ML19260D117

Safety Evaluation Supporting Amend 39 to License DPR-35

ML19260D117
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Site:	Pilgrim
Issue date:	01/08/1980
From:	Office of Nuclear Reactor Regulation
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0. 39 TO FACILITY OPERATING LICENSE N0. DPR-35 BOSTON EDIS0N COMPANY PILGRIM NUCLEAR POWER STATION, UNIT NO. 1 DOCKET NO. 50-293 I.

Power Level for Operability of Rod Worth Minimizer (RWM)

1.0 INTRODUCTION

By application dated March 10, 1977 Boston Edison Corrpany (BECo or licensee) proposed a change to the Pilgrim Unit 1 Technical Specification (TS) 3.3.B.3 to increase the power level, from 10% to 20%, below which the RWM must be operable.

2.0 EVALUATION The RWM restricts control rod selection to a preprogrammed pattern or order.

Use of the RWM prevents the selection of high reactivity worth control rods which, if upon withdrawal, become upcoupled and subsequently dropped out of the reactor core, could cause fuel damage. The basis for requiring RWM operability up to 10% of rated thermal power was an analysisl which indicated that above 10% power, even single operator errors cannot result in a dropped rod accident which could cause fuel damage, because of the lower rod worth when significant moderator voiding is present.

2 In the Fall of 1976 it was found that more recent analyses referenced by BECo in support of their Reload No. 3 application indicated that the RWM is assumed to be operable below 20% of rated thermal power to prevent fuel damage as a result of the postulated dropped rod accident.

BECo subse-quently committed to the use of the RWM during reactor operation below 20%

power and has administratively imposed this requirement, pending TS revision.

3.0 CONCLUSION

The proposed change is acceptable since it is consistent with the assumptions adopted in BECo's Reload No. 3 application which was approved on October 17, 1977 as part of License Amendment No. 27.

The language of the TS Bases 3

supporting this revision was modified to conform to the GE-STS,

1935 020 8002070 % 3 II.

Power / Flow Operating Map

1.0 INTRODUCTION

By letter dated December 28, 1977 BECo provided the staff with technical justification for operation limited by a rod block intercept line at power and flow conditions greater than the nominal 100% power / flow control line and less than the current rod block line. Operation in this manner pro-vides additional flexibility for power ascension while still complying with procedures to reduce pellet-clad interaction (PCIOMR's). The licensee believed that operation within the extended enve30pe is compatible with the current Technical Specifications. BECo, therefoPe, intended to make use of the added flexibility during power ascension methods and the effect of this change on abnormal operational transients has been considered for the Pilgrim Nuclear Power Station (PNPS).

2.0 EVALUATION All reactor safety analyses are based on power and flow constraints. These power / flow conditions are such that even with the occurrence of an abnormal operating transient, the core will be operated within safety limits.

BECo has provided the results of analyses and sensitivity studies to demonstrate that these criteria were met. These are discussed below.

2.1 Transients As shown in Reference 4, the three most limiting abnormal operational transients for PNPS are Turbine Trip Without Bypass (TTWOB), Loss of Feed-water Heater, and Rod Withdrawal Error (RWE). The transient analyses and sensitivity studies for the proposed change were perfonned with the same input parameters as those for the Reload 3 analyses. Because the end-of-cycle 4 (E0C4) scram reactivity insertion function is the most limiting condition, this curve was used for all analyses.

Each transient was analyzed at power / flow conditions of 100%/100%, 91%/75%, and 85%/61% (RWE was analyzed at only the first 2 points) to provide verification of transient behavior along the rod block intercept line to the point of rated power and flow. At the rated power / flow point the resultant transient behavior is the same as the previous analysis because the trip and rod block functions were not changed. The ACPR derived at the two lower values of power / flow are less than the ACPR for rated conditions for all transients except RWE.

2.1.1 Rod Withdrawal Error Since it is not apparent the RWE will not be the limiting transient at lower power levels along the rod block intercept line, the RWE was anlayzed along the rod block intercept line at the 91% power /75% flow, and the 100%/

100% point. At the 91%/75% point the RWE results in a aCPR that is the same as the value at the 100%/100% point. At points along the rod block intercept line between the 91%/75% point and the rod block intercept point 1935 021 (85% power /61% flow) the ACPR of the RWE may increase. However, similar analyses 7,8 have shown that any increase in the ACPR along this portion factor.

of the rod block intercept line can be accommodated by the Kf The Kf factor is normally used to provide margin for flow increase transients and will be at least 1.065 along the rod intercept line. The product of 1.065 and the MCPR operating limit is high enough to more than compensate for the potential increase in ACPR at the 85%/61% point.

