ML14175A844

Safety Evaluation Supporting Amend 87 to License DPR-23. SERs for Cycle 10 Reload Application Chapter 15 Events & Cycle 10 Reload Analysis - LOCA Analyses Encl

ML14175A844
Person / Time
Site:	Robinson
Issue date:	11/07/1984
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML14175A843	List: ML14175A842 ML14175A844 ML14175A845 ML20099J505
References
NUDOCS 8411290087
Download: ML14175A844 (96)

Text

o UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 87 TO FACILITY OPERATING LICENSE NO. DPR-23 CAROLINA POWER AND LIGHT COMPANY H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261

1.0 INTRODUCTION

By applications dated July 23, 1984 and August 1, 1984 and supplemental information dated August 8, and 20(2), 1984; September 7(2), 1984; and October 4, 12, and 22, 1984, Carolina Power and Light Company (the licensee) requested amendment to Facility Operating License No. DPR-23 for the H. B. Robinson steam Electric Plant, Unit No. 2 (the facility) to permit operation for Cycle 10 at full power (2300 Mwt).

Prior to the July 23, 1984 amendment request, documents in support of the forthcoming core reload were submitted by letter dated October 5, 1983. The supplementing letters provided information as follows:

1. August 8 and 20(2), 1984 provided confirmatory analysis in accordance with the July 23, 1984 application letter.

2. September 7, 1984 (84-366) provided Technical Specification (TS) changes resulting from our review due to clarifications, error corrections, and consistency within the TS. No significant changes were made.

6411290087 841107 PDR ADOCK 05000261 P

POR

-2

3. September 7, 1984 (84-400) resubmitted a K(z) curve for the Technical Specifications. Confirmatory analysis, since the original (July'23, 1984) submittal, necessitated minor revisions to the curve. This was a minor revision to an already submitted document resulting from a standard analytical process, therefore, it was not a significant change or a new submittal.

The amendment consists of:

a. Appendix A Technical Specification (TS) changes resulting from the analysis required for the Cycle 10 core reload,

b. Appendix A Technical Specification (TS) changes to allow the" performance of certain control rod drive evolutions when containment integrity in not intact, and

c. Appendix A Technical Specification (TS) change of an administrative nature such as deleting references to N-1 loop operation and clarifications within the TS and the FSAR.

The H. B. Robinson Unit 2 (HBR-2) plant has operated at reduced power since Cycle 8 in order to minimize degradation of the steam generators.

The steam generators have been replaced during the current outage and HBR-2 intends to operate at full power (2300 Mwt) during Cycle 10 and subsequent cycles., HBR-2 will install, during Cycle 10 refueling, Part Length Shielding Assemblies (PSLA's).

The PSLA's are designed to reduce the fast neutron flux to the pressure vessel weld seams by a factor greater than 7, thus, preventing the

vessel from reaching the pressurized thermal shock screening criteria prior to expiration of the current operating license.

In order to accommodate the rated power level, power distribution and the concurrent use of PLSAs the licensee requested thermal margin T

relief for Cycle 10 (and subsequent cycles) i.e., F = 2.32 and F AH 1.65 and corresponding revision of certain reactor protection system setpoints.

The peak discharge fuel assembly exposure is estimated at 44,000 MWD/MTU.

In support of these changes for Cycle 10 operation, the licensee submitted:

1. Document XN-NF-84-74, "Plant Transient Analysis For H. B.

Robinson Unit 2 At 2300 MWt With Increased F H."

The document presents the analysis of the SRP Chapter 15 transient and accident events and,

2. A revised LOCA Analysis,

3. A Cycle 10 core reload report including Technical Specifications modifications.

The Safety Evaluation Reports (SER) for Cycle 8 ana ycle 9 required the licensee (if it continued to rely on Exxon analyses) to develop a stand-alone Chapter 15 analysis methodology. As a consequence, Exxon Nuclear Company (ENC) developed a stand-alone methodology which is at present under staff review.

The licensee has requested and provided justification to defer submittal of the steam line break event until January 31, 1985.

The request was made to provide Exxon Nuclear Company time to develop an acceptable methodology for analyzing steam line break events.

Further Uetails concerning this request and the staff's independent analysis is contained in Attachment I of this SER.

S4.

2.0 EVALUATION Reload Fuel Design The H. B. Robinson-2 core consists of 157 fuel assemblies, each having a 15x15 fuel rod array. There are 204 fuel rods, 20 control rod guide tubes and one instrument guide tube.

Each fuel assembly has seven zircaloy spacers, inconel springs and zircaloy cladding. There are 65 fresh assemblies supplied by ENC including 12 PLSAs located at the core flats and especially designed to reduce the fast neutron flux to the lower girth weld seam of the pressure vessel.

The lower 42 inches of the PLSAs contain a 304 stainless steel column instead of fuel but otherwise are identical in design with all other assemblies.

The new assemblies of this reload are axially blanketed, except for the lower part of the PLSAs.

Fuel exposure for Cycle 10 has been based on a Cycle 9 exposure of 10,637 MWD/MTU and is estimated to be 10,820 MWD/MTU (312 EFPDs) with an estimated peak exposure of 34,705 MWD/MTU. The basic Exxon Nuclear design and design methods and the extended burnup mechanical design are described in Reference 2 an& 3, respectively. The mechanical design for the PLSAs is covered in Reference 4.

The 65 new assemblies (designated XN-7) include 36 which contain gadolinia bearing pins. The Cycle 10 fuel assembly design parameters are listed in Table 4.1 of Ref. 10.

Fuel Mechanical Design The mechanical design of the new reload assemblies is identical with that of previous reload assemblies with the following exceptions:

(a) they contain a natural uranium blanket (6 inches in length) at the top and bottom, and (b) the column insulator discs are no longer used.

The PLSA design is similarly identical to previous reload designs with the exception as noted that the lower 42 inches of fuel is replaced with 304 SS.

Because

-5 of the presence of the stainless steel the following aspects of the mechanical

-thermal design needed confirmation:

(a) thermal expansion effects for the stainless steel, (b) loss in assembly hold down capability due to the lower weight of the rods, (c) sensitivity to irradiation induced bowing and (d) the seismic stability of the lower weight assemblies. These issues have been discussed and the PLSA mechanical design has been approved.

(See topical report XN-NF-83-71, Reference 4).

Fuel Thermal-Hydraulic Design The thermal-hydraulic design of the Cycle 10 assemblies is identical to that of previous reloads, assuring compatibility.

The original Exxon analysis was documented in Ref. 7. The philosophy followed in the analyses was to choose a bounding fuel assembly power distribution at an exposure which has the worst radial peaking. Similarly to assure that the lowest value of the DNBR was accounted for, a bounding local power distri bution was used with the maximum radially peaked assembly.

The analysis-used a-5% lower plenum factor, 4.5% core flow bypass, a low estimate of the total reactor coolant system (RCS) flow rate, 6% steam generator tube plugging and an additional 3% reduction to account for flow measurement uncertainty. The results of the thermal-hydraulic analysis (which are given in ref. 7) are used as bases for the analyses of the anticipated operational occurrences.

Thermal-Hydraulic Analysis A bounding fuel assembly power distribution with the limiting values of FaH = 1.65 and F = 2.32 were utilized in order to assure the computation of the lowest DNBR results. The primary flow used in the safety analysis is based on an assumed 6% steam generator tube plugging and a 31 plant calorimetric flow measurement uncertainty.

This minimum primary flow will be assured by adjusting the low flow trip set points after the full power plant calorimetric is performed.

A Technical Specification change will be effected if required.

Core flow is based on a 4.50 core

flow bypass and a 5% lower plenum inlet flow maldistribution factor In this manner, the analysis considered all fuel types in a bounding manner and is acceptable.

Fuel Rod Bowing The core reload for Cycle 10 consists of Exxon fuel assemblies for which the hydraulic design is similar to the existing fuel.

According to the approved topical report XN-NF-75-32 (reference 19)

Exxon fuel must be reviewed for possible rod bowing penalty as a function of burnup.

For the H. B. Robinson Cycle 10 fuel design, burnup to 47,000 MWD/MTU are acceptable without penalty.

The calculated maximum design assembly burnup is 44,000 MWO/MTU and, there fore, no rod bow penalty needs to be applied.

NUCLEAR DESIGN Core Characteristics The H. B. Robinson-2 Cycle 10 is neutronically similar to Cycle 8; however, it differs in major respects from the previous reloads in that it will operate at a power level of 2,300 MWt and it incorporates 12 PLSAs to minimize fast flux irradiation of the lower girth weld seam of the pressure vessel.

In addition, the 65 new assemblies have a natural uranium axial blanket and utilize 4.0 w/o gadolinia. The average loading enrichment is 3.08 w/o in U-235 and the maximum is 3.34 w/o U-235. The estimated exposure of Cycle 10 has been based on a Cycle 9 exposure of 10,637 MWD/MTU and it is estimated to be 10,820 MWD/MTU or the equivalent of 312 EFPD.

The peak assembly exposure is estimated to be 34,705 MWD/MTU.

Power Distribution At full power (2,300 MWt) and equilibrium xenon conditions (100 MWD/NTU),

the calculated FAH = 1.48 and the peak F = 2.18, includiny a 4% and 5% measurement uncertainty respectively.

In addition F includes a 3%

engineering factor and an 11% allowance for operation with the Power Distribution Control-II (PDC-11) within t5% target bands.

For both, i.e.,

F H and FQ, the maximum value will occur at 5,000 MWD/MTU and their

-711 respective calculated values are 1.56 and 2.18. These values were calculated with approved methods (Ref. 5) and are within the limits of the Technical Specifications.

Reactiv Control Requirements The H. B. Robinson-2, isothermal temperature coefficient at Hot Zero Power (HZP) and Hot Full Power (HFP) conditions at beginning of cycle (BOC) and end of cycle (EOC) are shown in Table 6.1 of Reference 10.

The same table also compares the corresponding values of cycle 8 and cycle 10 for the boron worth (ppm/10 3 pcm) at HZP and HFP, the prompt neutron lifetime ( Psec) and the delayed neturon fraction.

The values for the two cycles are nearly identical.

The isothermal temperature coefficient at HFP and BOC10 is estimated to be -5.1 pcm/oF at a critical boron concentration of 1,002 ppm, and at HZP, BOC10 is -.7 pcm/oF at a critical boron concentration of 1,134 ppm.

At both extremes the value is negative.

The moderator temperature coefficient at HZP condition for the 8OC10 is estimated at.+1.0 pcm/oF, the value used in the transient analysis and requTred in the Technical Specifications is +5.0 pcm/oF (Ref. 8, 10).

Similarly at HFP the estimated value of the moderator temperature coefficients is -3.8 pcm/oF, well below the required value of 0.0 pc m/oF.

Control rod worths and shutdown margins have been calculated for cycle 10 and are summarized in Table 6.2 of Ref. 7. Control rod worths for cycle-10 are slightly higher than the corresponding values for cycle 8 as expected.

For cycle 10 the power distribution is higher in the core interior due to the PLSAs in the core flats.

The required shutdown margin in the two cycles is assumed the same, i.e., 1,770 pcm, thus the excess shutdown margin is 461 pcm for cycle 10 vs 565 for cycle 8 at EOC.

The corresponding values for BOC are 1,911 vs 1,554.

Hence, the shutdown margins for both cycles are comparable.

The control rod groupings shall remain the same.

Power distribution control is to be effected following the Exxon procedures known as Power Distribution Control, Phase 11 (PDC-11, Ref. 7).

The topical report and two supplements have been approved by the staff.

The analytical methodology for the neutronic calculations (Ref. 5) has also been approved. The reference neutronic analysis was performed using XTG (Ref. 11)a two-group, three dimensional coarse mesh code. The cycle power distribution is calculated using PDQ/HARMONY.

(Ref. 12, 13)

e-8 0

Because the results discussed above have been obtained using approved methods, and were used in the safety analyses with appropriate calculational uncertainties and are included in the Technical Specifications, we find these results to be acceptable.

