L-08-293, License Amendment Request No. 08-029, Credit for Containment Overpressure

License Amendment Request No.08-029, Credit for Containment Overpressure

ML083170522
Person / Time
Site:	Beaver Valley
Issue date:	11/07/2008
From:	Sena P FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-08-293
Download: ML083170522 (18)

Text

FENOC Beaver Valley Power Station NP.O.

Box 4 FirstEnergy Nuclear Operating Company Shippingport, PA 15077 Peter P. Sena 1//

724-682-5234 Site Vice President Fax: 724-643-8069 November 7, 2008 L-08-293 10 CFR 50.90 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Beaver Valley Power Station, Unit No. 2 Docket No. 50-412, License No. NPF-73 License Amendment Request No.08-029, Credit for Containment Overpressure Pursuant to 10 CFR 50.90, FirstEnergy Nuclear Operating Company (FENOC) hereby requests an amendment to the operating license for Beaver Valley Power Station (BVPS) Unit No. 2. The proposed amendment would modify the method used to calculate the available net positive suction head (NPSH) for the Unit No. 2 recirculation spray (RS) pumps as described in the Beaver Valley Power Station (BVPS) Unit No. 2 Updated Final Safety Analysis Report (UFSAR). The proposed change will revise the Unit No. 2 UFSAR to take credit for containment overpressure by allowing for the difference between containment total pressure and the vapor pressure of the water in the containment sump in the available NPSH calculation. The FENOC evaluation'of the proposed change is provided in the Enclosure to this letter. This license amendment request does not require any changes to the Technical Specifications.

To address Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors," a change in methodology is required for calculating the available NPSH for the Unit No. 2 RS pumps. Title 10, part 50, section 50.59(c)(2)(viii) of the Code of Federal Regulations (CFR) requires that a licensee not make changes that result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses, without first obtaining a license amendment pursuant to 10 CFR 50.90. FENOC has evaluated the proposed methodology change under 10 CFR 50.59 and determined that NRC approval of the change is required.

This change has been reviewed by~the Beaver Valley Power Station review committees.

The change was determined to be safe and does not involve a significant hazard consideration as defined in 10 CFR 50.92 based on the attached safety analysis and no

,U(lo significant hazard evaluation.

Beaver Valley Power Station, Unit No. 2 L-08-293 Page 2 The referenced letter informed the NRC that additional debris and chemical effects testing would be required for BVPS Unit No. 2 to comply with Generic Letter 2004-02.

This testing is presently in progress and is scheduled to be completed in December of this year. The referenced letter also states that a license amendment request to credit containment overpressure for BVPS Unit No. 2 would be submitted for NRC approval by November 9, 2008 and that a follow-up supplemental response to Generic Letter 2004-02 including the results of the BVPS Unit No. 2 debris and chemical effects testing (which will include credit for containment overpressure), results of downstream effects analyses (both in-vessel and ex-vessel), the effects on NPSH margins, and details of corrective actions would be submitted to the NRC by April 30, 2009. In order to meet the April 30, 2009 commitment, FENOC requests approval of this license amendment request by February 27, 2009. FENOC plans implementation of containment overpressure credit as part of the BVPS Unit No. 2 licensing basis upon NRC approval of this license amendment request.

There are no regulatory commitments contained in this letter. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager -

Fleet Licensing, at 330-761-6071.

I declare under penalty of perjury that the foregoing is true and correct. Executed on November -7, 2008.

Sincerely, Peter P. Sena III

Reference:

FENOC Letter L-08-257, Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors" - Request for Extension of Completion Date for Corrective Actions (TAC Nos. MC4665 and MC4666), dated August 28, 2008

Enclosure:

FENOC Evaluation of the Proposed Change

Beaver Valley Power Station, Unit No. 2 L-08-293 Page 3 cc:

Mr. S. J. Collins, NRC Region I Administrator Mr. D. L. Werkheiser, NRC Senior Resident Inspector Ms. N. S. Morgan, NRR Project Manager Mr. D.'J. Allard, Director BRP/DEP Mr. L. E. Ryan (BRP/DEP)

FENOC Evaluation of the Proposed Changes Beaver Valley Power Station Unit No. 2 License Amendment Request 08-029

Subject:

Credit for Containment Overpressure Section Title Page 1.0

SUMMARY

DESCRIPTION........................................................

