ML20012B147
| ML20012B147 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 02/26/1990 |
| From: | Mroczka E NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20012B148 | List: |
| References | |
| B13429, NUDOCS 9003130668 | |
| Download: ML20012B147 (35) | |
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(203) 665-5000 February,26, 1990 Docket No. 50-423 B13429 Re:
.U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Gentlemen:
Mill. stone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications Containment Pressure Introduction The Millstone Unit No. 3 containment structure is a cylindrical, painted carbon steel-lined, reinforced concrete structure which encloses the compo-nents :and major piping within the reactor coolant pressure-boundary.
The
. Millstone Unit No. 3 containment is a subatmospheric-type containment
_-In the dual containment plant concept for Millstone Unit No. 3, the-secondary con-tainment is comprised of the containment enclosure building and the associated supplementary leak collection and: release system (SLCRS)'provided to mitigate the. radiological consequences-of postulated accidents.
The containment is maintained at ' a subatmospheric pressure (typically 9.0 to 12.0 psia) during Modes 1 through 4 to limit the peak pressure-attained during a postulf.ed accident and to minimize radioactive. releases after an accident.- Northeast
-Nuclear Energy Company (NNECO) proposes to modify this operating pressure m
. inside containment. - Specifically, the pressure would remain at subatmospheric levels, but would be raised to slightly under atmospheric pressure (14.0 psia) during Modes 1 through 4.
The main purpose of - the proposed containment
. pressure increase is to reduce ~.the potential for ' personnel injury when entering containment due to crossing.the pressure boundary and due to oxygen
? deficiency while facilitating more timely access to containment for minor
. evolutions without requiring the plant to be in a cold shutdown.
It is noted
_Ra that-' containment entries are not performed on a routine basis.
- However, N8' containment entries are required to inspect for unidentified-reactor coolant oec system
- leakage, boron precipitation investigations (as per Generic 88 Letter 88-05), and plant start-up surveillances/ inspections.
The proposed increase in the containment pressure is based upon a new containment analysis So performed by Stone and Webster at the direction of NNECO. The new containment (o r 18 analysis demonstrates that it is safe to operate Millstone Unit No. 3 at a g@
containment pressure of 14.2 psia with 75'F service water temperature.
.However, operation at or below 14.0 psia is necessary to account for
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U.S. Nuclear Regulatory Commission B13429/Page 2 February 26,-1990 Backaround Millstone Unit No. 3 Technical Specifications require that the containment pressure be maintained greater than 8.9. psia (air partial pressure),- but less than atmospheric ~ (above the service water limit line of Figure 3.6-1 in the -
Technical Specification).
Experience has shown that the risk of injury to
)lant employees performing physical labor in subatmospheric containments is ligh. Comparatively, very little is known about health effects and stress on people in high-temperature, low-pressure environments. Many studies have been carried out on low-temperature, low-pressure environments such as those experienced by pilots, and the studies have been done on low-pressure environ-
-i ments in which people have had several days to acclimate themselves to the environment (e.g., mountain climbers).
Subatmospheric containment entries are different from the above examples in that personnel entering the containment are only allowed 10 to 20 minutes for the pressure equalization versus several hours to days for other low pressure conditions.
In addition, personnel are L
required to wear self-contained respirators, Rexnord " Bio-Packs" to supply J
supplemental oxygen.
Bio-Packs weigh approximately 40 pounds and reheat the l
breathing air as a result of the respirator design.
Also, the containment environment is warmer than most of the environments for which research has been done.
The average containment temperature is about 80*F and may have local hot spots in excess of 100*F.
The combination of low pressure and high temperatures has great potential for personnel injury during containment entries.
At Millstone Unit No. 3, 38~ personnel medical incidents have occurred due to containment entries at subatmospheric conditions during the past four years since the plant received its operating license.
In addition, the use of the " Bio-Pack" also causes the personnel working in containment to become less efficient.
Hence, radiation exposure to operators is a potential concern.
l In an effort to reduce the personnel injury risk associated with entering the L
containment during Modes 1 through 4, NNECO is requesting the NRC to review and approve a change in the Millstone Unit No. 3 containment operating pres-l-
sure to allow NNECO to operate the plant at power (during Modes 1 through 4) with a containment pressure of not less than 10.6 psia and not greater than 14.0 psia.
However, NNECO intends to maintain the Millstone Unit No. 3 containment pressure between 13.75 psia and 14.00 psia-during Modes 1 through 4.
Maintaining the containment pressure between.13.75 psia and 14.0 psia would allow plant personnel to enter the containment with a minimal pressure change and would not require the use of supplemental oxygen (" Bio-Packs").
The benefits of this change relate primarily to the reduced risk of injury to personnel from the reduced pressure boundary transition and eliminating the need to carry heavy (40-pound), awkward supplemental oxygen units in this hot environment. All injuries that have occurred to personnel in this environment have been related to these factors.
The personnel risk factors of the subatmospheric containment and benefits of maintaining containment pressure at or above 13.75 psia are shown below:
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- U.S. Nuclear Regulatory Commission B13429/Page. February 26, 1990 Factor Benefits Elimination of Supplemental
- 1. Reduced fatigue.
Oxygen Requirement
- 2. Reduced risk of heat stress.
- 3. Increased mobility / agility.
- 4. Increased visibility.
- 5. Enhanced communications (within a confined J
space).
- 6. Provides engineering control as a solution 1
. to a personnel safety concern.
- 7. Reduction of time required to perform tasks resulting in reduced ionizing radiation
~
exposure (ALARA).
- 8. Elimination of Immediately Dangerous. to 1
Life and Health (IDLH) atmosphere (oxygen deficient).
Reduction of Pressure Reduced risk of ear injury due to a decreased Differential (increased.
pressure differential during containment Containment' Pressure) entry.
The benefits to the.public include:-
L' A..
Reduced time for containment drawdowns which give (1) fewer airborne E
. effluents and (2) lower probability of-a LOCA with open containment air Lejector valves that subsequently fail to close.
l
'B.
More timely and more frequent containment entries to address potential L
safety issues such as RCS leaks.
- Both experience and risk evaluation have shown that working within a subatmo-spheric containment carries a high risk of injury to plant personnel.
Good L
practice requires _ the use of engineering controls to reduce or eliminate hazards to employees where possible.
This practice is consistent with the L
Occupational Safety and Health Administration's (OSHA)' standards.
Increasing the Millstone Unit No. 3 containment pressure yields a significant risk reduction to the plant personnel while maintaining safe operation within the acceptable limits of the accident analysis.
The pressure selected for analy-sis provides an operating band that permits the above benefits yet maintains the important advantage of a subatmospheric containment, of ensuring prompt detection of breaches in containment integrity.