SAFETY ANALYSES For the new set of parameters, namely the increased power level, FAH and F the values have been reviewed to determine which ones influence Q

the results of the transient and accident analyses. It was determined that the following events needed to be reanalyzed:

0 Excess Load a

Scram shutdown margin o

Steam generator tube rupture o

Loss of load o

Loss of normal feedwater o

3-pump coastdown 0

Locked rotor o

Uncontrolled rod 'withdrawal (subcritical or low power) o Uncontrolled rod withdrawal (at power) o RCCA misalignment o

CVCS malfunctions with decreasing boron concentration 0

Refueling 0

Startup o

RCCA ejection o

LOCA fuel damage limits The nominal plant rated operating conditions and the nominal core and fuel design parameters used in the accident analyses are listed in tables 15.0.2-1 and 15.0.2-2 of Ref. 8.

The axial power distribution used for transients which do not require power redistribution is shown in Figure 15.0.3-1 of the same reference. The nuclear enthalpy rise factor is 1.65, the axial peaking factor is 1.65, the total heat flux

peaking factor is 2.32, and the fraction-of power generated in the fuel is.974. Operating parameter ranges and reactivity coefficients used in the analyses are shown in Tables 15.0.4-1 and 15.0.5-1 of ref. 8.

A discussion of the event analyses vs the acceptance criteria will follow.

Excess Load Excess load event manifests itself whenever there is a rapid increase in the heat removal from the reactor coolant without a corresponding increase in thte reactor power.

This power-energy removal mismatch results in a decrease of the reactor coolant temperture and pressure. Hence, when the moderator temperature reactivity coefficient is negative an increase in power may occur. If there is a positive temperature coefficient the power will decrease and will not produce a challenge to the acceptance criteria.

This event constitutes a challenge to the Specified Acceptable Fuel Design Limits (SAFDL). The conditions for the minimum SAFOL margin are full power and maximum feedback at EOC.

The event initiator was a 10% step increase in turbine steam flow. The reactor control is assumed to be in automatic.

The secondary and primary pressures initially fall and the primary temperature will fall resulting in reactivity insertion from the moderator coefficient and a decrease of the pressurizer level.

Additional reactivity will be inserted from the control rods which initiate withdrawal to.

increase power responding to the load demand.

In about 80 sec a new steady state will be reached.

The minimum DNBR computed was 1.33 which is above the limit of 1.17. Hence, this transient is acceptable.

For the analysis the PTSPWR2 code was used to provide input to XCOBRA-IIIC.

This methodology is under review by the staff. However, the review has progressed to the point. where the portions pertinent to this application have been found to be acceptable Scram Shutdown Margin The particular quantity of interest here is the shutdown margin after trip. This is part of the inadvertent opening of a steam generator or reload safety valve. This event is most limiting at the end of cycle.

There is adequate shutdown margin at BOC10.

The required margin is 1,000 pcm and the excess margin is 1,911 pcm. (Reference 10).

The required analysis will be performed and be submitted for review during CY85 i.e., before the EOC10.

This is acceptable.

10 Steam Generator Tube Rupture A rupture of a steam generator tube will release primary coolant into the lower pressure secondary.

This event is similar (and bounded by) the inadvertent opening of a pressu izer PORV and was reported in reference

10.

In that analysis a single valve was assumed to have failed open at full power.

The maximum relief capacity was set at 288,000 lb/hr at the PORV pressure setting.

The initial conditions included 102% of rdted power (2,436 psia), a primary pressure of 2,220 psia, F = 1.65 and F A H Q

2.32.

No credit was taken for lowering of the primary coolant pressure.

The result of this calculation was a MDNBR value greater than the limit of 1.17 allowed in the XNB DNB correlation. Thus, an assumption of no fuel failure is acceptable for the evaluation of the radiological con sequences of the transient, which were reviewed and were also found to be acceptable.

Loss of Load The loss of load event is an undercooling transient that results from station disconnection fron the grid, turbine trip or electrical generator malfunctions.

Following the loss of load the main steam stop valve closes caus-ng a large mismatch between reactor power output and heat removal which in turn causes a secondary temperature and pressure increase.

As the primary to secondary AT decreases the primary to secondary heat transfer decreases and the primary temperature and pressure will rise.

Assuming that the reactor does not trip after the turbine trip the high primary pressure will trip the reactor and open the primary safety valves.

Energy fron the systen will also be removed through the steam generator relief valves.

The primary challenge of this transient is to the primary system overpressurization acceptance criterion (peak pressureil.10% of the design value) and the secondary challenge is to the SAFDL because of the increasing primary temperature.

The purposes of the analyses for this transient were to maximize the overpressurization and the SAFOL challenge, hence, the input parameters were biased to maximize the overpressure and minimize DNBR respectively.

-e11*

The analyses were carried out with the PTSPWR2 and the XCOBRA-IIIC methodology. The results indicate that for the maximum pressure case (from 102% of full power) a 270 F rise in the cold leg temperature occurs at 14.5 sec into the transient. The reactor trips at 6.80 sec (on high pressure) and the safety valves open at 6.83-sec.

The maximum pressure at pump discharge was computed at 2,661 psia which is below the 110% of the 2,500 psia (i.e., 2,750 psia) of the design pressure.

For the MDNBR case the value computed is 1.19 which is above the 1.17 limit. Because these values are within the NRC acceptance criteria of SRP section 15.2.1 we find these analyses acceptable.

Loss of Normal Feedwater Loss of feedwater coulo result from loss of feedwater pumps, feed control valve malfunction, loss of offsite power etc.

Loss of feedwater flow results in a decrease of steam generator water level, decrease in secondary system heat removal capability and increase in primary pressure and temperature. The reactor trips due to the steam generator low-low water level signal.

The objective of the analysis is to demonstrate the adequacy of the steam generator inventory and relief capacity and proper setpoint of the safety valves, to prevent primary pressure from exceeding 110% of the design pressure of 2,500 psia. In the analyses, the single failure event is the failure of the steam driven auxiliary feedwater pump to start and it is assumed that the primary pumps are on or tripped. (two different cases)

In addition the turbine trips with simultaneous closure of the turbine stop valve, the main fee'dwater valves are ramped closeG and the diesel generator is initiated with specified delay. The analyses were carried out with the Exxon code SLOTRAX.

(Ref. 14)

The results indicate that the primary pump off case challenges the primary pressure limit and the pumps on case the minimum steam generator inventory criterion. The pressure was estimated to reach the primary relief setpoint in 1.5 sec with a required relief valve rate of 215 lb/sec.

The rated rate is 240 lb/sec, therefore, pressure

012 will not rise above the valve setpoint. For the pumps-off case the minimum steam generator inventory is 75% of the initial amount.

The SAFDLs are bounded by the loss of flow event. We find that the results of the analyses meet the SRP Section 15.2.7 criteria and they are acceptable.

Three Pump Coastdown Loss of all three primary coolant pumps can result from loss of electric power to the pump motors.

Following loss of power the pumps will coast down and that is governed by the pump flywheel inertia.

Loss of primary coolant when-the reactor is at power will result in rapid coolant temperature rise and corresponding reduction in ONB margin or even ONB if the reactor is not tripped promptly. Loss of primary flow will result in temperature and pressure rise, which will be mitigated by the primary system safety valves.

This event results in a heatup transient and challenges the 110% of-design pressure criteria and the MDNBR criteria.

Reactor trip signals are basedon signals from pump motor power supply undervoltage or underfrequency and reactor coolant low flow.

In the analysis only the low flow trip is taken into account.

The transient is governed by the initial overpower DNb margin,.rate of flow.degradation, low reactor coolant flow trip setpoint, available shutdown reactivity and the moderator temperature coefficient. The objectives of the analyses are to investigate the overpresssurization limit and the MONBR value. The analysis methodology is based on the PTSPWR2 code providing the input to the XCOBRA-lI1 code.

For both cases analyzed it is assumed that. reactor control is in manual the power level is at 2,346 MWt, i.e., rated + 2%, reactor trip setpoint of low flow-3%, moderator temperature coefficient at 1.2 BOC and the pellet to clad heat transfer coefficient at the maximum value.

The results indicate a maximum primary bounding pressure of 2,574 psia which is lower than the 2,750 psia (110% of 2,500 psia) and the MUNBR at 1.21 which is greater than the allowed minimum value of 1.17.

These values meet the criteria of SRP Section 15.3.1 and 2 and, therefore, we conclude that the results are acGeptable.

- 13 Locked Rotor for a Reactor Coolant Pump

.This event is caused by an instantaneous seizure of a primary reactor coolant pump rotor, when flow through the affected loop diminishes rapidly.

A reactor trip signal on low primary flow will be initiated and the reactor will trip.

However, because of the reduced flow the primary temperature and pressure will rise.

Following turbine trip the secondary temperature (and pressure) will rise, further reducing primary to secondary heat transfer and further increasing primary temperature and pressure. Primary pressure will be relieved by opening of the safety valves.

The rapid rise of the primary temperature will cause a rapid decrease of the DNB and the primary pressure rise will challenge the primary boundary pressure limit.

The analysis method is based on PTSPWR2 and the XCOBRA-IIIC codes. The objectives of the analyses were to estimate the peak primary pressure, the extend of fuel failures and that the possible radiological consequences are-bounded by the limits of 10 CFR 100.

The initial conditions assumed (among others):

power at rated +2% (i.e., 2,346 MWt), maximup pellet to clad heat transfer coefficient and manual reactor control.

The characteristics of this transient are unique in the sense that the flow will be reversed in the affected loop due to the higher pressure in the reactor vessel and the effective core coolant flow will be about 60% of normal at 40 sec into the transient.

Increased temperature will cause increased power to 107.6% of rated due to the assumed positive moderator temperature coefficient.

The maximum pressure is reached at 3.51 sec of 2,516 psia which is less than the allowable value 2,750 psia. However the MDNBR.will be.90 i.e., less than the allowed by the XNB correlation of 1.17. The estimated fuel failures used in the evaluation of the radiological consequences are 55% of the total number of assemblies. The radiological consequences are bounded by those of the LOCA, and are within the Standard Review Plan guideline limits of 10 CFR 100.

The results of the analysis are acceptable.

14 Uncontrolled Rod Cluster Withdrawal This event results from an uncontrolled rod bank withdrawal and it can result in a rapid and large reactivity insertion. The maximum insertion rate is determined from the worth of the rdd bank.

We distinguish three sub-events depending on initial power, i.e., subcritical, low power and full power. The analysis is performed using the PTSPWR2 and XCOBRA-IIIC codes.

(a) Subcritical and Low Power In this case the malfunction will result in a rapid and large reactivity insertion which will be terminated by the low range setting of the power range flux trip. The reactivity insertion is countered initially (promptly) by the Doppler effect followed by rod insertion. Other trips for this event are:

source and intermediate range flux trips, inter mediate range rod block and low and high power range trip settings. In the subcritical case the objective of the analysis is to determine the fraction of fuel exceeding the MDNBR limit for input to the radiological consequences evaluation.

The conditions of the analysis maximized power and minimized core flow. A peak surface heat flux equivalent to 69% of full power was reached at 16 sec into the transient. The MDNBR value (for the XNB correlation) was 1.26, which is above the 1.17 limit.

Since there is no core damage, these results are acceptable.

(b) Power Operation Power operation for the purposes of this analysis is defined between 10-100% of rated power. In this transient the overpower AT is set to protect against MDNBR. The power range reactor trip protects against high power levels. The objective of the analysis is to examine the broad range of reactivity insertion possible and assure of the adequacy of the trip setpoints to meet the acceptance criteria i.e., on primary pressure and MDNBR. The cases analyzed included 10%, 60% and 102% of rated power and negative and positive reactivity feedbacks. The rated power reactivity insertions were found to be bounding with respect to MDNBR for the lower power ratings. From full power with positive

15 moderator temperature coefficient the nuclear high power trip (118% of rated power) is reached in 1.84 sec. and the MDNBR reaches 1.32 at 3.02 sec.

The pressure increases to a maximum of 2260 psia.

The results of the analysis indicate that neither of the acceptance criteria is violated, hence, the results of this transient are acceptable.

Rod Control Cluster Assembly Misalignment Analyses for this event (with its subdivisions) have been submitted in Ref.

15, (which is Supplement 1 to Ref. 8).

Rod Control Cluster Assembly (RCCA) misoperation includes, withdrawal of a single full length RCCA, static misalignment of a RCCA, a dropped full length RCCA and a dropped RCCA bank.

The conditions to be satisfied for the first are primary pressure, core coolability and radiological consequences for 10 CFR 100.

The other three should satisfy primary pressure limit and the MDNBR limit.

