ML20203B042
| ML20203B042 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/02/1999 |
| From: | Langenbach J GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20203B048 | List: |
| References | |
| 1920-98-20522, NUDOCS 9902100197 | |
| Download: ML20203B042 (17) | |
Text
,
i i,.
{
GPU Nuclear,Inc.
(..
Route 441 South NUCLEAR Post Office Box 480 Middletown. PA 17057-0480 Tel 717-944-7621 February 02, 1999 1920-98-20522 U.S. Nuclear Regulatory Commission Attention: Document ControlDesk Washington, DC 20555
Dear Sir:
Subject:
Three Mile Island Nuclear Station, Unit 1, (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request (TSCR) No. 274 Engineered Safeguards Feature (ESF) Systems I.cakage Lindi And-Response to Request for Additional Information (RAI), dated November 20,1998 - re: Control Room Habitability Evaluation References-List of prior submittals related to this TSCR are at the end of this cover letta.
In accordance with 10 CFR 50.4 (b)(1), enclosed is TSCR No. 274 (Enclosure 1), which supersedes TSCR No. 266 (Reference 1). The purpose of this TSCR is to expand the scope of systems and test requirernents for post-accident Reactor Building sump recirculation ESF systems; and, to increase the maximum allowable leahge of Technical Specification 4.5.4 for the applicable portions of the
)
ESF systems outside of containment which any have leakage in the Auxiliary Building.
The scope of systems is expanded by adding the Building Spray (BS) and Make-Lp (MU) systems i
. in addition to the Decay Heat (DH) Removal system presently addressed by Tech. Spec. 4.5.4. The new allowable leakage value is consistent with the revised dose consequence analyses performed
' for the bounding Updated FSAR (UFSAR) design basis accident (DBA), namely the Maximum Hypothetical Accident (MHA). This TSCR safety rialysis also describes changes to the Control Room (CR) Habitability Evaluation which address the subject NRC RAI above, and updates the CR l
Habitability Evaluation dose consequence results of Reference 2, submitted to NRC on March 24, 1998.
10000 0
- In addition, this TSCR also revises the Technical Specification 3.15.3 Bases for the Auxiliary and Fuel Handling Building Ventilation System (AFHBVS). The revisions to Technical Specification j
3.15.3 Bases for the AFHBVS serve to clarify syster, design requirements and accident analysis considerations.
~
9902100197 99020 PDR ADOCK 0 9
P PD_ft
. - _... _ _. _ -. _. _. _. _.....~.._ _ _
.~
e U.S. Nuclear RW-*n Commission 1920-98-20522-Page 2 l
dhanges to the Updated FSAR (UFSAR) required to reflect the revised analyses and clarification of the assumptions will be included in the next revision of the UFSAR to be submitted following the next refueling outage, pursuant to 10 CFR 50.71(e), assuming NRC approval of the license I
amendment associated with this TSCR.
Using the standards in 10 CFR 50.92, GPU Nuclear (GPUN) Inc. has concluded that these proposed changes do not constitute a significant hazards consideration, as described in the enclosed analysis paformed in accuid.c,ce with 10 CFR 50.91(a)(1).
GPUN Inc. requests review and approval ofGis TSCR by May 14,1999 in order to have sufficient time for preparation of the required procedure changes needed to perform the Technical Spec 3 cation 4.5.4 surveillances during the upcoming 13R refueling outage targeted to begin on I
Septensbar 10,1999.
Also enclosed is the Certificate of Service for this request certifying service to the chief executives of the township and county in which the facility is located, and the designated official of the Commonwealth ofPennsylvania's Bureau of Radiation Protection.
1 Sincerely,
{
)wfD ames W. Langenb h l
Vice President and Director, TMI JWilgmg File: 98159
Enclosures:
(1) TMI-l. TSCR No. 274 Safety Evaluation, No Significant Hazards Consideration, and Technical Specification Revised Pages (2) Certificate of Service for TMI-1 TSCR No. 274 Refemnces-1.
GPUN Inc. Letter No. 6710-97-2252, dated July 30,1997 -- TSCR No. 266, Accident Recirculation Leakage Limits 2.-
GPUN Inc. Letter No.1920-98-20145, dated March 24,1998 " Control Room Habitability Evaluation" 3.
