ML20140C911

Proposed Tech Specs,Modifying TS Section 15.3.3, Eccs,Acs, Air Recirculation Fan Coolers & Containment Spray, to Incorporate AOT Similar to Ones Contained in NUREG-1431,Rev 1

ML20140C911
Person / Time
Site:	Point Beach
Issue date:	06/03/1997
From:	WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20140C907	List: ML20140C911 NPL-97-0315, Suppl to TS Change Request 192 to Licenses DPR-24 & DPR-27, Modifying TS Section 15.3.3, Eccs,Acs,Air Recirculation Fan Coolers & Containment Spray to Incorporate Allowed Outage Times Similar to Those Contained in NUREG-1431,Rev 1
References
RTR-NUREG-1431 NUDOCS 9706100131
Download: ML20140C911 (29)

Text

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15.3.12 CONTROL ROOM EMERGENCY FILTRATION 1

Applicability Applies to the operability of the control room emergency filtration.

Objective i

To specify functional requirements of the control room emergency filtration during power operation and refueling operation.

Soecification 1.

Except as specified in 15.3.12.3 below, the control room emergency filtration system shall be operable at all times during power operation and refueling operation of either unit.

2.

a.

The results ofin-place cold DOP and halogenated hydrocarbon tests, conducted in accordance with Specif'::ation 15.4.11, on HEPA filter and charcoal adsorber banks shall show a minimum of 99% DOP removal and 99% halogenated hydrocarbon removal, b.

The results oflaboratory charcoal adsorbent tests, conducted in accordance with Specification 15.4.11, shall show a minimum of 90g% removal of methyl iodide. If laboratory analysis results for in-place charcoal indicate less than 90g% methyl iodide removal, this specification may be met by replacement with charcoal adsorbent which has been verified to achieve 90g% minimum removal and which has been stored in sealed containers, and retesting the charcoal adsorber bank for halogenated hydrocarbon

removal, The results of fan testing, conducted in accordance with specification 15.4.11, shall show c.

operation within i10% of design flow.

9706100131 970603 PDR ADOCK 05000266 P

PDR Unit 1 - Amendment No.-440 15.3.12 1 Jan=rj 25, P88 Unit 2 - Amendment NoA44

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Halogenated hydrocarbon testing of the charcoal adsorber bank shall be performed l

c.

after each complete or partial replacement of charcoal adsorbers or after any structural maintenance of the adsorber housing. Halogenated hydrocarbon testing shall be at design velocity 20%.

d.

Laboratory sample analysis ofin-place charcoal adsorbent shall be performed at least once per year for standby service or after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation and, as a minimum, shall be conducted at velocities within 20% of design,1.75 mg/m 2 inlet iodide concentration,95% relative humidity and 30 C (86"F).

Fans shall be tested at least once per year or after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of operation since the e.

previous test, and following fan maintenance or repair.

Basis The control room emergency filtration system is designed to filter the control room atmosphere and makeup air to the control room during control room isolation conditions. The control room emergency filtration is normally isolated and not in operation and testing more frequently than that specified is not required to insure operability or performance. If the efficiencies of HEPA and charcoal adsorbers are as specified, the resulting control room doses during accident conditions will be less than allowable levels in Criterion 19 of Appendix A to 10 CFR 50. The charcoal adsorbent laboratory sample analysis is performed in accordance with ASTM D3803-89. " Standard Test Method for Nuclear-Grade Activated Carbon."

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i Unit 1 - Amendment 409 15.4.11-2 December 23,1937 Unit 2 - Amendment 44-2

ATTACHMENT 2 TECHNICAL SPECIFICATIONS CHANGE REOUEST 192 ENVIRONMENTAL CONSEOUENCES OF A LOSS OF COOLANT ACCIDENT l

Introduction l

l A large pipe mpture in the RCS is assumed to occur. As a result of the accident, it is l

assumed that core damage occurs and iodine and noble gaa activity is released to the l

containment atmosphere. A portion of this activity is released via containment leakage to j

the outside atmosphere (Reference 1). Also, once recirculation of the Emergency Core l

Cooling System (ECCS) is established, activity in the sump solution may be released to the environment by means ofleakage from ECCS equipment outside containment in the auxiliary building. This section describes the assumptions and analyses performed to determine the amount of radioactivity released and the offsite and control room

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radiological consequences resulting from these releases.

Input Parameters and Assumptions The analysis of the large break LOCA radiological consequences uses guidance outlined in the Standard Review Plans (References 1 and 2). One hundred and two (102) percent of the current power level of 1518.5 MWt (1548.9 MWt) is assumed in the analysis. Both the offsite and control room doses are calculated based on the conservative assumptions J

listed in Reference 3.

Containment Leakage l

Following the large break LOCA, 50% of the core iodine activity and 100% of the core 1

l noble gas activity are assumed to be immediately released to containment when determining doses due to containment leakage. Fifty percent of the iodine released to containment is assumed to instantaneously plate out on containment surfaces. This leaves 25% of the core iodine activity and 100% of the core noble gas activity instantaneously available for leakage from the containment (Reference 3). This iodine is assumed to be 91% elemental,4% methyl and 5% particulate (Reference 3).