The withdrawal of a RCCA at power will insert positive reactivity and cause increased power generation. If the secondary does not respond to the increased power-production the temperature and pressure of the primary will increase. A single RCCA withdrawal will cause in addition severe power redistribution in its immediate neighborhood, hence, locally it is possible that some assemblies may experience boiling transition and some fuel failure.

The overtemperature AT trip will afford the.primary protection.

This transient is evaluated with the PTSPWR2, XCOBRA-IIIC codes and: power at 102% of nominal, radial peaking factor of 1.27 and conservative values for the moderator temperature and Doppler coefficients.

The system pressure is estimated to reach a peak value of 2,275 psia, however, the MDNBR will (locally) reach.64 i.e.,

below the limiting value of 1.17. The estimated fuel failures to be used for the evaluation of the radiological consequences are estimated to be 7.8% of the total number of the fuel assemblies. The radiological consequences of this evaluation are acceptable.

16 For the other three cases i.e., static misalignment, dropped full length RCCA and RCCA bank the PTSPWR2, XCOBRA-IIIC codes were used for the analyses.

Rated full power and the design values of the radial peaking factors and conservative values of the remaining paramefers were utilized. The results indicate that neither the pressure, MDNBR nor the peak pellet linear heat.

generation rate violate the coresponding criteria of:

110% of design pressurel1.17 and above 21 kw/ft, respectively and, therfore, the results are acceptable.

CVCS Malfunctions with Decreasing Boron Concentration Reactivity can be added to the reactor by feeding reactor makeup water to the RCS. The normal dilution procedure is subject to administrative procedures to prevent inadvertent dilution.

The method for the resolution of such event is to estimate the time from the initiation of the dilution to the time the adverse effect manifests itself.

This i.s not exactly what the SRP recommends but it is consistent with past and current practice. The events which were examined were dilution during refueling, cold shutdown, hot standby, startup and at power.

The minimum time t6loss of shutdown margin was 16.4 min. which is greater than 15 min and considered adequate for operator action to stop the dilution process.

The results are considered acceptable.

Inadvertent Loading of Fuel Assembly into an Improper Location This event could result from a misplacement of a fuel assembly (Ref.

10).

A mismatch of fuel assembly and fuel assembly location could result in higher power production in an assembly and cause it to exceed the maximum values of the peaking factors. Administrative procedures have been established to avoid such misplacement.

These procedures require core power distribution monitoring at several power levels to assure that no technical specification limits have been violated.

17 However, the licensee did not perform an analysis to evaluate the effects of such fuel assembly misplacement.

We approve of the Cycle 10 operation under the administrative procedures, however, we will require that the licensee submit such an analysis before the next fuel reload.

RCCA Ejection This transient is described in Ref. 10, and could result from the mechanical -failure of a control rod mechanisn pressure housing, resulting in the ejection of a RCCA. The result of such failure is very rapid reactivity insertion coupled with severe distortion of the power distribution.

The transient has been analyzed using the XTG an approved Exxon computer code (Ref. 16).

No credit was taken for the power peak flattening effects of the Doppler feedback or the moderator feedback. (where the value was negative).

The fuel pellet energy deposition resulting from the ejected RCCA was estimated for BC and EOC conditions.

Under hot full power conditions the maximum fuel pellet energy was estimated to be 165 cal/gm at BOC and 172 cal/gm at EOC. This results in less than the maximum allowed of 280 cal/gm as stated in Regulatory Guide 1.77 and is acceptable.

Chapter 15 Transient and Accident Events To support the Cycle 10 core reload with a return to full power (2300 Mwt) and increase FN H, the licensee submitted, "NX-NF-84-74, Plant Transient Analysis for H. B. Robinson Unit 2 At 2300 Mwt With Increased FN Document XN-NF-84-74 presents the analysis of the SRP A H Chapter 15 transient and accident events. The staff's SER is contained in Attachment I.

The licensee has requested and provided justification to defer submittal of the steam line break events until January 31, 1985 in

order to provide ENC time to develop an acceptable methodology for analysing'the consequences of postulated steamline break events, (see additional details in Attachment I to this SER).

Attachment A of Attachment I provides the staff's independent analysis of the steamline break event. Our analysis confirmed that the new steam generators with integral flow restrictors decreased the severity of the event when compared with the FSAR analysis.

The design basis analysis did not result in a calculated DNBR below the specified acceptable fuel design limit (SAFDL). The analysis for Cycle 10 showed the margin the SAFDL is significantly increased. Consequently, fuel integrity is maintained.

Based on the above discussion as expanded in Attachment I and the licensee's justifications, we find the licensee's request to defer the submittal of the steamline break event until January 1985 acceptable.

Based on our review of the Chapter 15 transient and accident events contained in Attachment I, we find them acceptable with the licensee's commitments contained therein and reiterated in the forwarding letter of SER.

Loss of Coolant Accident (LOCA) Analysis To support the changes described in the introduction (Section 1.0) for Cycle 10 operations, the licensee provided a revised LOCA analysis.

The ECCS evaluation model utilized to perform the LOCA analysis for HBR-2 is the revised Exxon Nuclear Company (ENC) evaluation model EXEM/PWR also submitted for review. The staff's review of the ENC

019 revised model as utilized for the Cycle 10 LOCA and the staff's evaluation for the LOCA analysis including codes and models is combined as a separate SER and contained in Attachment II to this SER.

Based on the Attachment II SER, we conclude that the LOCA analysis satisfies the requirements of 10 CFR 50.46 and that the evaluation model utilized satisfies the requirements of Appendix R to 10 CFR 50 and, therefore, is acceptable.

TECHNICAL SPECIFICATION CHANGES The licensee has proposed (References 1, 17 and 18) a number of changes to the Technical Specifications for the cycle 10 reload. The changes to the Technical Specifications can be associated with the major changes for the cycle 10 loading, i.e., return to the 2,300 MWt power level, the new DNB correlation, revision of the power distribution control and the deletion of N-1 loop operation.

A listing of all the proposed Technical Specification changes is given in Attachment I of References 1, 17 and

18. Our review and evaluation of each proposed change follows, with the numbering corresponding to that presented in References 1, 17 and 18.

Technical Specification 2.1, Figure 2.1-1 This figure shows the limits and the allowable combinations of thermal power, coolant pressure and coolant inlet temperature, under full flow conditions. The Technical Specification limits shown incorporate the new DNB correlation, the new thermal power level of 2,300 MWt the new hot channel factors and the combined steady state uncertainties. The changes indicated in Technical Specifications 2.1 (a) and 2.1 (d) and in the basis of 2.1 (TS 2.1 pp 1-5, attachment 8. Ref. 1 and the editorial changes of reference 18).have been reviewed and found to be in agreement with approved changes and, therefore, are acceptable.

20 Specification 2.3 This specification deals with the overtemperature and overpower.AT setting changes shown in pp 2.3-1 to 2.3-6 of attachment 8 ref. 1 and the editorial changes of ref. 18.

The reasons for the changes are: return to the 2,300 MWt power level, the new DNB correlation and deletion of N-1 loop operation.

The overtemperature and the overpressure limit settings form the largest part of the chanyes, the rest are either reference or editorial changes.

We have reviewed the proposed changes and we find that they consist of utilization of previously approved (Westinghouse) overtemperature% T and overpower AT analytical expressions with parameter values adjusted for the use of the new DNB correlation and the new power level of 2,300 MWt. We find these acceptable.

Srecification 3.1.1.1, LCO for CoolantPumps This specification deals with the limiting conditions of operation for the reactor coolant pumps.

The changes are indicated in pp 3.1-1 to 3.1-3b of attachment 8 Reference 1 and in the editorial changes listed on pp 3.1-1 and 3.1-2 of Ref. 18. The most important aspect of this specification is that no power operation is allowed without all primary coolant pumps being in operation. This condition and the conditions and actions specified for operation at or less than 2% of thermal power, have been reviewed and are acceptable.

Specification 3.1.1.2 Steam Generator This specification requires that at least two steam generators shall be operable whenever the average primary coolant temperature is above 350 0 F.

This specification is listed in pp 3.1-3 to pp 3.1-3b of attachment 8, reference 1. The-specification itself has not changed.

Some editioral and minor changes were made in the basis. These changes were reviewed and found to be acceptable.

Specification 3.1.3.1 Minimum Conditions for Criticality This specification defines +5.0 pcm/*F as the upper limit of the moderator temperature coefficient for the reactor to be critical at less thdn 50'. ot rated power and linearly deci easing to 0 pcm/*F at full power.

This Specification and its basis are listed on pp 3.1-11 and 3.1-12 of attachment 8, reference 1. The value of the moderator temperature coefficient has been changed from +2.0 pcm/*F. However, the transient analyses which involve heatup of the primary coolant supported the value of +5pcm/*F.

On this basis we find this specification acceptable.

Table 3.5-1.

Engineered Safety Feature System Initiation Instrument Setting Limits.

The table and the proposed changes are listed in pp 3.5-7 and 3.5-7a.

The changes are either of editorial nature or a consequence of returning to 2,300 MWt of rated power operation.

On this basis the changes to Table 3.5-1 are acceptable.

Table 3.5-3, Instrumentation Operating Conditions for Engineered Safety Features.

This table is partially listed on p. 3.5-10a and the changes refer to the footnotes and they are of editorial nature.

On this basis the changes to Table 3.5-3.,are acceptable.

Specification 3.6.1, Containment Integrity This specification deals with containment integrity and the circumstanceN under which operations can be performed in the reactor regarding Core reactivity changes. The proposed changes are listed on pp. 3.6-1 and 3.6-2 of attachment 8 ref. 1 and pp. 3.6-1 to 3.6-3 of ref. 17 and are due to the elimination of the part length rods.

The proposed changes define the conditions and the operations (tests) to be performed when the containment is not intact. However, the shutdown maryin duringi any of these tests will always be maintained at a level yreater than 1t of6k/k.

The boron dilution will be monitored duri.ng any such change.

Under these conditions the proposed changes are acceptable.

-22 Specification 3.6.2, Internal Pressure The proposed changes in this specification are listed on pp 3.6-1 and 3.6-3 of attachment 8 ref. 1 and on p 3.6-2 of ref. 17.

The changes are editorial and reference updating to the FSAR.

The proposed changes are acceptable.

Specification 3.6-3, Containment, Autonatic Isolation Trip Valves The proposed changes are listed on p.3.6-3 of attachment 8, ref. I and on pp 3.6-2 and 3.6-3 of ref. 17 and are of editorial nature and reference updating.

The proposed changes are acceptable.

Page 3.86, on Containment The proposed changer are reference updating and, therefore, are acceptable.

Specification-3.10.1.5 Control Rod Operation The proposed changes 4.ca listed on p. 3.102 of attachment 8 reference

1. The changes are due to returning to 2,300 MWt power level and are acceptable.

Specification 3.10.2 Power Distribution Limits The proposed changes in this specification are listed on pp 3.102 to 3.107a of attachment 8 in ref. 1. The proposed changes emanate from the return to 2,300 MWt rated power, the revised power distribution control and the deletion of Ni loop operation. The determination of FQ (Z) and FH under different conditions is specified in detail.

In addition alternate actions are specified regarding reactor power assuming that the required values of FQ (Z) or F6H cannot be satisfied.

Conversely, the allowable power level is estimated for given values of the peaking factors with the required uncertainty factors for engineering (FQE = 1.03) measure ment (Fu N = 1.05) and the instruments (FQa = 1.02) and their proper usage is

23.

specified.

The required frequencies for core power distribution maps and for the target axial flux difference are specified.

Alternate actions for the power level depending on the axial flux difference are also given.

Finally the calibration of the ex-core detectors is specified.

We have reviewed the proposed changes and found them acceptable for the operation of cycle 10 because they have been taken into account in the transient and accident analysis.

Specification 3.10.3, Quadrant Power Tilt Limits The proposed changes to this specification are listed on pp 3.10-7a and 3.10-7b of attachment 8, reference 1. The proposed changes reflect the return to 2,300 MWt rated power and the revised power distribution control.

The new specification is in compliance with the provisions of the standard review plan and therefore is acceptable.

Specification 3.10.8, Required Shutdown Margin The proposed-changes are indicated on p 3.10-12, pp 3.10-14 to 3.10-20 and pp 3.10-22 to 3.10-24 of attachment 8 reference 1.

The changes are due to:

return to 2,-30 MWt, the new DNB correlation, the new power distribution control, deletion of references to part length rods, updating of references and repagination.

