Text

Emergency License Amendment Request to Revise Technical Specification 5.3.2 to Allow Operation with 56 Full-Length Control Rod Assemblies for Cycle 20
	ML15343A347
Person / Time
Site:	South Texas
Issue date:	12/03/2015
From:	Gerry Powell; South Texas
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NOC-AE-15003315
	Download: ML15343A347 (29)
	v • d • e

Nuclear Operating Company South Teas Iroject Electric Generationg Station pEo. Box2A9 Wadsworth, Texas 77483 ,viv -

December 3, 2015 NOC-AE- 15003315 10 CER 50.90 10 CFR 50.91 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, .DC 20555-0001 South Texas Project Unit 1 Docket No. STN 50-498 Emergency License Amendment Request to Revise Technical Specification 5.3.2 to Allow Operation with 56 FulI-Lencqth Control Rod Assemblies for Unit 1 Cycle 20 Pursuant to 10 CFR 50.90 and 10 CFR 50.91(a)(5), STP Nuclear Operating Company (STPNOC) hereby requests an emergency license amendment to South Texas Project Operating License NPF-76. Currently, Technical Specification (TS) 5.3.2 requires the Unit 1 core to contain 57 full-length control rods. The proposed amendment would revise TS 5.3.2 to require the Unit 1 Cycle 20 core to contain 56 full-length control rods with no full-length control rod assembly in core location D-6.

In preparation for restarting Unit 1 during refueling outage 1RE19, STPNOC performed control rod drop time surveillance testing per TS surveillance requirement 4.1.3.4. During this testing, Control Rod D-6 in Shutdown Bank A did not function as expected. During troubleshooting activities, Control Rod D-6 was unable to be moved using normal methods. Subsequently, Control Rod D-6 was moved to the bottom of the core (i.e., fully inserted).

Unit 1 was cooled down to Mode 6 and the reactor head was disassembled and inspected. It was determined that the issue with Control Rod D-6 is confined to the Control Rod Drive Mechanism (CRDM).

In-situ replacement of the affected CRDM would be a first-of-a-kind activity in the United States requiring special tooling that is unavailable at this time. Therefore, STPNOC has decided to remove Control Rod D-6 for Unit 1 Cycle 20 and operate the unit with 56 full-length control rods.

Approval of this license amendment request is required for Unit 1 to enter Mode 5 and resume power operation.

The Enclosure to this letter provides a technical and regulatory evaluation of the proposed amendment. Attachments 1 and 2 to the Enclosure contain the proposed TS page markup and clean TS page, respectively.

STPNOC is requesting approval of the proposed amendment on an emergency basis pursuant to 10 CFR 50.91(a)(5) to allow Unit i to resume operation and is requesting approval by December 11, 2015. The TS change will be a permanent change, but will only be in effect for the duration of Unit 1 Cycle 20. Once approved, the amendment shall be implemented within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . STPNOC only requires NRC approval of the proposed change to TS 5.3.2; all design changes and supporting safety analyses discussed in this document were performed in accordance with the current licensing basis.

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STI: 34244822

NOC-AE-1 5003315 Page 2 of 3 The proposed amendment has been reviewed and approved by the STPNOC Plant Operations Review Committee and has undergone an independent Organizational Unit Review.

In accordance with 10 CFR 50.91, STPNOC is notifying the State of Texas of this license amendment request by transmitting a copy of this letter and enclosure to the designated State Official.

There are no commitments in this letter.

Ifthere are any questions or if additional information is needed, please contact Drew Richards at (361) 972-7666 or me at (361) 972-7566.

I declare under penalty of perjury that the foregoing is true and correct.

Executedon Oee~(5Z~

G. T. Powell Site Vice President amr/GTP

Enclosure:

Evaluation of the Proposed Change

NOC-AE- 15003315 Page 3 of 3 Cc:

(paper copy)

(electronic copy)

Regional Administrator, Region IV Morgqan, Lewis & Bockius LLP U.S. Nuclear Regulatory Commission Steve Frantz, Esquire 1600 East Lamar Boulevard Arlington, TX 76011-4511 U.S. Nuclear Regqulatory Commission Lisa M. Regner Lisa M. Regner Senior Project Manager NRG South Texas LP U.S. Nuclear Regulatory Commission John Ragan One White Flint North (O8H04) Chris O'Hara 11555 Rockville Pike Jim von Suskil Rockville, MD 20852 CPS Enerqy NRC Resident-Inspector Kevin Polio U. S. Nuclear Regulatory Commission Cris Eugster P. O. Box 289, Mail Code: MN116 L. D. Blaylock Wadsworth, TX 77483 Cramn Caton & James, P.C.