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F U.S. Nuclear Regulatory Commission B13429/Page 4
-February 26, 1990 Discussion
' A.
Existina Desian Basis The existing containment design is subatmospheric.
The design basis for the various loss-of coolant accidents (LOCA) inside containment was a subatmospheric containment operating pressure of 9.8 psia at a service water temperature of 75'F during Modes 1 through 4.
The actual contain-ment operating pressure has caused personnel injuries to those who enter containment at subatmospheric conditions during power operation.
The current containment leak rate (L ) is 0.9 weight percent per day, and the secondary bypass leakage from co$tainment is 0.01 L,.
The original design basis for the containment and all the LOCA analyses were completed by Stone and Webster using the LOCTIC computer program.
The major design basis accidents- (DBA) include containment pressure and temperature response following a LOCA, subcompartment pressurization following a LOCA, and combustible gas concentration following a LOCA.
Also, the containment pressure and temperature response following a main steam line break (MSLB) and control rod ejection accident (CREA) were analyzed for the original design basis.
The original design basis and associated radiological consequences were analyzed for the LOCA and include the exclusion area bourdary (EAB), low population zone (LPZ), Millstone Unit No. 3 control room, Millstone Unit No,' 2 control room, and technical support center (TSC).
The CREA radio-logical consequences were analyzed for the EAB and LPZ.
Dose calcula-tions for the emergency operations facility and the Millstone Unit No.1
-control room are bounded by the Millstone Unit No. 1 LOCA, even with the proposed changes and hence were not quantitatively analyzed, it is noted that original-radiological consequence analysis based on the subatmo-spheric design shows all containment leakage terminating within I hour.
Also, with the original design, there is no credit taken for containment spray iodine removal.
The original containment pressure analysis was previously revised and is presented in Amendment 17 to the Final Safety Analysis Report (FSAR).
Tne calculated peak containment pressure decreased from 39.4 psig to 36.1 psig.
NRC approval of this analysis is documented in Supplement 5 to the Millstone Unit No. 3 Safety Evaluation Report, NUREG-1031.
It is noted that the current Technical Specification does not reflect the containment peak pressure of 36.1 psig.
It was decided not to revise the Technical Specifications as the original analyzed value of 39.4 psig bounded the revised value of 36.1 psig.
a 79 U.S. ' Nuclear-Regulatory Commission B13429/Page 5
-February 26,.1990L B.
ProDosed Desian Chance
.The proposed design change will allow NNECO to operate Millstone Unit No. 3 with a maximum containment pressure of 14.0 psia during Modes 1 through 4.
The safety analysis for the proposed change was performed at.
a maximum allowable containment pressure. of 14.2 psia.
The proposed day to 0.'65 weight percent per day and an increase.9 weight percent per change also incorporates a reduction in L.from 0 in secondary contain-ment bypass leakage from the containment from 0.01 L to 0.042 L Pressure transmitters 3LMS*PT43A and 3LMS*PT43B will be 8 sed to monit8r.-
-pressure in containment.
Safety Assessment
~ NNECO has reviewed the proposed design change pursuant to 10CFR50.59, which accompanied the proposed license amendment, to assess the impact on - the accidents evaluated as the design. basis, the potential for creation of a.new unanalyzed event, and the impact on the margin of safety.
NNEC0 has deter-mined-that the proposed design change constitutes an unreviewed safety ques-tion (USQ) due to an increase in postaccident radiological consequences, but has determined the proposed design change to be acceptable and - safe.
The proposed design change also involves a change in the Millstone Unit No. 3 Technical - Specifications.
Therefore, pursuant to - 10CFR50.90, NNEC0 hereby
. proposes-to amend its operating license, NPF-49, by incorporating the attached changes into the Technical Specifications of Millstone Unit No. 3.
Along with this amendment request, supporting documentation is provided as follows:
o provides the safety evaluation for the design changes, o forwards the revised pages of the Technical Specifications.
o provides a description of the proposed Technical Specifica-tion. changes.
NNEC0 has reviewed the proposed Technical-Specification changes in accordance with 10CFR50.92 and has determined that the changes do not involve a significant hazards consideration.
The basis for this determination is discussed in Attachment 3.
Moreover, the Commission has provided guidance concerning the application of:
the standards in 10CFR50.92 by providing certain examples (March 6,
- 1986, 51FR7751) of amendments that are considered not likely to involve a signifi-cant hazards consideration.
Example (vi) provides that a significant hazards consideration finding is unlikely for:
A change which either may result in some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan; for example, a
s U.S. Nuclear Regulatory Commission B13429/Page 6 February 26, 1990 change resulting from a small refinement of a previously used calculational model or design method.
This example appears applicable to the proposed changes. The proposed changes result in an increase in the ceasequences of a LOCA and a CREA accident.
For the dose to the public (see Attachment 1), the following values were calcu-lated increases and the percentage that increase represented in relation to-allowable (Standard Review Plan) limits:
Receptor Part of Increase Percent Accident MC1110D of Body in Dose of Limit LOCA EAB Whole Body 2.700 rem 11.00%
LOCA LPZ Thyroid 15.500 rem 5.00%
LOCA LPZ Whole Body 2.000 rem 8.00%
CREA N Loop LPZ Whole Body 0.018 rem 0.20%
CREA N-1 Loop LPZ Whole Body 0.018 rem 0.30%
CREA N-Loop LPZ Thyroid 0.040 rem 0.05%
CREA N-1 Loop LPZ Thyroid 0.030 rem 0.04%
Based on the above, NNECO has concluded that the increases are a small per-centage of allowable limits and hence are not a significant increase in consequences of a previously analyzed accident.
All doses remain within the acceptable limits. As seen from Table 2 of Attachment 1, for the control room and TSC personnel, the percentage increase in doses is higher than for the public, particularly for the beta dose (Factors of 3 and 8 for Millstone Unit No. 2 control room and Millstone Unit No. 3 control room, respectively).
However, all doses remain within General Design Criterion 19 limits of 30 rem to the thyroid and skin and 5 rem to the whole body, Standard Review Plan Section 6.4 and the Millstone Unit No. 3 FSAR Section 3.1.2.19.
The overall conclusion, therefore, is that the control ;ooms and TSC will remain habitable with acceptably low risk to plant personnel.
The results of the proposed changes are " clearly within all acceptable criteria... specified in the Standard Review Plan."
Therefore, the proposed license amendment does not involve a significant hazards consideration.
Based upon the information contained in this submittal and the environmental assessment for Millstone Unit No. 3, there are no unacceptable radiological or nonradiological impacts associated with the proposed change and the proposed license amendment will not have a significant effect on the quality of the human environment.
The Millstone Unit No. 3 Nuclear Review Buard and the Millstone Station Site Operations Review Committee has reviewed and approved the proposed license amendment and has concurred with the above determination.