Figures 3.10-4 and 3.10-5 have been added.

We have reviewed the changes and found the values of HONBR, peak value of the linear power density, FAH and F to be within their approved ranges.

The method and the specified limits of the allowable deviation from the target flux difference are acceptable.

Therefore the proposed specification is acceptable.

Specification 3.11.2, Movable In-Core Instrumentation The proposed changes are indicated on pp 3.11-1 and 3.11-2 of attachment 8 of reference 1. The changes are due to returning to 2,300 MWt rated power, change of the number of the in-core instrumentation thimbles, and deletion of N-1 loop operation. The proposed change in the number of thimbles is an increase and is acceptable, likewise the changes due to the power level and N-i loop operation are also acceptable.

24 Specification 4.11.1 APDMS Operation The proposed changes are listed on pp 4.11.1 to 4.11.3 of attachment 8 in reference 1. The changes are due to returning to the 2,300 MWt level of operation, the change in the number of thimbles (see specification 3.11.1 above) and the improved power distribution control.

The changes have been reviewed and found to be acceptable.

Specification 5.3.1 Reactor Core The proposed changes are listed on p 5.3-1 of attachment 8, reference 1 and p 5.3-1 of reference 18.

The proposed changes allow the reconstitution of the partial length shield assemblies, specify the new total core fuel loading, specify the,.maximum' fuel enrichment and allow axial natural uranium blanket.

We have reviewed the proposed changes and find them necessary and acceptable for the operation of Cycle 10.

Specification 5.3.2 Reactor Coolant System The proposed changes are listed on pp 5.3-1 and 5.3-2 of attachement 8, reference 1. The changes are updating of references and editorial.

The proposed changes are acceptable.

3.0

SUMMARY

We have reviewed the information submitted on Cycle 10 for the H. B. Robinson Unit 2. This included the original and subsequent submittals which were additions and clarifications in response to questions generated by the review. We find the Cycle 10 operation acceptable for the fuel system mechanical design, nuclear desiqn.

thermal-hydraulic design and analysis, transients and accidents, the radiological consequence analysis for the locked rotor, the rod cluster control assembly withdrawal and the steam generator tube rupture; and the Technical Specifications proposed.

25 This approval is subject to the following conditions:

(a) confirmation of the scram shutdown margin analysis (to be submitted during CY85) and (b) submittal of a fuel misloading analysis (during Cycle 10).

4.0 FINAL DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION 4.1 On August 24, 1984, the Commission published in the Federal Register (49 FR 33764) a Notice of Consideration of Issuance of Amendment To Facility Operating License And Proposed No Significant Hazards Consideration Determination And Opportunity for Hearing. That notice addressed a change requested by the licensee in their letter dated July 23, 1984. The change requested involved changes to the Technical Specifications in support of their Cycle 10 core reload and return to full power (2300 MWt) with new steam generators. The reload application included special peripheral part length shielded fuel assemblies, which will be installed to accommodate the pressurized thermal shock program. To accommodate these assemblies, a low leakage core and return to full power (2300 MWt) with new steam generators, hot channel factor (F and H) limits and BOC moderator temperature coefficients are being increased and over temperature and overpowerAT setpoints are being reduced.

In support of these changes the licensee provided a safety evaluation for the Cycle 10 core reload, reanalyzed the Chapter 15 events and provided a LOCA analysis.

Because the Commission determined there was insufficient time for its usual 30-day notice of the proposed action for public comment, that notice established'-a period until.September 14, 1984 for comment, state

26 that a final determination on no significant hazards would be made before-issuance of the license amendment, and provided that if no significant hazards are involved, a subsequent notice of opportunity for a hearing would be published.

In our evaluation of the LOCA and fuel performance analysis we determined that the revised analysis for Cycle 10 reload of HBR-2 satisfies the requirements of 10 CFR 50.46 and that the evaluation model utilized satisfies the requirements of Appendix K to 10 CFR 50.

The application for Cycle 10 reload incorporated plant design changes resulting from steam generator replacement and justification of return to full power. In support of this the licensee submitted analyses of the Chapter 15 transient and accident events.

The Safety Evaluation Reports (SER) for Cycle 8 and Cycle 9 required the licensee to develop a stand-alone analysis methodology which does not infringe.upon other vendor's methods. The methodology is under our staff's final review and have not yet been approved. We have evaluated and discussed these items in detail in our SER Attachment I and our review has progressed sufficiently to conclude that analyzed events would not be significantly altered upon completion of review.

This conclusion is based upon code validation results, limiting boundary conditions applied to each event and benchmarking the computer codes with known experimental transients conducted at the LOFT facility.

The licensee's contractor has not finalized its methodology for evaluating the consequences of postulated steam line break events.

The licensee will reanalyze the SLB event for Cycle 10 by January 31, 1985.

Our bases for accepting the late submittal are as follows:

27 (1)

H. B. Robinson Unit 2 replaced its steam generators with a new model that incorporates an integral flow restrictor within the outlet nozzle. The flow restrtctor significantly reduces the consequences. of a major rupture of a steam line, (2) The limiting consequences of a large steam line break occurs at end of cycle (EOC) when the moderator coefficient is at its most negative value, and (3) The staff analysis of the steam line break event (guillotine break) showed ample margin to the acceptance criteria for H. B. Robinson Unit 2. (See Appendix A to Attachment I of our SER).

4.2 On October 5, 1984, the Coimmission published in the Federal Register (49 FR 39396) a Notice of Consideration of Issuance of Amendment to Facility Operating License And Proposed No Significant Hazards Consideration Determination And Opportunity for Hearing.

That notice specifically addressed a change requested by the licensee in their letter dated August 1, 1984.

Because the Commission determined there was insufficient time for its usual 30-day notice of the proposed action for public comment, that notice established a period until October 17, 1984 for comment, stated that a final determination on no significant hazards would be made before issuance of the license amendment, and provided that if no significant hazards are involved, a subsequent notice of opportunity for a hearing would be published.

The proposed change as requested by letter dated August 1, 1984,

28 involves changes to the Technical Specification to allow additional control rod evaluations while containment integrity is not intact but only while maintaining a shutdown margin of > 1%A4k/k.

In our evaluation we note that during the additional conditions and operations (tests) to be performed when the containment is not intact, the shutdown margin must be maintained at > 1 k/k as already required by Technical Specification 3.6..1c.

The boron dilution will be monitored during any such changes.

4.3 We have determined that the proposed change does not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility or a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction inee margin of safety.

5.0 Environmental Conclusions This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. ihe staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards considera tion and there has been no public comment on such finding. A portion of the amendment proposed was subsequently changed; the Commission has also made a final no significant hazards consideration finding with respect to the changed portion of this amendment. Accordingly, this

29 amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR Sec 51.22(c)(9).

Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

5.2 Safety Conclusion We have concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety 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 will not be inimical to the-common defense and security or to the health and safety of the public.

Dated: November 7, 1984 PRINCIPAL CONTRIBUTORS:

J. Guttman R. Jones G. Requa L. Lois M. Wohl S. WU

30 References A. B. Cutter, letter to S. A. Varga, "H. B. Robinson Steam Electric Plant Unit 2, Docket No. 50-261/Lic. No. DRR-23. Request for License Amendment Cycle 10 Operation" serial NSL-84-1 32 5, dated July 23, 1984.

2. XN-75-39, "Generic Fuel Design for 15x15 Reload Assemblies for Westinghouse Plants" Exxon, September 1975.

3. XN-NF-83-55, "Mechanical Design Report Supplement for H. B. Robinson Extended Burnup Fuel Assemblies" Exxon, August, 1983.

4.

XN-NF-83-71, "Mechanical Design Report Supplement for H. B. Robinson Part Length Sheild Assemblies" Exxon, September, 1983.

5.

XN-NF-75-27(A) "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors: Exxon, June 1975.

Supplement 1, September 1976, Supplement 2, December 1977 and Supplement 3, November 1980.

6. XN-NF-77-57(A) "Exxon Nuclear Power Distribution Control for Pressurized Water Reactors -

Phase II" Exxon, January 1978. Supplement 1, June 1979 and Supplement 2, September 1981.

7. XN-75-38, "H. B. Robinson Unit 2, Cycle 4 Reload Fuel Licensing Data Submittal"

Exxon, August, 1975.

8.

XN-NF-84-74, "Plant Transient Analysis for H. B. Robinson Unit 2 at 2300 MWt with Increased FAH" Exxon, August 1984.

9.

XN-NF-84-72, -H. B. Robinson Unit Large Break LOCA-ECCS Analysis with Increased Enthalpy Rise Factor:

Break Spectrum" Exxon, August 1984.

10.

XN-NF-83-72, "H. B. Robinson Unit 2, Cycle 10 Safety Analysis Report" Revision 2, Exxon, July 1984.

Supplement 1, Exxon July, 1984.

31

11.

XN-CC-28, Rev. 5, "XTG - A Two Group Three-Dimensional Reactor Simulator Utilizing Coarse Mesh Spacing.(PWR-Version)" Exxon, July 1979.

12. WAPD-TM-678, "PDQ7 Reference Manual" Westinghouse Electric Corporation, January 1967.

13.

WAPD-TM-478, "HARMONY:

System for Nuclear Reactor Depletion Computation" Westinghouse Electric Corporation, January 1965.

14.

XN-NF-82-91(P),

"H. B. Robinson-2, Loss of Feedwater Transient at 1955 MWt:

Model Description and Results" Exxon,

November, 1982.

15.

XN-NF-84-74, Supplement 1.

"Plant Transient Analysis for H. B. Robinson Unit 2 at 2,300 MWt with Increased F.

Supplement 1, Analysis of Control Rod Misoperation Events" Exxon Nuclear Company, September 1984.

16.

XN-CC-28 (NP)

(A),

Revision 5, " XTG - A Two Group 3 - Dimensional Reactor Simulator Utilizing Course Mesh Spacing (PWR Version)" Exxon Nuclear Company, July 1979.

17. A. B. Cutter, letter to S. A. Varge, "H. B. Robinson Steam Electric Plant, Unit No. 2, Docket No. 50-261/License No. DPR-23. Request for License Amendment Control Rod Drive Evolution" Serial NSL-84-186, dated August 1, 1984.

18.

A. B. Cutter, letter to S. A. Varga, "H. B. Robinson Steam Electric Plant, Unit No. 2, Docket No. 50-261/License No. DPR-23, Cycle 10 Operation", Serial: NLS-84-366, dated September 7, 1984.

19.

XN-NF-75-32(P)(A) "Computational Procedures for Evaluating Fuel Rod Bowing", Exxon October 1983, and Supplements 1, 2, 3 and 4.

ATTACHMENT I SAFETY EVALUATION REPORT FOR H. B. ROBINSON UNIT 2, CYCLE 10 RELOAD APPLICATION CHAPTER 15 EVENTS 15.0 Introduction And Analytical Techniques The Carolina Power and Light Company (CP&L) submitted XN-NF-84-74, "Plant Transient Analysis For H. B. Robinson Unit 2 At 2300 MWt With Increased FNH" in support of its Cycle 10 reload application for H. B. Robinson.

XN-NF-84-74 presents the analyses of the Chapter 15 transient and acci dent events. These analyses were performed by Exxon Nuclear Company, the fuel vendor for the H. B. Robinson plant.

The application for the Cycle 10 reload incorporated plant design changes resulting from steam generator replacements and justification for return to full power operation.

The analytical methodology and the computer models used in the safety analyses have not been approved. The Safety Evaluation Reports (SER) for Cycle 8 and Cycle 9 required the licensee (if it continued to rely on Exxon analyses) to develop a stand-alone analysis methodology which does not infringe upon other vendors' methods. As a consequence, Exxon Nuclear Company (ENC) developed a stand-alone methodology which is at present under staff review.

The computer programs used in the analyses are PTSPWR2, SLOTRAX and RELAP5. The RELAP5 computer program was submitted in response to NRC's small-break LOCA analysis concerns outlined in TMI Action Plan Item

-~

2 II.K.3.30 (NUREG-0737). The use of this code for mild transient cal culations, as applied in XN-NF-84-74, should be acceptable. This code has been developed and applied to transient analyses by the Office of Nuclear Regulatory Research, at the NRC. Generic approval for this code will result from the staff review of TMI Action Item II.K.3.30.

The staff's review of the PTSPWR2 computer program is nearing completion.

