ML20003E639

Forwards Response to Re Resolution of Open Items Described in NUREG-0680.Functional/verification Test Procedures for Restart Mods Will Be Submitted by 810501,in Startup Tech Specs

ML20003E639
Person / Time
Site:	Crane
Issue date:	04/01/1981
From:	Hukill H METROPOLITAN EDISON CO.
To:	Reid R Office of Nuclear Reactor Regulation
References
RTR-NUREG-0680, RTR-NUREG-680 L1L-100, NUDOCS 8104070453
Download: ML20003E639 (40)

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Metropolitan Edison Company Post Office Box 480 g

Middletown, Pennsylvania 17057 717 944 4041 Wnter's Direct Dial Number April 1, 1981 L1L-100

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R. W. Reid, Chief

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CC U. S. Nuclear Regulatory Commission y,;

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Dear Sir:

N'i Three Mile Island Nuclear Station, Unit 1 ('DC-1)

Operating License No. DPR-50 Docket No. 50-289 Open Item in NUREG 0680 The enclosed table and its attachments represent our response to your letter dated March 6, 1981 concerning the resolution of open items in the TMI-l Restart Safety i^ valuation Report NUREG 0680.

Functional / verification test procedures for the restart modifications will be described in the startup test specification to be submitted by May 1, 1981 (See L1L-054 dated February 28, 1981). The testing planned includes functional testing of the Reactor Protection System Diesel Loading Test and a 48-hour Emergency Feed Water Pump Endurance Test as indicated in l

NUREG 0680.

l Technical Specifications for the restart modifications listed below will be forwarded by May 1, 1981.

Anticipatory Reactor Trips Operability of PORV and Block Valve, Position Indications for PORV and Safety Valves, and Setpoints Containment Isolatiot Modifications 00l Instrumentation to Detect Inadequate Core Cooling j

Emergency Feedwater System Requirements l

810.40 7 0 U3 g

Metropoi tan Ecscn Company :s a Member of t e Genera! Puchc Ut&t es System

R. W. Reid April 1, 1981 TMI-1/IMI-2 Separation Setpoints Associated with ECCS analyses Sincerely,

/

H. D.

1 Director, TMI-l HDH:CWS:vj f Attachments cc:

B. H. Grier - w/o Attachments B. J. -Snyder - w/o Attachments

~

. D. D11 anni - w/o Attachments H. Silver - w/o Attachments L. Barrett - w/o Attachments 1

6

SER-(NUREG-0680) OPEN ITEMS ORDER 0 PEN ITEM DESCRIPTION LICENSE RESPONSE la'- Add'1 6 Commitment to complete EFW 2 hr. air supply Licensee will complete these modifications modification and reset the EFW pump steam prior to exceeding 5% reactor power.

(See chest safety valves.

attached revised page to Licensee's response to NUREG-0737).

4 Provide additional design detail for the Additional design detail concerning major long-term fuel Handling Building ESF venti-equipment location and capacity and indica-lation system, tion of the system autonatic actuation func-tions is attached.

(See revised Pages 4 and 5 and a revised system flow drawing for Re-start Report Supplemental 1 Part 2, Response to Question 52). We are also continuing to evaluate the criteria for this system in hope of achieving further simplification.

5 Implementation schedule and details for the A plot plan for this facility was provided low level solid radwaste storage facility.

on March 4,1981 to Mr. R. Jacobs (NRC).

Additional information is provided in the Re-start Report Supplemental 1, Part 2, Response to Question 53 as amended by Amendment 22.

2.1.2 Justification for applicability of the EPRI NUREG-0737, Item II.D.1 specifies that the Relief and Safety Valve Test Program to required justification be provided by Octo-TMI-1.

ber 1, 1981.

Detailed justification for TMI-l will be provided by this date.

It should be noted that EPRI has obtained input from the various NSSS vendors concerning relief and safety valve flows under accident conditions.

EPRI has enveloped these flow conditions with the flows used for the test program.

In ad-dition, the safety valves being tested by EPRI are the same make, model and oriface area as those installed at TMI-1.

The relief valve (PORV) being tested is also the same make and model as the THI-l PORV, but the ort face area is larger, which is conservative. Also see Licensee's letter dated March 3,1981 (Lil 036).

ORDER ITEM OPEN ITEM DESCRIPTION LI_CEN,SEE RESPONSE 2.1.3.a 1.

Justification for the safety valve el-1.

The subject justification was provided bow tap pressure sensor, with Restart Report Amendment 23 as Ap-pendix 2A.

2.

Study of safety and relief valve dis-2.

See the attached revised response to Re-charge tailpipe thermocouple response, start Report Supplement 1. Part 2, Re-sponse to Question 36, 2.1.4 Design details for containment isolation The design details were provided to R. Ja-modifications.

cobs (NRC) on March 9, 1981.

2.1.6.a-Leakage measurement program description and The scope of the TMI-l leakage reduction procedures, program is described in Restart Report Sec-tion 2.1.1.8.

Procedures for leak testing of Decay Heat Removal System (SP1303-11.16) and the Reactor Building Spray System (SP 1303-11.50) were forwarded to Mr. R. Jacobs (NRC) on March 26, 1981. These procedures are representative of procedures under de-velopment for the other systems included in the leakage reduction program. In addition to the individual leakage procedures, a pro-cedure is being prepared to evaluate the in-tegrated leakages of all of the systems in-cluded in the leakage reduction p.ogram. The preventive maintenance program will consist of performing the above procedures on a rou-tine basis (as specified in each procedure) and performing maintenance as necessary to maintain leakages within acceptance criteria.

The results of leakage measurements inspec-tions and maintenance will be reported to NRC annually in accordance with Technical Specification 6.9.1.B.3 (see cttached Page 6-13 from Technical Specification Change Request 100).

l l'

l.

ORDER OPEN ITEM DESCRIPTION LICENSEE RESPONSE l

2.1.6.b Plant shielding re.iew using final source See the attached revised Section 2.1.7.3 i

terms, and description of required modifi-of the Restart Report.

cations.