1 2.0 DETAILED DESCRIPTION........................................................

1

3.0 TECHNICAL EVALUATION

2

4.0 REGULATORY EVALUATION

4 4.1 Significant Hazards Consideration.................................................

5 4.2 Applicable Regulatory Requirements/Criteria.............................. 6 4.3 P re ce d e nt.................................................................................

8 4.4 C o nclusio ns...............................................................................

.. 8

5.0 ENVIRONMENTAL CONSIDERATION

8 6.0 R E FE R E N C E S............................................................................

9 Attachment 1

Markups of Unit No. 2 UFSAR Pages

Beaver Valley Power Station Unit No. 2 License Amendment Request 08-029 Page 1 of 9 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to change the method used to calculate the available net positive suction head (NPSH) for the Unit No. 2 recirculation spray (RS) pumps as described in the Beaver Valley Power Station (BVPS) Unit No. 2 Updated Final Safety Analysis Report (UFSAR).

Regulatory Guide (RG) 1.82, Revision 0, "Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident," establishes the current BVPS Unit No. 2 licensing basis of assuming 50% blockage of the containment sump screens. However, to address Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors" (Reference 1), a change in methodology is required for calculating the available NPSH for the Unit No. 2 RS pumps. Title 10, part 50, section 50.59(c)(2)(viii) of the Code of Federal Regulations (CFR) requires that a licensee not make changes that result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses, without first obtaining a license amendment pursuant to 10 CFR 50.90. FENOC has evaluated the proposed methodology change under 10 CFR 50.59 and determined that NRC approval of the change is required. Therefore, this license amendment request (LAR) is being submitted for NRC approval.

The proposed change will revise the Unit No. 2 UFSAR to take credit for containment overpressure by allowing for the difference between containment total pressure and the vapor pressure of the water in the containment sump in the available NPSH calculation.

The proposed methodology change does not require a change to the Technical Specifications, Technical Specification Bases or Licensing Requirements Manual.

2.0 DETAILED DESCRIPTION The method used to calculate available NPSH for the RS pumps is described in Section 6.2.2.3.2 of the Unit No. 2 UFSAR and includes an assumption that the containment pressure is equal to the vapor pressure of the water in the containment sump, consistent with Standard Review Plan Section 6.2.2. The proposed change will revise the UFSAR to take credit for containment overpressure by allowing for the difference between containment total pressure and the vapor pressure of the water in the containment sump in the available NPSH calculation. This change in methodology provides a more realistic calculation of the available NPSH for the Unit No. 2 RS pumps (removing an unrealistic, conservative assumption), and is consistent with the methodology used for Unit No. 1. The proposed changes to the UFSAR requested by this License Amendment Request (LAR) are described in Attachment 1.

Beaver Valley Power Station Unit No. 2 License Amendment Request 08-029 Page 2 of 9

3.0 TECHNICAL EVALUATION

3.1 Background

The current methodology for calculating available NPSH (referred to as NPSHA in the UFSAR) for the RS pumps at Unit No. 2 is described in Section 6.2.2.3.2 of the UFSAR as follows:

"Sufficient NPSH is available to the recirculation pumps during both the recirculation spray mode and the recirculation mode of low head safety injection. The following equation is used to calculate the NPSHA.

NPSHA = Pc + Z - Hf-Pv where:

Pc =Containment atmosphere total pressure Z

Elevation head of water above first stage impeller Hf Head loss from friction in pump suction pipe Pv =Vapor pressure of sump liquid (saturation pressure at liquid temperatures)

All preceding parameters are expressed in feet of head.