GPUN Inc. Letter No.1920-98-20561, dated October 6,1998 - POLESTAR Piur;&ary Report, Calculation PSAT 05656A.04, " Calculation ofThE-1 Engineered Safety Feature Component Leakage Iodine Release" 4.
GPUN Inc. Letter No. 1920-98-20526, dated October 15,1998 - LAR No. 276, Revised Accident Analysis AtmosphericDispersionFactors oc:
. AdministratorRegionI
~IMI Senior Resident Inspector TMI Senior Project Manager
- ~.. _.. _
e v
.j i
1 i
s ENCLOSURE 1 l
TMI-l TSCR No. 274 Safety Evaluation,'
3 1
No Significant Hazards Consideration and Proposed Technical Specification Revised Pages j
i i
l i
i 1
J l
',p i
i
F 1920-98-2055
)
Page 1 of13 1.
TECHNICAL SPECIFICATION CHANGE REOUEST (TSCR) NO. 274
)
5 GPU Nuclear (GPUN) Inc. requests that the following changed replacement pages be j
insented into the existing Technical Specifications:
l
~ Revised Technical Specification Pages: iii,3-62d, and 4-45 1
These pages are attached to this Enclosure.
II.
REASONS FOR CHANGES Table of Contents:
l The Table of Contents page iii is revised to reflect the new title of Technical Specification Section 4.5.4, namely: " Engineered Safeguards Feature (ESF) Systems Leakage."
Technical Specification 3.15.3 Bases 1
i Background and Description of Changes:
I The present description of the Waste Gas Tank Rupture (WGTR) accident in the TM1-1 UFSAR Section 14.2.2.6 states that the reactor coolant passes through purification l
demineralizers which remove 99% of the iodine, and implies that carbon filters in the
' Auxiliary Building with 90% efficiency would remove additional iodine prior to the radioactivity releases into the environment. However, the ' calculation of the radiological 1
consequences for this accident did not take credit for iodine removal via the purification j
demineralizers or the AFHBVS carbon filters.
This TSCR revises the Technical Specification Section 3.15.3 Bases for the Auxiliary and Fuel Handling Building Air. Treatment System (a.k.a., the Auxiliary and Fuel Handling Building Ventilation System - AFHBVS) to clarify the system design requirements and accident analysis considerations.
The proposed changes revise statements to the effect that the AFHBVS is not, credited in reducing off-site doses for the Maximum Hypothetical Accident (MHA) or the Waste Gas Tank Rupture (WGTR) accident. These Bases changes also clarify that the Fuel Handling i
Building Engineered Safety Feature (ESF) Air Treatment System is utilized for the Fuel Handling Accident in the Fuel Handling Building consistent with Tecimical Sp-cificatbn 3.15.4. ' Lastly, the' References section of Technical Specification 3.15.3 Bases is also i
updated to include appropriate changes in identified UFSAR sections.
1 j
1
- . 4 1920-98-2055 L
Page 2 of13 Technical Specification 4.5.4 and Bases:
Backaround and Description of Channes:
This TSCR No. 274 supersedes TSCR No.! 266 (Reference-1)' and its Supplement 1 (Reference 2) which resuhed in the issuance of License Amendment No. 205 needed to resolve a discrepancy between the previous Technical Specification 4.5.4. rate for Decay Heat Removal System Leakage of 6 gph and that value identified in the UFSAR accident analyses, i.e.,0.6 gph. Technical Specification 4.5.4 is being revised to: expand the scope of systems and tes' requirements for post-accident Reactor Building samp recirculation ESF Systems;.and,1e increase the maximum allowable leakage for the applicable portions
- - outside of containment. The revisions proposed serve to address specific items identified in NRC Inspection Report 96-201 and the TMI-l Licensee Event Report (LER)97-004, Rev.0.
The scope of systems is expanded by adding the Building Spray (BS) and Make-Up (MU) -
systems to the Decay Heat Removal (DHR) system, previously addressed by Technical Specification 4.5.4. ' The new allowable leakage value is consistent with the revised Maximum Hypothetical Accident (MHA) dose consequence analysis. The revised MHA analysis includes new and/or revised assumptions and conservatisms, as discussed below in the Safety Evaluation Justifying Changes. Therefore, the Specification 4.5.4 is being re-titled as " Engineered Safeguards Feature (ESF) Systems Leakage" and the total leakage rate is being increased from 0.6 gph to 15 gph. This TSCR also revises' the testing criteria previously applied to the DHR system by deletion of the alternative " hydrostatic test" method, and conforms the surveillance testing for all three systems to the same testing criteria. Consequently, conforming changes to the Bases are required as proposed to reflect the elimination of the alternate methodology.
