ML19319E237

Safety Evaluation Supporting Amend 4 to License DPR-54

ML19319E237
Person / Time
Site:	Rancho Seco
Issue date:	05/19/1976
From:	Office of Nuclear Reactor Regulation
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NUDOCS 8003310744
Download: ML19319E237 (9)

Text

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t SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 4 TO LICENSE NO. DPR-54 i

SACRAMENTO MUNICIPAL UTILITY' DISTRICT RANCHO SECO NUCLEAR GENERATING STATION DOCKET No. 50-312

1.0 INTRODUCTION

On December 27, 1974, the Atomic Energy Commission issued an Order for Modification of License (Reference 2) implementing the requirements of 10 CFR 50.46, " Acceptance Criteria for.

Emergency Core Cooling Systems (ECCS) for Light Water Nuclear Power Reactors." One of the requirements of the Order was that the licensee shall submit a reevaluation of ECCS performance calculated in accordance with an acceptable evaluation model which conforms with the provisions of 10 CFR 50.46. The Order also required that the evaluation model shall be accompanied by such proposed changes in Technical Specifications or license amendment as may be necessary to implement the evaluation results. As required by our Order of December 27, 1974, Sacramento Municipal Utility District (SMUD) (the licensee) submitted an ECCS reevaluation and related proposed Technical Specifications. The reevaluation and Technical Specifications were submitted July 8, 1975, in Reference 1 using the Babcock and Wilcox ECCS evaluation model as described in Reference 7 and discussed in Section 2.0 of this Safety Evaluation (SE). Also discussed in Section 2.0 are the results of our review of the plant-specific areas of single failures, long-term boron concentration, potential sub-merged equipment, partial loop operation, ECCS valve interlocks, and the containment pressure calculation. Section 3.0 provides the results of our review of the proposed Rancho Seco Power Distribution Technical Specifications. The ECCS review is summarized in Section 4.0.

By letter dated May 23, 1975 (Reference 13) the licensee proposed changes to his Technical Specifications concerning surveillance requirements for the structural integrity of the reactor building.

By letter dated March 10,1976 (Reference 14) the licensee proposed a Technical Specification change to update the SMUD organization chart. Both the proposed changes of Reference 13 and Reference 14 are evaluated in Section 5.0.

Sections 6.0 and 7.0 present staff conclusions and references, respectively.

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2.0 ECCS REEVALUATION The background of our review of the B&W ECCS evaluation model and its application to Rancho Seco is described in

.our SE for this facility dated December 27, 1974, issued in connection with the Order for Modification of License. The bases for acceptance of the principal portions of the evaluation model are set forth in our Status Report of October 1974 (Reference 5) and the Supplement to the Status Report of November 1974 (Reference 6) which are referenced in the Decc:nber 27, 1974 SE.

The December 27, 1974 SE also describes the various changes required in the earlier version of the B&W model. Together,_the December 27, 1974 SE and the Status Report and its Supplement' describe an acceptable ECCS evaluation model and the basis for our acceptance of the model. The Rancho Seco ECCS evaluation which is covered by this safety evaluation properly conforms to the accepted model. The licensee's July 8, 1975 submittal (Reference 1) contains documentation by reference to B&W Topical Reports of the revised ECCS%odel (with the modifications described in our December 27, 1974 SE) and a generic break spectrum appropriate to Rancho Seco

-(Referen,ce 7.and 8, respectively).

ThegenericanalysiginBAW-10103identifiedtheworstbreak size as the 8.55 ft double-ended cold leg break at the pump discharge with a C =1.0.

The table below summarizes the results D

of the loss of coolant accident *(LOCA) limit analyses whichfdetermine the allowable linear heat rate limits as a function of elevation in the core for Rancho Seco:

I Elevation LOCA Peak Cladding

, Max. Local Time of (ft)

' Limit Temperature (OF)

Oxidation Rupture (kw/ft)

(%)

(sec) 2 15.5 2002 3.92 12.25 4

16.6 2136 4.59 13.01 6

18.0-2146 5.46 15.55 8

17.0 2110 5.19 15.01 10*

16.0 1931 2.93 39.20

• See discussion-in text.