The particulate and elemental iodine are removed from the containment atmosphere by the action of the containment sprays. The organic form ofiodine is not easily removed from the containment atmosphere and is assumed to be removed only by radioactive decay and leakage. The elementaliodine spray coefficient of 20 hr is determined based on the model suggested in Reference 4. Credit is taken for this spray removal until a decontamination factor of 200 in the containment inventory of elemental iodine is reached l

(Reference 4). The particulate iodine spray coefficient of 6.02 hr is also determined based on the model suggested in Reference 4. Credit is taken for particulate iodine removal until containment spray has been terminated. At this time a decontamination factor of 11.2 in the particulate iodine inventory in containment has been reached.

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The Technical Specification containment design allowable leak rate of 0.4% by weight of containment air is used for the initial 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Thereafter the containment leak rate is assumed to be one-half the de sign value, or 0.2%/ day (Reference 3)

ECCS Equipment Leakage When ECCS recirculation is established following the LOCA, leakage is assumed to occur from ECCS equipment outside the containment. It is also assumed that 50% of the total core iodine is in the sump water being recirculated (Reference 2). Hence, the ECCS equipment leakage results in the release of a significant amount ofiodine activity to the outside environ.nent. For this activity release path, no credit is taken for plateout of elemental iodine on containment surfaces or for iodine removal by the atmosphere filtration system in the auxiliary building. The iodine release from this path is conservatively assumed to be 100% elemental.

The ECCS equipment leaks at a rate of 400 cc/ min for the control room analysis and conservatively this rate is doubled to 800 cc/ min for offsite dose analysis. This leak rate is conservatively assumed to continue at this constant rate from the time ECCS recirculation is established until 30 days following accident initiation. Ten percent of the iodine in the leakage is assumed to become airborne due to flashing (Reference 2).

There is no noble gas activity in the ECCS recirculation water.

Control Room Parameters The doses to personnelin the control room are determined for both of the activity release 3

paths discussed above. The control room ventilation system volume is 65,243 ft, the filtered makeup flow is 4950 cfm with no filtered recirculation, and the unfiltered inleakage flow is 10 cfm. The control room filter iodine removal efficiencies are 95%

elemental,95% organic and 99% particulate.

Attached Tables The thyroid dose conversion factors, breathing rates and atmospheric dispersion factors used in the dose calculations are given in Table 1. The core activities used in the dose calculations are given in Table 2. The control room assumptions and parameters are given in Table 3. The major assumptions and parameters used to determine the doses due to containment leakage are given in Table 4 and those for ECCS equipment leakage are given in Table 5.

Description of Analyses Performed The offsite thyroid and whole body doses, as well as the control room thyroid, whole body and beta skin doses, are determined using guidance outlined in the Standard Resiew Plan l

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(References 1 and 2). Both the containment leakage and the ECCS continuous leakage activity release pathways are included in this analysis.

l Acceptance Criteria The offsite doses must meet the guidelines of 10 CFR 100, or 300 rem thyroid and 25 rem whole body for the initial 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period following the accident at the Site Boundary (SB) and for the duration of the accident at the Low Population Zone (LPZ). The dose criteria for control room personnel following the accident are 5 rem whole body,30 rem thyroid from the inhalation of radioactive iodine (Reference 6), and 30 rem p-skin (or 75 rem p-skin with protective clothing) per Reference 5.

Results The offsite and control room thyroid, whole body, and beta skin doses due to the large break LOCA are given in Table 6.

Conclusions The offsite thyroid and whole body doses are within the current NRC acceptance criteria for a loss of coolant accident. The control room whole body dose is within the current NRC acceptance criteria for the control room. The control room thyroid dose exceeds the 30 rem limit; however, by the use of potassium iodide pills, the control room thyroid dose would be reduced to approximately 29 rem which is within the 30 rem limit. The use of potassium iodide pills to reduce the thyroid dose to the control room operator by a factor of10 is based on NRC Safety Evaluation Report dated August 10,1982 for resolution NUREG-0737 item III.D.3.4 - Control Room Habitability. The control room beta skin dose exceeds the 30 rem accelaance criterion; however, it is within the 75 rem acceptance criterion which is based on the use of protective clothing.

References

1. NUREG-0800, Standard Review Plan 15.6.5, Appendix A, " Radiological Consequences of a Design Basis Loss-of-Coolant Accident Including Containment l

Leakage Contribution," Rev.1, July 1981.

2. NUREG-0800, Standard Review Plan 15.6.5, Appendix B, " Radiological Consequences of a Design Basis Loss-of-Coolant Accident: Leakage from Engineered Safety Features Components Outside Containment," Rev.1, July 1981.

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3. Regulatory Guide 1.4, Rev. 2, July 1974, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors."

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4 NUREG-0800, Standard Review Plan 6.5.2, " Containment Spray as a Fission Product Cleanup System," Rev. 2, December 1988.