2.1.7.a Design description of long-term EFW modifi-See the attached discussion of progress cations.

toward completion of the long-term modifi-cations.

2.1.8.a Design and operational review of the Reac-The required review is described in Restart l

tor Coolant Sampling System and Containment Report Section 2.1.2.4, Amendment 23.

In Air Sampling System, addition, a discussion of the accuracy of the various analyses and installation of the equipment necessary to permit determination of hydrogen or desolved gases will be provided by January 1982.

2.1.8.b Procedures and evaluations for interim high-Design details of the final high-range radio-range radio-effluent monitors and design effluent monitors have been forwarded sepa-l details for the final monitors, rately to Mr. R. Jacobs (NRC).

It is expec-ted that the final monitors will be in place before criticality; therefore, no interim methods are planned.

If, by July 1,1981, it becomes apparent that the final monitors will not be in place, the interim method pro-cedures and description will be forwarded by August 1,1981.

2.1.8.c In-plant iodine nonitoring procedures and The procedures for in-plant iodine monitor-training.

Ing were forwarded March 26,1981 to Mr. R.

Jacobs. Training will be conducted on these procedures and will be completed by July 1, 1981. Emphasis will be placed on two aspects; insuring that exposures are maintained as low as reasonably achievable, aad the require-ments (frequency etc.) for sampling.

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ORDER ITEM OPEN ITEM DESCRIPTION LICENSEE RESPONSE 2.1.9 Transient and accident procedures.

See Licensee's letter dated February 28, 1981 (Lit-054).

'2.2.1.b Long-term training program for the Shift See Licensee's letter dated March 19, 1981 Technical Advisor.

(L1 L- 068).

Additional Design details and procedural guidelines The design details and procedural guide-

' Item No. 4 for'RCS high point vents, lines will be forwarded by July 1,1981 as specified by NUREG-0737.

' Long-Te rm Program outline to meet NUREG-0737 Item See Licensee's letter dated March 30, 1981

~

Order Item II.K.3.30 (Small Break LOCA Methods).

( L1L-0 89).

lNo. 2 t

i Long-Term Design detail for containment pressure, Design details for the containment water

-Order Item water level and _ hydrogen monitors, level and pressure monitors was forwarded j

No. 3 to R. Jacobs (NRC) on March 26, 1981.

De-sign details for the hydrogen monitor will e

be forwarded by June 1981.

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NRC fluREC Of80 (SER)

NRC SPECIFIC LICENSEE I.ICEllSEE C0k9 TENTS ITEll tilllinER ITEM AND SCllEDUI.E TECllNICAl.

3 Sil0RT TITI.E SCllEDULE Pilop0 sal, COHH11HENT REFERTNCES ADDiTIOrlAl. L I Ells (SER)

SER OLl?

1.

CUST level alarms Restart tiot

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liefore > 5% power R.R.

SI, P2, Alasms ulli not lie from Specified (See Comments)

Qlua independent power suppllen leef ore restatt.

Inst, will alarm na l o s r.

of power anil local tank reaalings will be taken.

2.

EFW 4fl hr. endierance tent Rrstart flo t Speci f f eil Before > 5% power R. II. S I, P2, Q 7 Stattup prorcilure t o be nubmltted anil con.Incted SI.R F8 Cl-8 1.

EfW Water sointce transfer He 5% power R. it. St. P2. Ql4 9ER 0 Cl-9 n.

7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> air Supp. I, l'. 7,

b.

Itene t EIU teishine Before > 5% oower Q ll 1.

Il naIetten 7.

lipni aile I:F-V 10A/it for Restant ! Not Sperfiled (m/10/al2 R l!.. SI. P2 414 Will not l'e r " m pl et eil

.it e.im h t e.ik env i e onmene t I"'I o n P 8eStast'elne t.

repil pmen t de I i s. y.

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SUPPLEMENT 1, PART 2 RESPONSE TO QUESTION 52, PAGF]

In addition to the modifications described above, a v?ntilation system to mitigate the consequences of a postulated fuel hand-ling accident in the FHB will be installed.

This new system will meet the requirements of Regulatory Guide 1.52, Revision 2.

This system, and intermediate modifications to the Auxiliary and Fuel Handling Building Ventilation System, are described be-low.

f The Auxiliary (Aux. Bldg) and Fuel Handling Bui.1 ding (FHB) Vent-ilation System will undergo extensive modifications which will be undertaken in two phases as described below.

Phase 1 Prior to restart, the TMI-l ventilation system will have been modified as shown in Attachment 2.

The following equipment will have been added:

1.

Damper U with interconnecting ductwork.

2.

Damper T In the event that contamination (radioactivity) is sensed in the ductwork, Dampers U and T will close, Fan F will trip, and vent-dlation will be via filter trains M and N.

Dampers U and T are seismic Category I, meet ANSI /ASME N509 Construction Classifi-l cation B, Leakage Classification II, and are designed to fail closed.

Phase 2 Prior to the next refueling, the following additional modifica-tions will be made, es shown in Attachment 3:

1.

Filter trains Q and R will be added in the Aux. Bldg. elev.

305 ft.

From the intake side of the filter trains, the composition of the train will consist of a prefilter, an l

electric heater, a HEPA filter, a charcoal filter, and a l

final HEPA filter.

The design of filter trains Q and R will meet the requirements of NRC's Regulatory Guide 1.52,

" Design, Testing and Maintenance Criteria for Post Accident f

l Engineered-Safety Feature Atmospheric Cleanup System Air Filtration and Absorption Units of Light Water-Cooled Nuclear l

(

Power Plants."

2.

Dampers S and V will be added.

3.

Fans G and H will be installed in Aux. Bldg. on elev. 305 l

ft. together will all connecting ductwork and dampers.

l

SUPPLEMENT 1, PART 2 RESPONSE TO QUESTION 52, PAGE 5 4.