Regarding our schedule for this proposed amendment, NNECO intends to fully implement the license amendment within 30 days of its issuance by the NRC as follows.
The existing limit of lower air partial pressure of 8.9 psia is
e.
U.S. Nud4ar Regulatory Commission B13429/Page 7 February 26, 1990 requested to remain in force for approximately three weeks to permit a gradual rise to 10.6 psia total pressure (the new limit of lower pressure).
The instrument air usage in containment results in a gradual increase in the pressure.
The Millstone Unit No. 3 FSAR will be revised within six months once the Staff issues the license amendment.
This revised FSAR will reflect the results of the analyses performed for the proposed design changes and the Technical Specification changes.
In accordance with 10CFR50.91(b) we are providing the State of Connecticut with a copy of this proposed amendment.
l Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY I
E JYMrocfka //
Ser(for Vice President j
i cc, Mr. Kevin McCarthy, Director i
Radiation Control Unit Department of Environmental Protection Hartford, CT 06116 W. T. Russell, Region I Administrator j
D. H. Jaffe, NRC Project Manager, Millstone Unit No. 3 W. J. Raymond, Senior Resident Inspector, Millstone Unit Nos. 1, 2, and 3 STATE OF CONNECTICVT) 1 ss. Berlin COUNTY OF HARTFORD Then personally appeared before me, E. J. Mroczka, who being duly sworn, did state that he is Senior Vice President of Northeast Nuclear Energy Company, a Licensee herein, that he is authorized to execute and file the foregoing information in the name and on behalf of the Licensee herein, and that the statements contained in~said information are true and correct to the best of his knowledge and belief.
IAbsh A
GLV Rotary Pu 1 4 Myremmla.on Eg!r :thrch3f,1993
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Docket No. 50 423 B13429 1
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1 Attachment I i
Millstone Nuclear Power Station, Unit No. 3 Safety Evaluation for the Proposed Design Changes--Containment Pressure i
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February 1990
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.l U.S. Nuclear Regulatory Commission B13429/ Attachment 1/Page 1 February 26, 1990 1
Attachment I Millstone Nuclear Power Station, Unit No. 3 i
Safety Evaluation for the Proposed Deslan Chanaes--Containment Pressure l
1 Although an explicit discussion addressing the criteria of 10CFR50.59 is not required to process the proposed license amendment, we have included the i
safety evaluation on the proposed design change to assist the NRC Staff in its review of this matter.
Introduction The existing containment design is subatmospheric.
The current maximum containment operating pressure at a service water temperature of 75'F during Modes 1 through 4 is 9.8 psia.
The proposed design change will modify the I
maximum allowable containment operating pressure during Modes 1 though 4 to 14.0 psia.
The analysis for the proposed change was performed at a maximum allowable containment pressure of 14.2 psia.
In addition, the proposed design i
change also incorporates a reduction in containment leak rate (L from 0.9 weight percent per day to 0.65 weight percent per day and an inch)ase in secondary containment bypass leakage from the containment from 0.01 L" to 0.042 L.
NNECO has reviewed the proposed design changes pursuani to a
l 100FR50 59 to assess the effect on the accidents evaluated as the design i
basis, the potential for creation of a new unanalyzed event, and the impact on the margin of safety.
I.
Effect on the Accidents Evaluated as the Desian Basis A. Accidents Affected l
The following accidents and associated consequences are potentially affected by the proposed design change.
- 1. Loss-of-Coolant Accident (LOCA)
- a. Containment pressure and temperature response
- b. Subcompartment pressurization l
- c. Combustible gas concentration L
- 2. Main Steam Line Break (MSLB) l Containment pressure and temperature response
- 3. Control Rod Eiection Accident (CREA) l.
E 3
U.S. Nuclear Regulatory Commission Bl:429/ Attachment 1/Page 2 Feoruary 26, 1990 B. Safety Systems / Structures Affected The design bases of the following systems / structure are potentially affected:
- 1. Containment Structure The current design bases include the following:
- a. The peak calculated containment pressure following a LOCA (36.1 psig) is below the containment design pressure of 45 psig.
1
- b. The containment pressure is reduced to less than atmospheric pressure in less than 60 minutes following a LOCA.
- c. After depressurization, the containment atmospheric pressure is maintained below atmospheric pressure.
1 With the proposed change, the second and third bases will be eliminated. These bases are not directly related to the integrity of the containment structure, but were imposed to establish the basis for fission product leakage from the containment.
These two bases permit the assumption of no fission product leakage from the containment after the LOCA.
Therefore, a new fission product leakage scenario must be considered in the radiological assessment.
This assessment is provided in the subsequent paragraphs of this attachment.
- 2. Quench Sorav System Since the current radiological analyses do not take credit for fission product removal by the quench spray system, it is not a design basis for the system.
This was originally assumed to simplify - the dose calculations, and was acceptable because the resultant doses were acceptable.
The radiological evaluation for the proposed change does, however, include the effect of fission product removal by the quench spray system.
The effectiveness of the system for fission product removal is based on the as-designed system.
Because the quench spray system was designed to provide adequate containment heat removal, and since cost system parameters that maximize heat removal also maximize fission product removal, the system is effective for fission product removal.
(Performance of the quench spray system as a fission product removal system is discussed further later in this attachment.)
The proposed change will therefore result in the addition of fission product removal as a design basis for the quench spray system.
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't U.S. Nuclear Regulatory Commission B13429/ Attachment 1/Page 3 t
February 26, 1990 I
- 3. Containment Recirculation System As with the quench spray system, the effect of the containment recirculation system on fission product removal was not included in the current radiological evaluation, but will be included in the evaluation for the proposed change.
The proposed change will, therefore, also result 9n the addition of fission product removal as a design basis for the containment recirculation system.
- 4. (.pateinment Air Recirculation (CAR) System The - containment air recirculation (CAR) system is designated as safety-related.
However, the system is not designed to operate post LOCA and is automatically shut down by a containment depres-surization actuation (CDA) signal (Final Safety Analysis Report (FSAR), Section 9.4.7.2).
Therefore, the CAR system and changes related to this design change have no effect on the design basis accident (DBA).
- 5. Containment Vacuum System The containment vacuum system reduces the containment pressure from atmospheric to subatmospheric conditions.
This is accomplished prior to plant start-up using a vacuum ejector.
Subsequent to initial depressurization, one of two containment vacuum pumps removes air from the containment atmosphere to maintain subatmo-s)heric conditions.
At the higher operating pressures pro)osed, tie pump will remove a denser air mixture resulting in a ligher mass flow rate.
The proposed change in containment maximum operating pressure from 9.5 psia (at 75'F) to 14.0 psia will result in less frequent operation of the vacuum pump in order to maintain the new subatmo-spheric pressure.