5. NUREG-0800, Standard Review Plan 6.4, " Control Room Habitability System", Rev.

2, July 1981.

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6. " Nuclear Power Plant Control Room Ventilation System Design for Meeting General f

Criterion 19," K. G. Murphy and Dr. K. M. Campe,13th AEC Air Cleaning Conference.

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l TABLE 1 DOSE CONVERSION FACTORS, BREATHING RATES AND ATMOSPHERIC DISPERSION FACTORS Thyroid Dose Conversion Factors "'

lootepe (rem / curie) 1-131 1.07 E6 i

l-132 6.29 E3 1

1-133 1.81 E5 l-134 1.07 E3 l

I-135

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3.14 E4 Time Period Breathing Rate

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(m'/sec) 1 l

O-8 3.47 E-4 8-24 1.75 E-4 j

24-720 2.32 E-4 Atmospheric Dispersion Factors

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Site Boundary 0 2 hr 5.0 E-4 j

l Low Population Zone l

O-8 hr 3.0 E-5 8-24 hr 1.6 E-5 l

l 24-96 hr 4.2 E-6 96-720 hr 8.6 E-7 Release from Release from Auxiliary --

Control Room

• Containment Building 0-8 hr 3.0 E 3 1.7 E-3 8-24 hr 1.9 E-3 1.2 E-3 24-96 hr 1.2 E 3 6.7 E-4 l

96-720 hr 4.8 E-4 2.3 E-4 "8 ICRP Publication 30 m Regulatory Guide 1.4

• Wisconsin Electric letter NPL 97-0041

• Wisconsin Electric letter NPL 97-0276 1

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CORE ACTIVmESM8 i

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Total Core Activity at Nuclide Shutdown (Ci) i l-131 4.13 E7 I-132 5.92 E7 I

l-133 8.45 E7 1

i 1 _134 9.30 E7 1-135 7.89 E7 Kr-85 5.07 E5 Kr85m 1.13 E7 Kr-87 2.16 E7 Kr 88 3.0 E7 Xe 131m 4.41 E5 Xe-133 8.36 E7 Xe-133m 2.63 E6 Xe-135 2.30 E7 Xe-135m 1.60 E7 Xe-138 7.04 E7 m These core activities are based on a core power level of 1548.9 MWt. The activities were updated as a part of the Point Beach fuel upgrade program.

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TABLE 3 CONTROL ROOM PARAMETERS Volume 65,243 ft*

Unfiltered Inleakage '

10.0 cfm I

Total Flow Rate 19800 cfm j

Filtered Makeup 4950 cfm j

Filtered Recirculation O cfm 1

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i-Elemental 95 %

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l Organic 95 %

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Particulate 99 %

1 Occupancy Factors

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1 0 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1,0 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 0.6 l

4 - 30 days 0.4 l

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TABLE 4 ASSUMPTIONS USED FOR LARGE BREAK LOCA DOSE, ANALYSIS CONTAINMENT LEAKAGE i

E Power (102%)

1549 MWt lodine Chemical Species Elemental 91 %

Methyl 5%

Particulate 4%

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w lodine Removalin Containment

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j instantaneous lodine Plateout 50 %

Containment Spray Start delay time 90 seconds i

o injection spray flow rate 1190 p m Duration of injection spray 65 minutes Spray removal coefficient Elemental 20 hr

Particulate 6.02 hr

Containment Net Free Volume 1.065E6 ft*

Sprayed Volume 475,000 ft 8 Unsprayed Volume 590,000 ft*

i Containment Mixing Containment Fan Coolers Start Delay Time 90 seconds Number of Units 2

Flow Rate per Unit 38,500 cfm Containment Leak Rete 0-24 hr 0.4%/ day

> 24 hr 0.2%/ day

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ASSUMPTIONS USED FOR LARGE BREAK LOCA DOSE ANALYSIS I

i ECCS EQUIPMENT LEAKAGE

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1 Power (102%)

1549 MWt lodine Activity in Recirculation Water 50% Core lodine Activity l

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100% Elemental Leakage Rate For Offsite Doses 800 cc/ min For Control Room Doses 400 cc/ min Leakage Duration From start of racirculation through 30 day duration Time of Recirculation Initiation 20 minutes Sump Water Volume 197,000 gallons lodine Flashing Fraction to Environment 10 %

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TABLE 6 LARGE BREAK OFFSITE AND CONTROL ROOM DOSES

1. Thyroid Doses Dose (Rem)

SB (0-2 Hrl LPZ (0-30 Day)

CR (0 30 Davl l

Containment Leakage 133.3 24.37 186.0 l

ECCS Equip, Leakage 57.12 37.0 106.7 Total 190.42 61.37 292.7

2. y-body Doses Dose (Rem)

(R,1Q-2 Hrl LPZ (0-30 Day)

CR [0-30) Day Containment Leakage (10) 0.72 0.07 0.012 Containment Leakage (NG) 2.52 0.38 1.354 ECCS Equip. Leakage 0.24 0.06 0.004 Total 3.48 0.51 1.37 l

