View source for 3F0198-12, Forwards 90-day Response to GL 97-04, Assurance of Sufficient NPSH for ECC & Containment Heat Removal Pumps, Dtd 971007.No Commitments Made in Submittal

{{Adams
| number = ML20198C749
| issue date = 01/05/1998
| title = Forwards 90-day Response to GL 97-04, Assurance of Sufficient NPSH for ECC & Containment Heat Removal Pumps, Dtd 971007.No Commitments Made in Submittal
| author name = Holden J
| author affiliation = FLORIDA POWER CORP.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000302
| license number = 
| contact person = 
| document report number = 3F0198-12, 3F198-12, GL-97-04, GL-97-4, NUDOCS 9801080045
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE
| page count = 11
}}

=Text=
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   ,            Ficrida Power
                ? A "ST 8",.'OI."d mn hnuary 5,1998 3F019812 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

==Subject:==
Respnse to Generic Letter 97-04, "Asrurance of Suflicient Net Positive Suction licad for Emergency Core Cooling and Containment Heat Removal Purnps"

==Dear Sir:==

Genetic Letter (GL) 97-04, " Assurance of Suflicient Net Positive Suction liead for Emergency Core Cooling and Containment IIcat Removal Pumps," dated October 7,1907, requested specific infonnation be provided to the Nuclear Regulatory Commission (NRC) wit                                                                    W n 90 days of the date of the GL Florida Power Corporation (FPC) is hereby submitting the requested infomtation relative to the Crystal River Unit 3 Nuclear Generating Plant (CR-3). The information is provided in the Attaciiment to this letter. No commitments are made in this submittal.
Should you have any questions or requ re additional infomation conceming this response, please contact Mr. David Kunsemiller, Manager, duelear Licensing at (352) 563-4566.
Sincerely,
[tb -
Amh J. Iloiden Director Site Nuclear Operations                                                                                                                                                             /
                                                                                                                                                                                              /
       .IJil.kdw Attachment xc: . Reponal Administrator, Region 11 Lf / f//
                                                                                                                                                                                    "~
NRR Project Manager Senior Resident inspector j)'****.'I.
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* 9801000045 980105 PDR     ADOCK 05000302 p                                    PDR CRifSTAL RIVER ENERCY COMPT.EX: 15760 W. Power Line Street
* Crystal River, Floritsa 34428 6708 e 0521795-6486 A ARwins Progress Company

l e                                                                                                                                                                              r i           U.S. Nuclear Regulatory Commission
           - 3F0198-12 Page 2 of 2 STATE OF FLORIDA COUNTY OF CITRUS John J. liolden states that he is the Director, Site Nuclear Operations for Florida Power Corporation, that he is authorized on the part or'said company to sign and fi!e with the Nuclear Regulatory Commission the information attached hereto; and that all cuch statements made and matters                                                                  4 set forth therein are true and correct to the best of his knowledge, information, and belief.
IL Jo    J. lioiden Director Site Nuclear Operations Sworn to and subscribed before me this                                                                     6 _Mday of        d4 a ve T           1998,by John J. Holden.                                                                                                                                                       y Signature ofNo$ary Public State of Florida USA ANN MCBRIDE Notary Public, State of Floride l
My Comm. Exp. 0ct. 25, IM9 C*""". no Cc mse                                                                               }y, ggn f77 cg,, Jg (Print, type, or stamp Commissioned Name of Notary Public)
Personally                   Produe::d Known
[_ -OR Identification l
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              ' U.S. Nuclear Regulatory Commission                                                                                                       Attachment -
3F019812                                                                                                                                 Page 1 of 9 ATTACilMENT RESPONSE TO GENERIC LE' ITER 97-04 Generic Letter (GL) 97-04 requested specific information regarding the available NPSH for the emergency core cooling (incleding core spray and decay heat removal) and containment heat removal pumps that meet either of the fewing criteria:
(1) pumps that :ake suction from the containment sump or suppression pool following a design-c-                          basis loss-of-coolant accident (LOCA) or secondary line break. or (2) pumps used in " piggyback" operation that are necessary for recirculation cooling of the reactor core and containment (that is, pumps that are supplied by pumps which take suction directly from the sump or suppression pool):
The following pumps at CR-3 meet the specified criteria and are, therefore, within the defined scope of GL 97-04:
APPL.lCABLE GidEIT11RIA                                                                                P_UA1E 1             Building Spray Pumps (BSP) l A/lB 1             Decay lleat Pumps (DilP) l A/lB 2              Makeup Pumps (MUP) I AllB/IC The requested information is provided below, as applicable, for the pumps within the defined scope of GL 97-04:
NRCINFORA(A110NREQUEST:
(1)      Specify the general methodok>gy used to calculate the head lou associated with the ECCS suction strainers.
FPCRESPONSE:
Building Spray Pump and Decay llent Pump Evaluation The NPSII evaluation looked at the worst case conditions from an NPSH standpoint, and combined different accident scenarios to be bounding. The following general assumptions were used to calculate the most limiting NPSH values for the BSPs and the DIIPsi
: 1. Minimum Reactor Building (RB) pressure
: 2. Minimum RB sump level
: 3. Maximum pump flow rates (this obtains the maximum piping losses)
    .                      4. Maximum vapor pressure expected at time of sump recirculation

