3F0499-01, Forwards Response to NRC 990304 RAI Re LAR 222,requesting Changes to CREVS & Vftp.Ltr Established No New Commitments

Forwards Response to NRC 990304 RAI Re LAR 222,requesting Changes to CREVS & Vftp.Ltr Established No New Commitments

ML20205M232
Person / Time
Site:	Crystal River
Issue date:	04/08/1999
From:	Holden J FLORIDA POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
3F0499-01, 3F499-1, TAC-MA0667, TAC-MA667, NUDOCS 9904150183
Download: ML20205M232 (9)

Text

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Florida Power 8, PJfa^AT

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/, il 8,1999 L1499-01 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 Response to Request for Additional Information - License Amendment Request #222 Related to Control Room Emergency Ventilation System (TAC No. MA0667)

References:

1. NRC to FPC letter,3N0399-02, dated March 4,1999, " Crystal River Unit 3

- Request for Additional Information - License Amendment Request 222 Related to Control Room Emergency Ventilation System (TAC No, MA0667)"

2. FPC to NRC letter, 3F0798-15, dated July 30,1998, " License Amendment Request #222, Revcion 1, Control Room Emergency Ventilation System and Ventilation Filter Test Program (TAC No. M91823)"

Dear Sir:

In Reference 1, Florida Power Corporation (FPC) was requested to provide additional information regarding License Amendment Request (LAR) #222 (Reference 2). LAR #222 proposes changes to the Improved Technical Specifications (ITS) for the Control Room Emergency Ventilation System (CREVS) and to the Ventilation Filter Test Program (VFTP).

Responses to the requests are provided in the attachment.

This letter establishes no new regulatory commitments. If you have any questions regarding this submittal, please contact Mr. Sid Powell, Manager, Nuclear Licensing at (352) 563-4883.

Sincerely, de I J. J. Iloiden

! Director, Site Nuclear Operations JJil/dah w \

Attachment _ ,

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.c: Regional Administrator, Region 11 NRR Project Manager Senior Resident inspector CRYSTAL RIVER ENERGY COMPLEX: 15760 W. Power une Street + Crystal Wer, Florloa 34428-6708 + (352)7FA486 9904150183 990400 .

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PDR ADOCK 05000302-P PDR , ,

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U.S. Nuclear Regulatory Commission Attachment 3F0499-01 Page 1 of 8 ATTACIIMENT Response to Request for Additional Information - License Amendment Request #222 Related to Control Room Emergency Ventilation System NRC Request 1.a

1. In Florida Power Corporation's (FPC's) license amendment request No. 222 dated July 30, 1998, FPC proposed a new action statement for Technical Specification 3.7.12, Control Room Emergency Ventilation System, that would allow operation to continue for up to seven days with a breach in the Control Complex IIabitability Envelope less than or equal to one square foot in excess of the limit. or 179 in2 . The CR-3 Control Room IIabitability Report only supports a breach of 35.5 in2 . The NRC staff requests the following additional information:

a. The radiological analyses included in the Control Room IIabitability Report determined that a breach of 35.5 in2 could be tolerated and not exceed the criteria of General Design Criterion-19. No analyses of control room doses were performed at the proposed breach size. Limiting parameter values'in the technical specifications are supposed to be based on the plant design basis analyses. The FPC qualitative analysis does not provide the staff with sufficient information to approve this portion of the amendment request. Please provide an evaluation of the radiological consequences of the design basis accidents occurring while a breach of the size proposed in Limiting Condition for Operation (LCO) action 3.7.12.B exists. This should consider the Design Bases Accidents addressed in FPC's Control Room liabitability Report. Personnel acions to seal the breach may be credited in Lis analgis, provided that the time to affe.t these actions is conservatively estimated and formal commitments are made to ensure that necessary personnel and equi pment are available at all times when LCO action 3.7.12.13 is in effect.

Response 1.a The doses due to the existence of an additional breach of one square foot have been determined for each of the accidents analyzed in the Control Room liabitability Report included with License Amendment Request (LAR) #222, Revision 1 (Reference 1). The methodology utilized for this estimate is discussed below.