Dampers U and T will be blocked open and will no longer respond to closure signals.

During normal operation, filter trains M and N (50% capacity) will be utilized together with fans A,C,E and F.

During refueling operation, l

in addition to the equipment normally operating, filter trains R or Q and fan G or H would also operate.

If contamination (radioactivity) is sensed in the ductwork or Elev.

347' area of the FHB, dampers S and H will close, fan F will trip and an alarm will annunciate in the control room and FHB.

This action will serve to separate the Bldg. and Ventilation systems and assure that all air leakage will be into the FHB.

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ATTAC!! MENT 3 TO QllESTION 52

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Filters:

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Dampers : U and T - Existing; S and V - New

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Alarms / Controls : Dotted Lines h

- Alarm - Control Room CR AUXILIARY AND FUEL llANDI.ING 11UILDING

@ LOC - Alarm - Local in FHB VENTILATION SYSTDI l'RIOR TO NEXT REFUELING

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SUPPLEMENT 1, PART 2 QUESTION:

36.

Provide a detailed descriptica of the backup capability provided for determining the position of the PORV and pressurizer safety valves beyond the differential pressure transmitters.

RESPONSE

The differential pressure transmitters will be the primary indi-cation that either the PORV or the safety valves are open.

In addition, the PORV position will be monitored by accelerometers, which are de-scribed in Section 2.1.1.2.

Another indication that either the PORV or the safety valves are open includes the temperature detectors on the discharge lines of these valves and Reactor Coolant Drain Tanks (RCDT) indications.

The RCDT level i.t recorded in the control room, and there is a high level alarm.

There are also control room indicators for RCDT temperature and pressure.

It has been determined that the ambient temperature in the vicinity of the tailpipe thermocouples influences them in such a way that it is i

difficult to determine PORV or safety valve position using them.

In or-der to correct this situation, additional thermocouple junctions will be installed in the long-term to subtract out the influence of ambient tem-perature.

Once modified, the tailpipe thermocouple system will respond in the following way:

Valve Opening:

The thermocouple indications will rise quickly in response to steam condensation inside the tailpipes.

t Valve closure:

The operator will plot tailpipe differential temperature (tailpipe

(

surface temperature minus ambient temperature) at one minute intervals.

l The observed response will be compared to the predicted response range l

for a fully closed valve.

If the temperature dois not decay at a rate l

between the minimum predicted temperature decay rate and the maximum predicted temperature decay rate, it can be inferred that the valve has not fully closad.

Expected response for a partially closed valve would be for the observed cooldown of the tailpipe to exceed.the maximum ex-pected cooldown rate initially.

Eventually, the observed cooldown rate would level off and track above the minimum expected cooldown rate for a closed valve.- If the' valve is only leaking, the observed response will remain above the expected cooldown rate continuously.

• A typcial temperature decay is gradual as shown on the attached figure.

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1.

A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures and their associated man re= exposure according to work and job functicns, (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), vaste processing, and refueling). The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements.

Small exposures totalling less than 20% of the, individual total dose need not be accounted for.

In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

(This tabulation supplements the requirements of Section 20.407 of 10CFR Part 20).

2.

The following infor=ation on aircraf t movements at the Harrisburg International Airport:

a.

The total number of aircraft movements (takeoffs and landings) at the Harrisburg International Airport for the previous twelve-month period.

b.

The total number of movements of aircraf t larger than 200,000 pounds, based on a current percentage estimate provided by the airport manager.

3.

The following information from the periodic Leak Reduction Program tests shall be reported:

a.

Results of leakage measurements, b.

Results of visual inspections, and c.

Maintenance undertaken as a result of Leakage Reduction Program tests or inspections.

4.

The following information regarding pressurizer power operated relief valve and pressurizer safety valve challenges shall be reported:

a.

Date and time of incident, b.

Description of occurrence, and c.

Corrective measures taken if incident resulted l

.from an equipment failure.

.C.

Monthlv Operating Reports. Rou?.ine reports of operating statistics and_ shutdown experience shall L.

Submitted on a monthly basis to the Office of Inspection and Enforcement, U. S. Nuclear Regulatory l

_ Commission, Washington, D.C. with a copy to the Region I Office no later than the fif teenth of each month following the calendar l

month covered by the report.

6.9.2 Reportable Occurrences Reportable Occurrences, including corrective actions and measures to prevent recurrence, shall be reported to the NRC.

Supplemental reports may be required to fully describe final resolution of an occurrence.

In case of corrected or supplemental reports, reference shall be made to the original report date. --(These. reporting requirements apply only to Appendix A Technical Specifications.)

Tech. Spec. Change Request No. 100 6-13 L}b-@?@

. _ - ~,

f 4

l 2.l'2.3 Plant Shielding Review

[

. 2.1.2.3.1 Introduction A shielding review has been conducted in response to USNRC Report -NUREG-0578 anc the September 13, 1979 letter from f

Darrel G. Eisennut of the NRC to all operating nuclear power plants (Ref.10). The requirements defined in the September 13 letter were subsequently clarified in' letters to all operating i

nuclear power plants from Harold R. Denton dated October 30, 1979 (Ref.' 11) and D. G.. Eisenhut dated September 5,1980 l

' ( Re f. 12), and NUREG-0737 (Ref. 14).

Among the requirements defined in the three NRC letters and NUREG-0737 is a requirement to conduct a review to determine l

whether post-accident radiation fields unduly limit personnel access to areas which will or may require occupancy to permit l

an operator to aid in the mitigation of or recovery from an accident (i.e., vital areas) or unduly degrade the proper i

operation of safety equipment. An evaluation of plant areas and 'the methods used in conducting the review for the Three Mile Island Nuclear Station, Unit 1 (TMI-1) is discussed below.

The results of the review have determined which plant equipment and-which plant areas requiring post-accident access need to be -

i modified. - A descript' ion of all the areas reviewed and their

- access requirements is presented in Section 2.1.2.3.4.