The new operating conditions are within the existing design conditions and the normal operating band for the
- pumps, The pumps also remove air from containment after a DBA.
Since the i.
system is not used until several weeks after a DBA, the system is not an engineered safety feature, is not safety related (reference FSAR Section 9.5.10.2), and does not impact the consequences of a DBA.
- 6. Containment Pressure Monitors The current method of setting and maintaining containment pressure is utilizing 3LMS*PT934, 935, 936, and 937.
These transmitters have a range of 0 to 60 psia and are used to perform containment isolation.
It was determined that based on the range of these l
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s U.S. Nuclear Regulatory Commission B13429/ Attachment 1/Page 4 February 26, 1990 transmitters, the instrument error was too large for pint person-nel to use these transmitters for maintaining the containment pressure within the desired range yet provide assurance that the pressure was below the proposed Technical Specifiction limit.
Currently, there are two narrow range transmitters, 3LMS*PT43A and B (8.5 to 13.5 psia), that provide indication on the main control board.
These transmitters will be changed to 8.5 to 14.5 psia which will permit control board indication in the new range of operation.
The computer readouts will be utilized to set and maintain containment pressure within the Technical Specification requirements.
C. Imonet of Chance on Performance of Safety Systems l
The proposed change has no effect on the performance of the quench spray and containment recirculation system for heat removal.
However, the proposed change necessitates adding fission product i
removal as a design basis for the quench spray and containment recir-I culation systems.
The performanch of these systems in this function t-was evaluated by determining iodine removal coefficients and decontam-ination factors.
The current design requirements assumed for heat removal were used along with methodology of ANSI /ANS 56.5 1979, l.
"American National Standard for PWR and BWR Containment Spray System l'
Design Criteria," to calculate these factors.
D. Imoact of Chance on Consecuences of Accidents
- 1. LOCA Containment Pressure and Temoerature Response l
The containment pressure / temperature response was reevaluated with the proposed maximum operating pressure of 14.2 psia.
This evalua-tion was performed using the same methods and computer models l'
described in Section 6.2.1 of the FSAR.
Two LOCA cases were l
reevaluated:
the hot leg double-ended rupture (DER) and the pump L
suction DER with failure of one Engineered Safety features (ESF) train. The hot leg DER is the limiting accident for peak contain-ment pressure, with 36.09 psig being the current maximum calculated l
pressure.
This was recalculated to be 38.57 psig with the increased initial containment pressure.
The pump suction DER with failure of one ESF train is the limiting accident for the long-term containment pressure, with current analysis calculating a depressurization time of 2,480 seconds and a margin below atmospheric pressure of 0.05 psig at 15,900 seconds.
This pressure transient was recalculated with the increased initial I
containment pressure.
The containment pressure remains above atmospheric pressure for the duration of the analysis, as shown in l
Table 1.
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l' U.S. Nuclear Regulatory Commission B13429/ Attachment 1/Page 5 February 26, 1990 1
l The revised containment pressure and temperature response for both the hot leg and pump suction DERs are shown in Figures 1 and 2, 1
respectively.
Figure 3 shows the long-term pressure transient i
following the pump suction DER.
i l
The containment )ost-LOCA temperature / pressure response is an input parameter for ot1er plant equipment and analyses, and the effect of the results of the reevaluation are as follows:
- a. Containment Desian Pressure The revised maximum containment pressure of 38.57 psig repre-
)
sents a 2.48 psig increase over the current maximum
)ressure.
This pressure is still well below the containment des < gn pres-sure of 45 psig.
The proposed change, therefore, has a minor effect on the maximum calculated containment pressure.
- b. Containment Leakaae Scenario l
The proposed change alters the scenario for fission product leakage to be used in the radiological evaluation.
The current radiological evaluation assumes leakage of fission products from the containment ends at I hour after the LOCA, after which time the containment pressure is subatmospheric.
The revised con-tainment pressure calculation results in the containment pres-i sure remaining above atmospheric pressure, indicating that continued leakage must be assumed.
The effect of this is discussed further later in this attachment.
- c. Emeroency Core Coolina System (ECCS) Performance The containment post-LOCA pressure response is an input to the evaluation of ECCS performance; but, because a lower initial containment pressure is limiting, the effect of an increased j
containment pressure is beneficial to ECCS performance.
The l
proposed change, therefore, has no negative effect on the performance of the ECCS.
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- d. Intearated leak Rate Test fiLRT)
The proposed increase in normal containment operating pressure (to 14.2 psia) yielded revised values for containment peak pressure (P ), L, and the bypass leakage limit fraction.
The revised valdes ah:
U.S. Nuclear Regulatory Connission B13429/ Attachment 1/Page 6 February 26, 1990 Containment Parameter Enaineerina Units Revised Value P,
(psig) 38.57 L
(wt%)/ day 0.65 L}a (SCFH) 2206.33 B pass Limit Fraction (wt%)/ day 0.042 L,
)
of L, i
The proposed change in containment operating pressure reduces the magnitude of the ILRT test pressure and acceptance criteria.
This is also true for the Type B and C testing as well.
The bypass leakage limit was increased from 0.01 L to 0 This increased bypass fraction was considered nelessary.042 L,becaulej of the difficulty in meeting such a low value for LLRT during i
the last refueling outage.
Failure to meet such a restricted low limit could result in additional unwarranted occupational exposure to repair valves with marginal leakage, i
Type A, B, and C testing acceptance criteria are all reduced by j
the same factor:
27.78 percent from their current values.
In terms of SCFH, the change in the acceptance criterion is
~1
-30.47 percent.
For the bypass leakage limit, it was increased j
by approximately +192.05 percent.
The net effect of this change will be the imposition of even tighter containment leakage limits (excluding the bypass leakage limit) upon the Type A, B, and C containment boundary systems and their components.
Containment integrity will be improved and provide additional assurance that the public will be pro-tected.
The bypass leak limit increase also meets the above require-ments.
The proposed changes in P and L will also require administrative revision of the Typ8 B and t leak rate testing i
procedures.
- e. Electrical Eauioment Oualification (EED) Service Conditions i
(1) The service conditions for qualifying equipment required to function in the postaccident containment environment are given in FSAR Appendix 3R.
The proposed change affects the following Millstone Unit No. 3 environmental qualification service condition parameters.
l-(a) Normal pressure and temperature inside containment.
(b) Accident pressure and temperature profiles inside
[
containment.
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V U.S. Nuclear Regulatory Commission B13429/ Attachment 1/Page 7 February 26, 1990 (c) Postaccident pressure and temperature profiles inside containment.
(d) Certain postaccident radiation total integrated dose (TIDs) outside containment, j
(2) The following discussion covers the impact on the qualifica-tion of equipment resulting from changes to normal enviren-1 ment and the impact of the change on the accident environ-j mental parameters and the post DBA operation parameters.