TSCR No. 274 also incorporates information derived during the development of the Control
- Room Habitability Evaluation performed to determine thyroid dose consequences submitted by GPUN Inc. letter dated March 24,1998 (Reference 3), as well 'as new iodine partitioning factors calculated by POLESTAR and submitted to the NRC by GPUN Inc. letter dated October 6,1998 (Reference 4). This analysis addresses those concerns raised in the Staff's Request for Additional Information (RAI) dated November 20, 1998 regarding the assumptions used in the TMI-l Control Room Habitability Evaluation. Accordingly, the Control Room habitability analysis using the MHA source term release provided by this TSCR revises that provided by GPUN Inc. in Reference 3, in that: (1) the Staff recommended 3-sigma approach has 1 een used for the calculation of atmospheric dispersion coefficients using ARCON96 methodology for the Reactor (Containment) Building to the Yard Intake pathway; (2) the release sources outside the Reactor Building containment structure are conservatively being treated as point sources, rather than diffuse sources, with no credit given to the dispersion effects created by other structures within the building complex; (3) the use ofiodine partitioning factors for Auxiliary Building ESF releases (as provided in the POLESTAR report, Reference 4); and, (4) the use of meteorological data consistent with other licensing applications, e g., License Amendment Request No. 276 (submitted by Reference 5) and TSCR No. 272.
lw
- Enclosure 1.
1920-98-2055 Page 3 of13 i
III.
S AFETY EVALUATION JUSTIFYING CHANGE HL1 Technical Specification 3.15.3 Baser Changes l
' The proposed changes to Technical Specification 3.15.3 Bases are being requested ;o reflect that no credit is taken for the use of AFHBVS carbon filters in the revised Maximum
. Hypothetical Accident (MHA) analysis.
This change request 'does not require any modifications to plant stmetures, systems, or components. The revised MHA analysis dose consequence results are below the 10 CFR 100 guidelines for both the exclusion area-j boundary (EAB) and the low population zone (LPZ).
The Control Room. dose 1
consequences for the revised MHA analysis results in a potential thyroid dose to an operator which is slightly higher that the current Standard Review Plan Section 6.4 guidelines (See -
Table 1); however, the resuhs do not exceed the permissible annual occupational exposum j
limit of 50 Rem to the thyroid as specified in'10 CFR 20.1201 (a) (ii),. In addition, self-contained breathing apparatus and potassium iodide are available to mitigate the potential dose consequences. Further, the T.S. 3.15.3 Bases changer have no affect on the Waste Gas Tank Rupture (WGTR) accident analysis as no credit was taken in the original analysis of record for the use of the'AFHBVS carbon filters. Hence, these proposed Bases changes have no impact upon' nuclear safety, or safe plant operations. The current UFSAR incorrectly implies that credit was taken for these filters in the Chapter 14 description of the WGTR safety analysis. The UFSAR error is scheduled to be corrected with the next UFSAR update pursuant to 10 CFR 50.71e.
l HL2 Technical Specification 4.5.4 ESF Isakage Csanges.
l l
The impact of the revised ESF Systems leakage rate of 15 gph on off-site dose consequences and Control Room habitability are determined by dose analyses performed considering the bounding MHA design basis accident of the UFSAR. The revised Control i
Room Habitability Evaluation dose consequence analysis now takes credit for the use of iodine partitioning factors,. and incorporates the conservative 3-sigma approach recommended by the Staff for the calculation of atmospheric dispersion factors (X/Qs) for l
the Reactor (Containment) Building to Yard Intake pathway. The MHA postulates a large L
release of fission products to the Reactor Building. The release is non-mechanistic and so a l
specific means for it to occur is not postulated. The intent of the safety evaluation is to demonstrate that the safety of the plant is assured with tne occurrence of a postulated large releaseof adioactivity.