2.1.2.3.2 Re f erences 5

1.

Midland Final Safety Analysis Report, Table 11.1-2, Total Core Fission' Product Activity Versus Time in Equilibrium Cycle.

2.

C. M. Lederer, J. M. Hollander and J. Perlman, " Table of l

Isotopes," Sixth Edition, John Wiley and Sons, Inc., 1967.

j 3.

A..Tobias, " Data for the Calculation of Gcama Radiation Spectra!and Beta Heating from Fission Products (Rev. 3),"

RD/B/M2669, 'CNDC(73)P4, Central Electricity Generating :

I

Board, Research Dept., Berkely Nuclear Laboratories, United p

Kingdom, June (1973).

1 l.

4.

E. D. Arnold ana B..F. Maskewitz, "SDC - A Shield Design

{

Calculation Code for Fuel Handling Facilities," ORNL-3041,

' March 1966..

, 5.

Reactor' Shielding Design Manual, ed. by T. Rockwell,.

Technical. Information Service, Dept. o f Commerce,.

Washington,' DC (1956).-

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H. Ono and A. Tsuruo, "An Approximate Calculation Method of Flux for Spherical and Cylindrical Sources with a Slab Shield," Journal of Nuclear Science and Technology, 2 (6),

pp. 229-235, June 1965.

7.

K. Shure and O. J. Wallace, " Compact Tables of Functions for Use in Shielding Calculations," Nuclear Science and Engineering, 56 pp. 89-94, January 1975.

8.

J. L. Kamphouse, " Cylindrical Source Shielding Equations Utilizing Compact Functions and Including Buildup," Nuclear Science and Engineering, Technical Note, 68, pp. 212-217, November 1978.

9.

U. S. Nuclear Regulatory Commission, "TMI-2 Lessons Learned Task Force Report and Short-Term Recommendations," USNRC Report NUREG-0578, July 1979, Recommendation 2.1.6.b.

10.

Letter from D. G. Eisenhut (NRC) to All Operating Nuclear Power Plants,

Subject:

Followup Actions Resulting from the NRC Staf f Reviaws Regarding the Three Mile Island Unit 2 Accident, dated September 13, 1979.

11.

Letter from H. R. Denton (NRC) to Ali Operating Nuclear Power Plants,

Subject:

Discussion of Lessons Short-Term Requirements, dated October 30, 1979.

12.

Letter from D. G. Eisenhut (NRC) to All Licensees of Operating Plants and Applicants for Operating Licenses and

. Holders of Construction Permits,

Subject:

Prr ' iminary Clarification of TML Action Plan Requirements, dated September 5,1980.

13.

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

14.

Clarification of TMI Action Plan Requirements Of fice of Nuclear Reactor Regulation. Division of Licensing, U.S.

Nuclear Regulatory Commission, NUREG-0737,Section II.B.2, October 1980.

15.

EDS Nuclear, Inc. Report No. 02-0370-1060, Revision 0, October 1980, Environmental Qualification of Class lE Electrical Equipment, Three Mile Island Nuclear Station -

Unit Q2e Report, (See response to.I&E Bulletin 79-01B of January 30, 1981, L1L 026).

2.1-38a Am.

2.1.2.3.3.0 Methods 2.1.2.3.3.1 Source Terms An isotopic core inventory for 310 ef fective full power days in an equilibrium cycle with a power level of 2552 MWt was utilized for the development of the source terms (Ref.1).

This is presented in Table 2.1-7.

This inventory is slightly conservative because the TMI-1 power level is 2533 MWt.

It was utilized in the absence of comparable information in the TMI-l FSAR.

The activity assumed for source term calculations was based on the following:

a.

Liquid Containing Systems:

100% of the core equilibrium noble gas inventory, 50% of the core equilibrium halogen inventory, and 1% of all others.

In determining the source term for recirculated, depressurized cooling water, it was assumed that the water contains no noble gases.

b.

Gas Containing Systems:

100% of the core equilibrium noble gas inventory and 25% of the core equilibrium halogen invent o ry.

Two liquid source terms were used in the evaluation. For sy-stems or portions of systems which will contain post accident fluid recirculated from the reactor building sump, the source term was based on diluting the isotopic inventory discussed in item a. above with the minimum expected volume of fluid in the bottom of the reactor building post accident. This volume includes that of the Borated Water Storage Tank (BWST) and the Reactor Pressure Vessel (PV). This is designated as the

" Recirculation" source.

For systems which will contain post accident fluid from the reactor coolant system which will not be diluted as noted above, the' source term was based on diluting the isotopic l

inventory in item a. above with the volume of fluid in the i

reactor coolant system. This is designated as the " Reactor Coolant" source.

The activity assumed for the containment gaseous source term calculations was based on 100% of the noble gas core inventory and 25% of the halogen core inventory. The containment airborne source term was based on diluting the isotopic inventory of item b. above with the air contained in the containment free volume. This is designated as the

" Containment Gas" source.

2.1-38b Am.

The inventories and source terms discussed above were calculat-ed for the time period immediately af ter the postulated acci-dent.

For other time periods, the decay parameters given in Refs. 2 and 3 were used to adjust the source terms for radioac-tive decay. The sources were converted to standard shielding source term format as a function of time af ter the postulated accident and were used as the basic input data to the shielding codes.

2.1.2.3.3.2 Calculation o f' Dose Rates Both SDC code (Ref. 4) and the Gilbert / Commonwealth developed SPOTl code were used in performing the dose rate calculations.

The SDC Code uses the methodology of Ref. 5 which represents the cylindrical source by an equivalent line source. The SPOTl Code uses the methodology originally presented in Ref. 6, which represents the cylindrical source by an equivalent cylindrical annular segment. This methodology was elaborated on by Shure and Wallace in Ref. 7 and is presented as utilized.in the SPOTl Code in Re f.

8*.

These codes give comparable calculational 1

results; therefore, the use of one code or the other was based on convenience.

