detemined assuming the activity is contained in the vapor space in the primary coolant system. '

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Systems Containing the Source l

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Systems assumed in your analysis to contain hign levels of radioactivity in a postaccident situation should include, but not be limited to, contain-l ment, residual heat removal system, safety injection systems, CVCS, containment spray recirculation system, sample lines, and gaseous radwaste systems'(or equivalent of these systems). If any of these systems or 3

others that could contain high levels of radioactivity were excluded, you should explain why such systems were excluded.

3.

Oose Rate Cirteria The design dose rate for personnel in a vital area should be such that the guidelines of GDC 19 should not be exceeded during the course of the accident. GDC 19 requires that adequate radiation protection be provided such that the dose to personnel should not be in excess of 5 rem whole

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body or its equivalent to any part of the body for the duration of tne accident. Zten detennining the dose to an operator, care must be taken to determine the necessary occupancy times in a specific area. For example, areas requiring continuous occupancy will require much lower dose rates than areas where minimal occupancy is required. Therefore, allowable dose rates will be based upon expected Occupancy, as well as the radioactive source tenns and shielding. However, in order to provide i

a general design objective, we are providing the following dose rate criteria with alternatives to be documented on a case-by-case bases. The recommended dose rates are average rates in the area. Local hot spots i

may exceed the dose rate guidelines. These doses are design objectives I

and are not to be used to limit access in the event of an accident.

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Areas Reouirino Continuous Occucancy : <15 mrem /hr (average over j

30 days). These areas will require full time occupancy during the i

i course of the accident. The Control Room and onsite Technical Support Center are areas where continuous occupancy will be required.

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The dose rate for these areas is based on the control room occupancy factors contained in SRP 6.4 b.

Areas Re'cuiring Infrecuent Access: GDC 19. These areas may require access on a irregular basis, not continuous occupancy. Shielding should be provided to allow access at a frequency and duratien estimated by the licensee. The plant radiochemical / chemical analysis laboratory, radwaste panel, motor control center, instrumentation

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locations, and reactor coolant and containment gas sample stations are examples wnere occupancy may be needed often but not cohtinuously.

IMPLEMENTATION DATE By January 1,1981, or full power operation, whichever is later, complete modifications to assure adequate access to vital areas following an accident resulting in a degraded core.

DOCUMENTATION REQUIRED Provide a summary of the shielding design review and a description of the results of this review and a description of the mcdifications made to the plant to implement the results of the review.

Include in your submittal:

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source terns used in the evaluation; W

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systems assumed in your analysis to contain high levels of radicactivity in a postaccident situation.

If any of the systems listed in CLARIFICATION item 2 were excluded, explain why such systems are excluded from review; 3.

specify areas where access is considered necessary for vital system operaticn after an accident. If any of the areas listed in the CLARIFI-CATION section above were not considered to be areas requiring access after an accident, explain why they were excluded; 4.

designation of the codes used for analysis, such as ORIGEN, IS0 SHIELD, QAD or others; 5.

'.the projected doses to individuals for necessary occupancy times in vital areas; and 6.

for other modifications, which allow access to areas where access would be useful but not vital, you should specify the anticipated modifications and the scheduled completion date for moaifications.

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ADDITIONAL ACCIDENT MONITORING INSTRUMENTATION (Containment monitors on'y)

(II.F.1)

Introduction The requirements asscciated with item II.F.1 should be censidered as advanced implementation of :artain requirements to be included in Regulatory Guide 1.97, Revisien 2 "Instrumentalien for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Envirens Conditions Curing and Following an Accident," and in other Regulatory Guides, which will be promulgated in the near future.

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ATTACHMENT 3 (II.F.1)

HIGH RANGE CCNTAIhEENT RADIATION MONITOR l

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POSITION i

In-containment radiation level monitors with a maximum range of 108 rad /hr l

l shall be installed. A minimum of two such monitors that are physically i

separated shall be provided. Monitors shall be developed and qualified to 1

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CLARIFICATION l

1.

Provide two radiation monitor systems in containment which are documented to meet the requirements of Table II.F.1-3.

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The specification of 108 rad /hr in the above position was based on a I

l calculation of postaccident containment radiation levels that included both particulate (beta) cod photon (gama) radiation. A radiation cetector l

that responds to both beta and gama radiation cannot be qualified to i

post-LOCA containment environment but gama sensitive instruments can be so qualified. In order to follow the course of an accident, a cortainment I

monitor that measures only gama radiation is adequate. The requirement was revised to provide.for a photon-only measurement with an upper range of 107 R/hr.

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If two subsystem monitors are required to span the required range in Table II.F.1-3, the ranges of the subsystem monitors shall overlap (i.e.,

upper value/ low value overlap) by at least a factor of ten (10).

4.

The monitors should be located in a manner as to provide a reasonable assessment of area radiation eenditions inside containment. The monitors shculd be widely separated so as to provide independent measurements and should " view" a large fraction of the containment volume. Monitors shculd not be placed in areas which are protected by massive shielding and should ce reasonably accessible for replacement, maintenance," or calibraticn. Placement high in a reactor building dome is not recommended.

5.

For SWR Mark III containments, two such montaring systems should be inside both the primary containment and the secondary containment.

6.

The monitors are required to respond to gamma photens witn energies icw as 60 kev and to provide an essentially flat response for gamma energies between 100 kev and 3 MeV, as specified in Table II.F.1-3.

Monitors that use thick shielding to increase the upper range will under-estimate post-accident radiation levels in containment by several orders of magnitude because of their insensitivity to low energy gammas and are not acceptable.