The maximum core-wide metal-water reaction for Rancho Seco was calculated to be' O.557 percent, a value which is below the allowable limit of 1 percent.

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, 'Ab'shown'inlthetabulation,'thecalculatedvaluesforthepeakclad temperature and. local metal-water reaction were below the allowable limits specified in 10 CFR 50.46-of 22000F.and 17 percent, respect-ively.- BAW-10103 has also shown that the core geometry remains amenable to cooling and that long-term core cooling can be established.

We noted during our review of BAW-10103 that the LOCA limit

' calculation ~at the 10-foot elevation in the core showed reflood

~ rates below 1' inch /second at~251 seconds in the accident (Section 7.2.5).

Appendix K.to 10 CPR 50.46 requires that when reflood rates are less than 1 inch /second, heat transfer calculations shall be based on the assumption that cooling is only by steam, and shall take into account any flow blockage calculated to occur as a result of cladding swelling

-or rupture as such blockage might affect both local steam flow and.

' heat transfer.. As indicated in References 5 and 6, a steam cooling

.model:for reflood rates less than 1 inch /second was not submitted by B&W for our review.- The steam cooling model submitted by B&W in BAW-10103.is therefore considered to be a proposed model change requiring our further review. -Accordingly, B&W was informed that until the proposed steam cooling model is reviewed, the heat transfer calculation at the '10-foot elevation during the period of steam cooling specified in BAW-10103 must be further justified. In lieu of using their proposed steam cooling model B&W has submitted the i

results of calculations at the 10-foot elevation using adiabatic heatup during-the steam cooling period, where this period is defined

-l by B&W as the time.when the reflood rate first-goes below 1 inch /second

-to the time that the 10-foot elevation is predicted to be covered by-water. The new calculated peak cladding temperature, local metal-water reaction and core-wide metal-water reaction at the 10-foot elevation are 19460F, 3.02%,'and.647% respectively. These values remain below the allowable limits of 10~CFR 50.46 and are acceptable.

Our review of plant-specific assumptions discussed in the following paragraphs regarding the Rancho Seco analyses addressed the areas of single failure criterion, long term boron concentration, potential submerged equipment, partial loop operation, ECCS valve interlocks,'and the containment pressure calculation.

2.1.

Single Failure Criterion Appendix K to110 CFR 50'of the Co M asion'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 occured. - ' Babcock and Wilcox hes

-assumed all contain= ant cooling systems operating to minimize

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-cont'ainment pressure and has separately assumed the loss of one diesel to minimize ECCS cooling.- We concluded in Reference 5 that the application of the single failure criterion was to be confirmed during subsequent plant reviews.

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'A review of Rancho Seco piping and instrumentation diagrams _ indicated that1the spurious actuation of certain motor-operated valves could

' affect the appropriate single failure assumptions. The licensee has indicated that_a spurious actuation ofLcore flooding tank (CFT) vent valves BV-26511 or HV-26512_vould result in a decrease in CET pressure. 'Accordingly,'a Technical Specification has-been added requiring these.normally closed motor-operated valves to have their

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power disconnected by opening and tagging the. breakers. This is acceptable to us. A review was also condue::ed of the electrical Jachematics'for ECCS motor-operated valves. It was determined that a single failure of valve interlocks could not affect the appropriate single failure assumptions.

