ML20198F433

Forwards Draft Sser Re Util ECCS Performance Reanalysis Based on B&W BAW-10102.Licensee Requested to Make Listed Mods to Tech Specs or Sys Design/Operation

ML20198F433
Person / Time
Site:	Washington Public Power Supply System
Issue date:	09/12/1975
From:	Baer R Office of Nuclear Reactor Regulation
To:	Moore V Office of Nuclear Reactor Regulation
References
CON-WNP-1087 NUDOCS 8605280506
Download: ML20198F433 (7)

Text

_._ . _ _ . _ _ _ . .

~ '

hc O 4 SEP 121975 Docket Nos.: 50 460 513 Voss.A. Moore. Assistant Director for LWR's, Group 2, RL

, THRU: Thomas M. Novak, Chief, Reactor Systems Branch, TR Plant Names - WPPSS Nuclear Plants, Unita 1 & 4

Licensing Stage . Canaemation Permit Docket Number: dM13 Milestone Number: 27-21 Responsible Branch- LWR 2-3 and Project Leader: T. Cox Requested Completion Data: August 15, 1975 Description of Reviews SER Supplement Technical Review Branch Involved: Reactor Systems Branch Review Status: Complete Pursuant to the requirements of the Connaission's regulations,10 CFR 50.46, the applicant incorporated the Babcock and Wilcox topical report BAW-10102 into his application by reference to comply with the ECCS

, performance reanalysis requirement. The analysis were based an the Babcock and Wilcom ECCS evaluation model, which the NRC staff has reviewed and concluded was in conformance with 10 CFR 50, Appendix K.

The staff reviewed references 2 through 5 and our findings are delineated in the enclosed draft Safety Evaluation Report.'

The licensee has been requested to modify the Technical Specifications or systesa design / operation as follows:

1) To satisfy the single failure criterion regarding specific motor operated valves, the circuit breaker for the motor operators of each of the valves will be locked open.

, 2) To achieve operation of the ECCS from the control room to prevent excessive boron concentrations in the core during the long term, post LOCA, the valves should be motor operated with control and indication in the control room. Flow indication should also be panttored in the control room.

! 3) The Technical Specifications shall not permit partial loop operation.

1 Drigi.nal signed by Robert! Baer, Section Leader i

Reactor Systems Branch Division of Technical Review Office of Nuclear Reactor Regulation

Enclosure:

e[

SER Supplement i i

• "'cee: See page 2. 0605280506 750912 -

PDR ADOCK 05000460

'""~'"'* E PDR_, , _.

oaes> . . .,

Foran AEC.)l4 (Rev. 9 5H AICM 0240 Tr u. e. eovammusm? Pasurine orrics e era.ase.t se

Voss A. Moore -2 i -

SEP 1g 1375 cc:

l S. Har.auer i R. Mainaman l A. Giambusso V. Stallo ,

! T. Cox f T. Novak W. T===i=a W. Mcdonald DISTRIBUTION:

! DOCKET FILE

NRR READING FILE RSB READING FILE P. FLING R. BAER 9 .

. ; l.,I 'n'

u. .

\ .

O Rus ra.

s, I

W I

INM 9v'v N 8ls.2 n.y7f

.. ,. /. .

RB e snn._ R._-._W_

.u ... .. .

. ._,........,_.1

.,mn &

u_ .

i

y .

- ~

DRAFT SAFETY EVALUATION WPPSS NUCLEAR PROJECTS NO. 1 5 4 DOCKET NOS. 50-460 AND 50-513 I. Introduction In their letter of June 19,1975 (reference 1), the applicant (Washington Public Power Supply Systen (WPPSS)) incorporated the Babcock and Wilcox topical report BAW-10102 (reference 2) into their application to construct WPPSS Nuclear Projects Numbers 1 and 4. Pursuant to the requirements of the Comaission's regulations, 10 CFR 50.46, this topical report was submitted by B&W to demonstrate compliance with the ECCS Final Acceptance Criteria for B&W 205 plants. The applicant has submitted additional information in references 3, 4 and 5.

In addition to the reviced LOCA analysis, the staff's review addressed the specific areas of minimum containment pressure, single failure criterion, effects of boron precipitation on long term cooling capability, and submerged valves within containment.