These codes were used to calculate dose rates at -the midplane laterally from cylindrical sources. Dose rates were calculated for explicit pipe segments containing various sources. Dose rates were calculated for shielded and unshielded conditions and as a function of time af ter the postulated accident. Dose rates rate from tankage was also calculated utilizing the

. shielding codes.

The dose rate was determined at a representative location within a given area and that dose rate was used as the general area dose rate. The criteria used for selection of a '

representative location was twofold: first, that the dose rate should be reasonably representative-of the area and secondly that the dose rate should be conservative for the area.

Usually, the main contribution of. the dose. rate comes from unshielded-piping in the immediate area. Other sources farther away -or behind shield walls usually contribute significantly-less to the general area dose rate.

As indicated above, midplane dose rate data versus lateral distance was calculated for explicit pipe segments containing various sources.' Then, the distance from a given pipe to the representative location within the area was decennined from the

. physical piping layout drawings. This distance was then used to determine the dose rate from the dose rate versus distance O

~

See Appendix 2B-for clarification of this methodology.

2.1-38c-

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data.

This being done for all pipes in the area, the dose rate contributions were added to obtain the total dose rate at the representative location within the area. Where significant, dose rate contributions from tankage or shielded pipes was also added to obtain the total dose rate for an area.

Denoting time at the inception of an accident as time T=0, and noting that some operations may be performed at times signifi-cantly af ter T=0, dose rates may generally be reduced by a fac-tor of 5 for T=8 hours, and by a factor of about 30 for T=5 days.

2.1.2.3.3.3 Calculation of Doses to Personnel During Post Accident Access to Vital Areas Personnel doses received in performing a specified operation in a given vital area were calculated as the sum of the doses re-ceived during travel to and from the vital area and the dose received while performing-the specified operation in the vital area.

The doses received during travel were determined by calculating dose rates at selected locations along the travel route (or at a single location if the dose rate along the travel route is relatively uniform) using the ce hodology discussed in Section 2.1.2.3.3.2 and multiplying the dose rates by the appropriate travel time for each selected location along the travel route.

It was assumed that the individual travels at a rate of 50 feet per minute along the travel paths as indicated in Section 2.1.2.3.4.

Doses received while performing a given operation were deter-mined by multiplying the dose rate for the given area by the time assumed to perform the operation. Dose rates for the given vital area were determined using the methodology discussed in Section 2.1.2.3.3.2.

Details with regard to explicit areas are given in Section 2.1.2.3.4 2.1.2.3.3.4 Acceptance Criteria Die acceptance criteria for personnel access was based on the following guidelines:

a)- The post accident dose rate in areas requiring continuous occupancy should not exceed 15 mr/hr (averaged over 30 days).

b) The post accident dose rate in areas which do not require continuous occupancy should be such that the dose to an individual during a required access period is less than 5 Res whole body or its equivalent.

2.1-38d Am.

c) The minimum radioactive terms used in the evaluation are to be equivalent to the source terms recommended in Regulatory Guide 1.4 as clarified by the references mentioned in Sec-tion 2.1.2.3.1.

These guidelines were design objectives and were not a basis for limiting access in the event of an accident.

The acceptability of equipment was determined based on the re-sults of the review of electrical safety equipment conducted in response to IE Bulletin 79-01B and has been reported separately (Ref. 15).

2.1.2.3.4 Results 2.1.2.3.4.1 Review of Plant Areas for Post Accident Access Requirements Areas which will or may require occupancy to permit an operator to aid in the mitigation of or recovery from an accident have been designated as vital areas. The control room, Technical Support Center (TSC), Operations Support Center (OSC), sampling station and sample analysis area, post-LOCA hydrogen control system motor control centers, instrument panels, emergency power supplies, security center and radwaste control panels, are included within this designation.

Specific designation and discussion of these and other areas where post-accident access might prove useful follows. All areas evaluated are identified by Roman numerals and are shown in Figures 2.1-12, -13, -14, -15, -16, -17 and 18.

A summary of the dose rate by area is given in Table 2.1-8.

This _ review is predicted on the fact that letdown of reactor coolant outside containment will not be employed then coolant

- a tivity is at unsatisf actory levels. Letdown will be automatically terminated via the containment isolation system and will not be re-established if activity levels are unacceptably high (refer to Section 2.1.1.5).

. Power operated vent valves in the reactor coolant system will -

i ensure that natural circulation and adequate core cooling can

[.

be maintained following a LOCA by ' venting the RCS to the containment atmosphere -(refer. co Section 2.1.2.2, RCS Venting). When letdown via the normal path to outside containment is not permitted, reactor coolant letdown can be accomplished through the RCS vents. Plant procedures will be modified to address post accident letdown as it relates to post accident shielding.

As a consequence of these actions high pressure injection and low pressure injection piping and components located outside the containment building will contain only the Recirculation 2.1-38e Am.

L

source following an accident. Also, there will be no accumulation of radioactive gas resulting from the accident in the waste gas system and components located outside containment.

The radiation dose rates identified herein are based on the conservative a.

mption that any systems and components that may contain the sources of paragraph 2.1.2.3.3.1 will contain those sources at T=0.

Actual recirculation of spilled reactor coolant may not occur until the clean water in the Borated Water Storage Tank is expended.

Areas Requiring Access Area I Area I is located in the Fuel Handling Building at the 306' elevation, as shown in Figure 2.1-12.

This area is part of the travel rcute to other areas in the Auxiliary Building.

A radiation dose rate of 0.03 Rem / hour in this area results from the reactor coolant sample inlet line routed through Area V as shown in the figure.

The time to travel through the area will be less than a minute with resultant exposure of less than 0.5 mrem.

Area II Area II is located on the 305' elevation of the Auxiliary Building as shown in Figure 2.1-12.