The revised MHA analysis for Off-Site and Control Room dose consequence determinations l
was performed using the following assumptions and analysis methodologies:
I=
L SOURCETERM-RELEASE ASSUMPTIONS:
'1, The current Updated FSAR isotopic core inventory for a power level of 110% of
~
2568 MWt, which is more conservative that the 102% specified in Standard Review Plan 15.6.5.HI.4.a.
4
.L y
+
-y
Encbsure 1 1920-98-2055
' Page 4 of13 l
- 2. '
The activity released from the core was based on ANSI /ANS 56.5-1979, "PWR and BWR. Containment Spray. System Design Criteria".
The fractional activities
+
released into the Reactor Building are as follows:
One hundred percent of the noble gases Elemental iodine:. 47.75% of the total iodine in the core. 50% of the elemental iodine released was assumed to immediately plate out.
= >. -
Organic iodine:.1% of the total iodine in the core.
F Particulateio:line: 1.25% of the totaliodine in the core.
3.
lodine removal from the Reactor Building atmosphere was modeled assuming two Reactor Building (RB) Containment Spray trains with only two out of three RB emergency cooling units (fans) in operation. The use of two spreys was found to -
give a higher dose consequence thar. one spray.
This can be explained by considering that the elemental iodine spray removal coefficient is-limited to a maximum of 10 hr , and ESF leakage would begin earlier with 2 spray pumps in operation., A volumetric gas flow rate of 58,000 cfm was assumed for the rate of mixing within the Reactor Building between the sprayed and unsprayed regions
. based on the capacity of twa cooling fans.
4.
Reactor Building (Containment) Spray iodine removal efficiencies for two spray trains were assumed to be:
Elementaliodine: 10 hf' 3
Particulate iodine: 6.06 hr Organiciodine: 0.0144 hr 5.
When determining the R8 atmosphere iodine concentration, credit was taken for Containment Spray caly until the time at which RB sump recirculation (28.5 minutes) begins. However, potential leakage from the Containment BS system was assumed to be a contributor to ESF leakage for the entire 30 days.
6.
A Reactor Building (Containment) leak rate of 0.1% was used based on the TMI-l Technical Specification limits for peak. reactor building pressure (55psig) during a LOCA for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then a 0.05% leakage rate was assumed for the remaining 29 days. In addition, no credit was taken for the AFHBVS filters in the revised MHA dose calculations, which is reflected in the proposed clarification in this TSCR for the Technical Specification 3.15.3 Bases.
OFF-SITE ATMOSPHERIC DISPERSION FACTORS:
7.
The Exclusion Area Boundary (EAE) and the Low Population Zone (LPZ) atmospheric dispersion coefficients were determined using Regulatory Guide 1.145 methods, and are the same as those provided in the License Amendment Request No. 276, dated October 15,1998 (Reference 5), as follows-l Interval X/Q l
EAB:
0-2 Hrs.
8.0E-4 sec / ra3 n,
' Enclosure 1 1920-98-2055
- Page 5 of13 LPZ:
Interval X/Q 0-2 Hrs.
1.4E-04 sec / m' 2-8 Hrs.
6.0E-05 sec / m' 3
8-24 Hrs.
3.9E-05 sec / m.
1-4 Days 1.6E-05 sec / m' 4-30 Days 4.0E-06 sec / m'
' CONTROL ROOM ATMOSPHERIC DISPERSION FACTORS:
8.
The atmospheric dispersion coefficients for Reactor Building (Containment) releases to the Control Room ventilation via the Yard Intake have been determined using the diffuse leak approach of the ARCON% code transformed into a continuous virtual point source, and the NRC recommended 3-sigma approach for conservatively estimating initial diffusion coefficients to establish the virtual point distance. The site meteorological data for 1992,1993,1995, and 19% (previously submitted to the NRC by Reference 5*) was used as input for determining the 95* percentile X/Q values, as follows:
Containment to Yard Intake Pathway-Interval X/Q 0-2 Hrs.
3.40E-04 sec / m' 2 - 8 Hrs.
2.25E-04 sec / m' 8 - 24 Hrs.
1.02E-04 sec / m'
.1-4 Days 7.61E-05 sec / m' 4-30 Days 4.99E-05 sec / m' 9.