1.1 ECCS Rennnivsis ,

The information submitted addressed the loss of coolant from small ruptured pipes (0.5 ft2 and smaller, reference 6) and major reactor coolant system pipe ruptures (reference 2). The analyses cubmitted were performed with an acceptabic cvaluation model which was in

~

conformity with 10 CFR 50, Appendix K and described in reference 7.

A spectrum of break sizes, configurations and locations were performed in accordance with the staff's " Minimum Requirements for ECCS Break Spectrum Submittals", dated April 25, 1975. The analyses identified the worst b,reak as the 8.55-ft 2 double-ended break at the pump discharge.

The table below summarizes the fesults of the LCCA limit analyses which determined the allowable linear heat rate limits as a function of elevation in the core.

• e

3 -

_ L LOCA Peak node temp. F Elev. limit, Maximum Local Time of ft. kW/ft Ruptured Unruptured Oxidation, % Rupture 2 -14.9 2097 1931 4.1 25.90 4 16.2 2002 2156 5.9 26.19 6 16.8 2017 2126 5.3 25.40 8 15.3 1763 2177 6.7 32.30 10 14.2 1880 2171 6.3 26.75 The maximum core-wide metal / water reaction was calculated to be 0.62%

which is belcw the allowable limit of 1%.

An shown in this tabulation, the calculated values for the peak clad temperaturc, and local ractal/ water reactions were below the allavable limits of 2200*F and 17%, respectively, specified in 10 CFR 50.46.

Reference 2 has also shown from results of analyses that the core geometry remains amenable to cooling and that long term cooling can be established.

The analyses of postulated breaks during partial lopp operation have not been provided by the applicant. The applicant has indicated that these analyses will be provided during the first quarter of 1976. Until the staff has reviewed these analyses, the Technical Specifications will not pcrmit partial loop operation.

1.2 Boron Build-Up During Long Term Post-LOCA Core Cooling The licensee amended in Reference 5 the operating procedures described in Reference 2 for the long term post-LOCA core cooling period and has' indicated that 'these procedures would prevent excessive concentration of boron in the core region. By amending the op, crating procedures for Mode 1, this Mode now conforms to the single failure criterion. The NRC staff has reviewed the proposed emergency procedures and the referenced analyses and conclude that the ECCS can be operated in a manner that would prevent excessive boron soncentration from' occurring provided this mode of operation can be achieved from the control room.

The primary mode of operation (Mode 1) consists of establishing suction from the decay heat drop line and the reactor building sump with one of the low pressure injection trains. In order to establish suction from the drop line, an operator must manually operate a throttling valve in the -

bypass line based on locally road flow indication. The staff util require that each of the throttling valves in both trains be notor-operated valves with control and indication from the control room. In addition, flow indication in the bypass lines shall also be provided in the control room. ,

i

r- ,

e 0, . L.

B bcock and Wilcox in Reference 2 had provided two additional back-up modes to Mode'l to control boron build-up in the core. These modes' consist of injecting dilute water from the reactor building sump into the hot leg by way of the pressurizer sprays (Mode 2) or the decay heat drop line (Mode 3). With Mode 1 conforming to the single failure criterion, it is the staff's understanding that Modes 2 and 3 are not required for long term cooling in WPPSS. Therefore Modes 2 and 3 have not been reviewed for WPPSS.

1.3 Single Failure Criterion Appendix K to 10 CFR Part 50 of the Commission's regulations requires that the combination of ECCS subsystems to be assumed operative shall be those available after the most damaging single failure of ECCS equipment has occurred.

Babcock and Wilcox has conservatively assumed all containment cooling systema operating for the independent containment calculation, and to assume the diesel failure for the ECCS calculation.

A review of the UPPSS piping and instrumentation diagrams indicated that the spurious actuation of specific motor operated valves could affect the appropriate single failure assumptio'ns. The staff identified the following motor-operated valves which did not satisfy the single failure criterion:

MOV f Location 1A One valve immediately outside the 1B containment wall in each of the lines to the core flooding nozzles.