(a) Normal Environment
- 1) Pressure and Temoerature j
The present EEQ program is based on a normal contain-j ment pressure range of 9.5 to 14.7 psia.
The pro-1 posed new operating pressure of 14.2 psia f alls within this range and will therefore not impact the existing normal pressure qualification of EEQ equip-1
- ment, j
The normal maximum average (NMA) temperature for each j
environmental zone is used to calculate the qualified life of EEQ equipment in that zone.
NNECO has determined that the effect of the proposed change in operating 3ressure on these NMA values for contain-2 ment will
>e negligible.
The net effect is expected to be a slight lowering of average ambient tempera-tures which is conservat.ive for equipment qualifica-tion purposes.
Therefore, the proposed change will not adversely impact the existing normal temperature i
qualification of EEQ equipment.
- 2) Other The proposed change will not alter any of the remain-ing normal environment parameters.
(b) Accident /Postaccident Environment
)
- 1) Pressure and Temperature The postulated DBA accident and post-DBA pressure and temperature response envelopes for all containment i
zones are profiled in the
- FSAR, Appendix 3B--, Appendix A, page 6.
As permitted in NUREG-0588, these qualification parameters are based on the time dependent pressure and temperature values
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U.S. Nuclear Regulatory Commission B13429/ Attachment 1/Page 8 February 26, 1990 for the design of the containment structure rather than on any ::pecific or composite accident calcula-tions.- Except for the long-term post-DBA pressure j
value, the calculated new pressure and temperature values are bounded by the' existing envelopes.
Therefore, the only potential impact on EEQ equipment will be in the post DBA region between I hour and t
one year.
I The pressure parameter is not considered an aging parameter which causes degradation of materials.
Equipment is type tested for the pressure parameter only for the period from initiation of the accident i
until the pressure service condition returns to essentially the same level that existed before the postulated event.
Acceptable thermal and radiation aging results are then presumed to demonstrate the preservation of the material properties required to maintain pressure withstand capability during the i
post-DBA period.
The calculated new post-DBA con-tainment pressure value of 1.75 psig is, for equip-ment qualification purposes, essentially the same as the preaccident level.
Therefore, the proposed increase in the post DBA containment pressure value will have no adverse impact on the qualification of EEQ equipment.
There will be no change in the pressure or tempera-ture accident environment parameters for zones other l
than those located in containment.
- 2) Radiation For equipment qualification, the accident design bases for the zone radiation environments (FSAR Section 3.118.5.1) are different for inside and outside containment locations.
The inside-containment radiation values are based on the direct consequences of a LOCA (reference FSAR Section 15.6) and will not change since the original distribution l
of fission product inventory (based on source terms per Regulatory Guides 1.4 and 1.89, and NUREG 0588) remains unchanged.
For outside-containment zones, the accident radiation values are typically based on radiation emanating from the containment structure plus exposure from recirculating emergency fluid I
systems as applicable (Reference FSAR para-graph 3.118.5.1).
However, the accident radiation l
environment parameters for Zones AB-16 and AB-39 are
V 2
U.S. Nuclear Regulatory Commission B13429/ Attachment 1/Page 9 February 26, 1990 i
calculated based on filter loads f rom operation of the supplementary leak collection and release system i
(SLCRS) and the auxiliary building ventilation system (ABVS), respectively.
Both of these systems are classified as ESF filter systems required to operate postulated (reference FSAR Section 6.5), and since the post-LOCA i
post-DBA fission product leakage levels are expected to increase with this change, the accident radiation parameter values have been reeval-usted.
The original calculated accident radiation values
)
used for Zones AB-16 and AB 39 were based on assump-tions that included containment never going subatmo-spheric post-LOCA and containment sprays not used.
This calculation also noted that the amount of containment LOCA leakage filtered by the ABVS would be small, and that all containment leakage is there-fore assumed to be filtered by the SLCRS (ZoneAB-16).
The DBA which yields the worst-case radiation environmental conditions for the ABVS filters (Zone AB-39) is an ECCS leak with all leakage being released through these filters alone.
A revision to this calculation was performed arior to
)
initial plant start up to incorporate the effect of the containment pressure going subatmospheric 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> post-LOCA and to add potential ECCS leakage in the ESF building to the SLCRS filter load.
The ABVS filter load was unchanged in this revision.
The net SLCRS filter load decreased for the containment subatmospheric case (with no credit taken for con-tainment spray) and is the accident radiation value presently used for Zone AB-16 in the equipment qualification program.
The accident radiation value for Zone AB-39, based on the ABVS filter load from I
the bounding ECCS leak, remained unchanged.
l The newly calculated containment spray reduction factor ccmbined with a postaccident positive contain-ment pressure condition yields a new accident radia-tion value for the SLCRS filters to be used for equipment qualification in Zone AB-16. This calcula-tion is based on a LOCA and considers the increased containment leakage while crediting the effects of the quench spray system.
Although the new accident radiation value is an increase over the present value, the existing equipment qualifications remain valid with adequate margin.
The accident radiation
i i
U.S. Naclear' Regulatory Commission B13429/ Attachment 1/Page 10 February 26, 1990 L'
value for Zone AB-39 remains unchanged.
Therefore, the proposed change will not adversely impact the j
existing accident radiation qualification of EEQ equipment.
- 3) Other L
E The proposed change will not alter any of the remain-ing accident environment parameters.
- 2. Subcomoartment Pressurization In general, the use of a minimum initial pressure in the subcom-1 partment analysis is conservative. Sensitivity studies reported in NUREG/CR-1199 have shown that peak AP is usually higher for a lower
)
initial pressure of the subcompartment.
The studies have also i
shown that a larger initial air mass in the subcompartment causes i
an increase in the predicted maximum AP.
Increasing the air compartment initial pressure has the effect of increasing the air i
mass.
However, the influence of the initial pressure is greater than that of the initial air content. Therefore, the net effect of i
increasing the initial pressure will be a reduction of the pre-l dicted maximum AP.
.l
- 3. Combustible Gas Concentration l
1 An increase in the containment operating pressure causes an increase in the mass of air in containment.
Because the rate of L
generation of hydrogen is unchanged, the concentration of hydrogen l
would be lower.
The proposed change, therefore, has no effect on
)
l the current evaluation of hydrogen generation and control.
The i
current hydrogen concentration limit in containment is 4 percent.
l
- 4. MSLB Pressure and Temperature RescoDie A review of the containment MSLB analysis (summary of results given in FSAR Table 6.2-22) indicates the maximum containment operating 1
pressure cases are governing for the containment pressure response and the minimum containment operating pressure cases are governing for the containment temperature response.