This expression can be simplified by making the conservative assumption that the vapor pressure of the pumped liquid is equal to the total containment pressure as follows:

NPSHA = Z - Hf" 3.2 Proposed Change The proposed change to the method used to determine available NPSH will eliminate the assumption that the vapor pressure of the pumped liquid is equal to the containment atmosphere total pressure. Therefore, the full equation, NPSHA = P, + Z - Hf-Pv, will be used to calculate available NPSH. Each term in the equation can be calculated explicitly using the licensed MAAP-DBA code on a transient basis for a given loss of coolant accident (LOCA) scenario. This allows for the calculation of a time history of available NPSH which can then be used to determine the minimum value that is compared to the required NPSH to demonstrate acceptable pump operation. This is the same methodology currently approved and in use at Unit No. 1.

Beaver Valley Power Station Unit No. 2 License Amendment Request 08-029 Page 3 of 9 3.3 Change Evaluation The current Unit No. 2 licensing basis, which assumes that the total containment pressure is equal to the vapor pressure of the pumped liquid, is unrealistic when sump vapor pressure is below the initial containment air pressure. For relatively cool sump water conditions, no credit is given for the sub-cooling of the sump liquid, which is analogous to assuming that some or all of the air leaked out of containment. Since head loss across a debris bed on the sump strainers is very sensitive to temperature (as well as flow), the head loss increases substantially at lower sump temperatures due to the increased viscosity. The combined effect of the increased viscosity and the unrealistic containment pressure assumption at cooler temperatures reduces the calculated value of available NPSH.

For small break LOCA cases when the sump temperature at RS pump start is approximately 120 0F, the associated sump vapor pressure would be 1.7 psia. In this case, if the vapor pressure is equal to total containment pressure, the containment pressure must be assumed to be 1.7 psia. Such an assumption is unrealistic given that the initial containment minimum air pressure prior to the onset of the loss of coolant accident (LOCA) will be at least 12.8 psia. The assumption therefore results in a penalty of approximately 25 feet of available NPSH when compared to a calculated value of available NPSH which includes the initial air pressure for the containment pressure term. Therefore, it is more realistic and appropriate to explicitly calculate the containment total pressure and sump vapor pressure, and then use these values in the calculation of available NPSH for the RS pumps.

The MAAP-DBA program used for the containment analyses has the capability of calculating each of the terms in the available NPSH equation. Therefore available NPSH can be explicitly calculated on a transient basis. Containment pressure, sump water inventory and sump temperature are calculated based on the thermodynamic conditions in containment. A volume versus height table is utilized to calculate the containment sump level based on the sump inventory. This establishes the static height of water above the pump suction at any point in time. Pump suction friction losses are calculated based on the hydraulic loss factors and the calculated pump flow. These losses also include the containment sump strainer head loss, including the contribution due to debris loading, which is calculated based on the total strainer flow and sump water temperature.

Analyses performed using MAAP-DBA demonstrate that even for those cases which result in the highest sump temperatures and lowest containment pressures, some level of sub-cooling remains in the sump throughout the transient. This conclusion is true even when all input parameters are biased to minimize containment pressure while maximizing sump temperature. One of the more significant conservative assumptions included in the available NPSH analyses is that the two streams from a double-ended guillotine break are mixed prior to flashing to steam at the containment pressure. This treatment of the LOCA mass and energy release maximizes the enthalpy of the liquid phase (which is transported to the sump) and minimizes enthalpy of the vapor phase (which is released to the containment atmosphere and increases its pressure). The net

Beaver Valley Power Station Unit No. 2 License Amendment Request 08-029 Page 4 of 9 effect of this retained conservative assumption alone is a reduction in the available NPSH by approximately 15 feet for the limiting case.

The proposed change in the Unit No. 2 methodology only impacts the reported results of available NPSH for the RS pumps. It will not impact the results of any other.

containment analyses such as peak containment pressure and temperature from a LOCA or main steamline break (MSLB), depressurization times, containment liner temperatures, etc. The reported sump water temperatures will also not be impacted since the release stream mixing assumption is already included in these calculations.

Analyses using the proposed methodology cannot be completed until head loss testing of the BVPS-2 strainers is complete; however, it is expected that crediting containment overpressure will be required to satisfy NPSH requirements for only a limited time following a LOCA (less than one hour). The available containment overpressure (and available NPSH) is conservatively calculated assuming parameters that are biased simultaneously in the conservative direction. Further, with respect to margin and equipment performance, testing of the RS pumps has demonstrated\\that the pumps are capable of stable operation at-conditions where available NPSH is reduced below the standard (3% reduction in head) requirement.