The atmospheric dispersion coefficients for a Reactor Building (Containment) release to the Control Room Ventilation Exhaust have been eve.luated using data sources from wind tunnel model experiments to provide a conservative estimate of X/Q values. This alternative method was used because geometry concerns make the ARCON96, corrected Gaussian model, unsuitable. The TMI-l model evaluates selective release points on the containment surfaces exposed to the environment. A fractional release is determined for each release point representative of a release area. Experimental surface concentration coefficients for buildings are used to detennine X/Q. The coefficient, K, is equal to (X/Q)Au, where: A is blockage area; and, u is wind speed. From a known blockage ' rea associated with wind direction a
and speed, the sector (X/Q)u value is determined. Using hourly site meteorological data for 1992,1993,1995, and 1996, hourly X/Qs are calculated and the applicable 95th, 90th, 80th, and 60th percentile values are determined for use in the dose assessment model. The resulting X/Q values at the respective intervals are as follows:
1*_1994 Metcomlogical data was not used hem >w the data did not meet the recovery requirements governed ty Regulatory Guide 1.23.
s V
1920-98-2055 Page 6 of13 Containment to Ventilation Exhaust Pathwav:
Interval X/Q 0-8 Hrs.
1.%E-03 sec / m' 8-24 Hrs.
1.37E-03 sec / m' l-4 Days 9.14E-04 sec / m' 4-30 Days 5.09E-04 sec / m' 10.
The atmospheric dispersion coefficients for Control Room ingress and. egress are conservatively based upon predictions of dispersion near the Control Building ventilation exhaust point. As no direct external pathway exists for this mode of dispersion to the Control Room the X/Q values are the same as the Containment release to Ventilation Exhaust above.
I 1.
The atmospheric dispersion coefficients for postulated releases from the Borated Water Storage Tank (BWST) were evaluated as a point source release, a conservative treatment of this release as it does not account for the effects of blockage by the large surrounding stmetures, e.g., Reactor (Containment),
Intermediate and Turbine buildings. The BWST is closer to the Control Building Ventilation System yard intake than the Reactor Building (76 meters compared to 91 meters); therefore, the conservative approach was to determine the X/Q for this release separate from any effects the surrounding structures may have on dispersion.
. Using the ARCON% Code to calculate the atmosphenc transport X/Qs, the resulting values are as follows:
1 BWST Release Pathway:
Interval X/Q 0-2 Hrs.
8.45E-04 sec / m' i
2-8 Hrs.
5.23E-04 sec / m' 8-24 Hrs.
2.49E-04 sec / m' l-4 Days 1.77E-04 sec / m' 4-30 Days 1.19E-04 sec / m' 12.
The atmospheric dispersion coefficients for postulated aeleases from the Auxiliary Building were determined using point source releases from the closest point to the yard intake. This approach is conservative because the enhanced effects of dispersion due to other structures within the plant complex are not credited resulting i
in higher concentrations at the yard intake than would actually exist. The resulting X/Q values at the respective intervals are as follows:
Auxiliary duildina Release Pathway:
Interval X/Q 0-2 Hrs 3.02E-03 sec / m' 2-8 Hrs 2.08E-03 sec / m' 8-24 Hrs 1.02E-03 sec / m' l-4 Days 6.63E-04 sec / m' 4-30 Days 4.37E-04 sec / m' m
]
p'(
\\
- Enclosure 1
~ 1920-98-2055
, Page 7 of13 -
' CONTROL ROOM-ASSUMPTIONS:
13.
The breathing rate for a Control Room operator was assumed to be
-- 3.47 x E-4 cu m/sec.
l14.
Exposure was based on 100% occupancy of the Control Room by an operator for the
' first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following release, followed by 60 % occupancy for the next 3 days, and 40 % occupancy for the remaining 26 days, per the SRP.
15.
Unfiltered in-leakage to the Control Room was assumed to be 10 cfm to account for
. ingress and egress per SRP 6.4
- 16. '
Hypothetical failures of the Control Room emergency recirculation dampers AH-D37 and AH-D39 for a maximum period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> were assumed, as this results in l
' the highest Control Room dose ' pursuant to Case 2 of the March 24,1998 Control l
. Room Habitability Evaluation, (Reference 3).
ESF LEAKAGE - IODINE PARTIONING FACTORS:
l' L
17.