The proposed change, therefore, has no effect on the containment temperature response.
l-The containment pressure response, however, will be increased by l
the proposed increase in operating pressure.
The current maximum calculated pressure is 31.49 psig for a full DER at hot standby (zero power), which is based on a containment i,
operating pressure of 12.342 psia (10.65 psia air partial pres-L sure).
An increase in the operating pressure to 14.2 psia l-(13.35 psia air partial pressure) will increase the resultant peak L-
)
4*
i
\\
l U.S. Nuclear Regulatory Commission j
B13429/ Attachment 1/Page 11 February 26, 1990 pressure following a main steam line break on the order of 3 psig, which is still well below the peak pressure following a LOCA.
The proposed change, therefore, has a minor effect on the containment l
pressure response to an MSLB.
I
- 5. Radioloaical Consecuences
- a. The following accidents and associated consequences are affected
)
by the above changes (1) LQCA 1
. Exclusion area boundary (EAB)--thyroid and whole body doses Low-population zone (LPZ)--thyroid and whole body doses Millstone Unit No. 3 control room--thyroid, whole body, and
=
skin doses 1
Millstone Unit No. 2 control room--thyroid, whole body, and skin doses 1
Technical support center (TSC)--thyroid, whole body, and I
skin doses (2) Control Rod E.iection Accident (CREA) l i
EAB--thyroid and whole body doses LPZ--thyroid and whole body doses The Millstone Unit No. 3 LOCA is not the limiting accident for the Millstone Unit No. I control room habitability calculations and it was qualitatively determined that the above changes do not change this conclusion. The LOCA doses
~
are limiting when compared to the control rod ejection accident (CREA) for control room and TSC habitability calculations and the above changes do not affect this conclusion.
Based on Standard Review Plan acceptance criteria and available models, skin dose is not calculated for the public, but was calculated for General Design Criterion 19 eompliance,
- b. Assumotions Figure 4 shows the pathways by which activity could be released from the Millstone Unit No. 3 containment following a LOCA or CREA.
The CREA also includes a pathway due to primary to-x-.-------. - - -
,a
--,r
1 4
U.S. Nuclear Regulatory Commission B13429/ Attachment 1/Page 12 February 26, 1990 secondary leakage which is not affected by the changes.
Cur-rently, with the containment pressure at < 10 psia, the contain-ment returns to subatmospheric conditions within I hour post-LOCA or CREA, and hence the containment leakage paths are l
assumed to terminate leakage at T - 60 minutes.
Changing the
)
allowable pressure to 14.2 psia results in a condition where the i
containment never returns subatmospheric, and hence leakage is
)
assumed to continue for the full 30 days over which accident dose calculations are performed.
(EABdosesarecalculatedover 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.)
This change in pressure in itself, without modifying other requirements or assumptions, would result in an increase in calculated accident consequences to levels in excess of 10CFR100 or General Design Criterion 19 limits, and hence would be unacceptable.
An effective increase by a factor of 3 in the allowable bypass leakage also increases the thyroid dose consequences as a larger fraction of the release is bypassing the emergency filtration systems.
Again, taken by itself, this change would result in unacceptable dose consequence.
In order to ensure a LOCA or CREA result in acceptable doses given the above two changes which increase the dose, other I
l changes in requirements or calculational assumptions were necessary.
The following lists all of the changes from the current FSAR dose analyses and the basis for the change.
1 (1)ContainmentLeakRateandDurationfl0CAandCREA)
(a) Current Analysis 0 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> -0.9 percent per day
> 1 hour-0-subatmospheric (b) Proposed Chanae EAB and LPZ 0-24 hours-0.65 percent per day 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s-30 days-0.325 percent per day (c) Control Room and TSC 0-1 hour-0.65 percent per day 1 hour-30 days 0.22 percent per day
i i
l U.S. Nuclear Regulatory Commission B13429/ Attachment 1/Page 13 february 26, 1990 j
(d) Basis j
The current analysis is based on the subatmospheric design of Millstone Unit No. 3, where all containment leakage terminates within I hour.
The proposed change in pressure results in the attached post-LOCA pressure curve Figure 1.
The pressure drops to approximately 4 psig within I hour, but never goes subatmospheric, varying between 3 and 5.5 psig for the entire 30 days.
To help compensate for the increased time release, the allowable Technical Specification leak rate, L is l
being reduced from 0.9 percent per day to 0.65 dr, cent per day.
Conservatively, for calculating the cose to i
the public, the Standard Review Plan guidance of assum-ing L for the first day and 0.5 L for the remaining i
29 da s was assumed.
For the codtrol room and TSC calcu ations, a more reasonable assumption was made which recognizes the fact that Millstone Unit No. 3 still has the containment cooling systems typical of a l
subatmospheric containment.
Thus, containment pressure is rapidly reduced compared to a standard atmospheric containment, for which the Standard Review Plan-guidance was developed. An evaluation was performed to determine the fraction of La leak rate which would exist from
'I hour to 30 days after the break.
The fraction was conservatively calculated to be 0.33 L.
Hence, the control room and TSC calculation assum% 0.33 L, from I hour to 30 days.
l (2) Containment Soray Iodine Removal fl0CA and CRE61 (a) Current Analysis No credit is taken for iodine removal.
(b) Proposed Chance Credit is taken for the iodine removal capabilities of the sprays.
Effective removal of coefficients were calculated.
Since Millstone Unit No. 3 Technical Specifications on sprays and additives are based on its design as a subatmospheric containment, no changes are necessary to ensure their operability and use for the removal rates assumed.
For the LOCA, sprays are assumed to be effective immediately.
For the CREA, a 10-minute delay is assumed to allow for manual initiation of sprays in case the automatic initiation set point is not achieved.
e
~.
,w
L d
U.S. Nuclear Regulatory Commission B13429/Attac h at 1/Page 14 February 26, 1990 (c) Enia Such credit could have been taken for the original LOCA and CREA calculations, but was not since the results were acceptable without the sprays.
(3)IodineSpecies(LOCAandCREA)
(a) Current Analysis 91 percent elemental 5 percent particulate 4 percent organic (b) Fronosed Chanae 95.5 percent elemental 2.5 percent particulate 2 percent organic (c) B1111 Standard Review Plan 15.6.5 provides the current iodine species breakdown for LOCA dose calculations.
- However, Standard Review Plan 6.5.2, Rev.1, specifies that the proposed breakdown be used when iodine removal by sprays is assumed.- Since credit for sprays is taken, the new percentage breakdown is assumed.
(4) Millstone Unit No. 3 Control Room and TSC X/Os (LOCA)
(a) Current Analysis Control Room and TSC values are assumed to be the same.
No credit was taken for the fact that the TSC is farther away from the release point.