To further demonstrate that there is an insignificant risk impact and that the current Unit No. 2 Probabilistic Risk Assessment (PRA) model adequately accounts for containment overpressure requirements, a study was done to determine the impact of operation of the RS pumps under accident conditions with a failure of containment isolation for systems which communicate directly with the containment atmosphere. The largest such penetration at Unit No. 2 is two (2) inches in diameter. The results of the study show that even with this largest penetration failure, there is no significant impact on the available NPSH. The estimated frequency of a containment isolation failure coincident with a large break LOCA is approximately 6.2 E-9. The change in risk associated with crediting containment overpressure is insignificant. This factor therefore does not need to be considered in the Unit No. 2 PRA model.

4.0 REGULATORY EVALUATION

FirstEnergy Nuclear Operating Company proposes to revise the method used to calculate available net positive suction head (NPSH) for the Beaver Valley Power Station (BVPS) Unit No. 2 recirculation spray (RS) pumps which take suction from the containment sump. Net positive suction head'is defined as the sum of containment pressure and static head of the sump fluid minus the sump fluid vapor pressure and suction piping friction losses. The method described in the Unit No. 2 Updated Final Safety Analysis Report (UFSAR) for calculating available NPSH conservatively assumes that containment pressure is equal to the vapor pressure of the sump fluid, allowing both of these terms to be neglected. The proposed change, and subsequent amendment, will allow available NPSH to be calculated as the sum of containment pressure and static head of the sump fluid minus the sump fluid vapor pressure and suction piping friction losses.

Beaver Valley Power Station Unit No. 2 License Amendment Request 08-029 Page 5 of 9 Title 10, part 50, section 50.59(c)(2)(viii) of the Code of Federal Regulations (CFR) requires that a licensee not make changes that result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses, without first obtaining a license amendment pursuant to 10 CFR 50.90.

FENOC has evaluated the proposed methodology change under 10 CFR 50.59 and determined that NRC approval of the change is required.

4.1 Significant Hazards Consideration FirstEnergy Nuclear Operating Company has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The change to the method used to calculate available NPSH for the RS pumps will not affect the probability of an accident because the RS pumps are not used during normal plant operations and cannot initiate an accident.

Successful operation of at least one train of RS pumps is required in order to demonstrate that containment and fuel cladding design basis limits are not exceeded. The design basis accident currently assumes a breach of the reactor coolant pressure boundary. There is no impact to the fuel cladding since the proposed change does not affect performance of the emergency core cooling systems. Successful operation of the RS pumps depends on adequate NPSH being available to support RS pump performance. The change in the methodology will result in an increase of the NPSH available to the RS pumps as calculated in the safety analysis. This will increase the calculated NPSH margin because the required NPSH to the RS pumps will not change due to the methodology change. Because the available NPSH remains adequate, with margin to NPSH requirements, acceptable RS pump performance will be assured and the design basis limits for containment pressure and fuel cladding will not be exceeded and the consequences of an accident will not be increased.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Beaver Valley Power Station Unit No. 2 License Amendment Request 08-029 Page 6 of 9

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The change to the method used to calculate available NPSH for the RS pumps will not create the possibility of a new accident because the operation of the plant or the RS pumps is not changed. The RS pumps are not used during normal plant operations and cannot initiate an accident. A different kind of accident will not be created because the proposed calculation method will produce an NPSH value that will ensure proper operation of the pumps and will not result in any new failure modes of the RS pumps.

Therefore, the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The change to the method used to calculate available NPSH for the RS pumps will not involve a significant reduction in a margin of safety because the change does not reduce the NPSH margin to the RS pump required NPSH.

The only controlling numerical value pertaining to available NPSH of the RS pumps that is established in the UFSAR is a lower limit specified in the UFSAR, referred to as the required NPSH for the RS pumps. The required NPSH limit will not be altered as a result of the proposed calculation method, and the required NPSH will continue to be maintained under the applicable accident scenario.

Therefore, the proposed amendment will not involve a significant reduction in a margin of safety.