It contains the liquid and waste gas control panels and the Decay Heat Removal Pumps remote oilers. Also, Area II is in the travel route to Areas III and IV on this elevation.

l l

The source of radiation in this area is tne Auxiliary Building HVAC exhaust duct which runs the length of the area about 15 feet above the floor. The Jose rate from this source is less than 0.1 rem / hour at T=0

.f all of the 0.1 percent per day con-tainment leakage goes thr sugh this duct.

A dose rate of less than

.0 rem / hour at T=0 results from an assumed I gpm makeup system liquid leakage inside the Auxiliary Building. This dose rate assumes a partition factor of 1 for the noble. gases and 0.1 for the halogens. The 1 gpm leak rate greatly exceeds the anticipated leakage in the Auxiliary l

Building after implementation of the leakage reduction program (reference NUREG-0578 Item 2.1.6a and NUREG-0737 item III D.1.1).

2.1-38f Am.

l

l r

t Post-accident access to Area II, may be desired to perform liquid and gas waste transfers and read waste gas decay tank pressure. Occupancy time to perform these activities are 10 minutes and 1 minute respectively, during which radiation exposures of less than 183 and 18.3 mrem respectively, would be experienced by Area II occupants.

Access to Area 11 may also be required to add oil to the Decay Heat Removal Pumps bearings via the remote oilers. Occupancy time to perform this operation is 5 minutes with a resultant L

exposure of less than 91.5 mrem.

Areas III and IV Areas III and IV are located on the 305' elevation of the Auxiliary Building as shown in Figure 2.1-12.

Area III

- contains engineered safeguards motor control centers (MCC's) 1A and IB, and Area IV contains the reactor coolant pump seal injection filter station.

Post accident occupancy of Area III may be required to reset circuit breakers in MCC 1A and IB and occupancy of Area IV may be required to open valve MU-V198 to by pass the seal injection filters.

A radiation dose rate of 6.3 + E3 Rem / hour at T=0 in Area III and 8.0 + E3 Rem / hour in Area IV results from the following Makeup and Purification system piping and components: reactor coolant pump seal injection piping, high pressure injection piping to cold legs 'C' and

'D', and the seal injection filters and valves.

- Assuming access is not required ~ to Area III until eight hours af ter an accident (i.e. T = 8 hrs) the dose rate would be 1.3 +

E3 rem /hr resulting in a dose of 100 rem for a five minute stay time to reset breakers.

The dose to an operator staying in~ Area IV for five minutes at T = 0 to open the valve will be 666 rem..

9

[

The estimated radiation dose'for travel to and from' Areas III and IV is negligible in comparison to the dose received during area occupancy.

' Areas XI and XII Areas XI and III are located on the 281' elevation of the Auxiliary Building,. reference Figure!2.1-13.- Within these 2.1-38g Am.

i

___m areas are valves requiring post-accident access for boron pre-cipitation control and continued operation of the decay heat system if the postulated accident was a break in one of the low i

pressure injection lines with the decay heat pump in the unfaulted line unavailable.

These operations currently require access to and manual operation of valves DH-VISA and B, DH-Vl9A and B and DH-V38A l

and B to achieve proper system alignment and flow control.

Valves DH-V15A and B and DH-V19A and B are locked open. Valves DH-V38A and B are locked closed. These valves are all located in the decay heat pump pits and DH-V19A, B and DH-V38A, B have reach rod extensions for operation from elevation 281'.

l Access may also be required post-accident to open and/or I

throttle air operated Decay Heat Closed Cooling System valves

- DC-V2A and B, and DC-V65A and B to achieve reactor coolant temperature control.

i i

A dose rate in Area XI of 7.0 + E3 rem /hr at T=0 emanates from Decay Heat Removal System piping associated with the cross-over

~

lines from the decay heat coolers to the makeup pumps via i

valves DH-V7A and B, the piping legs back to MU-V14A and B, and the ' sources located in Areas XIII and XIV.

[

. A dose rate in Area XII.of 2.0 + E4 mrem /hr. at T=0 results l

from decay heat piping associated with injection from the decay heat. pumps via DH-V4A and B, Reactor. Building sump crossover via DH-V12A and B, the piping leg back to DH-V14A and B, and i.

the decay heat drop line to D3-V12A and B via DH-V3.

l The occupancy. time to unlock and realign the manual valves is estimated at 5 minutes, and to operate the powered valves is 7

also 5 minutes. - The resultant dose-to an operator is 580 rem for-Area XI and.1700 rem for Area XII. The doses for travel to and from Areas XI and XII are negligible in' comparison to the dose received while performing the required operations in these-areas.

1 L

i Areas XIII and XIV I

L

- Areas XIII and XIV, as identified in Figure 2.1-13, are the locations of Decay Heat Removal System valves DH-V12A and B, and DH-V64' on the 281' elevation of the Auxilia y Building.

.These are manual valves which are locked closed. Post accident access is requiredifor opening DH-V12A 'and B for boron -

. precipitation control and to line ty: the Decay Heat System to take suction from the reactor coolant hot leg, if. desired.

. Access to DH-V64 is required post-accident 'for boron

. precipitation control.

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A radiation dose rate of 1.7 + E4 Rem / hour at T=0 in Area XIV emanates from decay heat removal piping associated with injection from the decay heat removal pumps via valves DH-V4A and B; reactor building sump crossover via DH-V12A and B; and the decay heat dropline to DH-V12A and B via DH-V3.

A radiation dose rate of 2.1 + E4 Rem / hour at T=0 in Area XIII emanates from the came piping sources as for Area XIV.

For an estimated occupancy time of 5 minutes to unice.k and j

manually open either DH-V12A, DH-V12B, or DH-V64 thc resultant I

radiation dose to the operator is 1.4 + E3 and 1.8 + E3 rem for areas XIV and XIII, respectively.

The extimated radiation dose for travel to and from Areas XIII and XIV via Areas VI, VII and VIII is 165 Rem.

Access could also made via area X.

Area'XV Area XV, located on the 281' elevation of the Auxiliary Build-ing as shown in Figure 2.1-13, contains the engineered safe-guards valve motor control center (HCC) IC.