Iodine release fractions from ESF leakage were based on SRP Section 15.6.5, Appendix B guidance that if the calculated flash fraction is less than 10%, or if the l;
water is less than 212' F, then an amount ofiodine smaller than 10% ofthe iodine in the leakage may be used ifjustified based upon actual sump pH history and j
ventilation rates. This was calculated in the proprietary Polestar Calculation No.
j L
PSAT 05656A.04, Rev. O. (Reference'4) based on the following considerations:
{
i
?
Evaluation of the elemental iodine concentration in the ESF liquid. This is a function of the core inventory ofiodine, the iodine release from the core during the MHA, the total ESF liquid mass, liquid density, and the ESF liquid pH.-
Evaluation of the iodine concentration in the Auxiliary Building gas space.
e-Evaluation of the iodine release to the environment based on the volumetric flow' of gas from the Auxiliary Building gas space to the environment.
The iodine partitioning between the liquid and bulk gas is based on equilibrium conditions. That is the fraction of the 12 in the liquid is assumed to partition
[.,
instantaneously to the bulk gas phase. This is very conservative as it neglects any
[
transient effects and it neglects the resistance to gas transport across the gas
[
boundary layer.-
i' e
The ESF component leakage is assumed to remain at the Reactor Building sump liquid temperature. This will maximize the partitioning ofiodine to the bulk gas.
Neglecting the heat transfer from the leaked liquid to the Auxiliary Building room wall surfaces and structures, as well as evaporative heat transfer, is 4
conservative.
2 j'
)
l l
t j
v e w, r
-c -w r
.-w--
+
I
4
+
1920-98-2055 Page 8 of13
' * - Maintaining a 30 gph ESF leakage rate for 30 days, which is very conservative since the Building Spray and Make-Up systems will be shutdown in less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when reactor coolant system and Reactor Building pressure are reduced.
The Auxiliary Building internal volume was treated as a single well-mixed volume. This conservatively neglects the vaults and cubicles, which are areas with the greatest potential for leakage but having the least possible air exchange with the environment.
The ESF system design and operation ensures that leakage resulting from the failure of a pressure boundary component can be isolated. The Auxiliary Building DHR and BS vault drains are monitored such that significant leakage would result in a Control Room alarm and isolation, for which no credit was
- taken, For input to the dose calculation, the fraction ofiodine released from ESF leakage has been time weighted over intervals corresponding to the X/Q intervals as shown below:
Fraction Filter i
Released Efficiency (E)
Interval (F)
(E = 1 - F) 0-2 hrs 2.92E-02 0.9708000 2-8 hrs 2.15E-02 0.9785000 8-24 hrs 1.39E-02 -0.9861000 j
24-24.5 hrs 2.04E-04 0.9997960 1
24.5 hrs - 4 days 8.65E-03 0.9913500
]
4-30 days 5.59E-03 0.9944100 j
ESF LEAKAGE ASSUMPTIONS:
18.
The analysis includes the effects ofincreased Engineered Safeguards Feature (ESP) component leakage into the Auxiliary Building, without taking credit for AFHBVS carbon filters. The analysis was performed using a total assumed leakage rate of 30 gph, twice the proposed TMI-l TSCR revision of 15 gph in Section 4.5.4 pursuant to the NRC Standard Review Plan (SRP) Section 6.4 guidelines. The 15 gph ESF leakage acceptance criterion was established based on the results of system leakage inspections throughout plant history and the acceptability of the resultant MHA dose consequence.
19.
The effects of ESF leakage through boundary valves to tanks vented to atmosphere at a rate of 3 gpm were considered as a release path. This leakage can occur to the BWST and the Sodium Hydroxide Tank. This analysis conservatively assumes that all such leakage goes to the BWST, since the pH of the water remaining in the Sodium Hydroxide Tank would provide far greater retention ofiodine in the liquid phase than would occurin the BWST.
i t
.m 1920-98-2055-Page 9 of13 A
- 20. ' - ESF leakage through boundary valves to tanks vented to the atmosphere is assumed s
to be 180 EP (3 gpm) for the first five hours, then reduced to 102 gph until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1
h and then reduced to % gph for the remaining 29 days. The 3 gpm acceptance criteria:for ESF boundary valve leakage is based on the capability of leakage
- detection tests and the acceptability of the resultant dose consequence of the MHA analysis with this proposed criterion. These boundary valves are tested in accordance with, the Technical Specification 4.2.2 required inservice inspection and IST program.-
Analysis Results and Conclusion.s; The STARDOSE computer program was used to calculate the Control Room and offsite dose consequences. STARDOSE is proprietary to Polestar Applied Technology'and is fully j
consistent with applicable Code ofFederal Regulations and Regulatory Guides. The computer program was prepared under the Polestar Quality Assurance Program and is intended for use in applications covered by Appendix B to 10 CFR 50 and 10 CFR 21.