Peter Nemeth City of Austin Cheryl Mele John Wester Texas Dep~t. of State Health Services Richard A. Ratliff Robert Free

Enclosure NOC-AE-1 5003315 Page 1 of 22 ENCLOSURE Evaluation of the Proposed Change

Subject:

Emergency License Amendment Request to Revise Technical Specification 5.3.2 to Allow Operation with 56 Full-Length Control Rod Assemblies for Unit I Cycle 20 1,.0 Summary description 2.0 Detailed description 2.1 Proposed amendment 2.2 Control Rod D-6 issue 3.0 Technical evaluation 3.1 System description 3.2 Current licensing basis 3.3 Impact on the safety analysis 3.4 Field work required to remove Control Rod 0-6 from service 3.5 Evaluation of potential design impacts 3.6 Adequate level of safety 4.0 Regulatory evaluation 4.1 Applicable regulatory requirements/criteria 4.2 Precedence 4.3 No significant hazards consideration determination 4.4 Conclusions 5.0 Environmental consideration 6.0 References Attachments:

1. Technical Specification Markup

2. Clean Technical Specification Page

Enclosure NOC-AE- 15003315 Page 2 of 22 1.0 Summary description This evaluation supports a request to amend Operating License NPF-76 for South Texas Project (STP) Unit 1 by adding a footnote to Technical Specification (TS) 5.3.2, "Control Rod Assemblies," to allow the Unit 1 Cycle 20 core to contain 56 full-length control rods with no full-length control rod in core location 0-6, in lieu of the current requirement of 57 full-Length control rods. An STP operating cycle is nominally 18 months.

In preparation for restarting Unit I during refueling outage 1REI9, Control Rod D-6 in Shutdown Bank A did not function as expected. Subsequent inspections have determined that there is deformation of the rod holdout ring in the Control Rod Drive Mechanism (CRDM) for Control Rod D-6 which requires replacement of the CRDM. In-situ replacement of the affected CRDM would be a first-of-a-kind activity in the United States and would require special tooling that is unavailable at this time. STP Nuclear Operating Company (STPNOC) has decided to remove Control Rod D-6 and operate Unit 1 with 56 full-length control rods for Cycle 20.

STPNOC is requesting approval of the proposed amendment on an emergency basis pursuant to 10 CFR 50.91(a)(5) to allow Unit i to resume power operation following refueling outage 1 RE1 9. Approval of the proposed amendment is requested by December 11, 2015, to support Unit 1 entry into Mode 5 and resume operation.

2.0 Detailed description 2.1. Proposed amendment The proposed amendment would revise TS 5.3.2, "Control Rod Assemblies," to add a footnote to permit operation with 56 full-length control rods during Unit 1 Cycle 20 in lieu of the requirement to contain 57 full-length control rod assemblies. STPNOC performed a thorough review of the TS and has determined that no other TS changes are required.

STPNOC only requires NRC approval of the proposed change to TS 5.3.2; all design changes and supporting safety analyses discussed in this document were performed in accordance with the current licensing basis.

2.2. Control Rod 0-6 issue On November 18, 2015, in preparation for restarting unit 1 following refueling outage 1RE19, STPNOC performed control rod drop time surveillance testing. During this testing, Control Rod 0-6 in Shutdown Bank A did not function as expected. During subsequent troubleshooting activities, Control Rod 0-6 was unable to be moved using normal methods. Control Rod D-6 was later able to be moved to the bottom of the core (i.e., fully inserted). Unit 1 was subsequently cooled down to Mode 6 and the reactor head was disassembled. All 57 Unit 1-CRDMs were inspected to determine the extent of condition. It was determined that the issue with Control Rod 0-6 is due to deformation of the CRDM rod holdout (RHO) ring which is used during rapid refueling operations (see Section 3.1 for a description of rod lockout and the rapid refueling feature at STP). No similar deformation has been observed on the other 56 Unit 1 CRDMs.

Previously, on November 11, 2012, during Unit 1 refueling outage 1RE17, Control Rod 0-6 failed to ;fully insert into the core when dropped during rod unlocking operations. While performing testing on November 12, 2012, Control Rod 0-6 dropped to the bottom of the core.

Following further testing and evaluation, Control Rod 0-6 performed as designed; the control rod passed all surveillance testing during the following operating cycle (Cycle 18). During the

Enclosure NOC-AE-1 5003315 Page 3 of 22 events in 1RE17, hydraulic forces caused the RHO ring to rise in position behind the stationary gripper latches simultaneous with the latches engaging the control rod drive shaft. The engagement of the stationary gripper latches concurrent with the falling control rod shaft provided a significant outward force on the latches resulting in deformation of the RHO ring.