From Ventilation Vent From Containment 0-8 hours 4.75(-3) 1.92(-3 8-24 hours 2.30 -3) 1.51(-3 1-4 days 1.03-3) 4.24(-4 4-30 days 1.28 -4) 4.80(-5 i
____ _ _____ _ ______ _ _ _____ _ _ j
1 i
i i
I U.S. Nuclear Regulatory Commission B13429/ Attachment 1/Page 15 February 26, 1990 (b) Prorosed Chance Control Room From Ventilation Vent From Containment 0-8 hours 2.24(-3 8.08(-4 8 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1 40(-3 5.49 -
i 1-4 days 5.08(-4 1.95 -
j 4-30 days 9.68(5) 2.75 -
Technical Sunoort Center From Ventilation Vent From Containment j
0-8 hours 1.20(-3) 4.85 -
8 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 7.98(-4) 3.22 -
l l-4 days 4.75(-4) 1.92 -
4-30 days 7.45( 5) 3.01(-5 I
(c) Basis l
1 The current X/Qs were calculated by Stone and Webster and used an NRC formula to derive a wind speed.
The proposed X/Qs used actual Millstone meteorological data to determine the wind speed in accordance with Regula-tory Guides 1.145.
Additionally, the current X/Qs for the TSC conservatively assumed the same values as the control room (the control room is closer to the vent and containment).
The proposed values use the actual distance to the TSC air intake and include a TSC occu-pancy factor of 0.6 for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in lieu of the control room occupancy factor of I for the first day.
This reduced occupancy factor is based on our r
emergency plan requirement to establish two 12-hour shifts for support personnel.
The 0.6 factor accounts for shift overlaps.
(5) Bypass Leakaae Fraction (LOCA and CREA)
(a) Current Analysis L
0.01 L, = 0.009 percent per day 1
(b) Proposed Chance l
0.042 L, = 0.028 percent per day L
, o.
)
U.S. Nuclear Regulatory Commission i
B13429/ Attachment 1/Page 16 February 26, 1990 (c) Bu it The current value is restrictively low because of the failure of the analysis to take credit for iodine removal by sprays.
Because of the effectiveness of the s) rays, even though the bypass leakage is increased in tie proposed chalige, there is a significant reduction in the calculated EAB thyroid dose.
This increased bypass fraction was considered necessary because of the diffi-culty in meeting such a low value for LLRT during the last refueling outage.
Failure to meet such a restric-i tively low limit could result in additional unwarranted occupational exposure to repair valves with marginal i
leakage.
(6) Plate-Out Coefficient (LOCA and CREA) 1 i
(a) Current Analysis l
An immediate reduction of a factor of 2 for iodine i
plate-out in the containment is assumed.
']
(b) Proposed Chance An immediate reduction due to plate-out is no longer assumed.
Instead, a plate-out removal coefficient is calculated in the model.
(c) Basis This change is per the guidelines of Standard Review Plan Section 6.5.2 on the use of sprays for iodine removal.
With sprays operating, plate-out is not as effective' and the immediate two-factor reduction is no longer appropriate.
7.
0- to 1-Minute Bvoass of Secondary Containment (LOCA and CREA)
(a) Current Analysis The entire 0.9 percent containment leakage is assumed to bypass the filtration systems during the first minute of the accident, i
o a>
k i
U.S. Nuclear Regulatory Commission B13429/ Attachment 1/Page 17 February 26, 1990 (b) Proposed Chanae Other than the fixed bypass fraction (0.042L no leakage is assumed to bypass the filters duri$g), the first minute.
(c) Ruh i-Secondary containment negative pressure less than 0.25 in water gauge (WG) is achieved within 45 seconds.
According to the guidance in Standard Review Plan 6.5.3, if negative pressure is achieved within the first minute, no bypass needs to be assumed.
This is appro-priate as it would take more than.one minute for fuel to fail and the activity to be transported through both the primary and secondary containments.
- c. Results Table 2 presents the results of the dose calculations for the LOCA. The current results are designated "0LD" on the table and are from the FSAR and Stone and Webster calculations.
The "NEW" column presents the results with the above assumption changes as calculated by NNECO.
There are some minor model differ 2nces between Stone and Webster and NNECO, but these are insignificant in terms of the final results and the magnitude of the changes as a result of the assumption changes.
For the EAB, the thyroid dose was reduced from 238 to 150 rem.
The whole body dose increased slightly from-16.8 to 19.5 rem.
For the LPZ, both the thyroid and whole body doses increased by about a factor of 2.
All values remain well within the 10CFR100 limits of 300 rem to the thyroid and 25 rem to the whole body.
For the dose to operating personnel in the Millstone Unit No. 2 and Millstone Unit No. 3 control rooms end TSC, the thyroid dose i
went down in all cases.
The whole body dose increased by approximately a factor of 2 in the control rooms and decreased slightly in the TSC.
The beta dose in the Millstone Unit No. 2 and Millstone Unit No. 3 control rooms had the biggest increase (factors of 3 and 8, respectively), and the TSC beta dose increased slightly.
All doses remain within the General Design Criterion 19 limits of 30 rem to the thyroid and skin and 5 rem to the whole body.
Table 3 presents the EAB and LPZ doses from the CREA.
Doses were calculated for both the N and N-1 loop conditions.
The j
proposed changes only affect the dose contribution from the
i U.S. Nuclear Regulatory Commission j
B13429/ Attachment 1/Page 18 j
February 26, 1990 primary side.
Secondary side dose contributions are presented to show the change in the total dose consequence.
For the EAB, the thyroid dose decreased by a factor of 2 and the whole body dose was slightly lower for both the N and N-1 loop analysis.
For the LPZ, the thyroid dose only increased by about 5 percent and the whole body dose increased by 50 percent from 0.03 rem to 0.046 rem.
All doses remain well within the stan-dard review plan Section 15.4.8 acceptance criteria of 75 rem to the thyroid and 6 rem to the whole body.
1 The proposed design ' changes will not have any impact on the probability of occurrence of any DBA.
Hcwever, the proposed i
changes do increase some of the calculated dose consequences of
)
the LOCA or CREA. As such, they constitute an Unreviewed Safety Question.
Since the overall results remain within the accept-able limits of 10CFR100, General Design Criterion 19 and the i
Standard Review Plan acceptance criteria, NNECO considers these changes acceptable and safe.
II.
Potential for Creation of a New Tvoe of Unanalyzed Event There are no new failure modes associated with the proposed changes.
A relatively small change in the containment operating pressure has no
)
direct effect'on systems and components.
As discussed above, all of the DBAs which could be adversely affected by the proposed changes were reanalyzed.
These analyses showed that all of the acceptance criteria remain satisfied.
Therefore, the plant response is not significantly altered as a result of the proposed changes and the proposed changes cannot be considered to create a new accident.