9 l'     .
U.S. Nuclear Regulatory Commission                                                       Attachment 1F019812                                                                                 Page 2 of 9 The LOCA scenario used is such that the break is in the Reactor Coolant System (RCS) piping at a high enough elevatian that no RCS volume is available for flooding. The break is also of a size that would result in the highest expected flows (including instmment error) from the Decay Hect Pumps (DHPs), the Building Spray Pumps (BSPs), and the Makeup Pumps (MUPs) in the " piggyback" mode of operation. Although Low Pressure injection (LPI)is in operation (which would require that RCS pressure be below the Core Flood (CF) system actuation point), no credit is taken for the CF tank inventory. The maximum expected temperature with no pressure in the RB (212" F) is used at the time of sump recirculation which willincrease the afTect of vapor pressure. This is a conservative assumption because the duration of a temperature at 212" F is short, and as temperature drops so will the affect of vapor pressure. Details regarding the methodology uted to calculate NPSH values follow.
The basic equation for calculating available NPSH (NPSH.) is:
NPSH. = h. - h, + h - hr.
where:
: h. = absolute pressure on the surface of the liquid supply level.
h, = head corresponding to the vapor pressure of the liquid at the temperature being pumped.
h, = static height that the liquid supply levelis above the pump centerline.
hr. = all suction line losses including entrance losses and friction losses.
Additional detail regarding each term is prov;ded below.
: h. = abrolute pressure on the surface of the liquid supply level This absolute pressure is made up of:
: 1. Atmospheric pressure present prior to the postulated loss of coolant accident (ho.).
: 2. Increase in containment pressure that results from the postulated loss of coolant accident (ht.oca).
,                 .onsistent with Regulatory Guide 1.1, " Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps," FPC applies the regulatory position that the emergency core cooling and containment heat removal systems should be designed so-that adequate NPSH is provided to system pumps assuming maximum expected temperatures of pumped fluids and no increase in containment pressure from that

U.S. Nuclear Regulatory Commission                                                   Attachme.nt 3F0198-12                                                                            Page 3 of 9 present prior to postulated loss of coolant accidents (LOCAs). This results in hun = 0 ft; therefore, h. = h.i = 33.9611 (14.7 psi).
Note: The partial pressure of air would be higher than 14.7 psi due to the elevated temperatures, but is not credited for conservatism.
hy, = head corresponding to the vapor pressure of the liquid at the temperature being pumped FPC uses satursted fluid at a temperature of 212" F when drawing santion from the sump.
This is the expected temperature at the time of swapover from the Borated Water Storage Tank (BWST) to the Reactor Building (RB) sump asauming no pressure in the RB. Fluid temperature is expected to decrease soon after swapover which decreases the effect of vapor pressure. At a fluid temperature of 212 ' F, hy . = 33.96 ft (14.7 psi). This equals the pressure present prior to the postulated LOCA (h.im = h,) and leaves only static head above the pump centerline to provide adequate NPSH.
h., = static height that the liquid supply level is above the pump centerline j
FPC calculates a minimum static height using the following assumptions:
: 1. The minimum Improved Technical Specifications (ITS) BWST level. The BWST is the primary contributor to RB sump level.
2      Initial BWST temocrature is assumed to be at its highest allowable temperature to minimize the afTect ofincreased level due to fluid expansion.
: 3. Inventory iam the makeup tank contributes to the available inventory (very small percentage).
: 4. The minimum RB flood level assumes saturated fluid at 212" F which takes into account evaporation of water due to the elevated temperatures.                         t
: 5. Additional iaventory unavailable for recirculation is accounted for from cavities that act as holdup volumes and prevent contribution to the static beight.
: 6.      No credit is taken for the inventory in the Core Flood Tanks.
: 7. No credit is taken for RCS inventory that would contribute to the available static height due to flow through the break.
hr, = all suction line losses including entrance losses and friction losses FFC calculates suction line and friction line losses using the following assumptions:

U.S. Nuclear Regulatory Commission                                                   Attachment 3F019812                                                                             Page 4 of 9
: 1. The pipe and fittings are 304 stainless steel. The pipe roughness is assumed to be that of clean commercial steel pipe. This is a reasonaMe assumption because typically there are not extended periods of flow through this section of pipe.
: 2. The length of piping accounted for is from the sump intake to the pump inlet.
: 3. Flow rates for the BS, DH, and MU pumps were assumed at the maximum flow rates when taking suction from the RB sump, with worst case instrument errors added. Credit fer Emergency Operatmg Procedures (EOP) is taken to throttle flows to the designed flows when taking suction from the RB sump.
: 4. Vendor data for all of the valves in :he suction lines was used in the NPSH calculation including the check valves just upstream of the BS pumps. Flow coeflicient (C,) values were provided at the apphcable flow rates to account for the insuflicient flow to maintain the check valves in the full open position.
Suction head losses were calculated using Reference 10. In addition, FPL p-dormed a test and evaluation an the system head losses for the BS system. The results of the test clearly show the calculated line losses to be conservative when compared to the actual results from the test (Reference 3).
The following approach was used for sump screen losses:
: 1. An assum9 tion of I fVsec through the screen is used to calculate the head loss, which accounts for an area reduction of approximately 53% from blockage of debris that may be present. This area reduction is consistent with Regulatory Guide 1.82, Rev. O, " Sumps for Emergency Core Cooling and Containment Spray Systems,"
which was used as a guide in the calculation of sump screen losses- Regulatory Guide 1.82, Rev. O states that, "The available surface area used in determining the design coolant velocity should be based on one-half of the free surface area of the fine inner screen t conservatively account for partial blockage."
: 2. The hydraulic loss through the inner sump screen accounts for the actual open area of the screen.
: 3. The .esultant head loss through the inner sump screen using the above approach is on the order of 0.0M ft Makeup Pump Evaluation The NPSH evaluation for the MU pumps was done separately from that of the SS and DH pumps since, when the MU pumps are m the " piggy-back" mode of operation, they are subject to the discharge pressure of the DH pumps. As a result, some simplifying assurrptions can be made.

4 U.S. Nuclear Regulatory Commission                                               Attachment 3F0198-12                                                                        Page 5 of 9 MU pump NPSil values were calculated using the following assumptions:
: 1. RB pressure was assumed to be 0 psig. No credit was taken for RB overpressure.           -
: 2. MU pump flow rate was assumed to be at a maximum value (600 gpm).
: 3. DilP ficw rate was maximized (this obtains maximura piping losses and minimum pressure rise across the Dil pumps).
The analysis was performed assuming that sli three MU pumps have the same NPSil characteristics    A calculation of each individual MU pump NPSil value was not performed because the pumps are similar and the NPSil margin is greater than 200 feet.
NRCINFORAfA110NREQUEST:
(2)    Identify the required NPSH and the amilable NPSH.
FPC RESPONSE.
The required NPSil and the available NF3!I values for the BSPs, DilPs and the MUPs are identified in the following table. The available NPSil valt.u were calculated using the methodologies outlined in the above response to NRC Information Request (1) and result in the calculation of the miriimum values for the available NPSil and the NPSil margin.
The required NPSil values were determined from f H applicable pump curves. Ilowever, it should be noted that the required NPSil value used for the calculation of the BSP IB eva.lable margin is a bounding value based on the most limiting BSP impeller.
MINIMUM        MINIMUM REQUIRED                   AVAILABLE      MARGIN NPSil (fl 1];0)
NPSil(f1 ll:0j (fUb Q BSP1A                    12.5                       14.5           2.0 BSPIB                    12.5                       13.8           1.3 DilP1A                   13.5                       15 0           1.5 DIIP IB                  13.5                       14.9           1.4 M UP 1A/lB/lC            34.0                      ~240          >200 NRCINFORAfAllONREQUEST:
(3)     Specify whether the current design-basis NPSH analysis differs from the most recent analysis reviesced and approved by the NRCfor schich a safety evaluation vras issued.