On March 17, 2014, during Unit 1 refueling outage 1RE18, Control Rod 0-6 was unable to be locked out in preparations for a rapid refueling. Visual inspection of the Control Rod D-6 CRDM determined that the RHO ring was damaged and was moving with the stationary gripper pole, causing increased stationary gripper latch closure times. The decision was made to continue with a non-rapid refueling and the refueling outage continued without the control rods being locked out. Following further evaluation, it was determined that the rod drop function and the ability to step Control Rod 0-6 was not impaired; the only affected function was the ability to lock out Control Rod 0-6 to perform a rapid refueling.

The need for this license amendment could not be avoided or predicted. During Unit 1 Cycle 19, the D-6 CRDM was monitored during monthly control rod exercise testing and the CRDM timing traces did not indicate further degradation affecting CRDM performance. During 1 RE1 9, cold rod exercises were satisfactorily performed for all control rods, including Control Rod D-6. Due to an issue with Reactor Coolant Pump seal leakage, Unit 1 was cooled down to Mode 5 for repairs and heated back up. The current issue with Control Rod 0-6 was discovered during rod testing at normal operating pressure and temperature following the Reactor Coolant Pump seal repairs.

The replacement of a CROM of similar design and installation configuration has not been performed in the United States. In-situ replacement of the 0-6 CROM would require special modified tooling similar in nature to the original manufacturing tooling which currently does not exist. The planning associated with a CRDM replacement activity would require fabrication of mockups to test the effectiveness of the tooling, methods, and procedures. This planning and preparation process is expected to require a lead time on the order of several months.

Consideration was given to operating Unit 1 Cycle 20 with Control Rod 0-6 fully inserted in the core. This option is~not considered viable for the following reasons:

The impact on core power distribution would likely require operation at a reduced power level;

The core would be susceptible to radial xenon oscillations that would challenge operator responses; and

Uneven depletion of fuel assemblies would have a significant impact on the core design for future fuel cycles with regard to fuel economy and safety/operating margins.

Therefore, STPNOC has determined that the best option is to safely operate Unit 1 Cycle 20 with Control Rod 0-6 removed. STPNOC is currently evaluating future repair options to restore Control Rod 0-6 to its original function during the next refueling outage, 1RE2O, in Spring of 2017.

3.0 Technical evaluation 3.1. System description Unit 1 currently contains 57 full-length control rod assemblies divided into four control banks (Control Banks A, B, C, 0) and five shutdown banks (Shutdown Banks A, B, C, 0, E). Of the nine banks, Control Bank 0 is used for short-term control during normal at-power operation. The remaining control banks are normally used for reactor startup and shutdown. The shutdown

Enclosure NOC-AE- 15003315 Page 4 of 22 banks provide additional negative reactivity to meet shutdown margin requirements. During Modes 1 and 2, the shutdown banks are fully withdrawn from the core in accordance with TS 3.1.3.5 and as specified in the Core Operating Limits Report (COLR). Control Rod D-6 is located in Shutdown Bank A and is located in the core as shown in Figure 1.

Figure 1, Control Rod Locations R P N M L K J H G FE D C B A SC 1 SB 2 B SB C B 3

4 SD!

SD B

SA SB C

SE

~ B SA t.....i........

S SCA A S 6 BC A C S 7

8 C E A D A S 9

10 B A C A CB 11 SDA A S 12 D SE SA D SA 13 14 1111 SA JSC~

B ISBI C JSB B

SD J]

15 SC SB SB SD Control Number of Bank Number of rods Shutdown Bank rods A 8 SA 8 B 8 SB 8 C 8 SC 4 D 5 SD 4 SE 4 Note that the STP Updated Final Safety Analysis Report (UFSAR) refers to control rods as Rod Cluster Control Assemblies (RCCAs). Generally, "RCCA" refers to the group of individual neutron absorber rods fastened at the top end to a common spider assembly and 'control rod" refers to the entire assembly including the control rod drive shaft. However, for the purposes of this submittal, the terms "control rod" and "RCCA" can be considered to be synonymous.

Enclosure NOC-AE- 15003315 Page 5 of 22 Each control rod is moved by a CRDM consisting of a stationary gripper, movable gripper, and a lift pole. Three coils are installed external to the CRDMs to electromechanically manipulate the CRDM components to produce rod motion. In the STP installation, a fourth coil is installed as part of the Rod Holdout Control System (discussed below). The CRDMs are magnetic jacking type mechanisms that move the control rods within the reactor core by sequencing power to the three coils of each mechanism to produce a stepping rod motion. Rod positioning is achieved through a timed sequence of stationary, movable, and lift coil current. At each point in time during rod positioning, the control rod is being held by either the stationary gripper or movable grippers. Should both sets of grippers be de-energized simultaneously, the corresponding control rod would drop into the core. The primary function of the CRDMs is to insert or withdraw control rods in the core to control reactivity and provide the required shutdown margin.