To further minimize the potential for a water hammer 'due to the discharge of ECC water into a dry line we require that valves SFV-25003 or SFV-25004 be left in the open position during normal operation (depending on normal makeup alignment). This maintains at least one ECCS'. train. filled with a continual supply of water from the borated' water storage tank (BWST) due to the available static head built into the design. Such a configuration will also eliminate the need for one automatic safety action in the event of a LOCA; that is, the automatic opening of this valve to provide ECC water to the -low pressure injection ?(LPI) and building spray

_ pumps.. In~ addition, we have added a Technical Specification whereby a monthly procedure of Lventing the existing ECCS puun castings is required to ensure that no air pockets have formed. Such venting must also be performed Prior to any ECCS flow tests. The licensee

has 'also agreed to -investigate the feasibility of venting high points in ECCS lines.-

2.'2 Containment Pressure-The ECCS containment pressure calculations for Oconee Class

- plants were performed generically by B&W for reactors of this

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. type as described in Reference 8.

The NRC staff reviewed B&W's

' evaluation model and published the results of this review in

Reference 5 and 6.-

We concluded that B&W's contMn=ent pressure model was acceptable for ECCS evaluations. We required that jus- -

ntification of the plant dependent _ input parameters used-in the

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containment analyses be~ submitted for our review of each plant.

Justifications for the containment input data was' submitted for l:

. Rancho Seco on October 15,1975 (Reference 11). ' This -justification

-allows comparison of the actual containment parameters for Pancho n

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Seco.with those assumed in Reference 8.

The licensee has evaluated the contafement net-free volume, the passive heat sinks, and operation of the containment heat-removal systems with regard to the conservatism for the ECCS analysds. This evaluation was based on as-built design-

.information.

The catainment heat removal systems were assumed to' operate at their.==N=

capacities, and minimum operation values for the~ spray water and service water temperatures were assumed. The i.,

containment pressure analysis was demonstrated to be conservative for

Rancho Seco.

We have concluded that the plant-dependent information used for the ECCS containment pressure analysis for Rancho Seco 1s conservative and, therefore, the calculated containment pressures are in accordance with Appendix K to 10 CFR 50 of the Commission's regulations.

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i 2.3 Lona-Term Boron Concentration The NRC staff has reviewed the proposed procedures and the

- systems-designed for preventing excessive boric acid buildups in the reactor vessel during the long-term cooling period after a LOCA.

The licensee has agreed to implement procedures for Rancho Seco which would allow adequate boron dilution i

during the long-term and which will employ a concept similar to that described in BAW-10103. We have noted that a failure of a diesel will affect each of the proposed dilution modes. The licensee 4

has indicated that the, controllers for all the pertinent valves are located in the auxiliary building, thus' enabling the operator to connect power to the vailves with. jumper cables. Based on calculations by B4,W, which taka credit for natural circulation through the vent valves over 30 days is available before forced circulation is necessary;- therefore, the licensee's backup procedure to a power i.

-failure is acceptable.

As initia' lly proposed by the licensee, dilation Mode 1 was to be first

-attempted which would establish suction from the reactor vessel outlet pipa through the; piping from the hot legs to the decay heat pumps with'one LPI string. It is our position that Mode 1 should not be' attempted.as a method to control boron concentration in the core during long-term' cooling. References'3 and 8 state that success of Mode 1 is'not ensured because of the possibility of gas or-steam entrainment in.the~ decay heat pump suction norsle. Such gas I

or steam entrainment could result in severe damage to the decay heat removal pump.

Long-term heat removal requirements can exist for long~ durations (days 'or months) after the accident and continuous 1

operation of-one train of the decay heat removal system is required.

In the event of equipment" malfunction in this train, no method is-available to remove the' decay heat if the other train has been

-l previously damaged. Therefore, since initiation of Mode 1 is not allowed, Modes 2 and 3. (as proposed in Reference 3) must be single k

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r failure proof.in combination. To employ a single failure proof dilution concept, the licensee will make a modification to the existing design.during the next refueling outage. This modification will employ a hot leg drain mode in combination with a backup hot

. leg injection mode. Several valves inside containment will be modified i

to allow remote operation of hot leg injection from the control room.