This code has been significantly modified since its application to Cycle 8 and Cycle 9 reloads. The code has been benchmarked with several LOFT experimental transients, with a RELAP5 analysis, and with an operating plant transient. Our review has progressed sufficiently to conclude that the analyzed events submitted in XN-NF-84-74 will not be significantly altered upon completion of review.

The analytical methods (by which the licensee applies a computer program for a specific event) is documented in XN-NF-84-73(P), "Exxon Nuclear Methodology For Pressurized Water Reactors Analysis Of Chapter 15 Events."

This methodology report is still being developed by Exxon and undergoing staff review. Our review of both XN-NF-84-73(P) and XN-NF-84-74 concludes that the calculated results for H. B. Robinson Unit 2 would not be appreciably altered upon our completion of the methodology review. This conclusion is based upon the code validation results and the limiting boundary conditions applied to each event.

ENC has not finalized its methodology for evaluating the consequences of postulated steam line break events.

However, by incorporating an integral flow restrictor within the nozzles of the steam generators, the

3 consequences of a postulated steam line break event is significantly reduced. In addition, the limiting operating conditions for a postulated steam line break is at end of cycle (EOC).

At this time in operating cycle, the moderator density or temperature coefficient is at its most negative value. This maximizes the potential for return to power from an over-cooling event.

In order to confirm that no fuel failure is anticipated to occur, the staff performed its analysis of a steam line break event for H. B.

Robinson. Results of the staff's analysis is documented in Appendix A to this report. CP&L has committed to provide reanalyses of the steam line break events for H. B. Robinson. We require this submittal, including documentation of the methodology, by January 31, 1985. It is our understanding that the analyses will be performed with RELAP5. We require a copy of the RELAP5 input deck for our review.

The loss of feedwater-event was analyzed with the SLOTRAX computer code.

SLOTRAX is under staff review. Our review indicates that SLOTRAX under predicts the pressurization of the primary system for the loss of feed water event. However, the insurge of primary coolant into the pres surizer is conservatively calculated by the homogeneous equilibrium model in SLOTRAX. The licensee, applying the conservative pressurizer inflow, performed a hand calculation of the peak pressure by assuming isentropic compression of the steam. This analysis is conservative. We require the licensee to provide code validation of SLOTRAX by November 30, 1984. This has not been submitted to the staff as part of the SLOTRAX documentation.

4 The following sections address the specific events analyzed in XN-NF-84-74.

15.1 Increase In Heat Removal By The Secondary System 15.1.1 Feedwater Malfunctions That Result in a Decrease in Feedwater Temperature The licensee concluded in technical report XN-NF-83-72 that the excess load event, documented in Section 15.1.3 of XN-NF-83-74, bounds the consequences of the decrease in feedwater temperature event. We find the licensee's assessment acceptable.

15.1.2 Feedwater System Malfunctions That Result in an Increase in Feedwater Flow The licensee concluded in technical report XN-NF-83-72 that the excess load event, documented in Section 15.1.3 of XN-NF-84-74, bounds the overcooling response of the decrease in feedwater temperature event. In addition, the rod withdrawal event, documented in XN-NF-84-74, bounds the reactivity insertion response of the decrease in feedwater temperature event. We find the licensee's assessment acceptable.

15.1.3 Increase In Steam Flow (Excess Load)

Section 15.1.3 of XN-NF-84-74 evaluates the Excess Load Event for H. B. Robinson 2. The maximum step increase in load demand was 10% from full power operation. This was stated to be the

5 maximum capacity of the turbine steam regulating valves from the most degraded DNBR condition.

The Excess Load Event is classified as a Condition II event, an Anticipated Operational Occurrence. The acceptance criteria for this event is that the primary system pressurization remains below 110% of design values; that the DNBR not decrease below 1.17 when applying the XNB correlation; that the radiological consequences be less than 10 CFR 20 guidelines; and that the event should not generate a more serious plant condition without other faults occurring independently.

In assessing this event, the licensee performed two analyses.

One analysi5.minimized the moderator temperature feedback and the second analysis maximized the contribution of the moderator feedback. The conclusions of these analyses showed a negli gible difference between the resulting minimum DNBR for the two cases. The analysis with minimum reactivity feedback resulted in a minimum ONBR of 1.331. The analysis with the maximum reactivity feedback resulted in a minimum DNBR of 1.332. Since the calculated minimum DNBR did not decrease below 1.17, no fuel failure was predicted to occur.

6 The similarity of the minimum DNBR for both events is attri buted to the similarities of the thermal-hydraulics during the initial 45 seconds.

During this time interval the minimum primary system pressure decreased to 2205 psia, and the core power and core inlet temperature (decreasing by 40 F) behaved similarly for both analyses. Differences in plant responses occurred following the time of minimum DNBR. For the maximized feedback event, the DNBR remained relatively constant near the minimum value as the primary system pressure increased and leveled off at a slightly higher value. The minimum feedback event, however, continued to increase in pressure and in DNBR.

This is attributed to the less negative (zero) moderator temperature coefficient. The primary system pressure achieved a peak of 2390 psia. This is well within 110% of the primary system design pressure.

15.1.3.1 Conclusion For The Excess Load Event The licensee demonstrated conformance to the acceptance criteria for the Excess Load Event, as it applies to H. B.

Robinson Unit 2. The methodology used in analyzing the Excess Load Event is acceptable. The applicant used the PTSPWR2/Mod 1 (1984 version) computer program to calculate the thermal hydraulic systems and core heat flux responses. This code is undergoing staff review and an SER is anticipated by end of

7 calendar year 1984. We have reasonable assurances that upon completion of our review of PTSPWR2/Mod 1, any modification or restrictions placed upon the code would have negligible impact on this analyzed event. We therefore find the analysis of the Excess Load Event acceptable.

15.1.4 Inadvertent Opening of a Steam Generator Relief or Power Operated Relief Valve The licensee concluded in technical report XN-NF-83-72 that the excess load event, documented in Section 15.1.3 of XN-NF-84-74, bounds the consequences of an inadvertent opening of a steam generator relief valve. The excess load event results in symmetric cooldown of all 3 steam generators. The open atmospheric relief valve results in asymmetric cooldown of the primary system.

Exxon Nuclear Company, the fuel vendor for H. B. Robinson, is developing a new methodology for evaluating steam line break and stuck-open atmospheric relief valve events. This metho dology will account for asymmetric thermal-hydraulics within the reactor vessel.

This methodology and analysis will be submitted by January 31, 1984 and will be used to confirm that the excess load event is bounding. We find the licensee's response acceptable.

15.1.5 Steam System Piping Failures The analysis of a postulated steam line break or an inadvertent opening of a steam generator relief or safety valve requires the modeling of thermal-hydraulic asymmetry within the reactor vessel.

Previous H. B. Robinson analyses for these events were performed by Westinghouse and by Exxon Nuclear Company.

The analyses performed by Exxon Nuclear Company were determined unacceptable for previous Cycles. The reason was primarily due to insufficient justification for neglecting asymmetry in the thermal-hydraulics within the reactor vessel.

Exxon Nuclear is developing its analytical methodology for steam line break analysis. This methodology will use the RELAP5 computer program and model the asymmetric thermal-hydraulics for these events.

For CycleIO, CP&L replaced the steam -generators at H. B.

Robinson. These generators have integral flow restrictors designed within their outlet nozzles. The restrictors decrease 2

2 the minimum cross sectional flow area from 4.7 ft to 1.4 ft These flow restrictors significantly reduce the consequences of a postulated steam line break event.

In addition, the limiting operating conditions for a major rupture of a steam line is at end of cycle (EOC).

At this time, the moderator density or temperature coefficient is at its most negative value. This provides the greatest potential for return to power.

9 To confirm that no fuel failure would occur, the staff per formed its analysis of a steam line break event for H. B.

Robinson. Results of the staff's analysis is documented in Appendix A to this report.

15.1.5.1 Conclusion For The Steam Line Break Events We have reviewed the licensee's justification for delaying submittal of the steam line break events and find them ac ceptable. The staff's analysis of the steam line break event for H. B. Robinson Unit 2 showed ample margin to the specified acceptable fuel design limits (SAFDL). Consequently fuel integrity should be maintained.

CP&L has committed to provide reanalyses of the steam line break events for H. B. Robinson. We require this submittal, including documentation of the methodology, by January 31, 1985. It"Ts our understanding that the analyses will be performed with RELAP5.

We require a copy of the RELAP5 input deck for our review.

15.2 Decrease In Heat Removal By The Secondary System 15.2.1 Steam Pressure Regulator Malfunction That Result in Decreasing Steam Flow This event is not applicable to H. B. Robinson Unit 2 since it has no steam line pressure regulators.

4

.1.1g 15.2.2 Loss of External Electrical Load Section 15.2.2 of XN-NF-84-74 evaluates the Loss of External Electrical Load event for H. B. Robinson 2. This analysis assumes an instantaneous loss of generator load. Offsite power is not affected for this event and is therefore available for reactor coolant pump operation.

The loss of load event was analyzed twice. In one case, the event was initiated at the limiting conditions for assessing peak primary system pressurization. The second case was initi ated at limiting conditions for minimum DNBR considerations.

The loss of load event is classified as a Condition II event, an Anticipated Operational Occurrence. The acceptance criteria for this event is that the primary system pressurization remains below 110% of design values; that the DNBR not decrease below 1.17 when applying the XNB correlation; that the radio logical cnsequences be less than 10 CFR 20 guidelines; and that the event should not generate a more serious plant condition without other faults occurring independently.

The analysis of this event was initiated by an instantaneous loss of generator load. The turbine stop valves closed as the turbine tripped. A reactor trip was not credited from the turbine trip. The isolation of the secondary system led to its pressurization. The secondary dump valves were assumed not to function.

0 11 The analysis which challenged the primary system overpres surization resulted in a peak pressure of 2661 psi.

This pressurization is well below 110% of the primary system design pressure. The event was initiated at 102% of ratedspower.

Conservative multipliers were assumed for the Moderator and Doppler reectivity coefficients. The initial pressurizer water level was biased high and the pressurizer pressure was biased low. The pressurizer spray and PORVs were assumed inoperative.

These biases were predetermined based on sensitivity studies to be documented within XN-NF-84-73(P).

The analysis which maximized the challenge to the fuel design limit (minimum DNBR) was biased by increasing the core inlet temperature; decreasing the pressurizer pressure; and crediting operation of the pressurizer sprays and PORVs. This tended to minimize system pressurization. Consequently, the minimum DNBR analysis resulted in a peak primary system pressure of 2310 psi, or 351 psi lower than for the peak pressurization event.

This analysis resulted in a minimum DNBR of 1.19, which is greater than the fuel design limit (for the XNB correlation) of

.1.17. As a result, fuel integrity is maintained.

15.2.2.1 Conclusion for the Loss Of External Electrical Load Event The licensee assessed the consequences of a loss of external electrical load event with respect to challenging the primary system pressure response and the fuel design limits. These

12 were presented as two bounding analyses using the PTSPWR2/MOD1 (1984 version) computer code. The results of these analyses are found acceptable.

The PTSPWR2/MOD1 computer code and methodology (documented in XN-NF-84-73(P)) are under staff review. The sensitivity studies which determined the limiting operating conditions (biases) for this event have not been submitted in the XN-NF-84-73(P). We require the licensee to submit these results prior to December 31, 1984.

Our review of the PRSPWR2/MOD1 computer program is nearing completion. We anticipate issuing an SER by December 31, 1984.

We have reasonable assurances that upon completion of our review of PTSPWR2/MOD1, any modification or restrictions placed upon the code would have negligible impact on these analyzed events..We therefore find the analysis of the loss of external electrical load event acceptable.

15.2.3 Turbine Trip The licensee concluded in technical report XN-NF-83-72 that the turbine trip event is not required to be analyzed since it is bounded by the loss of load event, Section 15.2.2 in XN-NF-84-74. We find the licensee's assessment acceptable.

13 15.2.4 Loss of Condenser Vacuum and Other Events Resulting in Turbine Trip The licensee concluded in technical report XN-NF-83-72 that the subject events are bounded by the loss of load event and need not be analyzed. We find the licensee's assessment acceptable.

15.2.5 Inadvertent Closure of Main Steam Isolation Valves (MSIVs)

The licensee concluded in technical report XN-NF-83-72 that the subject event is bounded by the loss of load event and need not be analyzed. We find the licensee's assessment acceptable.