4 l
U.S. Nuclear Regulatory Commission                                                 Attachment 3F0198-12                                                                          Page 6 of 9 FPCIMSPONSE:
The current CR-3 design-basis NPSH analysis coes differ from the most recent analysis reviewed and approved by the NRC for which a safety evaluation was issued. References 3,4 and 5 represent the most recent analyses revie ved and approved by the NRC for which a safety evaluation was issued. In addition, Section 6.4.2, " Summary of Post-Accident Recirculation" and Table 6-12, " Post-Accident NPSH Requirements" of the CR-3 FSAR have been revised since the time of the mos: recent SER (Reference 6) to reflect the current NPSH values. Revision 24 to the FSAR will revise Table 6-12 to reflect the most current NPSH values consistent those provided in the FPC response to NRC Information Request (2) above. Details follow regarding the CR-3 licensing basis for the BSPs and the DHPs when in the recirculation mode of operation.
Building Spray Pumps Licensing Basis On July 5,1974, the NRC issued a Safety Evaluation Report to CR-3 (Reference 1).
Section 6.2.2 of the SER, " Containment Heat Removal Systems," states that, "NPSH requirements for the reactor building spray pumps can be met during the post-accident p                   recirculation phase by throttling pump flow during recirculation from 1500 gpm to 1200 gpm." The SER stated that, "We h .ve reviewed the containment heat removal systems far conformance to the GDC Noa. 38, 39, and 40, and Regulatory Guide 1.1, " Net Positive Suction Head for Emergency Core Cooling and Heat Removal System Pumps" dated November 2,1970. We conclude that the systems meet the requirements of these criteria and are acceptable."
It should be noted that this SER acknraledged that CR-3 was not designed and constructed to the General Design Criteria published in 1971 (10 CFR 50 Appendix A),
stating that, "This facility was designed and constructed to meet the AEC's GDC, as originally proposed in July 1967. The Commission published the revised G" in 1971 just before the FSAR was Fled. We conducted our technical review ag              .ae present version of the GDC and we conclude that the plant design acceptably conforms to the current criteria."
Prior to the issuance of the CR-3 SER, the NRC published Regulatory Guide (RG) 1.79, "Preoperational Testing of Emergency Core Cooling f>ystems for Pressurized Water Reactors," (Reference 2) describing an acceptable test program for Emergency Core Cooling Systems (ECCS)in Pressurized Water Reactors (PWR). By {{letter dated|date=February 11, 1976|text=letter dated February 11,1976}} (Reference 3), in response to specmc NRC questions relative to CR-3 compliance with RG 1.79, FPC submitted a copy of the 13abcock & Wilcox Line Loss Calculations for Reactor Building Spray and Decay Heat Removal Pump Suction Piping During the Recirculation Mode from the Reactor Building Sump.
In addition, on October 11,1976 (Reference 4), FPC submitted a copy of the report 3
entitled, " Evaluation of Calculated System Head Losses and Reactor Bui: ding Sump Test       ,
l

U.S. Nuclear Regulatory Commission                                                                              Attachment 3F0198-12                                                                                                       Page 7 of 9 System liead Losses Relative to Crystal River Unit No. 3." This report presented the results of a design review and evaluation performed to compare the calculated fluid frictionallosses with the reactor building sump flow test data relative to the DHP and BSP suction systems from the building sump at CR-3.
As documented in Reference 5, "FPC made modifications to the Reactor Building Spray System to preclude a runout condition on the containment spray pumps or to prevent exceeding the margin available for net positive suction head to the pumps." In Section 6.2.2 of Supplement No. 3 to the CR-3 SER (Reference 6), the NRC concluded that, based on the review of the FPC modifications to the reactor building spray system and the results of the preoperational tests, the modifications will preclude pump runout conditions and will prevent exceeding the margin available for net positive suction head and are, therefore, acceptable.
In Reference 7, the NRC Integrated Performance Assessment Process (IPAP) inspection team concluded that, "the cavitation free operation of the building spray pump IB during post-LOCA recirculation phase is questionable due to various non-conservative assumptions made in the calculation." The assessment team idemified this issue as Unresolved item UR-96-201-02.
FPC subsequently (Reference 8) stated its plans for addressing the BSP-1B NPSH issue, as well as other design-related issues, prior to restarting the plant from the current forced outage. Reference 8 stated that, "FPC plans to conduct factory testing and/or modify the pump impeller to improve the margin between required and available NPSil" In Reference 9, the NRC confirmed its understanding that FPC would resolve the design issues delineated in Reference 8, including the BSP-1B NPSH issue.
To resolve the BSP-1B NPSH issue, FPC sent all four BSP impellers (including 2 spares) to the pump vendor where a modi 6 cation to the impellers was performed.                                           Post modification testing demonstrated acceptable results of a reduction in the required NPSH, with no impact on head capacity performance. The post modification margin for the BSPs, when taking a suction from the RB sump, is as identified in the response to NRC Information Request (2) above.
Decay IIcat Pumps Licensing Basis On July 5,1974, the NRC issued a Safety Evaluation Report to CR-3 (Reference 2).
Section 6.3 of the SER, " Emergency Core Cooling Systera (ECCS)," states that, "When a predetermined amount of water in the borated water storage tank has been injected, or-receipt of a low-level alarm for the BWST, suction will be transferred manually to the containment sump for the recirculation mode of operation provided by the LPIS [ Low Pressure Injection System). The ECCS will then provide the long-term core cooling requirements by recirculating the spilled reactor coolant collected in the containment sump back to the reactor vessel through the core flooding line nozzles." The SER concluded                                  ,
I.