For example, a conservative increase of 0.02 ACPR can be estimated for the rod block inter-cept point (85%/61%), and the corresponding compensation due to Kf would be about 0.08 in increased initial CPR. This compensation has been previously found acceptable (Reference 5) and is applicable for PNPS.

The APRM rod block setpoint is selected to allow for failed instruments for the worst allowable power profile.

It is demonstrated that even if the operator ignores all alarms during the course of this transient, the rod blocks will stop rod withdrawal when the CPR is 1.06 (the CPR safety limit).

it powers and flows lower than the 85%/61% condition withir, the proposed cerating envelope, a RWE results in smaller aCPR values.

The use of tne 5 esent Kf factors limits the control rod positions such that the resultirg M.'PR's are conservative and bound the ACPR due to a RWE. The consequences of the RWE transient decrease at lower flows and the effective MCPR required by the use of the Kf values become increasingly conservative. Thus, the MCPR of an RWE from the rod block intercept line or the rod block line will be greater than that for the RWE from rated conditions. The analyses pre-sented by BECo, and previous similar analyses reviewed and approved by the staff, show that this mode of operation is acceptable for all future cycles when transients that affect the entire core are considered (e.g.,

turbine trip without bypass, loss of feedwater heater). However, transients such as the RWE which affect only local portions of the core appear to be sensitive to the particular core configuration. Since the RWE may be the limiting transient at power levels below 100% rated along the rod block intercept line, similar analysis and justification should be provided for future cycles, if this new method of power ascension is to be used.

2.1.2 Peak Pressure Margin (25 psi Below Lowest Set Safety Value)

An analysis of the transient which involves main steam line isolation valve (MSIV) closure with high flux scram is used to evaluate compliance with the ASME pressure vessel code. The GE design criteria for adequacy of the safety valve capacity is a 25 psi margin between the peak vessel pressure and the ASME Boiler and Pressure Vessel code limit of 1375 psig based on a postulated MSIV closure transient with an indirect scram. BECo presented the results of analyses of the postulated transient, which showed the peak vessel bottom pressure at the rod block intercept point (85%/61%) is 1281 psig, 32 psi below that for the 100%/100% point and 94 psi below the code limit. At the intermediate point (91%/75%) on the rod block intercept line, the peak vessel bottom pressure is 1290 psig, also less than the 100%/100% point.

1935 022 2.1.3 Operating MCPR Limits for Less Than Rated Power and Flow A statistical analysis was perfomed to determine the part-load safety limit MCPR requirements along the APRM rod block line (Reference 6).

The results of the analysis show a small increase (<.01 at the rod block intercept point) in the safety limit MCPR requirement for part-load conditions due to increase in uncertainty of flow measurement.

However, this small increase in the part-load safety limit is more than compensated for by Kf factor-based operating MCPR limit for part-load conditions.

(For PNPS, the operating MCPR increases by about 0.08 due to the Kf factor, compared to a 0.01 decrease due to flow measurement uncertainty. ) The Kf factor is determined such tnat any inadvertent increase in core flow results in a MCPR greater than or equal to the safety limit MCPR at 100% power. Therefore, the safety limit will be satisfied when the plant is operating in accordance with the new power-flow operating envelope.

2.1.4 Xenon Transients The typical FSAR transient analysgs assume equilibrium xenon conditions.

To investigate the change in MCPR from non-equilibrium conditions, a localized xenon transient is simulated. On the basis of actual rod swap observations, a reactivity swing from equilibrium to peak of -0.05 ak/k was conservatively established to model the effect of xenon. This reactivity swing is input as a local reactivity change to the BWR siru-lator and associated change in the MCPR is calculated. This analysis is perfomed for both the rods-in and rods-out core configuration. This conservative, xenon induced, reactivity change through the extremes of rod configurations resulted in a ACPR of -0/09.7 Thus, the safety limit will not be violated for any potential xenon transient.

2.2 Thermal-Hydraulic Stability Analysis Themal-hydraulic stability analyses are presented in Reference 9.

Pre-viously in order to eliminate staff concerns on this topic PNPS precluded natoral circulation operation. With this elimination of natural circula-tion as a normal mode of operation the results of the stability analysis are acceptable (decay ratio <0.5).

2.3 Accidents The staff has reviewed the FSAR accident analysis for PNPS and agrees with the licensee that the change in power ascension methods and conditions will not significantly affect the consequences or probabilities of any accident sequence.