In addition, the hot leg drain mode will be modified to allow operation and flow indication from the control room. During the interim, the licensee has installed temporary piping to allow hot leg injection without the need to rely on certain manual valves inside 1

. containment. Both existing modes contain manual valves outside the containment which must be operated using local handwheels. To minimize operator radiation exposure, the licensee has agreed to operate these valves prior to the shift to the recirculation mode.

We have reviewed'the operating procedures associated with this interim proposal and conclude that until the more permanent mod-ification is implemented during the next refueling outage, this proposal is acceptable. A final review of the permanent installation and associated operating procedures will be conducted prior to startup

-after refueling.

2.4 Submerged Valves The licensee has conducted a review of equipment arrangement to determine if any valve motors insi6 the containment'will become submerged following a LOCA. Based oa this revieu, no valves were identified which would be flooded and which would affect short-term or long-tern ECCS functions or contahant isolation.

2.5 Partial Loop Analyses Existing Technical Specifications require reduced reactor power when i

operating with less than four reactor coolant pumps on the line.

To allow an operating configuration with less than four reactor coolant pumps on the line (partial loop), we require an analysis of the predicted consequences of a LOCA occurring during the proposed partial loop operating mode (s). The licensee submitted an analysis for partial loop operation with one idle reactor coolant pump (three pumps operating) in Reference'9. This analysis concluded that the worst break was the 8.55 ft2 guillotine at the reactor coolant pump discharge, with Cp=1.0.

The worst 1 esk selected was located in the active leg of the partially idle loop. Placing the break at the discharge of the pump in an active cold leg of the partially idle loop -(instead of at.the discharge of the pump in an active cold leg of the fully active loop) yields the most degraded positive flow through the core during the first half of the blowdown and results in higher cladding temperatures.

The maximum cladding temperature for the one-idle-pump mode of operation was 17660F.

Our view of all input assumptions and conclusions resulted in a set of inquiries which were answered in References 4 and 10.

The results of a new analysis was submitted to reflect a more appropriate

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value of initial. pin pressure. The. original partial loop analysis in Reference 9 used an initial pin pressure of 1600 psi. As was demonstrated in-the time-in-life sensitivity study _in Reference 8, the worst pin pressure for this analysis should have been 7F' psi. The maxinum cladding temperature for the. reanalysis is 170 - 7, a value which is t-lwithin the criterion ~of 10 CFR 50.46. Therefore, this analysis may

.be used to support the licensee's proposed operation with one idle reactor. coolest pump.

l Since an analysis of ECCS cooling performance with one idle reactor coolant pump in each loop has not been submitted, power operation in this configuration-has been limited by Technical Specifications to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />..

-Single loop ' operation-(i.e., operation with two idle pumpa in one loop) is prohibited by Technical Specifications.

3.0 POWER DISTRIBUTION 4

We have reviewed the proposed Technical Specification in Reference 1.

Changes in the allowable heat generation rates as a fune*: ion of height in~the core have been accommodated by revision of the p,,wer-imbalance Specifications (Figure 3.5.2-2).

Only minor changes have been made in the Rod Position Limit Specification.

The reduction in allowable heat generation rate (kw/ft) in the lower half of the core has been accommodated by reducing the-allowable negative. imbalance at full power ay approximately 4%.

The. increase in allowable heat generation rate at 'the top of the core (as compared to that permitted by the Interim Acceptance Criterion) permits' relaxing the positive axial imbalance by3 approximately 3%..On the basis of our review, we find the 1

Technical Specification changes proposed in Reference 1 to be acceptable. We clarified proposed Specification 3.5.2.4 A(1) by slight wording changes, however.

4.0 ECCS

SUMMARY

We have completed our review of the Rancho Seco ECCS performance reanalyses and have concluded:~

a. The proposed Technical Specifications are based oc 4 LOCA analysis. performed in accordance with Appendix K to 10 CFR 50.

b. The'ECCS minimum containment pressure calculations were performed in accordance with Appendix K to 10 CFR 50.

c. The single failure criterion will be satisfied. This conclusion ric based on the additional Technical Specification requirements-

described in subsection 2.1 of this Safety Evaluation.