A radiation dose rate of 25 Rem / hour at 7 = 0 in this area comes from the unshielded reactor coolant sample recircula-tion line routed above MCC IC.

The rate conservatively assumes a pure undiluted reactor coolant sample is drawn at T = 0.

Access to Area XV may be required af ter an accident to reset thrown circuit breakers in MCC 1C.

Occupancy time for resetting circuit breakers is estimated to be two minutes resulting in a radiation dose of 0.83 Rem to the operator.

Area XVI Area XVI comprises two associated plant functions; the Nuclear Sampling Room ( Area XVIA) and the Radiochemistry Laboratory (Area XVIB) on the 306' elevation of the Control Building as shown in Figure 2.1-15.-

The radiation sources for these areas come primarily from the reactor coolant sampling and recirculation lines running into the Nuclear Sampling Room and direct radiation from the Containment Building. The reactor coolant sampling and recirculation source is conservatively assumed to contain pure, undiluted reactor coolant at T = 0.

2.1-38i Am.

The Containment Building direct radiation source is less than 0.5 mrem / hour. The radiation dose rate at T = 0 in the Nuclear Sampling Room is 276 rem / hour.

Post-accident access to these areas may be required to draw and prepare a reactor coolant sample for analysis.

Refer to section 2,1.2.4 for the discussion of doses received while obtaining a reactor coolant sample.

Area XVIII Area XVIII is located in the Intermediate Building as shown in figure 2.1-16.

This area is the location of the emergency feedwater pumps and associated piping and components. Post accident access to this area may be desirable for inspection and maintenance of equipmenc.

A dose rate at T=0 of 0.5 mrem /hr. in Area XVIII results from radiation from the containment building.

Area XX Area XX, located on the 305' elevation of the Intermediate Building as shewn on Figure 2.1-16, contains the hydrogen recombiners.

Access to Area XX would be required during the first six days following an accident for installation of the second recombiner. Start-up of a recombiner would be required -

approximately 12.6 days af ter the LOCA.

The dose rate in Area XX comes from the Containment Building and the operating recombiner.

At-T = 0 the dose rate (without a recombiner operating) is 30 mrem /hr. At T = 12.6 days the dose rate, with a recombiner in operation, is 1.56 rem /hr.

The radiation dose rate for travel to Area XX or the hydrogen recombiner control panels, located in a room directly below the recombiner on the 295' elevation, ie negli~ible.

g Areas XVII, XXI and XXII l

Areas XVII, XXI and XXII are the Operations Support Center, l

Technical Support Center, and Control Room as shown in Figures i

2.1-L5,12.1-17, and 2.1-18 respectively.

The Control Room is on the 355' elevation of the Control Building. The Technical Support Center is located on the 322' elevation of the Control Building. The Operations Support Center is located in the Health Physics Laboratory on the 306' elevation of the Control Building.

2.1-38j Am.

l The radiation dose rates for these areas comes primarily from the reactor coolant sampling and recirculation lines running into the Nuclear Sampling Room (Area XVIA on Figure 2.1-15) and secondarily from direct radiation from the Containment Building. The reactor coolant sampling and recirculation source is conservatively estimated to contain the Reactor Coolant Source at T = 0.

The Containment Building direct radiation source is less than 0.5 mrem / hour.

The radiation dose rates in the Technical Support-Center, ( Area XXI) the Control Room ( Area XXII); and the Operations Support Center ( Area XVIIB) and the monitor area ( Area XVIIA) are 480, 1.7,18 and 340 mrem / hour, respectively. The radiation dose rate in these areas would be negligible if reactor coolant sampling was not assumed at T = 0.

Area XXIV Area XXIV is the diesel generator building. Access to Area XXIV is required post-accident for operation / maintenance of the diesel generators.

The radiation dose rate in area XXIV comes from the Containment Building. At T=0 it is less than 10 mrem / hour.. The estimated dose accumulated during travel from the Control Building through the Turbine Building to area XXIV is negligible.

Security Access Center t

The security access center is on a direct line of sight with the containment at a distance of about 285 feet. The post-accident dose rate at this location is 750 mr/hr at T=0.

It will take approximately ten days for the dose rate to decay to 15 mr/hr. The dose rate average over 30 days.is less than t

15 mr/hr.

4 Areas for Which Post Accident Access is Not Required 4

The remaining areas of Figures 2.1-12 through 2.1-18 are those for which post accident access is not necessary. - Access to these areas is not needed because; 1.

These areas contain equipment and piping necessary for accident mitigation but do not require personnel access or, 2.

These areas do not contain any components for which

' post accident operator access is necessary or desirable.

These areas are-listed in Table 2.1-6.

2.1-38k.

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2.1.2.3.4 Review of Radiation Af fects on Electrical Equipment The design review of electrical equipment has been performed in accordance with NRC IE Bulletin 79-01B and has been submitted seperately (ref. 15).

2.1.2.3.5 Conclusions A.

The TMI-l plant shielding review has identified areas where post-accident radiation will preclude operator access.

Table 2.1-8 lists the plant areas evaluated and identifies the corresponding dose rates.

Areas where calculated doses preclude post-accident access without appropriate modifications are:

Areas III and IV-Makeup and Purification System seal injection filter bypass valve MU-Vl98 and MCCs lA and IB Areas XI and XII -

Decay Heat Removal System Valves DH-VISA & B, DH-V19A & B DE-V38A & B for Boron precipitation control and Decay Heat Closed Cooling System valves DC-V2A & B and DC-V65A

& B for decay heat removal temperature control.

Areas XIII and XIV - Decay Heat Removal System valves DR-V12A & B, DH-V64 for boron precipitation control and sytem operation..

B.

Dose rates in the following areas exceed 15 mrem /hr but the dose rates, averaged over 30 days are less than 15 mrem /hr.

Area XVIIB Operations Support Center Area XXI Technical Support Center Security Access Center C.

The calculated dose rates for the. remaining areas of the plant where post-accident access might be desirable are low enough to allow access for the expected period of occupancy..