15.2.6 Loss of Non-Emergency AC Power to the Station Auxiliaries The licensee concluded in technical report XN-NF-83-72 that the subject event is bounded by the loss of load event and need not be analyzed. We find the licensee's assessment acceptable.

15.2.8 Feedwater System Pipe Break The licensee concluded in technical report XN-NF-83-72 that the spectrum of steam line break events bounds the consequences of feedwater line break events. This was attributed to the high elevation of the feedwater nozzle. Consequently, mostly steam would be discharged out the break. This was the design basis of the plant and we find the licensee's assessment acceptable.

14 15.3 Decrease in Reactor Coolant Flow 15.3.1 Loss of Forced Reactor Coolant Flow Section 15.3.1 of XN-NF-84-74 evaluates the Loss of Forced Reactor Coolant Flow for H. B. Robinson Unit 2. This event was simulated as a loss of electric power to all of the reactor coolant pumps. Offsite power was assumed available.

The loss of forced reactor coolant flow is classified as a Condition II event, an Anticipated Operational Occurrence. The acceptance criteria for this event is that the primary system pressurization remains below 110% of design values; that the DNBR not decrease below 1.17 when applying the XNB correlation; that the radiological consequences be less than 10 CFR 20 guidelines; and that the event should not generate a more serious plant condition without other faults occurring independently.

The licensee has concluded that there exists no active single failure which would result in a more severe overpressurization or lower DNBR for this event. The licensee addressed the concern of overpressurization and minimum DNBR with two calculations. The calculation for maximizing the system pressurization response assumed a high reactor system initial pressure, a high pressurizer level, disabled PORVs, minimum reactor coolant flywheel inertia, high moderator reactivity temperature coefficient, low Doppler reactivity coefficient, and maximum heat transfer coefficient across the fuel gap.

A 15 The calculation for minimizing the DNBR assumed low initial primary system pressure, low pressurizer level, PORVs avail ability, minimum flywheel inertia for the reactor coolant pumps, increased core inlet temperature, high moderator reactivity temperature coefficient, low Doppler reactivity coefficient, and high gap conductance within the reactor fuel rods.

The above biases on operating conditions were determined as part of the methodology development, to be documented in XN-NF-84-73(P). These studies have not been transmitted to the NRC for review. We require. the licensee to submit these studies by December 31, 1984.

The analyses were initiated with a pump coastdown from the above operating conditions. The DNBR rapidly decreased with decreasing-coolant flow. The reactor.coolant temperature then increased (80 F) and expanded into the steam region of he pressurizer. Upon a low coolant flow indication (87% flow from the loop flow detectors), the reactor tripped. The reactor was assumed on manual control to prevent rod insertion upon an increase in coolant temperature. The reactor power reached 105%. The peak primary system pressure, for the maximum pressurization calculation, was 2582 psi.

This is well below 110% of design. For the minimum DNBR biased calculation, the peak primary system pressure was 2304 psi, or 278 psi lower.

The minimum DNBR for this-event decreased to 1.19.

16 15.3.1.1 Conclusions for the Loss of Forced Reactor Coolant Flow Event The licensee assessed the consequences of a loss of reactor coolant flow event with respect to challenging the primary system pressure response and the fuel design limits. These were presented as two bounding analyses using the PTSPWR2/MOD1 computer code. The results of these analyses are found accept able. The peak primary system pressurization was well.below 110% of system design and the minimum ONBR was above 1.17 when applying the XNB critical heat flux correlation. As a con sequence both primary system and fuel integrity are maintained.

Both the PTSPWR2/MOD1 computer code and methodology of implementation (documented in XN-NF-84-73(P)) are under staff review. The sensitivity studies which determined the limiting operating conditions (biases) for this event have not been submitted as part of XN-NF-84-73(P). We require the licensee to submit-these results prior to December 31, 1984. Our review of the PTSPWR2/MOD1 computer code is nearing completion. We anticipate issuing an SER by December 31, 1984. Our review has progressed sufficiently such that we have reasonable assurances that upon completion of our review of PTSPWR2/MOD1, any modification or restrictions placed upon the code would have negligible impact on these analyzed events. We therefore find the analysis of the loss of reactor coolant flow event acceptable.

17 15.3.2 F:low Controller Malfunction The H. B. Robinson Unit 2 plant has no primary coolant flow controllers. Therefore, this event is not applicable to H. B.

Robinson Unit 2.

15.3.3 Reactor Coolant Pump Shaft Seizure (Locked Rotor)

Section 15.3.3 of XN-NF-84-74 evaluates the consequences of a locked rotor event for H. B. Robinson Unit 2. The event was initiated by an instantaneous seizure of a rotor from one of the primary system reactor coolant pumps.

The locked rotor event is classified as a Condition IV event, a Postulated Accident. The acceptance criteria for the locked rotor event is that the radiological consequences be less than 10 CFR 100 guidelines; the event should not cause a conse quential loss of the required functions of the systems needed to cope wTt the reactor and containment systems; the radially averaged fuel enthalpy be less than 280 cal/gm; all fuel rods which experience a minimum DNBR below the specified acceptable fuel design limit (SAFDL, 1.17 for the XNB critical heat flux correlation) are assumed to fail; and the primary system pressure should not exceed 110% of design.

Two analyses were presented for this event. One analysis maximized the system pressurization and the other minimized the DNBR. Both calculations were initiated by an instantaneous seizure of a rotor from one of the primary system reactor

.9 18 coolant pumps. A reactor trip was initiated by a low flow signal from the affected loop.

As the flow decreased, the primary coolant temperature began to rise. With increasing coolant temperature the primary system liquid expanded into the pressurizer, which led to primary system pressurization.

Reverse flow in the affected loop occurred one second into the event. This was attributed to continued operation of the two remaining pumps.

The locked rotor calculations resulted in a core flow reduction to 60% of nominal. This occurred 4.0 seconds into the event.

The analysis, which biased the reactor operating conditions to minimize the DNBR, was initialized with a high core inlet temperature; low pressurizer level; low pressurizer pressure; high moderator reactivity temperature coefficient; low Doppler reactivity coefficient; and a high gap conductance to maximize the heat flux at the fuel pin surface. In addition, the PORVs were assumed operational to minimize system pressurization.

The analysis which biased the operating conditions to maximize primary system pressurization was initialized with a high core inlet coolant temperature; a high pressurizer level; high pressurizer pressure; high moderator reactivity temperature coefficient; and a high fuel gap conductance. For-this analy sis, both the pressurizer and secondary system PORVs were assumed disabled. The system, for this analysis, pressurized to 2524 psia. This is well below 110% of design.

19 15.3.3.1 Conclusions for the Reactor Coolant Pump Shaft Seizure (Locked Rotor) Event The licensee assessed the consequences of a seized or locked rotor event for H. B. Robinson Unit 2. Two analyses were performed. One challenged the primary system pressurization response and the other challenged the fuel design limits. Both analyses used the PTSPWR2/MOD1 (1984 version) computer code.

The results of these analyses were found acceptable.

With the above biases in operating conditions, a reactor trip signal on low coolant flow was generated 1.25 seconds into the event. As a result of the positive moderator coefficient, reactor power increased to 107.6% of rated. The minimum DNBR of 0.9 occurred shortly after reactor trip (2.17 seconds into the event). All fuel pins which experienced a DNBR below 1.17 were assumed to fail.

The licensee calculated the radiological consequences to be less than 10% of 10 CFR 100 limits.

The PTSPWR2/MOD1 computer code is a one-dimensional representation of a nuclear steam supply system. Since the primary system is in a non-compressible state, a potential exists for asymmetric flow distribution across the core. A request was made to the Office of Nuclear Regulatory Research (RES) at NRC to assess the multi-dimensional fluid characteristics of a locked rotor event. In response, RES conducted a generic evaluation of a locked rotor event using

20 the TRAC/PF1 computer program. Results of this evaluation showed negligible asymmetry of the coolant flow distribution across the reactor core.

As a consequence of the one-dimensional hydraulic characteristics of the locked rotor event, the PTSPWR2/MOD1 computer code should be appropriate for such application. The PTSPWR2/MOD1 computer code and methodology of implementation (documented in XN-NF-84-73(P)) are under staff review. The sensitivity studies which determined the limiting operating conditions (biases) for this event have not been submitted in XN-NF-84-73(P). We require the licensee to submit these results prior to December 31, 1984.

Our review of the PTSPWR2/MOD1 computer code is nearing completion. We anticipate issuing an SER by December 31, 1984.

Our review-has progressed sufficiently to acquire reasonable assurances that upon completion of our review, any modification or restrictions placed upon the code would have negligible impact on these calculations. We therefore find the analysis of the locked rotor event acceptable.

15.3.4 Reactor Coolant Pump Broken Shaft The licensee concluded in technical report XN-NF-83-72 that the locked rotor event bounds the consequences of the broken shaft event and Peed not be analyzed. We find the licensee's assessment acceptable.

21 15.3.4 Reactor Coolant Pump Broken Shaft The licensee concluded in technical report XN-NF-83-72 that the locked rotor event bounds the consequences of the broken shaft event and need not be analyzed. We find the licensee's assessment acceptable.

15.4.6 Chemical and Volume Control System Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant Section 15.4.6 of XN-NF-84-74 evaluates boron dilution events for H. B. Robinson Unit 2. The events analyzed were for the following reactor modes of operation:

(1) Refueling, (2) Cold shutdown with 3% delta rho shutdown margin and vessel filled to the centerline elevation of the hot legs (required for RHR mixing), (3) Cold shutdown with 1% delta rho shutdown margin and the primary system (excluding the pressurizer) filled with coolant, T4) Hot shutdown, (5) Startup and (6) Power operation.

The rate of dilution of primary system coolant is limited by the capacity of the charging pumps. This corresponds to an addition of 230 gpm of unborated water. For the cold shutdown mode of operation with emptied steam generators, the maximum dilution rate is limited to the capacity of one charging pump, or 77 gpm.

22 0

.0 The time for operator action was determined by solving the differential equation for fluid dilution.

The critical boron concentration and boron worth were determined with the XTGPWR

.computer code.

The boron dilution event is classified as a Condition II event, an Anticipated Operational Occurrence. The acceptance criteria for this event is that the primary system pressurization remains below 110% of design values; that the DNBR not decrease below 1.17 when applying the XNB correlation; that the radiological consequences be less than 10 CFR 20 guidelines; and that the event should not generate a more serious plant condition without other faults occurring independently. If operator action is required to terminate the transient, the following minimum time intervals must be available between the time when the alarm announces that dilution is occurring and the time -f-loss of shutdown margin:

a. During Refueling:

30 minutes.

b. During Startup, cold shutdown hot standby, and power operation:

15 minutes.

15.4.6.1 Conclusions for the Boron Dilution Events The licensee assessed the minimum time available for operator action to mitigate the consequences of a boron dilution event.

The licensee has determined that during refueling, the operators have in excess of 30 minutes to respond and mitigate the dilution process after receiving alarm indications. We find this acceptable.

()123 During startup, cold shutdown, and hot standby operating conditions, the licensee calculated that the operator has in excess of 15 minutes to respond and mitigate the dilution event. We find this acceptable. The dilution event at power operation is bounded by the consequences of the rod withdrawal events. The consequences for these events showed that fuel integrity is maintained (MDNBR is greater than 1.17). We find this acceptable.

15.5 Increases In Reactor Coolant System Inventory 15.5.1 Inadvertent Operation Of Emergency Core Cooling System The licensee concluded in technical report XN-NF-83-72 that the subject event need not be analyzed for Cycle 10 reload. The licensee argued that the shutoff head of the high head safety injection pumps is 1500 psia, which is well below the trip actuation setpoint of 1850 psia. With regards to the pres surized thermal shock issue, the licensee has an ongoing program, which includes installing part length shielding fuel assemblies to meet the screening criteria for RTNDT. We find the licensee's assessment acceptable.

15.5.2 CVCS Malfunction that Increases Reactor Coolant Inventory The licensee concluded in technical report XN-NF-83-72 that the subject event need not be analyzed since it is bounded by other events and previously addressed in the updated H. B. Robinson Unit 2 FSAR. We find the licensees assessment acceptable.