3A One valve at the DER pump discharge in 3B each of the lines to the core flooding nozzles. .

The licensen has indicated that the circuit breaker for the motor operatorc of each of these valves will be locked open and this will be required by the Technical Specifications.

The appliebnt has identified five valves that would be submerged following a LOCA. Only the submergence of the EHR letdown isolation valves could inhibit the safety function of these valves. The licensee has relocated these- valves to an elevation of 412 feet positioning the motor operator at an elevation of 417 fect'. The applicant indicated that the maximum sump flood icyc1 is 410 feet.

6

O e 1.4 ECCS Containment Pressure Evaluation The ECCS containment pressure calculations for WPPSS 1 & 4 were done using Babcock and Wilcox ECCS evaluation model. Although B&W containment pressure model was approved by the staf f, we required that justification of the plant-dependent input parameter used in the analysis be submitted for our review of each plant.

Justification for the containment input data were submitted for WPPSS 1 & 4 in reference 5. This justification consists of a comparison of the' actual containment parameters for WPPSS 1 & 4 with those assumed by B&W in BAW-10102. The assumptions made by Washington Public Supply System for the containment net free volume, and operation of the containment heat l

removal systems were conservatively selected for the ECCS analysis. Passive heat sinks were determined from guidelines contained in the Branch Technical Position CSB 6-1.

The PPPSS 1 & 4 contain::ient data were shown to result in a containment pressure slightly louer (less than 1' psi) than that obtained using the B&W assumptions of BAW-10102 for the first 60 seconds af ter the accident

! and higher thereafter. The applicant has concluded that littic or no difference in peak cladding temperature would result between the two containment pressure calculations.

We have concluded that the plant-dependent information provided by Washington Public Supply System for WPPSS 1 & 4 is conservative for ECCS analysis.

Therefore, thecalculatedcontainmegpressuresare in accordance with Appendix K to 10 CFR Part 50.46 ot the Commission's regulations.

EYAtt!ATION CONCLUSTONS The staff has completed its review of the WPPSS Nuclear Projects 1 and 4

~

ECCS performance reanalysis and has concluded:

1) Compliance to the acceptanec criteria of 10 CFR 50, Appendix K has been demonstrated.

2) The single failure criterion will be satisfied provided that the proposed locking out of power to the specified motor operated valves as n'oced is implemented in accordance with the Technical Specifications.

3) The procedures to control boron concentration build up during the post-LOCA, long term cooling period is acceptable providing the system is modified to permit operation of the system from the control room as noted above.

4) The ECCS minimum containment pressure calculations were performed in accordance with Appendix K of 10 CFR Part 50.

' o, , . ' , '

k, e

REFERENCES '

1. Letter from J. J. Stein to A. Giambusso - G01-75-122, "WPPSS Nuclear j- Projects No3. 1 & 4 Submittal of Information Demonstrating Compliance J

with'the ECCS Final Acceptance Criteria," June 19, 1975.

[ 2. Lowe, R. J., agi al., "ECCS Evaluation of.B&W's 205-FA NSS Revision 1,"

Babcock and Wilcox Company, BAU-10102, July 1975.

3. Letter from N. O. Strand to A. Giamhusco - C01-75-147, "WPPSS Nuclear-Projects Nos.1 & 4 Schedule for Submittal of Additional Information for ECCS Analysis," July 11, 1975.

4. Letter from N. O. Strand to A. Giambusso - G01-75-150, "WPPSS Nuclear Projects Nos.1 & 4 Submittal of Additional ECCS Information," July 18, 1975.

5. Letter from N. O. Strand to A. Giambusso - G01-75-153, "WPPSS Nuclear Projects Nos. 1 & 4 - Submittal of Additional ECCS Information,"

July 25, 1975.

6. Jones, R. C.,

jdi eQ.., "Multinode Analysis of Small Breaks for B&W's 205 - Fuel Assembly Nuclear Plants with Internals Vent Valves,"

BAW-10074, Babcock and Wilcox, November 1973.

7. Dunn, B. M., et al., "B&W's ECCS Evaluation Model," BAU-10104, Babcock and Wilcox, May 1975. .

9 b

e 9

• e g

Q .

L i

.