1 l
The radiological dose consequences of the revised MHA analysis are increased from those values identified in the UFSAR for both the EAB 'and LPZ; however, the resultant doses are still well below the 10 CFR 100 guideline limits; see Table 1. In addition, the 304ay-Control Room habitability dose consequences were evaluated by incorporating the additional MHA release paths and applying the calculated partitioning fraction of iodine based upon the calculations of actual sump pH history and air exchange rates as permitted J by SRP guidance, (Reference 4). The principal assumptions and methods as discussed with i
Staff remain applicable to this new Control Room evaluation. The revised Control Room thyroid doses as shown on Table 1 are slightly higher than the current NRC Standard Review Plan 6.4 guidelir.es for the control room, i.e.,42 Rem vs. 30 Rem; however, self-l contained breathing apparatus and potassium iodide are available to reduce iodine uptake j
and mitigate subsequent dose consequences from such exposure. Furthermore, in light of l
the permissible annual occupational exposure limit of 50 Rem to the thyroid specified in 10 CFR 20.1201 (a) (ii), the Control Room thyroid dose determined as a result of the revised MHA analysis is considered to be acceptable. Therefore, it is concluded that the proposed changes to the Technical Specification Section 4.5.4 and its Bases do not adversely affect nuclear safety or safe plant operations.
6 I
4 m.--
, +
y
'gd-v
!F LF.
)
1920-98-2055:
l
Page 10 of13-i l
IIL3 Technical Specification 4.5.4 Testing and Bases Changes j
I l-This TSCR also revises the testing criteria previously applied to the DHR system by
[
l.
deletion of the alternative " hydrostatic test" method, and conforms the surveillance testing l
. for all three systems to the same testing criteria. Consequently, conforming changes to the l
l Bases are required to reflect the elimination of the alternate methodology.
These i
i
' administrative changes have no impact upon nuclear safety or safe plant operation, as an i
adequate margin of safety is maintained by the use of the existing DHR surveillance testing l
method, which will be used for all three ESF Systems. The test method prescribed in the j
l revised Technical Specification 4.5.4.2 for the applicable portions of the DH, BS and MU l
_ systems ensures that the testing results account for the highest pressure within that system l
during the sump recirculation phase of a design basis accident. This change serves to l
eliminate any potential confusion that may be caused by having two test methods instead of l
one, and therefore provides greater clarity and consistency ofimplementation.
i r
l l
4 i
2 t
n 6
(
_ -(
e 2
.u.
l 1920-98-2055 Page 11of13 TABLE 1 COMPARISON OF POST ACCIDENT MHA DOSES THYROID
. : ;. 3.,g syv:y; m~~y..
fhk{j[$;.ik&h h? k&? l w
Exclusion Area Boundary 189 213 300 (2-hour consequence)
Low Population Zone -
l 8.8 163 300 (30-day consequence) l Control Room l
N/A 42.3 30*
j WHOLE BODY
, n ~ y.. :..~:,
- ya.w!?
,,...J j
n A
h).bf Af.0_.
- f..{,' _ ','
S's
,~
yy g l
Exclusion Area Boundary **
7.6 3.6 25 (2-hour consequence)
Low Populatiori Zone 0.21 l
1.37 25 (30-day consequence) l Control Room N/A l
0.53 5*
i SKIN g
g
? -
Exclusion Area Boundary N/A N/A N/A i
(2-hour consequence)
Low Population Zone N/A N/A N/A (30-day consequence)
Control Room N/A 8.05 30*
NRC Standard Review Plan, NUREG-080% 3ection 6.4 The existing UFSAR analysis conservatively used whole body dose conversion factors that included both beta and gamma radiation. The revised analysis more appropriately uses whole body dose conversion factors from ICRP-30 that are based on only gamma radiation.
i
g a,
b:
- Encl 5sure 1 I
I
- 1920-98-2055 I
g
- Page 12 of13 -
p9 IV.'