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3. CR S-Skin Dose

• Dose (Rem) 30 Day Containment Leakage (10) 0.12 Containment Leakage (NG) 43.02 ECCS Equip. Leakage 0.04 Total -

43.18 l

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NG = Noble Gases i

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ATTACHMENT 3 1

l TECHNICAL SPECIFICATIONS CHANGE REOUEST 192 ADDITIONAL INFORMATION Post Accident Containment Atmosphere Sample Containment post-LOCA pressure response documented in PBNP Final Safety Analysis Report (FSAR) Chapter 14, indicates that a pressure rating of 15 psig for the containment atmosphere post-accident sample system (PASS) is sufficient to allow j

sampling and analysis within three hours of the decision to sample as required by NUREG 0737, Item II.B.3. Sample pump leakage at this pressure has been evaluated and it has been determined that at the existing leakage levels, dose rates to the personnel taking the sample could be as high as 2200 Rem /hr - extremities and 80 Rem / hour -

whole body at two hours after accident initiation. These dose rates would result in a person exceeding the GDC 19 dose limits of 75 Rem extremities, 5 Rem whole body for obtaining a sample, as specified by NUREG-0737, Item II.B.3.

Previous information submitted in support of this Technical Specifications change request contained in a letter dated September 30,1996, presented a new containment pressure response curve such that sampling would not be possible for approximately 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> after the initiation of the accident due to pressure being above the 15 psig continuous rating of the sample pump. The September 30,1996 submittal also presented information that the sample pump was capable of operating at up to 23 psig. This was based on an informal evaluation which assessed the difference in density between air and the post-accident containment atmosphere density.

During the Integrated Leak Rate Test (ILRT) performed on the Unit 2 containment in March 1997, the containment atmosphere sampling system was restored to operation when containment pressure was below 5 psig as stipulated by the ILRT procedure.

Upon restoration, the sample pump tripped. This called into question the ability to take a containment atmosphere sample.

The post accident containment atmosphere sample system obtains a sample of the air from the containment air radiation monitoring system. The primary purpose of the system is to monitor the containment atmosphere for radioactive particulates and noble gases. During normal operation the system draws a sample from the containment atmosphere utilizing a motor-driven sample pump. Air is drawn through a dual chamber sampler assembly. Air passes through a filter in the first chamber where trapped l

particulates are monitored by a beta scintillation detector. The air is then routed through l

a second sample chamber where it is monitored by a beta scintillation detector for noble l

gases. After passing through the two chambers and sample pump the air is returned to i

containment.

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.The system is isobted automatically on a containment isolation or containment ventilation isolation signal. These signals also trip the sample pump. For post accident containment atmosphere sampling, after resetting the containment isolation signal if i

necessary, the system is unisolated and aligned so that the pumps draws a sample from the containment atmosphere through a sample septum. A sample of the containment atmosphere is drawn from the septum using syringes and transported to a remote l

I location for analysis. The system is purged with service air.

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The current plan for resolution of this discrepancy is as follows:

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1. The containment atmosphere sample systems will be upgraded by installation of an j

eductor and nitrogen source for taking a containment atmosphere sample and j

purging the sample lines. An eductor will be used in lieu of the sample pump for post j

accident sampling purposes. The system will be upgraded to be capable of obtaining 2

an atmosphere sample at up to the containment design pressure of 60 psig. This will I

enable a se ple to be drawn within an hour following accident initiation based on the revised con ' t nment pressure profile provided in with this submittal.

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2. Modifications will also be cooleted to reduce personnel time in the sample hut -

while aligning the system and e taining the sample. This and the modification to install an eductor is expected to i ' duce the dose to within GDC 19 limits.

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3. Procedure changes will also be com!dered if necessary to reduce dose within limits.

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4. The above modifications will be completed prior to startup of both units from the present outages.

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5. Evaluation of the isolation provisions for the containment atmosphere and reactor 1

coolant sample systems is continuing. The isolations will be modified if determined to be necessary.

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6. Options are being evaluated for the long-term to further upgrade the systems and

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reduce dose.

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Containment Sump Volume The assumed refueling water storage tank (RWST) available volume has changed which in turn affects the sump volume. The RWST available volume has been reduced to accommodate revised net positive suction head calculations and RWST level instrument uncertainty.

The net positive suction head analysis for the containment spray pump shows that at least 9% oflevel in the RWST will maintain adequate suction pressure. The revised RWST level instrument uncertainty is 3%. The Technical Specification minimum volume in the RWST is 275,000 gallons (per TS 15.3.3) which corresponds to 95% level in the RWST.

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.Therefore, the RWST level change for drain down is now 92% to 12% based on the inclusion of the 3% instrument uncertainty conservatively at the initial and final level points.

Previous Value The sump volume used in previous calculations was 223,635 gallons. This volume included a RWST water volume based on the RWST being drained from 95% level to 6%

level.

l New Value j

The analytical limits for RWST drain down are now 92% and 12%. This has reduced the amount of water available. The new value used when calculating the dose due to ECCS i

leakage is 197,000 gallons.

Containment Spray Duration The change in RWST levels also affects the containment spray duration.