3.0 TECHNICAL SPECIFICATIONS 1935 023 In order to verify that the reactor is maintained within the analyzed bounds, the staff recommends Technical Specifications similar to those provided for Millstone Nuclear Power Station Unit No.1 by Amendment No.

52 to DPR-21 dated July 11, 1978. The licensee has agreed to such Technical Speci fica tions.

4.0 CONCLUSION

S PNPS has shown that the change in power a3cension methods do not signifi-cantly affect the consequences for any transient or accident previously analyzed and accepted by the NRC. On this basis we find this proposal acceptable with the provisions that (1) a Technical Specification be added to limit the power / flow conditions to within the analyzed limits as suggested in the text, and (2) a RWE analysis must be performed or appropriate justification provided for future cycles whic.h ensures that the change in critical power ratio (ACPR) for a RWE fron conditions on the rod block intercept line does not exceed the ACPR far an RWE from 100% power /100% flow conditions.

III.

Core Spary and Containment Cooling System

1.0 INTRODUCTION

By letter dated Novamber 13, 1979, BECo proposed changes to the Technu 7i Specifications for PNPS. The proposed changes would allow the suppresslin chamber (torus) to be drained and removal of up to one control rod drive mechanism (CRDM) without compromising core cooling capability. The change is required to permit modifications to be installed in the torus that will bring it into confonnance with the Mark 1 Containment Long Term Program acceptance criteria gd is similar to a previous change approved for Millstone Unit No. 1 The capability to remove a CRD mechanism with the torus drained will permit CRD maintenance and Mark 1 modifications to proceed in parallel, thus reducing the overall outage duration.

The core spray (CS) system is one of the low pressure emergency core cooling systems (ECCS) along with the low pressure coolant injection (LPCI) system, which protects the core in case of a loss-of-coolant accident (LOCA) when the reactor pressure is low. Draining the torus will remove the normal source of water for both CS pumps and the residual heat removal (RHR) pumps when in the LPCI mode. The Containment Cooling System (Suppression Pool Cooling) will also be deactivated with no water in the torus. The proposal includes permitting a single control rod to be withdrawn and its CRDM to be removed while the torus is drained.

In this situation, if the seal on the velocity limiter were to fail, a leak from the bottom of the core at a rate as high as 300 gpm could develop.

It is therefore necessary that an alternate source of water Se available if CS/LPCI is required. The licensee proposed the condensate storage tanks and the refueling cavity and dryer / separator pool as alternate sources of water. To supply this water to the core, both core spray systems must be available when work is being done that has the potential for draining the reactor vessel.

1935 024 2.0 EVALUATION In reviewing the proposed new TS 3.7.A.3 and 3.5.F.5, we have considered the possible ways in which water could be lost from the reactor vessel.

These include (1) leakage past the seal on the velocity limiter of the control rod, (2) inadvertent operation of valves or pumps in such a way that water flows from the core, or (3) the break of a line connected to the vessel.

The first case, draining of the reactor vessel at a rate up to 300 gpm if the control rod blade seal is unseated, is well within the capability of the core spray systems, assuming a single failure.

The second mechanism for a loss of reactor cooling water is a possible error in which a pump is started or a valve is opened such that there is a decrease in the amount of available water to protect the core. The licensee has addressed this mechanism and has concluded that sufficient controls are in effect to preclude the possibility of this mechanism existing when no work is being done which has the potential for dr:ining the reactor vessel. We concur. Thus, no requirements are placed on the operability of CS and Containment Cooling Systems with the torus drained, unless work is being done which has the potential for draining the reactor vessel. The licensee has analyzed this mechanism for the case when work is being done and concluded that the rate of coolant loss by inadvertent operation of any single pump or valve would be less than the leakage past the control ' od flow limiting seal. Therefore, the core would be protected against a ioss of water of this magnitude when work is being performed which has the potential for draining the reactor vessel.

The third mechanism for loss of reactor water is a pipe break.

Because the system is not pressurized in the refueling mode, the probability of a significant break is negligibly low and was not considered further.

A total capacity of 224,000 gallons of water is available above the reactor vessel in the refueling cavity and another 165,000 gallons are available in the dryer / separator pool. The requirement of a minimum level at 114 feet elevation will assure at least 350,000 gallons of water are available.

With no action, a leak from the reactor vessel to the drywell would result in flooding approximately 120,000 gallons into the drywell, at which point the torus would begin to fill by flow into the vent pipes from the drywell. At a leak rate of 300 gpm, sufficient water will accumulate in the torus to provide minimum NPSH requirements for the LPCI and/or core spray pumps after about 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />, assuming the torus is intact.