Therefore, the proposed changes will not create the possibility of an accident of a different type than any evaluated previously in the safety analysis report.
Ill. Imoact on the Marain~of Safety The proposed changes would cause an increase in the maximum containment pressure following a design basis LOCA by 3 psi.
The resultant 38.57 psig pressure is well below the containment design pressure of 45 psig. The containment pressure reduces to less than 50 percent of the peak containment pressure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the postulated accident, thus satisfying the Standard Review Plan Section 6.2.1. A.
The calculated radiation doses in the EAB, LPZ, and to operating personnel remain well within the 10CFR100 limits, General Design Criterion 19 limits, and the Standard Review Plan acceptance criteria for the postulated LOCA and CREA.
I U.S. Nuclear Regulatory Comission B13429/ Attachment 1/Page 19 February 26, 1990 Sumary and Conclusions Based on the foregoing assessment, the proposed design changes herein are considered an Unreviewed Safety Question (USQ) due to an increase in post-accident radiological consequences, but are acceptable and safe.
4 U.S. Nuclear Regulatory Commission 813429/ Attachment 1/Page 20 February 26, 1990 Table 1 Millstone Unit No. 3 Lona-Term LOCA Pressure and Temoerature Pump Suction DER Minimum Time _.___
Pressure (osia)
ESF Temoerature ('F) 11 seconds 31.49 246.2 110 seconds 26.98 235.8 171 seconds 27.09 236.0 325 seconds 24.94 231.0 1800 seconds 8.65 171.2 2990 seconds 4.73 139.9 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 4.19 133.2 1 day 4.07 125.8 2 days 3.43 117.7 30 days
- 1.75 90.0 l
l
- Extrapolated from end of analysis at 22.75 days l
l l
l l
l
,. ~.
s o
U.S. Nuclear Regulatory Commission B13429/ Attachment 1/Page 21 February 26, 1990 h
Table 2 Millstone Station Doses frem) Due to Millstone Unit No. 3 LOCA(I)
EAB (0- to 2-Hour Dose)
Old New Thyroid Whole Body Thyroid Whole Body 2.38(+2) 1.68(+1) 1.50(+2) 1.95(+1)
LPZ (0- to 30-Day Dose)
Old New Thyroid Whole Body Thyroid Whole Body 1.61(+1) 1.59(+0) 3.16(+1) 3.54(+0)
Millstone Unit No. 3 Control Room 01d New Thyroid
- Whole Body Beta Thyroid Whole Body Beta 2'.98(+1) 1.8(+0) 3.2(+0) 2.60(+1)(2) 3.05(+0) 2.45(+1)
Technical Support Center Old New Thyroid Whole Body Beta Thyroid Whole Body Beta 2.50(+1) 1.90(+0) 2.10(+1) 7.37(+0) 1.42(+0) 2.49(+1) l L
l (1) The Standard Review Plan Acceptance Criteria for EAB and LPZ dose limits are (1) 300 rem to the thyroid and (2) 25 rem to the whole body.
General Design Criteria 19 limits for the Millstone Unit Nos. 2 and 3 control room and the Technical Support Center are 30 rem to the thyroid and skin and 5 rem to the whole body.
(2) Millstone Unit No. 2 low wind speed:
2.8(+1) is now limiting for thyroid.
1
i l
a-
- e a
U.S. Nuclear Regulatory Commission 813429/ Attachment 1/Page 22 February 26, 1990 Millstone Unit No. 2 Control Room Old New Thyroid Whole Body Beta Thyroid Whole Body Beta 2.48(+1)(3) 2.09(-1)(4) 2.67(+0)(5) 1.84(+1)(2) 4.81(1)(3) 8.29(+0)(4) i h
I t
(3) Millstone Unit No I steam line break:
2.624(+1) is limiting for
. thyroid.
(4) Millstone Unit No.1 steam line break was limiting [4.20(-1)). Millstone Unit No. 3 LOCA is now limiting.
-(5) Millstone Unit No. I steam line break was limiting [5.87(+0)]. Millstone Unit No. 3 LOCA is now limiting.
r j
c l
U.S. Nuclear Regulatory Commission B13429/ Attachment 1/Page 23 february 26, 1990 Table 3 Millstone Unit No. 3 i
Doses f rem) Due to Control Rod E.iection Accident (I)
{
EAB 0 to 2 Hours frem) j N Loon Old New Thyroid Whole Body Thyroid _
Whole Body I
Primary 1.20+1) 2.70(-1) 4.79(40) 2.55(-l Secondary 1.70 -1) 2.10(-2) 1.70(-1) 2.10(-2 Total 1.22 +1) 2.91(-1) 4.96(+0) 2.76(-l N-1 Loop 01d New Thyroid Whole Body Thyroid Whole Body l
Primary 1.30(+1) 2.80 -
4.79(+0 2.55(-l Secondary 2.00 -2 4.40 -
2.00(-2 4.40(-3 Total 1.30 +1 2.84 -
4.81(+0 2.60(-l LPZ 0 to 30 Days frem)
N Looo Old New e
Thyroid Whole Body
_ Thyroid Whole Body Primary 7.50(-1) 3.10(-2 7.90(-1 4.52(-2 Secondary 9.10(-3) 1.10(-3 9.10(-3 1.10(-3 1
Total 7.59(-1) 3.21(-2 7.99(-l 4.63(-2 N-1 Looo Old New Thyroid Whole Body Thyroid Whole Body Primary 7.60(-1) 2.70(-2) 7.90-1) 4.52(-2)
Secondary 1.10(-3) 2.30(-4) 1.10 -3) 2.30(4)
Total 7.61(-1) 2.72(-2) 7.9) -1) 4.54(-2)
(1) The Standard Review Plan acceptance criteria for the thyroid dose is 75 rem and for the whole body dose is 6 rem.
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it.oi t
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i ne t
'0 nigee I
nrm
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1 d aaa eh r
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e m
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Eu np n
i Tst o Osoin Nant o eo.mG amamm L
n
- 4.a f,
1 L
Figur@ 4
'l
]
Ventilation Vent 95 %
Secondary Containment or Auxiliary Building
- Filter Efficiency q
Filtered Leakage 0.6222%/ day (from both s arayed and-unsprayec regions)
Unsprayed Region 5
Re7!n L
B pass Leakage b.0278%/ day (from both.
r sprayed and
-- g unsprayed regions)
L i
i L
L Containment Sump b
50 % I
(-
Leak Rate = 104cc/hr 10% 1 Flash For conservatism, all leakage that is filtered is assumed to ESF Building L criginate in the Auxiliary Building and is thus discharged via the ventilation vent. Leakage into the secondary containment would go to SLCRS and to Unit 1 stack. MP1 stack releases would result m lower doses.-
0
=
.