(

124 15.6 Decrease In Reactor Coolant System Inventory 15.6.1 Inadvertent Opening Of A Pressurizer Safety Or Power Operated Relief Valve In technical report XN-NF-83-72, the licensee referenced the FSAR design basis analysis of an inadvertent opening of a pressurizer safety valve. The H. B. Robinson Unit 2 licensing basis acceptance criteria for this event is as for postulated accidents. However, the licensee performed an analysis which demonstrated that DNBR would not decrease below the specified acceptable fuel design limit (SAFDL). The calculated minimum DNBR was 1.33, well above the 1.17 SAFDL for the XNB critical heat flux correlation.

We find the licensee's assessment acceptable.

15.6.2 Steam Generator Tube Rupture Section 15.6.3 of XN-NF-84-74 evaluates the Steam Generator Tube Rupture event for H. B. Robinson Unit 2. This event is initiated with an instantaneous rupture of a steam generator tube, relieving primary system coolant to the shell of the steam generator.

The steam generator tube rupture event is categorized as a Condition IV event, a Postulated accident. The acceptance criteria for this event are as follows:

25 (1)

For a postulated accident with an assumed pre-accident iodine spike in the reactor coolant and for.the postulated accident with the highest worth control rod stuck out of the core, the calculated doses should not exceed the guideline values of 10 CFR 100, Section 11.

(2)

For the postulated accident with equilibrium iodine concentration for continued full power operation in combination with an assumed accident initiated iodine spike, the calculated doses should not exceed 10% or 2.5 rem and 30 rem, respectively, for the whole-body and thyroid doses.

Challenge to the specified acceptable fuel design limits (SAFOL), or fuel integrity, for the steam generator tube rupture event is bounded by the analysis of the inadvertent opening of a pressurizer relief valve (Section 15.6.1).

The analysis of the inadvertent opening of a pressurizer relief valve showed that the minimum ONBR did not decrease below the SAFDL. Consequently, fuel integrity is maintained.

The licensee applied the H. B. Robinson design basis methods for calculating radiological releases for Cycle 10. The only variation in the method was a reanalysis of the primary to secondary coolant break flow for the new steam generator (the steam generators for H. B. Robinson Unit 2 were replaced).

26 The analysis assumptions for this event assumed loss of offsite power which resulted in steam relief directly to the atmosphere through a stuck open PORV. Operator action at 30 minutes into the event was credited to isolate the affected steam generator.

The.RELAP5/MOD1 computer program was used to calculate the primary to secondary flow characteristics and the flow out the atmospheric dump and POR valves. Several break locations were evaluated for limiting conditions. The limiting break location was determined to result adjacent to the hot leg with cold leg fluid temperature conditions. The RELAPS model nodalization of the steam generator was acceptably detailed. The primary system was modeled as a stand-alone steam generator between the hot and cold legs. The reactor vessel was not modeled. To conservatively bound the possible break and atmospheric release rates, conservative primary system boundary conditions were employed: These included maintaining'a constant primary system pressure of 2280 psia and temperature of 536.2oF. Sensitivity studies were performed with a boundary temperature of 614.6 0F and combination of 614.60 F at the hot leg and 536.2 OF at the cold leg. The lower temperature case resulted in the maximum flow out the tube.

In addition to the primary to secondary heat transfer, the licensee incorporated an additional energy boundary condition to the secondary system equivalent to 1/3 of the core generated power, including the energy generated by the primary coolant

27 pumps, plus the energy equivalent to 100'F cooling of the primary system. This assumption maximized the mass transferred through the PORVs out to the atmosphere.

To confirm the acceptability of the RELAP5 break flow model, the licensee benchmarked the calculated flow rate with the Moody and Henry/Fauske break flow models. The comparison validated the conservatism of the RELAP5 calculation.

We find the method for calculating break flow characteristics acceptable.

15.6.2.1 Summary for the Steam GeneratorTube Rupture Event The licensee performed a radiological assessment of a postu lated steam generator tube rupture event. The licensing design basis assumptions were used in this assessment. This assump tion credTded operator action to isolate the faulted steam generator 30 minutes into the event.

The contaminated mass entering the atmosphere was conserva tively calculated. Since the maximum allowable Tech Spec primary system activity has not been modified since the last FSAR update, the same activity was applied to the analysis for Cycle 10.

The consequential dosage for this event was calculated at 0.6 rem whole body and 3.4 rem thyroid. These are well within the 10 CFR 100 guidelines. We find the analysis of the steam generator tube rupture event acceptable.

A-1 APPENDIX - A CONFIRMATORY STEAM LINE BREAK ANALYSIS IN SUPPORT OF THE H. B. ROBINSON UNIT 2, CYCLE-10 RELOAD APPLICATION I.

INTRODUCTION The previous H. B. Robinson Unit 2 steam line break analysis was performed by Exxon Nuclear Company using the PTSPWR2 computer program. Exxon Nuclear Company is the fuel vendor for Carolina Power & Light.Company (CP&L), the licensee of H. B. Robinson Unit

2.

The PTSPWR2 computer program is a one-dimensional analytical representation of a nuclear steam supply system (NSSS). The program assumes ideal thermal-hydraulic mixing of the coolant entering the reactor vessel from the affected and intact steam generators. In addition, the moderator and Doppler reactivity feedback are obtained from average core thermal-hydraulic conditions.

Proprietary experimental data obtained by the NSSS vendors have shown significant thermal-hydraulic asymmetry of the fluid states within the reactor vessel for expected steam line break condi tions.

Consequently, the staff requested (as part of the generic review of the PTSPWR2 computer program) Exxon Nuclear Company to

A-2 refine its analytical methods to account for asymmetric influences and demonstrate acceptability of the PTSPWR2 results.

Exxon Nuclear Company is revising its analytical methods to address the above concerns.

The licensee has committed to provide reanalyses of the steam line break event for Cycle 10 by January 31, 1985. This commitment is acceptable.

The bases for accepting a late submittal of the steam line break event are as follows:

(1) H. B. Robinson Unit 2 replaced its steam generators with a new model that incorporates an integral flow restrictor withim-the outlet nozzle. The flow-restrictor significantly reduces the consequences of a major rupture of a steam line, (2) The limiting consequences of a large steam line break occurs at end of cycle (EOC) when the moderator coefficient is at its most negative value, and (3) Staff analysis of the steam line break event (guillotine break) showed ample margin to the acceptance criteria for H.B. Robinson Unit 2.

£

?

A-3 The following documents the staff's analysis of a postulated steam line break event for H. B. Robinson Unit 2.

II.

MODEL DESCRIPTION The computer code used for analyzing the steam line break (SLB) event was RELAP5/MOD1.5 Cycle 39. An input deck of a generic 3-l'oop Westinghouse plant was modified by data supplied by CP&L and ENC to model the H. B. Robinson plant.

II.1 NODALIZATION The model nodalization used for the H. B. Robinson SLB calcula tions is shown in Figs. A-1 and A-2. This nodalization represents the major components and flow paths of the H. B. Robinson 3-loop nuclear steam supply system.

The model consists of two loops. The intact loop is a lumped representation of two loops containing the unaffected steam generators. The pressurizer is connected to the unaffected loop.

The affected loop contains the steam generator with the faulted steam line.

Except for the upper head region, the rector vessel was divided into two parallel channels proportioned 2:1. the model incorpo rated cross flow junctions in the upper and lower plena to simu late thermal-hydraulic coupling between the two core channels.

[-V 3Mf9I3

• uwel~via 6UtPON 9*IUOW/90dV (SSSN) walsKS (tddnS weaiS jppq lt 06(3 91~

Ct 00(

3 Utz)

I 1

11013wih 60t 1

001 9IUI ZOE)3 0D~3 L$02

A,5 cl 70 Co a"

IT4o1 12 5

2r

£3 4

43 3

j 3

w4 0 25 Z

Split Vessel RELAP5/MODI.5 Noding Oiagram.

FIGURE A-2

A-6 The amount of coupling was experimentally predetermined. Heat slabs representing the primary system metal masses in the vessel, pressurizer and steam generators as well as the metal in the primary coolant piping were included in the model.

11.2 MAJOR ASSUMPTIONS AND INITIAL CONDITIONS The following major assumptions and initial conditions were used:

1.

The system initial conditions prior to initiation of the SLB event are listed in Table A-1.

2.

A uniform power profile was used.

A power fraction of 0.1667 was assigned to each of the six axial core regions. In addition, these power fractions were weighted 2:1 between the intact core and the affected core regions.

3.

Point kinetic reactivity feedback as a function of four parame ters wafcalculated by a control system. The method is similar to that applied by Westinghouse, the reactor vendor for H. B. Robinson.

(a)2 Moderator Density Reactivity Feedback The moderator density reactivity, as documented in Table A-2 was provided by Exxon Nuclear Company. Each of the six volumes within a core channel provided one-sixth of the total moderator reactivity feedback for that channel (uniform axial weighting). The overall moderator reactivity was given by weighting the affected and

A-7 TABLE A-1 STEAM LINE BREAK ANALYSIS INITIAL CONDITIONS Parameter Zero Power Core Power 27.75 MW Core mass flow 29,166 lb/s Core T 0.64 0F Cold"Teg temperature 550.oF Primary pressure 2251 psia Secondary pressure 1004 psia Secondary mass 135,000 lb/steam generator Steam/Feed flow Boron concentration 0 ppm

A-8 TABLE A-2 MODERATOR DENSITY REACTIVITY Moderator Density Reactivity (Lb/ft3)

($)

43.93

-3.71 46.73 0.00 49.35 3.35 51.51 6.19 53.88 8.38 5-67 10.08 57.51 11.41

A-9 unaffected core channels in accordance with the Westing house methodology.

(b) Doppler Reactivity Feedback The Doppler contribution to total reactivity was divided into two parts. One-part represented the power coeffi cient at constant moderator temperature (Table A-3),

while the other part accounted for the variation in the moderator temperature.

(c) Control Rod Insertion The control rods, with the exception of the single most reactive rod, have a reactivity worth of -3.61 $.

This reactivity was assumed to be linearly inserted with.0.2 sec. delay at time of reactor trip.

(d) Boron Reactivity Feedback A core average boron concentration calculated by the RELAP5 control system, was used for the reactivity feedback. It was assumed that the HPI system initiated 13 seconds after a generated SI signal.

It was also assumed that borated water did not enter the primary coolant system until the HPI lines were purged of its initial inventory. The clearing of the lines was

A-10 TABLE A-3 DOPPLER POWER REACTIVITY Core Power Density Reactivity

(% of Rated)

Cs) 0 0.00 5

0.65 10

-1.18 20

-1.96 30

-2.61 401 3.61

A-11 assumed to take 30 seconds. This was based upon a line volume of 30 ft3 and an injection rate of 1 ft3/sec.

The boron worth is given in Table A-4. The initial boron concentration in the boron injection tank was specified as 21000 ppm. It was conservatively assumed that this concentration decreased exponentially to 10 ppm over a period of 120 sec. This is conservative since the makeup water flowing into the tank is borated at approximately 2000 ppm.

4. The trips and setpoints used in the SLB calculation are listed in Table A-5.

5. The SI injection systems represents a single high pressure injection train. Injection temperature was set at 1200F.

6.

All of he main feedwater was diverted to the affected steam generator during the initial 10 sec of the transient. The flow was assumed constant at 3861.1 ibm/sec. The temperature of the feedwater was assumed at 1200F.

A-12 TABLE A-4 BORON REACTIVITY COEFFICIENT Moderator Density Boron Coefficient (Lb/ft3)

($/PPM) 43.93 0.020 46.73 0.022 57.51 0.028

A-13 TABLE A-5 STEAM LINE BREAK ANALYSIS TRIPS AND SETPOINTS Trip Setpoint

1. High steam flow 450 lb/s

(

40% nominal)

2.

Low steam line pressure 615 psia SI signal

3.

Low Tavg 5430F

4.

Low primary pressure 1780 psia

5. Safety injection (1 and (2 or 3) or 4 of the above trips

A-14

7. The reactor coolant pumps remained in operation at a constant speed throughout the transient.

8. A recirculation model was added to the steam generators so that a conservative (perfect) separation could be calculated.

By calculating only steam flow out the break, the energy removed from the system is maximized.

III.

BENCHMARK ANALYSIS H. B. Robinson's original steam generators did not have an inte gral flow restrictor incorporated into their outlet nozzles. The original FSAR analysis, therefore, modeled the break area as 4.6 ft2.

To benchmark the H. B. Robinson RELAP5 model with the FSAR results, a calculation was performed which assumed a 4.6 ft2 break area (The cross sectional area for the flow restrictor is 1.4 ft2).