Previous Value Previously, the spray duration was calculated to be 1.37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> based on the 95% to 6%

RWST level change.

New Value The following assumptions and inputs are used in the new calculation of spray duration:

Initial RWST level is assumed to be 92%, which is 95% minus 3% for instrument uncertainty.

A containment spray pump fails at the outset of the accident, leaving 2 trains of RHR and SI operating in conjunction with I spray pump.

- At 60% RWST level, the train "A" SI and RHR pumps are secured in preparation for containment sump recirculation.

- At 34% RWST level, containment sump recirculation begin's. After this point, only the containment spray pump draws from the RWST.

At 12% RWST level, the containment spray pump is secured.

RWST contains 2833 gallons per % level.

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RHR pump flow rate is 1950 gpm per pump.

SI pump flow rate is 1100 gpm for one pump operating, and 1700 gpm for two pumps i

operatmg.

Containment spray flow is 1555 gpm per pump.

The containment spray duration based on the above inputs is 69 minutes. This result bounds the spray duration used in the large break LOCA radiological analysis of 65 minutes, including the 90 second delay assumed in the analysis for spray initiation.

Large break LOCA simulations performed using the Point Beach Nuclear Plant simulator have shown that the containment spray duration assumed in the analysis is consistent with times achieved using the emergency operating procedures. Spray duration on the order of 50 minutes could occur when the containment sump suction valve for the train aligned for containment sump recirculation was the single failure. Based on this scenario, all four containment fan coolers would be assumed to operate, which promotes containment mixing of the unsprayed volume of the containment into the sprayed region and more rapidly reduces radio-iodine levels in the containment atmosphere. A sensitivity analysis performed shows that this scenario is less limiting for control room dose.

Containment Spray Removal Coefficients The elemental and particulate spray removal coefficients have no+ changed from those provided in the April 2,1997, supplement to Technical Specifica ions Change Request

-192.

Decontamination Factor i

The calculated decontamination factor has changed due to the decrease in containment sump volume. The calculated value still exceeds 200, therefore a value of 200 is used for the calculation oflarge break LOCA doses. The decontamination factor of 200 is the maximum value used for analysis purposes. Based on the fact that a value of 200 was -

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used in the earlier calculations, the decontamination factor used in the analysis is not being changed due to the new sump volume.

Dispersion Factors Previous Values Previously, the Control Room dispersion factors were calculated using the equations given in NUREG/CR-5055 " Atmospheric Diffusion for Control Room Habitability Assessments." The following values do not include occupancy factors.

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. Control Room X/Q's for Containment Releases [sec/m']:

2.le-3 (0-8 hr) l 1.3e-3 (8-24 hr) 8.3e-4 (24-96 hr) 3.3e-4 (96-720 hr)

Control Room X/Q's for Auxiliary Building Releases [sec/m']:

1.0e-3 (0-8 hr) 7.0e-4 (8-24 hr) 3.9e-4 (24-96 hr) 1.3e-4 (96-720 hr) l New Values The Control Room dispersion factors have been re-calculated using equations given in K.G. Murphy and K.M. Campe " Nuclear Power Plant Ventilation System Design for Meeting General Criterion 19," 13th AEC Air Cleaning Conference. The values below do not include occupancy factors.

Control Room X/Q's for Containment Releases [sec/m']:

3.0e-3 (0-8 hr) 1.9e-3 (8-24 hr) 1.2e-3 (24-96 hr) 4.8e-4 (96-720 hr)

Control Room X/Q's for Auxiliacy Building Releases [sec/m']:

1.7e-3 (0-8 hr) 1.2e-3 (8-24 hr) 6.7e-4 (24-96 hr) 2.3e-4 (96-720 hr)

Bnis For releases from the containment building and other release points at the Point Beach Nuclear Plant to the control room ventilation intake, the x/Q value is calculated through use of the methodology given in K.G. Murphy and K.M. Campe " Nuclear Power Plant Ventilation System Design for Meeting General Criterion 19," 13th AEC Air Cleaning i

Conference. The equation for a diffuse source and point receptor is used.

This equation is used when activity is assumed to lei from many points on the surface of the containment building in conjunction with a sk.gle point receptor. The use of this equation is also appropriate for a point soure.: - point receptor where the difference in elevation between the source and receptor is greater than 30 percent of containment height. The equation described above was used for both containment releases and for auxiliary building stack releases because the containment release was from many points on j

the surface of the containment building and the difference in elevation between the 5

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. auxiliary building vent stack and the control room ventilation intake is greater than 30

- percent of the containment height. The auxiliary building vent stack orientation with j

respect to the control room intake is 58.8 meters distance,111 feet 9 inches height for the control room intake,168 feet height for the auxiliary building vent stack, and the down-wind direction sectors from the intake to the vent are SSE, S, SSW, SW. The s/d ratio is i

approximately 1.5 based on s = 58.8 meters and d = 39.7 meters. The 39.7 meter length is j

based on the shortest side of the containment facade.

i Using the methodology described in Regulatory Guide 1.145 and Murphy /Campe, the x/Q value that is exceeded five percent of the time for the Point Beach Nuclear Plant was calculated using the hourly site specific meteorological data measured during the 1991

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through 1993 calendar years. For each hour of meteorological data, the corresponding f

x/Q value at the control room ventilation intake distance from the containment building was calculated. These x/Q values were then used to construct a cumulative probability

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distribution of x/Q values for the site. This probability distribution is described in terms of j

probabilities of exceeding a given x/Q value during the total time.