Based on the above, FirstEnergy Nuclear Operating Company concludes that the proposed methodology change does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.2 Applicable Regulatory Requirements/Criteria The change described in this license amendment request is consistent with the following regulations and regulatory guides.

Beaver Valley Power Station Unit No. 2 License Amendment Request 08-029 Page 7 of 9 General Design Criterion (GDC) 35, "Emergency Core Cooling," states:

A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling -is prevented and (2) clad metal-water reaction is limited to negligible amounts.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

GDC 38, "Containment Heat Removal," states:

A system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other. associated systems, the containment pressure and temperature following any loss-of-coolant accident and maintain them at acceptably low levels. Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

GDC 38, "Containment Heat Removal" and Standard Review Plan (SRP) Section 6.2.2, "Containment Heat Removal Systems" relate to the capability of the containment system to accomplish its safety function. The SRP indicates that the spray system should be designed to accomplish this without pump cavitation occurring and that a supporting analysis should be presented in sufficient detail to permit the staff to determine the adequacy of the analysis and should show that the available NPSH is greater than the required NPSH.

Regulatory Guide (RG) 1.82, Revision 3, "Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident," Section 1.1.2, states, in part, that debris that could accumulate on the sump screen should be minimized. The RG also indicates that predicted performance of the emergency core cooling systems (ECCS) and the containment heat removal pumps should be independent of the calculated increases in containment pressure caused by postulated LOCAs in order to ensure reliable operation under a variety of possible accident conditions. However, for some operating reactors, some credit for containment accident pressure may be necessary.

The above criteria continue to be met, or are not impacted, by the proposed change to the available NPSH calculation methodology.

Beaver Valley Power Station Unit No. 2 License Amendment Request 08-029 Page 8 of 9 4.3 Precedent The precedent for the proposed change is Unit No.1 that has used the proposed method to determine available NPSH since initial plant licensing, with some refinements approved as part of Unit No. 1 License Amendment 28, dated August 27, 1980 (Reference 2). Along with various plant modifications, the major refinements used to determine available NPSH to the RS pumps consisted of assuming: 1) spray thermal effectiveness of 100%; 2) low initial containment pressure, 3) high initial containment temperature, and 4) using the pressure flash method instead of the temperature flash method in the containment analysis for evaluation of NPSH. The Unit No. 1 methodology has since been reviewed by the NRC as part of the BVPS response to NRC Generic Letter 97-04 (Reference 3), and the Containment Conversion (Reference 4) and Extended Power Uprate (Reference 5) license amendments.

4.4 Conclusions Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment will change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Beaver Valley Power Station Unit No. 2 License Amendment Request 08-029 Page 9 of 9

6.0 REFERENCES

1.

Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors,"

dated September 13, 2004

2.

NRC letter dated August 27,1980, "SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 28 TO FACILITY OPERATING LICENSE NO. DPR-66," Unit No. 1 Amendment 28

3.

Duquesne Light Company Letter L-98-001, dated January 6, 1998, "NRC Generic Letter 97-04"

4.

NRC letter dated February 6, 2006, "BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2 (BVPS-1 AND BVPS-2) - ISSUANCE OF AMENDMENT RE:

CONTAINMENT CONVERSION FROM SUBATMOSPHERIC TO ATMOSPHERIC OPERATING CONDITIONS, (TAC NOS. MC3394 AND MC3395)," Amendments 271 and 153

5.

NRC letter dated July 19, 2006, "BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2 (BVPS-1 AND BVPS-2) - ISSUANCE OF AMENDMENT REGARDING THE 8-PERCENT EXTENDED POWER UPRATE, (TAC NOS.

MC4645 AND MC4646)," Amendments 275 and 156 4

Markups of Unit No. 2 UFSAR Pages Beaver Valley Power Station, Unit No. 2 License Amendment Request No.08-029 The following is a list of the affected pages:

Table 1.8-1 Page 1 of 91 6.2-52

• 6.2-53 6.2-54

• No Change. Page provided for context only.

BVPS-2 UFSAR Rev.

4-*TBD TABLE 1.8-1 USNRC REGULATORY GUIDES RG No.

1.1, Rev.