The break was simulated by an instantaneous opening of two flow paths, one connected to each side of the guillotine break (steam generator secondary). The primary break path (connected to the affected steam generator) was sized at 4.6 ft2, to simulate the unrestricted rupture of a main steam line. The second break path

A-15 was sized at 1.485 ft2, to simulate the flow through the restrictor of the broken steam line. Flow from this valve was terminated 10 seconds after break initiation, the assumed closure time of the steam isolation valves (MSIVs).

The event was initiated at 100 seconds (after obtaining steady-state initial conditions). Results of the reactivity, power level, primary pressure, and primary coolant temperatures are shown in Fig. A-3 through A-6, respectively.

The results from the original design basis FSAR analysis are shown in Fig. A-7. Comparisons between the staff calculation and the FSAR design basis analysis are in good agreement.

IV.

DETERMINATION OF THE LIMITING BREAK The following cases were analyzed to determine the limiting break location and conditions for H. B. Robinson Unit 2 Cycle 10:

Case 1:

Break between the flow restrictors with offsite power available. The blowdown areas are 1.388 ft2 (affected) and 1.485 ft2 (intact).

Case 2:

Break downstream of both flow restrictors with offsite power available. The blowdown areas are 1.388 ft2 (affected) and 2.776 ft2 (intact).

TOTAL REACTIVITY ome a*

-24 0-44 MCI 1

8 1520/04/84 0

68 lea ISO 2011 261 Time (s)

H.B.ROBINSON CYCLE 10 RELOAD STEAM LINE BREAK ANALYSIS BREAK UPSTREAM OF FLOW RESTRICTORS' FIGURE A-3

RATED POWER LEVEL FULL POWER = 2300 MW )

e.z hi

a.

IL 6.4 L

(n ftd (L

0..

gso l ea 15e 28g 250 Time (s)

H.B.ROBINSON CYCLE 10 RELOAD STEAM LINE BREAK ANALYSIS BREAK UPSTREAM OF FLOW RESTRICTORS" FIGURE A-4

PRESSURISER PRESSURE

0.

U)

(L Is" Hj&AO"~

0*

65*3 9/94/44 0

so toS 165 208 268 Time (s)

H.B.ROBINSON CYCLE 10 RELOAD STEAM LINE BREAK ANALYSIS BREAK UPSTREAM OF FLOW RESTRICTORS c

FIGURE A-5

I AFFECTED HL 2 AFFECTED CL 3 INTACT HL 4 INTACT CL 5 AVERAGE (L

0 a

388 HAMMJH I.5 89.57.83.

sa/84/94 0

so Ias 1S 288 250 Time (s)

H.B.ROBINSON CYCLE 10 RELOAD STEAM LINE BREAK ANALYSIS BREAK UPSTREAM OF FLOW RESTRICTORS FIGURE A-6

CORE HEAT FLUX STEAM FLOW REACTOR COOLANT REACTOR COOLANT AVERAGE 0

% of 2200 MWt PRESSURE, PSIA TEMPERATURE, F

REACTIVITY I

I Ui H

H N

N H

0 0

UL 0

1-n w

-r--

41 Ln Un o

O 0

N 01 00 0

0 0

0 0

0 0

LJ 0

UL 0

i

'0o 0

0 0

0 0

0 0

0 0

0 0

0 0

0 0

0 0

C oE o

I on I

a 00 II tdt 0

0EH 6

F-3 oD 0

f:0 00N j

W Hn H

0 I

CD 00 or

A-21 Case 3:

Break downstream of both flow restrictors with loss of offsite power at break initiation.

Case 4:

Break downstream of both flow restrictors with loss of offiste power at time of reactor trip.

All cases assumed initial hot shutdown conditions. This maximized the liquid mass within the steam generator shell, minimized the core generated decay heat and thereby maximized the overcooling for the event. The peak return to power and time of occurrence for each case is listed in Table A-6.

The results of the steam line break studies showed that the limiting condition occurs for the break downstream of the flow restrictors.(greatest cross sectional flow area) with offsite power available. Results for this case are shown in Figures A-8 thru A-19.

V.

DESCRIPTION OF THE LIMITING SLB EVENT As described in the previous section, the limiting steam line break (SLB) event occurred for a break postulated downsteam of the flow restrictors with offsite power available. The sequence of events is given in Table A-7. Various responses in the NSSS are shown in Figures A-8 through A-19.

A-22 TABLE A-6 MAXIMUM RETURN TO POWER FOR CASES 1-4 Maximum Return to Power Time

(% of 2300 MWt)

(sec)

Case 1 19.2 48.0 Case 2 22.4 47.4 Case 3 13.1 51.2 Case 4 13-6 50.6

A-23 TABLE A-7 SEQUENCE OF EVENTS FOR THE LIMITING SLB Time (Seconds)

Event 0.00 Break initiation 0.02 High steam flow signal 3.58 SI signal initiated all feedwater diverted to affected steam generator 3.80 Reactor Trip Initiated 5.60 Low steam line pressure signal 10.00 MSIV closed 13.58 Feedwater stopped 14.60 Reactivity becomes positive 16.58 HPI initiated 19.00 Peak reactivity 46.60 SI boron enters core 47.40 Peak Core Power

TOTAL REACTIVITY 2

10

-2

-4 HAGNUM 1.5 11.47.43.

00/96/84 B

6o 1ea 150 288 268 Time (s)

H.B.ROBINSON CYCLE 10 RELOAD STEAM LINE BREAK ANALYSIS BREAK DOWNSTREAM OF FLOW RESTRICTORS FIGURE A-8

I BORON 2 SH-UTDOWN 0 POWER FUEL 4 MOD.TEMP.

5 I40D.DENS.

EIDo n

u oOR Ar5 9

a 1 c4 1 9A 4 9 9 C9 114743 a

9 e

26 5

Tie-s H..OISNCYL 0RLA h1A84J~~FIUR A.-9.743 Q/68

RATED POWER LEVEL4 CFULL POWER 2300 MW) 6.3 1

1.7.3 2h826 Tie s

H..OISNCYL 0RLA STA LN BEK NLYI BRA ONTRA FFO RSRCOL FIUR A1

PRESSURISER PRESSURE 20" (I) hAOMIJ 1.56 11.47.43.

09/06/94 0

60 le6 156 268 2s8 Time (s)

Fi.B.ROBINSON CYCLE 10 RELOAD STEAM LINE BREAK ANALYSIS BREAK DOWNSTREAM OF FLOW RESTR:CTORS' FIGURE A-li

PRESSURIZER LEVEL to I-5 H

.56 11.47.4 99/86/84 so 5ea 110S 2N8 250 Time (s)

H.B.ROBINSON CYCLE 10 RELOAD STEAM LINE BREAK ANALYSIS BREAK DOWNSTREAM OF. FLOW RESTR7ICTORS' FIGURE A-12

AFFECTED HL 2 AFFECTED CL 3 INTACT HL 4 INTAPT CL 6 AVERAGE u w IAVM-M0 Ma 15 ul MW 9t~tt1.5 t 11.47.48.

8/60 60 Sto8o1 200 268 Time (s)

H.B.ROBINSON %CYCLE 10 RELOAD STEAM LINE BREAK ANALYSIS BREAK DOWNSTREAM OF FLOW RESTRICTORS FIGURE A-13 Ge

I AFFECTED LOOP CL FLOW RATE 2 INTACT LOOP CL FLOW RATE 20~

2 aZ t to"

)T 60" hAUNMt 1.5 11.47.43.

9/6/84 I

5 I.

53 09 2

Time (s)

H.B.ROBINSON CYCLE 10 RELOAD STEAM LINE BREAK ANALYSIS BREAK DO~WNSTREAM OF FLOW RESTRICTORS J

FIGURE A-14

BREAK FLOW RATE IS c

ie is tMOA~bt 1.5 11.47.43.

6D 5N Is @50 28 258 Time (s)

H.B.ROBINSON CYCLE 10 RELOAD STEAM LINE BREAK ANALYSIS BREAK DOWNSTREAM OF FLOW RESTRICTORS FIGURE A-1 5

INTACT STEAM LINE FLOW RATE In.

5.

2S" tL Beo e

168 288 269 Time (s)

H.B.ROBINSON CYCLE 10 RELOAD STEAM LINE BREAK ANALYSIS BREAK DOWN'STREAM OF FLOW RESTR:CTORS' FIGURE A-16

I AFFECTED SG MASS Xle 2 INTACT SG MASS

0)

MNGNM 1.5 11.47.43.

OQ/S6/94 Sso Is@

151 2a"

~

Tinme (s)

H.B.ROBINSON CYCLE 10 RELOAD STEAM LINE BREAK ANALYSIS BREAK DOWNSTREAM OF FLOW RESTRICTORS FIGURE A-17

I AFFECTED SG TEMPERATURE 2 INTACT SG TEMPERATURE UL.

40 h41LhI 14.3.i/49 soIO28 S

Tim (s

hi.OBNO CYCL_10_ELOA STA IN RA AAYI BRA ONTEMO FO ETIiOS STEAM LIEURE AKAALSI

I AFFECTED SG PRESSURE 2 INTACT SG PRESSURE ILI HA"S3.5 I1.47.48.

90/9 4 a

go la 150 2ee 258 Time (s)

H.B.ROBINSON CYCLE 10 RELOAD STEAM LINE BREAK ANALYSIS IiI BREAK DOWNSTREAM OF FLOW RESTRICTORS FG E

FIGUR A-19

A-36 The reactivity for the limiting case is shown in Figure A-8 and A-9. The initial reactivity begins at zero. $3.61 of negative reactivity is inserted by the control rods between 0.2 and 2.4 seconds. As the primary system is cooled, positive reactivity is inserted by the moderator and Doppler feedbacks. Criticality occurs at 14.6 seconds. At 19 seconds the reactivity peaked at

$0.58. At 46.6 seconds, boron enters the core through the emer gency core cooling system (ECCS).

The reactor power response is shown in Figure A-10. The power peaked at a value of 22.4% of rated power (2300 MWt) at 47 sec onds. Competing effects between the boron and Doppler reactivi ties led to some oscillatory power response.

Heat removaffrom the primary system led'to a rapid decrease in primary system pressure (see Figure A-11). The depressurization rate is significantly reduced as the reactor vessel voids within the upper head. The pressurizer is depleted of liquid inventory at the same time as the depressurization rate decreased (see Figure A-12).

The hot leg and cold leg coolant temperature for both the affected and intact loops, along with the average core coolant temperature, are shown in Figure A-13. As noted by the decreasing coolant temperatures, the energy removed by the steam generators exceed

A-37 the energy generated by the core. The addition of borated ECC water assures a steady decline in reactor power. As the primary system coolant temperature is decreased, the primary system flow increases. This is in response to the increasing density. Figure A-14 shows that primary system flow as a.function of time.

Figure A-15 and A-16 describe the break flow characteristics for the affected and intact steam generators, respectively. The blowdown of the steam generators is the primary forcing function for the SLB transient. At 150 seconds into the event (250 seconds of plot time), the break flow from the affected steam generator decreased to 577 ibm/sec. This is equivalent to the 20% of rated steam flow. The rapid isolation of the intact steam generators result from closure of the MSIVs.

The fluid inventory, temperature and pressure for the intact and affected steam generator shells are shown in Figures A-17, A-18, and A-19. The increase in mass to the affected generators (during the initial 10 seconds of the event) comes from the addition of main feedwater. As subcooling is decreased, the secondary temper ature in the downcomer begins a momentary increase. As the affected steam generator continues its blowdown, the secondary coolant temperature and pressure steadily decrease.

A-38 VI.

SUMMARY

The staff analysis of a steam line break event for H. B. Robinson

'Unit 2, Cycle 10, confirmed that the new steam generators with integral flow restrictors decreased the.severity of the event when compared with the design basis FSAR analysis. The design basis analysis resulted in a 39% return to power. This was confirmed by the benchmark analysis conducted in this review (see Section III to this Appendix).

The design basis analysis, with its 39% return to power, did not result in a calculated DNBR below the specified acceptable fuel design limit (SAFOL). Since the H. B. Robinson Unit 2 analysis for Cycle 10 resulted in a return to power of only 22.4% (approxi mately 50% of the design basis calculation), the margin to the SAFOL is signfficantly increased. Consequently, fuel integrity is maintained.