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Control Room Ventilation and Loss of Offsite Power i

j The operation of the control room ventilation system has not been evaluated previously for situations of the loss of offsite power. Although, in recent years the control room filter fan and recirculation fan have been included on the emergency diesel generator loading i~

tabulation during the recirculation phase of a loss of coolant accident (FSAR Table 8.2-2).

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The control room ventilation system for Point Beach does not automatically restart after a loss of offsite power. Operator action would be required to restore operation of the j

control room ventilation system after a loss of offsite power.

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The analyses for radiological consequences in the control room are based on operation of the system in Mode 4, which is unfiltered recirculation with filtered make-up. It is judged that the radiological effects in the control room of the stopping and subsequent restar, of the system after a loss of offsite power would not be greater than the doses associated i

with Mode 4 operation of the system post-accident, based on the following:

1. The control room would start from positive pressurization because the system normally runs in Mode I which is a positive pressudzation mode.

2. During the loss of ventilation, the air inside the control room would heat-up and expand, which would continue to enhance pressudzation and outflow and tend to minimize in-leakage.

3. The control room would normally be closed which helps maintain pressurization and

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reduces in-leakage.

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b

4. The control room ventilation system damper positions would be expected to fail or automatically transfer the system into Mode 4 which provides for filtered make-up. If any in-leakage through the normal control room intake occurred, it would be filtered at the same or higher efficiency assumed in the analysis.

5. Unfiltered radionuclides such as noble gases would not be drawn into the control room by the control room charcoal filter fan.

Recirculation Loop Leakage 1

Recirculation loop leakage is currently assumed to be 400 cc/ min for the control room dose analysis and 800 cc/ min for offsite dose analysis. The higher leakage rate is assumed for the offsite dose analysis to provide additional conservatism for this analysis. The rsMogical consequences of the LOCA (as reported in the PBNP FSAR section 14.3.5) were based solely on containment leakage after the LOCA. ECC.S leakage was not specifically included in the PBNP FSAR analysis. The original Technical Specifications for Point Beach contained the following basis information:

"The limiting leakage rates from the Residual Heat Removal system are judgment values based primarily on assuring that the components could operate without mechanical failure for a period on the order of 200 days after a Design Basis Accident.

The test pressure (350 psig) achieved either by normal system operation or by hydrostatically testing, gives adequate margin over the highest pressure within the system after a design basis accident. Similarly, the pressure test for the retum lines from the containment to the Residual Heat Removal System (60 psig) is equivalent to the design pressure of containment. A Residual Heat Removal System leakage of 2 gal /hr will limit off-site exposures due to leakage to insignificant levels relative to those calculated for leakage directly from the containment in the Design Basis Accident. The dose calculated as a result of this leakage is 7.7 mr for a 2 hr exposure at the site boundary."

Estimation of recirculation loop leakage is contained in section 6.2 " Design Evaluation,"

of the PBNP FSAR. This information has existed in the PBNP FSAR in basically its present form since the beginning of operation for PBNP. It is not apparent that the recirculation loop leakage shown in the FSAR table 6.2-11 was utilized in any radiological analyses. Provisions for detection and isolation of a seal failure in the Residual Heat Removal pump are also described in the PBNP FS AR section 6.2. Again, it is not apparent that leakage from the R.esidual Heat Removal system during recirculation was ever utilized in any radiological analyses other than that provided in the original PBNP Technical Specifications basis described above.

Primary Auxiliary Building Vent The Loss of Coolant Accident radiological analyses and results were described in a letter dated February 13,1997. These analyses did not account for filtration of the emergency 7

. core cooling system (ECCS) leakage although the introduction to that submittal stated the following, "It is not necessary to model the ECCS passive failure release path because Point Beach has a filtered ventilation system in the auxiliary building." This statement was based on interpretation of the guidance contained in NUREG-0800, Standard Review Plan 15.6.5, Appendix B, " Radiological Consequences of a Design Basis Loss-of-Coolant Accident: Leakage from Engineered Safety Features Components Outside Containment,"

Rev.1, July 1981. Previously, the assumption for passive failure has not been included in the radiological consequences of the LOCA for PBNP. Additionally, the utilization of the Primary Auxiliary Building (PAB) vent system in lieu of the assumption of a passive failure in the ECCS system has not been included in any previous analyses or requirements for Point Beach. After further consideration of the utilization of the PAB vent in this manner, it has beenjudged that application of this Standard Review Plan guidance is not warranted and is inconsistent with the current licensing basis for PBNP.