2.1.2.3.6 Planned Modifications A summary of planned modifications is given in Table 2.1.9.

2.1-381 Am.

Table 2.1-6 Areas for Which Access is Not Required Area Number Description Figure 2.1-XX IX Makeup Pump Compartments 13 XIX Leak Rate Air Dryer Area 16

& Piping Compartments XXIII Decay Heat Pits 14 l

1 i

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d TABLE 2.1-7 ISOTOPIC CORE INVEMTORY 4

Total Cora Total Core Inventory in Inventory in Isotope Curie s Isotope Curies Br 84 1.57 +E7 Xe 133 1.27 +E8

(

Br.85*

2.19 +E7 Xe 135m 3.26 +E7 Kr 83m.

9.25 +E6 Xe 135 2.09 +E7 Kr 85m 2.19 +E7 Xe 138 1.17 +E8 Kr - 85 5.30 +E5 I

129*

1.80 +E0 Kr 87 4.00 +E7 I

131 7.35 +E7

'Kr 88 5.60 +E7 I

132 8.62 +E7 Rb 88 5.64 +E7 I

133 1.28 +E8 Sr 89 7.42 +E7 I

134 1.60 +E8 Sr 90 3.99 +E6 I

135 1.27 +E8 Sr 91 9.72 +E7 Cs 134 1.27 +E6

.Sr 92 9.50 +E7

.Cs 13 6 8.02 +E5 Y

90 3.96 +E6 Cs 137 4.99 +E6' Y

91 9.85 +E7 Cs 138 1.23 +E8 Mo 99 1.28 +E8 Ba 137m 4.67 +E6 Ru 106 2.29 +E7 Ba 14 0 1.25 +E8 Xe 131m 4.38 +E5 La 140 1.27 +E8 i

Xe 133m

-3.07 +E6 Ce 14 4 7.50 +E7 l

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• Deleted as insignificant for. subsequent calculations.

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TABLE 2.1-8 RADIATION DOSE RATE BY AREA Figure i

Reference Dose Rate at T=0 Area (2.1-XX)

MREM /HR I

12 3.0 + E1 II 12 1.0 + E6 III 12

, 6.3 + E6 IV 12 8.0 + E6 V

12 1.3 + E4 VI 13 1.5 + E4 VII 13 8.8 + E4 VIII 13

.2.1 + E6 IX 13 1.9 + E6 X

13 1.7 + El XI 13 7.0 + E6 XII' 13 2.0 + E7 XIII 13 2.1 + E7 XIV 13 1.7 + E7 XV-13 2.5 + E4 XVI 15

-2.8 + ES

. XVIIA 15 3.4 + E2 XVIIB 15 1.8 + El' XVIII.

16 0.5 + EO

- XIX 16 6.1 + E3 XX 16 1.4 + ES XXI 17 4.8 + E2

. XXII 18 1.7 + EO XXIII 14 4.0 + E7 XXIV 16 1.0 + E1 l.

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i i i-1 STATUS OF LONG TERM MODIFICATIONS [ FOR UPGRADING OF THE TMI-l j. EMERGENCY FEEDWATER (EFW) SYSTEM 1. C,avitating Venturi i 1 l Engineering and calculations necessary to size and purchase i' this equipment are complete. A purchase order is expected to be issued in May 1981 and no delivery difficulties are i. anticipated. ] 2.. ' Redundant Control and Block Valves The block valves have been selected and orders are expected to be placed by June 1981. The control valves were t selected to be compatible with the existing air operated control valves since upgrading of the existing valve operators was planned. Recent information from the vendor indicates that upgrading of the existing valves is more i difficult than previously indicated and may require the -involvement of test facilities not-under the control of the vendor and may cause sched.le problems. We are evaluating the use of. control valves supplied by other vendors; however, the-delivery schedules for the alternatives may be as long:as two years. In addition, the compatibility of these alternative designs would also require a redesign of 4 the control systemLand the air supply. Selection of the ' valve design is expectedLto be made by May 1981:and delivery schedules will be known at that time. t s 3. Addition of'OTSG level and-Feedwater/ Steam d/p To The Safety Grade EFW Pump Auto-Start System The:Feedwater/ Steam d/p initiation signal is being reviewed-F for reliability since inadvertent. actuation may contribute to overcooling events.. It is possible that-when the review .is-complete, this signal will be determined.to be ~ 'y. unnecessary due to the installation of automatic start on i low OTSG level.

The automatic start on low OTSG level is delayed due to'the

_need to obtain qualifiedLlevel, transmitters.. The . transmitters have been. ordered and.are currently undergoing r f-_ Lqualificationitesting. -(See also item 4 below) [ -t p .TheJdesign' details are complete but are'on-hold pending. further1 review of the Feedwater/SteamLd/p signal.noted above. E r-

...a 4. Safety Grad,e OTSG,I? vel Instrumentation and Control Safety grade le t indicators independent of ICS/NNI qualified to IEEE-323 (1971), will be installed for restart. This system will be replaced with a safety grade system qualified to IEEE-323 (1974), in the long term. The long term level indication system is being used to generate the signals necessary for the safety grade automatic OTSG 1evel control system. The control system is currently being designed; however, the level transmitters have been ordered and are currently undergoing qualification. We are participating with severcl other utilities in an effort to obtain qualified differential pressure transmitters to meet the requirements of IEEE-323 (1974). Quarterly each utility receives an allotment of transmitters. The number of transmitters needed by each utility depends on the number of modifications required to be implemented and may exceed the quarterly allotment. This modification is competing with item 3 above for level transmitters. We are currently reviewing the designs for this modification and item 3 in the_ hope of achieving further simplification. The design is expected to'be completed during the first quarter of 1982. 5. Condensate Storage Tank Level Indication Low Low level alarms are planned to replace the existing control grade system in the long term. The level transmitters have been ordered and the design is under development with with engineering expected to be complete by November 1981.}}