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- 00CUMENTATION RE0VIRED Provide a description of the two high range containment monitors required and specify the' location of these monitors inside containment.

The description of the monitors should include:

1.

name of manufacturer and model number of the monitors; 2.

verification that the monitors meet the specifications of Table II.F.1-3; 3.

verification that the monitors are or will' be operable on January 1,1981, or prior to full power operation, whichever is later, and, 4.

a plant layout drawing showing.the location of the monitors.

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TABLE II.F.1-3 HIGH RANGE CONTAINMENT RADIATION MONITOR REQUIREMENT The capability to detect and measure the radiation level within the reactor contair. ment during and folicwing an accident.

RANGE 1 rad /hr to 108 rads /hr (beta and gania) er alternatively 1R/hr to 107 R/hr (gama only) (overlap with nornal radiation I

monitor (s) range by a factor of ten (10)).

RESPCNSE 60 kev to 3 MeV photons, with + 20% accuracy for photons of l

0.1 MeV to 3 MeV.

REDUNDANT A minimum of two physically separated monitors (e.g.,

monitoring widely separated spaces within centainment).

RELIABILITY Per Regulatory Guide 1.97, Revisicn 2, Table 1, Instrument Category 1.

.SPECIAL CALIBRATION In-situ calibration by electronics signal substituticn is

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I acceptable for all range decades above 10 R/hr.

In-situ calibration for decades below 10 R/hr shall be by means of

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_ TABLE II.F.1-3 (Continued)

SPECIAL ENVIRONMENTAL QUALIFICATIONS -

Vendors shall calibrate and type test representative specimens of detectors on at least one point in each decade of range from 1 R/hr up to 106 R/hr. Venders shall provide certification of calibration of each detector for at least one point per decade of range between 1 R/hr and / R/hr.

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1 IMPROVED IM-PLANT IODINE INSTRUMENTATION UNDER ACCIDENT CCNDITIONS (III.D. 3. 3)

POSITION Each licensee shall provide equipment and associated training and precedures for accurately determining the airoorne iodine ccncentration in areas within the facility where plant personnel may be present during an accident.,

CLARIFICATION Effective monitoring of increasing iodine levels in the buildings under accident conditions must include the use of portable instruments using sample media which will collect iodine selectively over xenen (e.g., silver zeolite) for the following reasons:

1.

The pnysical size of the auxiliary / fuel handling building precludes locating stationary monitoring instrumentation at all areas where airborne iodine concentration data might be required.

2.

Unanticipated isolated " hot spots" may occur in locations where no stationary monitoring instrumentation is located.

2 3.

Unexpectedly high backgr.ound radiation levels near stationary monitoring instrumentation after an_ accident may interfere with filter radiation readings.

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The time required to retrieve samples after an accident ray result in high personnel exposures if these filters are located in high dose rate areas.

1 Licensee shall have the capability to remove the sampling cartridge to a low background, low contamination area for further analysis.

Tnis area should be l

l ventilated with clean air containing no airborne radionuclides which may

-contribute to inaccuracies in analyzing the sample.

Normally counting rooms in auxiliary buildings will not have sufficiently low backgrounds for such analyses following an accident.

In the low background area, the sample should i

first be purged of any entrapped noble gases using nitrogen gas or clean air free of noble gases. The licensee shall have the capability to measure accurately the iodine concentrations present on these samples under accident conditions. There should be sufficient samplers to s:mple air from all vital areas.

00CUMENTATION REQUIRED S

Provide description of the in plant airoorne radioiodine sampling and analysis l

l systems specifying the number and types of samplers, sample media, sample flushing methods, and sample analysis equipment type and location.

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REFERENCES 1.

U.S. Nuclear Regulatory Commission, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," USV<C Report NUREG-0578, July 1979.

2.

'Lettar from D.- G. Eisenhut, NRC, to All Operating Nuclear Power Plants,

Subject:

Followup Actions Resulting from the NRC Staff Reviews Regarding the Three Mile Island Unit 2 Accident, dated Septemoer 13, 1979.

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Letter from D. B. Vassallo, NRC, to All Pending Operating License Applicatants,

Subject:

Followup Actions Resulting from the NRC Staff Reviews Regarding the Three Mile Island Unit 2 Accident, dated Spetember 27, 1979.

1 4.

Letter frcm H. R. Denton, NRC, to All Operating Nuclear Power Plants,

Subject:

Disscussion of Lessons Learned Short-Term Requirements, dated I

October 30, 1979.

5.

- Letter frcm D. B. Vassallo, NRC, to All Pending Operating License Applicants,

Subject:

Discussion of Lessons Learned Short-Term Requirements, dated Novemcer 9,1979.

6.

Letter frem D. G.. Eisenhut, NRC, to All Power Reacter Licensees,

Subject:

Clarification of-NRC Site Requirements for Emergency Response Facilities at:Each Site, dated April 25, 1980.-

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o 7.

Letter from D. G. Eisenhut, NRC, to All Power Reactor Licensees,

Subject:

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Five Additional TMI-2 Related Requirements to Operating Reactors, dated May 7, 1980.

8.

U.S. Nuclear Regulatory Commission, "NRC Action Plan Developed as a Result of the TMI-2 Accident," USNRC Report NUREG-0660. Vols.1 and 2, May 1980.

9.

U.S. Nuclear Regulatory Commission, "TMI-Related Require.ments for New Operating Licenses," USNRC Report NUREG-0594, June 1980.

10.

U.S. Nuclear Regulatory Commission, " Instrumentation for Light-Water-Cooled Nuclear Power Plan.ts to Assess Plant and Environs Conditions During and Following an Accident," Regulatory Guide 1.97, Revision 2.

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