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.dm The proposed interim procedures for long-term cooling after a'LOCA are acceptable to us. The implementation of more permanent procedures during the next refueling outage is required to provide assurance that the ECCSLcan be operated in a manner which would prevent excessive boric acid concentration.from occurring.

.The proposed mode of reactor operation with one idle, reactor

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coolant pump is supported by a LOCA analysis performed in accordance with Appendix K to 10 CFR 50. Operation with one idle pump in each loop is restricted to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

5.0 JREACTOR BUILDING SURVEILLANCE AND SMUD' ORGANIZATION r

By reference.13 the licensee proposed to change the Technical

. Specifications concerning surveillance of the reactor building.

The reactor building is prestressed by a post-tensioning system consisting of vertical, horizontal, and dome tendons. Surveillance includes measurement of lift off forces for the tendons and removal of strands-for corrosion inspection. In order to remove strands from;the horizontal tendons, but not the vertical and dome tendons, detensioning is necessary. The ' existing Technica1' Specification erroneously indicates that detensioning will'be performed on vertical and done tendons as'well, which is neither necessary, pra tical.

nor included in Regulatory Guide 1.35.. The proposed Tec~.aical Specification change corrects the error in an! acceptable manner.

By reference-14 the licensee proposed Technical Specificatiun changes updating the SMUD organization chart in Figure 6.2-1.

"he proposed chart ' describes ' major corporate organizational components but -does not affect the facility organization or the review boards, and is-consistent with our requirements. We conclude that this change does not' affect the administrative control of the facility.

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6.0 CONCLUSION

We have concluded, based on the considerations discussed above, that:. (1)_there is reasonable assurance that the health and safety c

of.the public will not be~. endangered by operation in the proposed manner, and '(2) such activities will' be conducted in compliance with}the Commission's regulations, and the issuance'of this amendment

- willinot be inimical to the conunon defense and secucity or to the

- health and safety of the public.

Dated:

May119, 1976~

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' REFERENCES-

1. : Letter from E. K. Davis to Mr. Angelo Giambusso dated

-July 8, 1975.

2. '" Order for Modification'of Licensee," Letter from A. Schwencer

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to Mr. E. K. Davis dated. December 27, 1974.

3.

Letter from J. J. Mattimoe to Mr. Robert A.~ Purple dated MayL8, 1975.

4.

Letters from J. J. Mattinne e, Mr. Robert W. Reid dated

. November 17, 1975 and NovenLer 25, 1975..

5.

" Status Repot.c by the Directorate of Licensing in the Matter of Babcock and Wilcox ECCS Evaluation Model Conformance to 10 CFR 50, Appendix K," dated October 1974 6.

" Supplement 1 to the Status Report by the Directorate of

~ Licensing in;the Matter of Babcock and Wilcox ECCS Evaluation Model Conformance to 10 CFR 50, Appendix K,"' dated November 13, 1974.

J L7. B. M. Dunn, et al., "B&W's ECCS Evaluation Model," BAW-10104, Babcock and Wilcox, May 1975.

8.

R. C. Jones, et al., "ECCS Analysis of B&W's 177-FA Lowered-Loop NSS," BAW-10103, Babcock and Wilcox, June, 1975.

9.

Letter from J. J, Mattimoe to Mr. Angelo Giambusso dated August 1, 1975._-

'10.

Letter from Kenneth E. Suhrke to Mr. A. Schwencer dated December 15, 1975.

11. Letter from J. J. Mattimoe to Mr. Robert W. Reid dated October 15, 1975.

12. Letter from E. K. Davis to_ Mr. Angelo Giambusso dated March 12, 1975.

13. Letter from E. K. Davis to Mr. Angelo Giambusso dated May 23, 1976.

14._ Letter from J. J. Mattimoe to Director, Division of Reactor Licensing dated March 10, 1976.

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