Detailed dose calculations were performed for both the Maximum liabitability Accident (MilA) with a loss of offsite power (LOOP) and the MHA without a LOOP for the Control Room liabitability Report. Both of these calculations included additional computer calculations of the dose as a function of breach size, such that the breach margin could be determined. Therefore, these calculations include sufficient information to determine the

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' Attachment U.S. Nuclear Regulatory Commusion 3F0499-01 Page 2 of 8 iberemental dose as a function of an incremental breach size. This dose per breach size was extrapolated to determine the dose due to an additional one square foot opening.

The ratio of the new total dose, including the breach margin and one square foot breach, to the base dose was then determined for the MHA calculations. The ratio was then applied to the other accidents to determine an estimated total dose. For the Fuel Handling Accident (FHA),

the ratio of total to base dose is assumed to be the same as for the MHA since normal ventilation is isolated before plume arrival in both of these accidents. Therefore, the only input of activity is due to in-leakage and an equivalent increase in the in-leakage area would have the same relative effect on the dose. For the Steam Generator Tube Rupture (SGTR) and Letdown Line Rupture (LLR), not all of the dose is due to in-leakage because there is an initial input of activity from normal ventilation makeup. The initial input should not be affected by an additional breach, so only the fraction of the dose due to in-leakage would increase by the ratio calculated for the MHA. However, for simplicity, the total dose was conservatively determined by applying the MHA ratio to both contributions.

The estimated total doses based on this methodology as compared to the doses calculated in the Control Room Habitability Report are provided below. The estimated doses include the dose from the base calculation, the dose from the breach margin (which is the breach size that would bring the thyrcid dose up to the 30 REM guideline for the limiting accident), and the dose from an additional breach of one square foot.

By comparison with the doses presented in the Control Room Habitability Report (i.e., no additional one square foot breach assumed), it is observed that the thyroid dose increased by a much larger factor than the whole body and skin dose. The reason the thyroid dose is more sensitive to breach size than the whole body or skin dose is due to the differences in activity removal mechanisms for iodine (thyroid dose) and noble gases (whole body and skin dose).

Activity removal from the control room air occurs as a result of decay, cleanup via filtered recirculation or out-leakage. Out-leakage flow rate is assumed to be equal to in-leakage flow rate during the course of the accident in order to maintain the same pressure differential across the Control Complex Habitability Envelope (CCHE). For iodine, the primary means of removal from the control room air is filtration through the emergency filtration system.

Therefore, an increase in the assumed in-leakage rate, which also results in an equal increase l in the assumed out-leakage rate, will significantly impact the input of iodine. This will have a small effect on the overall removal term, which is dominated by the filter removal factor. For )

noble gases, there is no filtration; thus, the only removal mechanism other than decay is out-leakage. Therefore, an increase in assumed in-leakage will be compensated for by increased removal term due to the corresponding increase in out-leakage.

l With the low sensitivity of the whole body and skin dose to in-leakage rate, all whole body and {

-skin doses remained within the respective 5 REM and 30 REM limits as specified by the application of NUREG-0800, Standard Review Plan (SRP) Section 6.4, " Control Room Habitability System," to General Design Cri.eria (GDC) 19. However, for five of the  ;

analyzed accidents, the calculated thyroid dose is greater than the current limit of 30 REM as j i

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U.S. Nuclear Regulatory Commission . Attachment 3F0499-01 Page 3 of 8 specified by'SRP Section 6.4. The Committed ' Dose Equivalent (CDE), as defined by 10 CFR 20, for the thyroid dose was calculated for each analyzed accident. The CDE for each' accident was well below the 5 REM limit.

Thyroid dose to the control room operators may also be controlled by the use of potassium iodine (KI) pills. Provisions exist in the Emergency Plan Implementing Procedures for -

considering administration of KI to personnel based on projected dose. A projected dose of 25 REM to an individual has been established as the threshold for considering administration of KI, but it may be administered at lower doses as deemed appropriate by Medical and Dose -

Assessment Personnel staffing the emergency response organization.