We have reviewed the list of modifications to be performed in the torus,10 and conclude that the capability to store at least 140,000 gallons (mini-mum NPSH) will not be degraded. Thus, torus integrity will be adequate for this function.

The postulated leak could continue unabated for at least 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> befc r2 3 5 02 5 water level would drop below the upper edge of tne reactor vessel.

Therefore, there is sufficient time to take other emergency measures, if necessary.

In addition, there is a low water alarm for the refueling cavity and automatic activation of the available low pressure ECCS upon a low-low water level in the reactor vessel.

Although no credit is given for the condensate storage tanks (CTS) in the above analysis, TS 3.5.F.5.b will require at least 200,000 gallons of useable water in the CST aligned with a suction path to the CS system.

The CST has a low level alarm to alert the operator in time to take corrective action. The CST will be available for ECCS with the torus drained.

3.0 CONCLUSION

The proposed Technical Specifications will provide adequate assurance that the core will remain covered in the refueling mode with the torus drained.

Protection is provided for (1) a leak of water past a velocity limiter on a control rod during CRDM maintenance, (2) a break in the reactor system piping, and (3) an inadvertent error in the opening or closing of a valve or starting a pump in such a way that water is lost from the reactor core. The protection is provided by assuring that adequate sources of water and methods of supplying this water are avail-able. With the torus drained, the primary source is the refueling cavity and dryer / separator pool water and the backup source is the CST. This water would be supplied by the Core Spray System or LPCI.

IV.

Minimum SRM Count Rate Requirements

1.0 INTRODUCTION

By letter dated November 21, 1979, Boston Edison Company proposed an amendment to the Technical Specifications for the Pilgrim Nuclear Power S ta ti on.

The effect of the amendment would be to allow the count rate in the Source Range Monitor (SRM) channels to drop below 3 counts per second (cps) when the entire reactor core is being removed or replaced.

The present Technical Specifications require that a count rate of at least 3 cps be maintained whenever one or more fuel assemblies are present in the core.

2.0 DISCUSSION During any core alteration, and especially during core loading, it is necessary to monitor flux levels.

In this manner, even in the highly unlikely event of multiple operator errors, there is reasonable assur-ance that any approach to criticality would be detected in time to halt operations.

I9[35 O2b The minimum count rate requirement in the Technical Specifications accom-plishes three safety functions:

(1) it assures the presence of some neutrons in the core, (2) it provides assurance that the analog portion of the SRM channels is operable, and (3) it provides assurance that the SRM detectors are close enough to the array of fuel assemblies to monitor core flux levels.

Unloading and reloading of the entire core leads to some difficulty with this minimum count rate requirement. Whea only a small numoer of assemblies are present within the core, the SRM count

• ate will drop below the minimum due to the small number of neutrons being produced, and due to attenuation of these neutrons in the water and control blades separating the fuel from the SRM detectors.

Past practice has been to connect temporary " dunking' chambers to the SRM channels in place of the normal detectors, and to locate these detectors near the fuel.

Besides being operationally inconvenient, dunking chambers suffer from signal variations due to their lack of fixed geometry. Moreover, the use of dunking chambers increases the risk of loose objects being dropped into the vessel.

3.0 EVALUATION 3.1 Minimum Flux in the Core A multiplying medium with no neutrons present forms the basis for an accident scenario in which reactivity is gradually but inadvertently added until the medium is highly supercritical. No neutron flux will be evident since there are no neutrons present to be multiplied. The introduction of some neutrons at this point would cause the core to undergo a sudden power burst, rather than a gradual startup, with no warning from the nuclear instrumentation.

This scenario is of great concern when loading fresh fuel, but is of lesser concern for exposed fuel.

Exposed fuel continuously produces neutrons by spontaneous fission of certain plutonium isotopes, photo-fission and photodisintegration of deutrium in the mod:rator. This neutron production in exposed fuel is normally great enough to meet the 3 cps minimum for a full core after a refueling outage with the lumped neutron sources removed.

Thus, there is assurance that a minimum flux level will be present as long as some exposed fuel is present. We therefore find the proposed amendment to be acceptable from the point of view of minimum flux provided the words " spiral reload" in proposed specification 3.10.8.4, pg. 203, are interpreted to mean " reload of fuel which has previously accumulated exposure in the reactor." We do not find the emendment to be applicable to the loading of a new core containing only fresh fuel.

Such a loading must use lumped neutron sources and dunking chambers to meet the normal 3 cps minimum count rate.