0 UFSAR Reference Section 6.2.2, 6.3 NET POSITIVE SUCTION HEAD FOR EMERGENCY CORE COOLING AND CONTAINMENT HEAT REMOVAL SYSTEM PUMPS (NOVEMBER 2, 1970)

Beaver Valley Power Station Unit 2 meets the intent of Regulatory Guide 1.1 for providing adequate net positive suction head (NPSH) for emergency core cooling and containment heat removal systems pumps with the following alternatives:

The containment pressure and sump vapor pressure are calculated explicitly on a

transient basis and used to calculate the av*i]1h~e NP*I4 for t-he Rec~rcm~ation grrav rnumns as described in Section 6.2.2.3.2.

The -viapor presdurro of th4=wat=

in the sump is assumoe to be equal to the.

.ntainmont pressure for the rocir'ulation phasc.

The statie head and suetien line prooosure drop is eonsidorod for-the emorgonoy eoro cooling system, in additien to assoufing tha the vapor pressure of the water in.

he

-oumfp is oqual to h

eontainmont prossure.

Tho vapor prtotlure of the.

.ump water ann. t o.d h

eentainmoint total pressure, thoroforo, assufming they are equal gives the limiting low valuo of available NPSH.

RG No.

1.2, Rev.

0 UFSAR Reference Section 5.3 THERMAL SHOCK TO REACTOR PRESSURE VESSELS (NOVEMBER 2, 1970)

The guidance provided by this regulatory guide regarding Section 5.3 thermal shock to the reactor pressure vessel is followed for Beaver Valley Power Station - Unit 2 with the following clarification:

Paragraph C.3 The vessel design does not preclude the use of an engineering solution to assure adequate recovery of the fracture toughness

'properties of the vessel material.

If additional margin is

needed, the reactor vessel can be annealed.

This solution was shown to be feasible by EPRI' program RP1021-1, "Feasibility and Methodology for Thermal Annealing an Embrittled Reactor Vessel."

RG No.

1.3, Rev.

2 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT ACCIDENT FOR BOILING WATER REACTORS (JUNE 1974)

This regulatory guide is not applicable to Beaver Valley Power Station

- Unit 2.

1 of 91

No Change. Provided for context only.

BVPS-2 UFSAR Rev.

16 impaired and approach flow velocities are low enough to prevent entrainment of most small particles.

System design allows for 50 percent plugging or loss of function of one-half of the sump.

Figure 6.2-128 illustrates plan and elevation views of the containment sump screen assembly.

The maximum elevation of water inside the containment following a LOCA is 708. ft-l in, which is approximately 12 ft-l in above the top of the screen assembly.

The design of the containment sump is in accordance with Regulatory Guide 1.82, with the following exceptions and justification for each:

1. The recirculation spray pumps take a suction from a single sump.

However, the sump is physically separated into two halves by trash bars and mesh screen.

The redundant recirculation pump suctions are located in separate halves of the sump.

2.

A portion of the containment floor slopes down toward the

sump, but a raised lip is

provided, which directs normal floor drainage to the segmented section of the containment sump and will prevent small debris from being swept directly into the sump.

Heat is transferred to the structural steel and concrete inside the containment by conduction and condensation.

The energy absorption rate by the passive heat sinks is shown on Figure 6.2-129.

Heat is also removed from the containment atmosphere by the quench and recirculation sprays and is transferred to the containment sump water.

The energy in. the containment sump water is transferred to the service water via the recirculation spray coolers.

The energy removal rate by the recirculation spray coolers from sump water is shown on Figures 6.2-130 and 6.2-130A.

Minimum ESF is assumed for this calculation.

6.2.2.3.2 Net Positive Suction Head Available to Recirculation Pumps Sufficient NPSH is available to the recirculation pumps during both the recirculation spray mode and the recirculation mode of low head safety injection.

The following equation is used to calculate the NPSHA.

NPSHA

= Pc + Z -

Hf

-Pv where:

Pc Containment atmosphere total pressure Z

Elevation head of water above first stage impeller 6.2-52

BVPS-2 UFSAR Rev. 4r& TBD

ý I

Hf

=

Head loss from friction in pump suction pipe Pv

=

Vapor pressure of sump liquid (saturation pressure at liquid temperatures)

All preceding parameters are expressed in feet of head.