FPC believes that the additional one square foot breach allowed by the action statement is acceptable based on the following reasons:

• Design basis calculations performed at the design basis condition breach limit of 35.5 in 2allowed by the Limiting Conditions for Operation (LCO) show that the SRP Section 6.4 application of GDC-19 limits of 5 REM whole body, 30 REM thyroid, and 30 REM skin dose are satisfied.

• Considering the frequency of a catastrophic reactor accident involving fuel damage, the probability of this type of event occurring during the relatively short allowance time in Condition B (7 days) is very small.

. The consequences (dose) to the control room operators for the analyzed accidents considering the increased breach allowance of one square foot allowed by Condition B are within the SRP Section 6.4 application of GDC-19 limits for whole body and -

skin dose.

• The increase in consequences (dose) to the control room operators for the analyzed '

accidents considering the increased breach allowance of one square foot allowed by Condition B is acceptably low for thyroid dose when considering the' calculated CDE as defined by 10 CFR 20.

. Additional protection is afforded to the control room operators with' the existing provisions for administration of KI pills.

• There is no increase to the site boundary radiological dose for this proposed change and, therefore, no increase in consequences to the public.

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i U.S. Nuclear Regulatory Commission ~ . Attachment 3F0499-01 Page 4 of 8 RADIOLOGICAL DOSE RESULTS OF ADDITIONAL BREACH OF ONE SQUARE FOOT Radiological Dose (REM)

Analyzed Accident -

With Breach Margin CRHR' Plus Additional One Square Foot Breach Thyroid / CDE 18.6 69 / 2.1 With LOOP Whole Body 0.4 0.62 Skin 15 21.8 MilA2 Thyroid / CDE 19.6 73.2 / 2.2 Without LOOP Whole Body 0.7 1.17 Skin 18 27.4 Thyroid / CDE 6.61 24.5 / 0.7 With LOOP Whole Body 0.0224 ' O.04 Skin 1.62 2.4 LLR' Thyroid / CDE 12.4 45.9 / 1.4 Without LOOP Whole Body 0.0558 0.09' Skin 4.04 6.1 Thyroid / CDE 9.6 35.5 / 1.1 SGTR' Whole Body 0.006 0.01 Skin 0.35 0.5 Thyroid / CDE 11.9 44 / 1.3 Without release filtration Whole Body 0.049 0.08 Skin 3.48 5.2 Fila 3 Thyroid / CDE 2.98 11 / 0.3 With release filtration Whole Body 0.049 0.08 Skin 3.48 5.2 8

Control Room liabitability Report (Attachment B of Reference 1) 2 Calculated by extrapolation 2

Calculated with largest ratio of Mil A events (Thyroid - 3.7. Whole Body - 1.7. Skin - 1.5)

NRC Request 1.b

b. On Page 6 of 13 of the submittal, FPC states that "This position is similar to one approved for the Waterford-3 Technical Specifications by License Amendment No.

115, issued October 4,1995." The use of Waterford-3 as a precedent does not appear valid due to differences in the configuration of the control rooms (e.g.,

Waterford employs a zone filtered pressurization design with two widely separated intakes) and differences between the proposed language of the Crystal River Unit 3 (CR-3) technical specification and that approved for Waterford. Please explain why

4 h . U.S. Nuclear Regulatory Commission . Attachment .

3F0499-01 Page 5 of 8

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FPC believes that the Waterford proceeding is a valid precedent for the proposed CR-3 amendment 'given these differences.

Response 1.b FPC is aware of the differences between CR-3 and Waterford 3 in both the configuration of the CC11Es and the language of the referenced Technical Specifications. The statement 'on page 6 of 13 of Reference 1 provided a reference to a previously approved license amendment with :

similar provision (not position as stated in the.NRC request). The similarity exists in the Technical Specification Action that permits breaches in the CCHE for a period not tol exceed seven days.