The use of the PAB vent filtration of ECCS leakage was included in the analysis of control room radiological consequences in the Point Beach response to the NUREG-0737, item III.D.3.4. In submittals to the NRC penaining to NUREG-0737, item III.D.3.4, the PAB vent filtration was not specifically identified for exemption from ECCS passive failure, but it was used to filter the 25,000 cc/hr leakage assumed from ECCS systems during the recirculation phase of a LOCA. As stated previously, the analyses submitted in support of this Technical Specifications change do not account for filtration of the emergency core cooling system leakage. This is more conservative than the previous analyses in support of NUREG-0737, item III.D.3.4, because these new dose analyses show that acceptable doses can be achieved without operation of the PAB vent filtration.

Prior to the submittal of analyses in a letter dated Febmary 13,1997, the radiological consequences of the LOCA (as reported in the PBNP FS AR section 14.3.5) were based solely on containment leakage after the LOCA. ECCS leakage was not specifically included in the PBNP FSAR analysis. It has been included in these analyses in support of this Technical Specifications change request to show the cumulative effects and demonstrate that the amounts of radioactivity released due to a postulated LOCA do not exceed the limits specified in 10 rFR 100.

The PAB vent stack remains the source of emission for ECCS leakage assumed in the analyses provided with this submittal.

Direct Radiation The direct radiation dose due to the plume outside of the control room is calculated using the computer program QAD-CGGP "A Combinatorial Geometry Version of QAD-PS A, A Point Kernel Code System for Neutron and Gamma-Ray Shielding Calculations Using the GP Buildup Factor." The inputs are consistent with those used in the large break LOCA dose calculations. It is assumed that a control room operator is located 10 feet inside the control room window for 75% of the occupancy tirr.: and 5 feet inside the control room window for 25% of the occupancy time. This assumption is the same as that used in the 8

l previous plume dose calculations detailed in a Wisconsin Electric letter to the Nuclear Regulatory Commission letter dated September 4,1984, " Additional Response to NUREG-0737, Point Beach Nuclear Plant."

The 30 day whole body dose from the radionuclides within the control room for the large break LOCA event is 1.37 rem, therefore the direct dose to operators from radiation outside the control room must be below 3.63 rem in order to remain within the 5 rem total whole body dose limit. With portable lead shielding in place in front of the control room door and window, the direct dose is approximately 3 rem over the 30 day duration of the event. The portable shielding, and the procedures for using it are already in place.

Therefore, the 5 rem whole body dose limit is met for PBNP.

3 I

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9

ATTACHMENT 4 i

TECHNICAL SPECIFICATIONS CHANGE REOUEST 192 BEVISED CONTAINMENT INTEGRITY ANALYSIS AND MAXIMUM DBA TEMPERATURE EVALUATION Additional changes to the postulated design basis accident (DBA) Temperature and 1

Pressure profiles were identified during finalization of the Westinghouse report I

documenting the development of containment integrity temperature and pressure profiles for a DBA. These changes are minor and werejudged to have insignificant affect on the Environmental Qualification analysis previously presented for TSCR 192.

These changes included:

i 4

- Results that show superheated conditions during the initial blowdown phase (saturated conditions established at about 7.5 seconds). The superheat results from using an 1

improved version of the Westinghouse containment analysis computer code COCO.

- The initiation of containment sump recirculation was assumed to occur at 2000 seconds compared to approximately one hour assumed in the previously submitted analysis.

j The earlier recirculation time (2000 seconds) was chosen to conservatively bound the initiation ofcontainment sump recirculation.

j 1

The revised DBA analysis results are provided in the attached graphs. The revised peak temperature is 291'F which decreases to 278 F at approximately 7.5 seconds. The previous peak temperature was postulated to be 280 F. This change has no impact since the environmental qualification of all equipment envelopes the 291 F.

The revised temperature profile graph also includes the saturated steam temperature based on the partial pressure of steam in containment. The saturation temperature is considered to be a more representative temperature profile for equipment and structures inside the containment which would be subjected to condensation heat transfer.

Additionally, the revised DBA temperature profile shows a 1 F (242 to 243 F) increase in temperature during the one to five hour point of the post-containment spray period. The equipment atTected by this change have been previously addressed in Wisconsin Electric's

" Supplement to Evaluation of EQ Impact of Wisconsin Electric Power Company Technical Specification Change Request 192" provided to the NRC on April 16,1997.

There is no impact by this 1 F change.

It should also be noted that the revised postulated DBA Peak Pressure decreased by <1 psig from the previously submitted information. This change has no impact on the environmental qualification.

I 1

i

.The revised per.k postulated DBA temperature of 291*F is enveloped by existing EQ reports, as noted below. The smallest margin to the peak temperature of 291 F is 5 F for Westinghouse ESF Cables, evaluation item No. 36 below.

Evaluation New Quall8ed Margin (F)

ItemDescription a ppar.w.

Item No.