This cxprss.ion can be implified by making the c.nscr.ativ.

assumoptin that the vapor press~ure of the pu.. pe.

liquid is equal t.

the total containment pre..ur. as f.llw...

NPSIIA Z

H*

It should be noted that NPSH is referenced to the inlet of the first stage impeller.

Table 6.2-59 compares DBA LOCA NPSHA to required NPSH for recirculation pump at start of recirculation spray, assuming minimum ESF.

The DBA LOCA NPSHA is conservatively low; the actual NPSHA will always be higher for the following reasons:

1.

As time progresses, containment water level used to calculate the elevation head will continually increase due to quench spray and break effluent release inside containment.

2.

The sumfp water is saubcoled with respect to the satur I

1 efmneratuare crroaconelna to containmonra atmoeatricr pressu-ro.

J.- ---

.2 I

at ion

42.

Spray and condensed water hold up at different floors and cavities inside containment is conservatively calculated and subtracted from the sump water inventory.

Condensed water film on heat sink surfaces is also subtracted from the sump water inventory.

4ý3.

The time delays for spray, drainage, and condensed water to reach the sump are considered.

Fudrtheormore, the NPSH is

.C

-i O0 11 eqgui rement of 1S5 ft= is based on a 9aumfp watr 0

I:

"*'t-L2[Ilpert-i~

el th eulsa~

--era

-h te recirculation sprcay m~ay be abeve 24:9 0F.

Aeeording to Figu-re 61 of the Hydraulic institute Standards, Thirteenth Edition

1975, the NPSH reeluired is roduccd at this higher water temperature by morece than a.5S ft.

For these reasons, i.e.,

conservatively low NPSHA and conservatively high NPSH required, a

higher margin between available and required NPSH is assured.

6.2-53

Rev. 4 TBD BVPS-2 UFSAR The results shown are net significeantly affected by breale location or-initial ntainment eonditions T..

Te a.umptin that the vapor pressure of the lieluid in the sump is equal to the containment tota pressure olimainates any effocts en the calculation of availability of

ECCS, spillage te.mp.ratur*.,

initial ntainm.nt pressure and temperature, RWST t..

mperaturo..,

and

.rvic. water temperature.

Sensitivity analyses are used to determine the limiting input parameters, break locations and single failures which predict the minimum NPSH available.

6.2.2.3.3 Iodine Removal by Containment Spray System The spray nozzles are selected to provide adequate iodine removal capability and containment coverage.

The resulting iodine removal coefficients are evaluated in Section 6.5.2.

6.2.2.3.4 Failure Analysis A

failure modes and effects analysis (FMEA) to determine if the instrumentation and controls (I&C) and electrical portions meet the single failure criterion, and to demonstrate and verify how the GDC and IEEE Standard 279-1971 requirements are satisfied, has been performed on the QSS and RSS.

The FMEA methodology is discussed in Section 7.3.2.

The results of this analysis can be found in the separate FMEA document (Section 1.7).

6.2.2.4 Inspection and Testing Requirements Preoperational tests are performed on the containment depressurization system as described in Section 14.2.12.

Section 6.6 describes the in-service periodic inspection and system pressure tests.

6.2.2.5 Instrumentation Requirements Control switches with indicating lights are provided in the main control room for the quench spray pumps.

These pumps can be started manually or automatically.

While in the automatic mode of operation, a diesel loading sequence signal combined with a CIB signal, or a CIB signal without a loss of power, will initiate the QSS.

These pumps can be stopped manually after CIB is reset.

These pumps deliver cold water containing NaOH from the RWST and CAT to the spray headers inside the containment.

Control switches with indicating lights are provided in the main control room for the quench spray pumps suction and discharge MOVs.

A CIB signal being present provides a signal to open these valves even though the quench spray pumps suction and discharge valves are normally open during plant operation.

When a

CIB 'signal is not present and a

respective quench spray pump is not' running, these valves can be closed manually, Control switches with indicating lights are provided in the main control room for the chemical injection pumps.

These pumps may be 6.2-54