The justification for the proposed CR-3 ITS change is provided in Reference I with additional '

information provided in the responses contained in this submittal. FPC believes that the.CR-3

. ITS change is warranted on its own merits and is not dependent on the previous approval of the Waterford 3 license amendment. The Waterford 3 amendment was referenced as a similar provision to facilitate NRC Staff review.

As stated in Reference 1, the proposal for an allo <ance of less than or equal to one additional square foot of breach margin, or its equivalent in-leakage, for up to seven days would provide significantly increased flexibility.in planning and scheduling. work activities. As provided in the analysis described in Response 1.a above, the estimated dose to CR-3 control room-operators with the existence of the preposed one square foot bleach cllowance open for the 30 day radiological accident period has been calculated. The calculation demonstrates that the conservatively calculated dose in accordance with 10 CFR 20 would not exceed the equivalent of 5 REM to the whole body. Therefore, the configuration of the CR-3 control room, even with an additional one square foot breach, provides adequate protection for plant operators.

The breach allowance is intended to be used for two purposes. The first is to accommodate maintenance and modification of the habitability envelape during routine plant evolutions.

Such activities are controlled in accordance with existing administrative procedures which require authorization and tracking of breaches to assure total breach size is not exceeded.

Accordingly, these oreaches can be closed in a timely manner following a radiological accident and, thus, would significantly reduce the potential dose contribution. No credit for closing.

these breaches is taken in the dose calculations discussed in Response 1.a.

The second use is to permit habitability envelope leakage integrity tening (required once per operating cycle) to be performed during plant operation without the tMeat of immediate plant shutdown for a small additional breach area. In the event of a test failure revealing leakage

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2 equivalent to' breaches up to 179 in , the location of the breches would not be known and could not be closed immediately. The risk cf allowing the leakage :quivalent to a one square

- foot breach to exist for seven days once per operating cycle for leaktge testing is considered to be insignificant. In the event a design basis radiological accident was to occur, doses resulting -

' from increased in-leakage are acceptable and would not require evacuation of the control room.

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U.S. Nuclear Regulatory Commission - Attachment 3F04994)1 Page 6 of 8 The calculations for CR-3 show that for the sources of toxic gas administratively limited to be '

stored onsite, and without automatic isolation of the control room outside air intake,' toxicity limits will not be exceeded in the control room within two minutes of nasal detection. The outside air intake rate used in those calculations is 5700 cfm. The in-leakage rate equivalent to a one square foot breach, with normal Control Complex and Auxiliary Building ventilation still in service, is insignificant compared to the normal supply flow. Note that the radiological calculation assumed an additional 1154 cfm through the.one square foot breach with a 0.2 inch water gauge pressure differential due to emergency mode of Auxiliary Building. ventilation system operation. With normal Auxiliary Nilding ventilation operating, there is expected to be a lower differential pressure with co.. spondingly lower in-leakage. Furthermore, the construction of the CCllE is such that only one wall and the roof are directly exposed to the outside atmosphere. The wall has no penetrations exposed to the outside, and the roof penetrations are sealed against the weather and possible standing water. All other penetrations -

into the habitability envelope are located in either the Turbine, Intermediate, or Auxiliary .

baildings. These structures would slow the buildup of toxic gas in the control room due to breach in-leakage. Therefore, the contribution to toxic gas concentration in the control room due to the existence of breaches is insignificant.

FPC is requesting the seven day allowance to permit performance of tracer gas testing of the habitability envelope to determine unfiltered in-leakage during power operation, without the threat of immediate plant shutdown. Experience with the performance of a tracer gas test at CR-3 and data reduction to determine in-leakage, indicates that approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is required per test. Due to the large volume of the CR-3 habitability envelope (~365,000 cubic feet encompassing 5 elevations), examination of the envelope for sources of leakage would require approximately 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Seven days would permit two rounds of testing.

Considering the frequency of a catastrophic reactor accident involving fuel damage,'the probability of this type of event occurring during this relatively short allowance time is very small. Additionally, the consequences to the control room operators for the analyzed accidents considering the increased breach allowance has been shown in Response 1.a to be acceptably low. There is no increase to the site boundary radiological dose for this proposed change and, therefore, no increase in consequences to the public. Based on these reasons, the risk increase for this allowance is considered to be insignificant.