U.S. Nuclear Regulatory Commission                                                      . Attachment 3F0198-12                                                                                Page 8 of 9        ;
4 "that th'e performance of the ECCS is in accordance with the Commission's Interim Acceptance Criteria" for the performance of the ECCS for light-water coaled nuclear power reactors.                                                                                  ,
I Test Description C.3.b of RG 1.79, "Prec serational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors" (Reference 1), includes a description of an              -
acceptable recirculation- test for Low Pressure Safety injection (LPSI) Systems that             i
                          , includes the demonstration of the capability to realign valves-and ' injection pumps to           :
recirculate from the containment floor or sump into the reactor coolant system. -Test Description C.3.b.(2) of RG 1.79 includes verification that the available net positive suction head for LPSI is greater than that required at accident temperatures.
By {{letter dated|date=February 11, 1976|text=letter dated February 11,1976}} (Reference 2), in response to specific NRC questions
                          - relative to CR-3 compliance with RG 1.79, FPC submitted a copy of the Babcock &
Wilcox Line Lors Calculations for Reactor Building Spray and-Decay. Heat Ren oval Pump Suction Piping During the Recirculation Mode from the Reactor Building Sump. In addition, on October 11,1976 (Reference 3), FPC submitted a copy of the report entitled,
                             " Evaluation of Calculated System Head Losses and Reactor Building Sump Test System Head Losses Relative to Crystal River Unit No. 3." This report presented the results of a
                          -design review and evaluation performed to compare the calculated fluid frictional losses with the reactor building sump flow test data relative to the DHP and BSP suction systems from the building sump at CR-3.
Section 6.3 of Supplement No. 3 to the CR-3 SER indicated that FPC modified the ECCS to include automatic flow control; err in the discharge of the DHPs to preclude the need
                           'for the operator to control pump runout which could result in insuflicient NPSH for the pumps. The NRC concluded that, in recognition of the modifications and preoperational tests, the ECCS is in accordance with Appendix K to 10 CFR Part 50, with the exception of the analysis with regard to return to nucleate boiling after critical heat flux conditions have been reached.
NRCINFORMATIONREQUEST:
(4)^      Specify uhether coruainment overpressure (i.e., containment pressure above the mpor pressure of the sump or suppression poolfluid) um credited in the calculation of amilable NPSH. - Specify the amount ofowrpressure neededand the minimum overpressure amilable.
FPCRESPONSE:
                         . Containment overpressure was not credited in the calculation of the available NPSH at
                          . CR-3.
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1

U.S. Nuclear Regulatory Commission                                                

==Attachment:==

13F0198 l?                                                                          Page 9 of 9 i
N4CINR)RMA110NREQUEST:
i (3)     When containment overpressure is credited in the calculation of available NPSH, conprm that an appropriate containment pressure analysis was done to establish the minimum containmentpressure.                                                                  ,
                 . FPC RFSPONSE:
As indicated in FPC response to NRC Information Request (4) above, containme.nt               i overpressure was not credited in the calculation of the. available NPSH at CR-3.
Therefore, Information Request (5)is not applicable to CR-3.
i P9FERENCES:
: 1.      NRC to FPC letter, 3N0774-01, dated July 5,1974, " Safety Evaluation by the Directorate of.
Licensing"
: 2.      Regulatory Guide 1.79, "Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors" i
: 3.      FPC to NRC letter,3F0276-06, dated February 11,1976
: 4.      FPC to NRC letter,3F1076-03, dated October 11,1976
: 5.      FPC to NRC letter,3F1276-07, dated December 10,1976 L
: 6.      N AC to FPC letter,3N1276-10, dated December 30,1976, " Supplement No. 3 to the Safety Evaluation Report" 1                 7.     .NRC to FPC letter, 3N0896-12, dated August 23,1996, " Crystal River Unit 3 Integrated Performance Assessment Process (IPAP) Final Assessment Report"
: 8.    . FPC to NRC letter, 3F1096-22,- dated October 28,1996, " Crystal River Unit' 3 Forced Outage" 9l      NRC to FPC letter,3N0397-07, dated March 4,1997, " Confirmatory Action Letter" 10.'    " Internal Flow Systems," D. S. Miller, Second Edition b
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