Possulated Maximum Test Report l

DEA Maximum Temperature Temperature (F)

(F) 1 291 340 49 Electrical Penetration PEN-RLK4-l Assemtdy 1641 l

2 291 341 50 Rh Firemes tilXLPE N-91055 Cable (Analysis of Test Reports 5804 & 5805) 3 291 445 154 Conex Electrical Penstrabon Conex IPS-AssemtWy and Penetrabon 1420 3 puces 4

291 432 141 Conax RTDs and ConaxIPS475 Them e e 5

291 432 132 Conex RTDs ConaxIPS 798 6

291 442 151 WCSF-N Spices Wyle 58442-1 i

)

7 291 324 33 Fan Cooier Motor, WCAP 7829 Thermainshc Epoxy ineuwian,spkm and j

hencents i

8 291 400 180 m Neutron Report No.010 Flux Montonng System l

l 9

291 315 24 Limtorque Valve Actuators Limnarque Report B0058 2

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• Evolustion New QueBAed Mergin (F) tem Deecription a p aha.

item No. -

Pdh Mealsmsn Test Report DBA Mammum Temperature Temperature (F)

(F) 10 291 340 40 Karte HTKInsulated FR Karte 9/12/80 IN Cable Report 4

1 1

10A 291 300 9

Karte FR insuleled FR Karte9/12/80 m Cette Report I

1 i

11 291 370 79 Namco EA-180 Umt Namco Report SWlches No.105 4

4 12 291 372 74 Namco EA-180 Umt Namco Report l

Swechee No.155

.1 4

13 291 430 130 Crosby Uftind sting Swech Crosby Report j

Assemt*

4245

+

1 1

14 2 31 420 129 Fodoro N-E 10 Series Wyle Report j

Pressure Trenamitters 45502-4 i

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l 15 291 412 1 21 Trenesmence DeleweiGems Wyle Report 4

LevelTrenommer 45700 2 i

1 i

4 2

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1 1

16 291 350 50 Eladrical Conductor Seal ConerIPS4 Assemtsse i

7 17 291 340 40 Ubon Veem Connedorand loomedk Nov American Boe Flex condut 78 1

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4 4

,s Evaluation New QuaRAed Margin (F)

Item Descripton Arr eawa u

item No.

Postulated Maximurn Test Report DBA Maximum Temperature Temperature (F)

(F) 18 2 91 385 94 Target Rock Model7/CC.

Target Rock 001 Solenced Veno Report No.

2375 19 291 310 19 Hydrogen Monitoring System EOSensor Report EXO-QTR-101 20 291 306 15 CEC PressureTransducer EOSensor and Tayco RTD Repyt EXO-OTR 119 21 291 415 124 Lalon Veem Thermocouple NTS Report Connectors

$58-1854 22 291 385 94 Anaconde Cable Franidin Report F-C4989-1 23 291 385 94 Anaconde Flemo-Guard FR-Franidin Report EPinstrumert ControlCable F-C4tKE2 24 291 385 74 Raychem Nudeer Grade Wyle Report Catie Spkas 1785042P 25 291 341 50 Rockbestos Frewellli Rockbestos ChemicesyCrees Linked Report 5804 P@ ethylene CatWes 25A 291 342 51 Rockbestos Firewellli Rockbestos irradletion Cross Linked Report 5805 P@sthylorm Cables 26 291 450 150 ASCO NP 1 Solenced VaNos ASCO Report AQR47368 i

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Evaluation New QuaBAed Margin (F)

Item Description AppReable item No.

Mw Maximum Test Report DBA Maximum Temperature Temperature (F)

(F) 27 291 346 55 Rockbestos Coaxiel Cable Roch Report QR.

6802 28 Not required for NA NA Rome Cable NA TSCR,see 97318-2 I

29 291 450 159 Rosemount Transmitter Rosemourt Model1154 Report D8400102 4

29A 291 420 129 Rosemount Model1154 wf Rosemount 1159 Remote Seal Report 7/D92

)

1 298 291 446 155 RosemountTransmeer Rosemount Condut Connector Seal Report Model353C D6300200 30 291 1800 1409 incore Thermocoupion ISA MC96.1 1975 r

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31 291 320 29 General Atomic High Range General Atomic Radiation Monitor RD-23 Report E-254-960 j

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32 291 420 169 Gamme.Metrice Neutron Gamma-9 Monitoring Syelem Cable Matrice Report 040 33 291 347 56 Westinghouse Penetrabone Weelinghouse Report PEN.

TR-7749 34 291 307 16 Umitorque MOVs Westinghouse ReportWCAP.

7410-L 5

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Evaluation New QuaMed Margin (F) learn Description Aph llam No.

Pew Maalmum Test Raport DBA Maximum Temperattre Temperature

(*F)

~

M 36 291 307 16 MOV Lutricant Weehnghouse ReportWCAP.

7410-L 36 291 296 5

Engneered Safety Features Weebnghouse 4

% and Spilces PeportWCAP.

l 3

74104., Vol.11 37 291 324 33 Contenmort Accidert Fan Wesunghouse Motor Bearing Lubricart Report WCAP.

7722 38 291 324 33 Cortninment Accadert Fan Wesenghouse Lubricant ReportWCAP.

7722 39 291 345 54 C*onite Cabie Franidin Report F-C3894

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