NRC Request 2

2. FPC's analysis of the fuel handling accident assumes, as an initial condition, that the control room emergency ventilation system (CREVS) is in the emergency (filtered) recirculation mode prior to moving irradiated fuel assemblies. Proposed adtion 3.7.12.D

- (which applies during movement of irradiated fuel) contains a note requiring the CREVS be placed in the emergency recirculation mode if automatic transfer to emergency recirculation is inoperable. Please explain why this statement is required since the accident analysis assumes that the CREVS is in emergency recirculation before irradiated fuel is moved.

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1. .- U.S. Nuclear Reg $latory Commission Attachment

3F0499-01 -

Page 7 of 8 Ilespon$e 2 FPC's analysis of the FHA is described in Section VII.3 of the Control Room Habitability'-

Report. Piocedural requirements ensure that the CREVS is placed in.the recirculation mode prior to and bring any irradiated fuel movement. This requirement is' for recirculation mode (unfiltered), and not emergency (filtered) recirculation mode as stated in the NRC request.

The FHA ~ analysis assumes that the recirculation filter is manually actuated after a thirty-minute delay.

Prior to 1996, the control room habitabil'ity calculations had only been performed for the :

design basis accident (DBA) loss of coolant accident (LOCA) scenario. There was no analysis of the control room dose for a FHA. Control room habitability calculations for other accidents, including the FHA, were performed in 1996-1998. It wa only these recent calculations that demonstrated the need to have the CREVS in recirculation prior to plume arrival in order to meet the dose guidelines for a FHA. The compensatory action taken was to irevise operating procedures to ensure that CREVS was placed in recirculation prior to and during irradiated fuel movement or heavy loads over irradiated fuel.

Prior to these revisions, there were no procedural controls or Technical l Specification requirements to place the CREYS in recirculation prior to fuel movement. Therefore, the Note in the Technical Specifications served a purpose. With the implementation.of the recent

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procedural controls, which are more reestrictive than the Technical Specifications, the' Note-serves no purpose as the system would already be in recirculation whenever irradiated fuel is L moved. The Note does not conflict with the procedural requirements and, therefore, has been left in the Technical Specifications as future changes to the analysis could result in times when the system may not have to be placed in recirculation (e.g., when moving irradiated fuel that has decayed for more than 30 days).

NRC Request 3

3. Item M of Table 1 of the Control Room Habitability report appears to contain an error.

T1. mixing rate is specified as 46,400 cfm. The mixing rate is usually specified as being eg' 1 to two times the volume of the unsprayed region per hour. -For CR-3 this is 23,200 r . FPC calculation M9'7-0110 Rev. 2 appears to have used 23,200 cfm as the mixing ute. Please resolve this apparent discrepancy.

Response 3 Item M of Table 1 reflect ithe mixing rate (46,400 cfm) that was used in the current analysis for a LOCA without a LOOP. Th: analysis is documented in Calculation M-97-0137,

- Revision 4, " Control Room Habitabila, Analysis Considering LOCA without LOOP."~ The-value for the mixing rate shoul:1 have been 23,200 cfm as was used in the current analysis for a ,

LOCA with a' LOOP (M-97-0110, Revision 4, " Control Room Dose Analysis and Maximum

Infiltration Following La LOCA- with - LOOP"). However, a comparison performed in 4 .

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U.S. Nuclear Regulatory Commission Attachment-3F0499 Page 8 of 8 M-97-Oi37 duplicated the base case of M-97-0110,~ Revision 0, and showed that the minor

- differences in assumptio'ns and models, along with the higher mixing rate of 46,400, resulted in a less than 2 percent change in results.' Therefore, the change in mixing rate is considered-to have negligible difference in the final results.

Reference

1. FPC to NRC letter, 3F0798-l'5, dated _ July L-30, 1998, " License Amendment .

Request #222, Revision 1, Control Room Emergency Ventilation System and' Ventilation Filter Test Program (TAC No. M91823)"

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