1935 027

_g-With the agreement of the licensee, we have therefore modified TS 3.10.B.4 to read:

"During spiral reload, each ggntrol cell shall have one assembly with a minimum exposure of 1000 MWD /t.' 8 3.2 SRM Operability Specification 4.10.B requires a functional check of the SRM channels, including a check of neutron response, prior to making any alteration to the core and daily thereafter. This would be sufficient for~ core unloading and reloading, except that the more extensive fuel handling operations involved imply a greater possiblity of SRM failure.

During spiral unloading and reloading, Proposed Specification 3.10.B.4 would increase this frequency to every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or, as an alternative, allow some exposed fuel to be loaded adjacent to the SRM detectors to provide a minimum 3 cps count rate continuously. We agree that this increased testing is sufficient.

3.3 Flux Attenuation The four SRM detectors are located, one per quadrant, roughly half a core radius from the e. enter. Although these are incore detectors and thus very sensitive when the reactor is fully loaded, they lose some of their effectiveness when the reactor is partially defueled and the detectors are located some distance from the array of remaining fuel.

GE's spent fuel pool studies have shownl2that 16 or more fuel assemblies (i.e., four or more control cells) must be loaded together before criti-cality is possible.

In spiral loading sequences in the Pilgrim core, an array containing four or more control cells will be at mos' tw control cells (i.e., about t.,o feet) away from an SRM detector.

We have pre-viously examined the sensitivity loss in such a case on another dccket,13 and found it to be at most one decade of sensitivity (i.e., about one fifth of the SRM's logarithmic scale). As in Reference 13, we found this to be acceptable.

4.0

SUMMARY

AND CONCLUSION At this point, we have examined ai; these safety issues and found the proposed amendment to be accepable provided it is understood that spiral reload will include a significant quantity of exposed fuel.

With the change described in 3.1 above, we find this amendment to be acceptable.

ENVIRONMENTAL CONSIDERATION 1935 028 We have determined that the amendment does not authori.e a change in effluent types or total amaunts nor an increase in power level and will not result in any significant environmental impact.

Having made this determinacion, we have further concluded that the amendment involves an action which is insignificant from ohe standpoint of environmental impact and pursuant to 10 CFR Section 51.5(d)(4) tnat an environmentai impact statement, negative declaration, or environmental impact aopraisal need not be prepared in connection with the issuance of the amendment.

CONCLUSION We have concluded, based on the considerations discussed above, that:

(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a sig-ificant decrease in a s?.fety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) sucr activities will be t.onducted in compliance with the Commission's regulations and the issuance of the amendment will not be inimical to the comon defense and security or to the health and safety of the public.

Dated:

January 8, 1980 1935 029 References 1.

NED0-10527 " Rod Drop Accident Analysis for Large Boiling Water Reactors" March 1972.

2.

Supplement 4 to NED0-20360 Revision 1 General Electric BWR Generic Reload Application for 8x8 Fuel, April 1, 1976.

3.

General Electric Standard Technical Specifications NUREG-0123 Revision 2 dated August 1979.

4.

General Electric Boiling Water Reactor Reload No. 3 Licensing Submittal for Pilgrim Nuclear Power Station Unit 1, May 1977 (NED0-21462-01).

5.

Memo to X. Goller from V. Stello, " Review of Millstone Unit 1 Reload 3 TAR 1775," dated October 30, 1975.

6.

Final Safety Analysis Report, Millstone Nuclear Power Station Unit 1, Docket No. 50-245, License No. DPR-21.

7.

Millstone Point Nuclear Power Station, Unit 1, Load Line Limit Analysis License Amendment Submittal, June 1976 (NED0-21285),

8.

"Nine Mile Point Nuclear Power Station Unit 1, Load Limit Line Analysis,"

NED0-24012.

9.

NED0-24058 " Pilgrim Nuclear Po,;ar Station Load Line Limit Analysis" September 1977.

10. BEco letter #79-92 dated May 9,1979 regarding Schedule

- the Implementation and Resolution of the Mark I Containment Loag Term Program.

11. Amendment No. 46 to DPR-21 for Millstone Nuclear Power Station Unit No. I dated March 10, 1978.

12. General Electric Standard Safety Analysis Report, 251-GESSAR, Section 4.3.2.7, pg. 4.3-27.

13.

" Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 27 to Facility Operating License No. DPR-63,"

Docket No. 50-220, enclosed with letter, T. A. Ippolito (NRC) to D. P.

Dise (Niagara Mohawk Power Corporation), dated March 2,1979.

1935 030