NO SIGNIFICANT HAZARDS CONSIDERATION !
L
~
~
GPUN has determined thit this TSCR poses no significant hazards consideration as dermed f
. by 10 CFR 50.92.
i:
1.
Operation of the facility in accordance with the proposed l amendment would not involve a significant increase in the probability or consequences of an accident L
.previously evaluated. No physical. modifications wh:ch would change stmetures, systems, or components are proposed by this TSCR for surveillance changes in Technical Specification 4.5.4 and its Bases. The proposed increase in the ESF-Systems leakage rate acceptance limit has no affect on the performance of ESF
. systems during a DBA. The proposed changes are supported by a revised MHA I
dose calculation using updated X/Q values and calculation assumptions. The MHA dose consequence analysis yields dose results that are below the 10 CFR 100-guidelines for both the EAB and LPZ. The calculated Control Room Habitability Evaluation does not exceed the permissible annual occupational exposure limit of 50 Rem to the thyroid as specified in 10 CFR 20.1201 (a)(ii). In addition, the potential thyroid exposure can be mitigated by the availability of self-contained breathing apparatus and potassium iodide. Therefore, the changes would not involve a significant increase in the consequences of accidents previously evaluated.
2.
Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any p,reviously L
evaluated. This TSCR does not involve any physical modifications that would affect structures, systems, or components, nor does it involve any changes in plant operation.
3.
Operation of the facility in accordance with the proposed amendment would not i
L involve a significant reduction in a margin of safety. This TSCR does not involve i
changes to Technical Specification defined Safety Limits, Limiting Conditions for
(
Operation, and does not involve any change to safety system setpoints for operation.
Therefore, the proposed changes do not involve a significant reduction in a margin i
ofsafety.
l
- V; ENVIRONMENTAL CONSIDERATION GPUN has determined that the changes to the TMI-1 Technical Specification Section 3.15.3 Bases and Section 4.5.4 and its Bases involve no significant chuge in the amount or type of l
any effluent that-may be released off-site, and that there is no significant increase in
{-
individual or cumulative occupational radiation exposure.
1 VI.
lMPLEMENTATION It is requested that the license amendment authorizing this change become effective upon issuance. Procedure changes associated with implementation of this license amendment will i:
be completed within 120 days ofreceipt of NRC's approval of this change.
i-y y r
v,w
+-
n g
a
-a w-
i f
' Ev.clo*ure h 192 M 8-2055 Page 13 of13 VII. REF(TENCES:
l.
GPUN Inc. Letter No. 6710-97-2252, dated July 30,1997 - TSCR No. 266, Accident Recirculation Leakage Limits 2.
GPUN Inc Letter No. 6710-97-2425-1, dated September 19,1997 - TSCR No. 266,
~ Supplement 1 3.
- GPUN Inc. Letter No. 1920-98-20145, dated March 24,1998 " Control Room Habitability Evaluation" 4.
GPUN Inc. LetterNo. 1920-98-20561, dated October 6,1998 - POLESTAR Proprietary Report, Calculation PS AT 05656A.04, " Calculation ofTMI-l Engineered Safety Feature Component Leakage lodine Release" 5.
GPUN Inc. Letter No. 1920-98-20526, dated October 15,1998 - LAR No. 276,
. Revised Accident. Analysis Atmospheric Dispersion Factors i
t; t
METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY AND PENNSYLVANIAELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION, UNIT 1 Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request (TSCR) No. 274 COMMONWEALTH OF PENNSYLVANIA
)
) SS:
COUNTY OF DAUPHIN
)
This TSCR is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station, Unit 1. As a part of this request, proposed replacement pages for the Appendix A Technical Specifications are also included. All statements contained in this submittal have been reviewed, and all such statements made and matters set forth therein are tme and correct to the best of my knowledge.
GPU NUCLEARINC.
BY:
d>>f4 Vic# resident and Direp, TMI Swom and subscri before me thisd/ dayof el4,,2 w.1999.
/
G V N o p blic
,I Notarial Seal Suranne C. Maklosik. Notary Public(\\
MC miss o xhi es ov 2 1999 Member,Pennsylvanta Associationof Notanes