Mechanically, each control rod location includes a guide tube, which is an assembly that houses and guides the control rod through the upper internals.

Unlike other pressurized water reactors in the United States, the STP CRDM design has a Rod Holdout Control System which allows STP to perform rapid refueling. In a rapid refueling, all control rods are held in the fully withdrawn position by the CRDMs and the reactor vessel head and upper internals are lifted together in one polar crane lift. This saves outage critical path time because the reactor does not have to be fully disassembled and shuffling RCCAs among fuel assemblies in the spent fuel pool prior to reload is not required. Holding the control rods in the fully withdrawn position using continuous energization of the gripper latches was not considered to be fully reliable, therefore an electromagnetically-actuated self-locking mechanism was developed to hold the control rods in place. One of the components of the rapid refueling feature is a separate rod holdout ring below the stationary gripper latch assembly which is positioned to hold the stationary gripper latch closed on the control rod drive after electrical power is removed. After withdrawing the control rods into the reactor vessel upper internals, the rod holdout coil is energized, the RHO ring moves up into a notch behind the closed stationary gripper latches, the stationary coil and then the rod holdout coil are de-energized, and the control rods become mechanically locked in place. After refueling, when the reactor vessel head is set on the vessel flange, the control rods are unlocked and fully inserted.

3.2. Current licensinqi basis As described in UFSAR Section 4.2.2.3.1, Rod Cluster ControlAssembly, the RCCAs are divided into two categories: control and shutdown. The control groups compensate for reactivity changes due to variations in operating conditions of the reactor; i.e., power and temperature variations. Together, the control and shutdown groups provide adequate shutdown margin.

As described in UFSAR Section 4.3.2.4.12, Rod Cluster Control Assemblies, only full-length assemblies are employed in this reactor. The RCCAs are used for shutdown and control purposes to offset fast reactivity changes. The allowed control bank reactivity insertion is limited at full power to maintain shutdown capability. All shutdown RCCAs are fully withdrawn before withdrawal of the control banks.

3.3. Impact on the safety analysis The removal of Control Rod 0-6 is considered a permanent plant change for Unit 1 Cycle 20 and impacts the nuclear design characteristics for this reload core design. As such, the reload design change process has been applied to determine the nuclear design changes and impact to core and fuel performance, as well as impact to the accident analyses described in UFSAR Chapter 15. The same process that is used for each new fuel cycle has been applied for the Unit 1 Cycle 20 D-6 redesign. This process involves determining the nuclear design changes

Enclosure NOC-AE- 15003315 Page 6 of 22 associated with core operation with RCCA D-6 removed, then evaluating the affected nuclear design parameters against a set of limiting values contained in the Reload Safety Analysis Checklist (RSAC). The RSAC process is used to determine ifthe change in core design adversely impacts the bounding key safety parameters assumed in the Chapter 15 safety analysis. In addition to the RSAC process, the impact on Departure from Nucleate Boiling (DNB) due to the change in power distribution attributable to the new core design is also evaluated.

WCAP-9272 provides the reload safety evaluation methodology. The evaluation is documented in the Reload Safety Evaluation to confirm the acceptability of safe operation with the new core design. There were no changes in methods used to perform the core reload design change process for Unit 1 Cycle 20 with RCCA 0-6 removed. In addition, there were no changes in the Unit 1 Cycle 20 fuel assembly core loading pattern as a result of RCCA D-6 removal. Results of the evaluation are described below.

Since the shutdown RCCAs are fully withdrawn from the core while at power, they have a negligible effect on core power distribution. Since Control Rod D-6 is in a shutdown bank, accidents where the core power distribution is a key assumption for the at-power operating condition are not impacted. Since the power distribution is not impacted, DNB is not impacted for at power events. In addition, DNB for zero power events such as the rod withdrawal from subcritical described in UFSAR Section 15.4.1 where the shutdown RCCAs are fully withdrawn is also not impacted.

The removal of Control Rod 0-6 impacts other parameters assumed in the UFSAR Chapter 15 analysis. These parameters are:

Available shutdown margin and most reactive stuck rod worth;

Boron worth when all RCCAs are inserted;

Rod worth of the adjacent RCCAs when all RCCAs are inserted;

The trip reactivity as a function of time; and

The most positive moderator density coefficient.

Other parameters assumed in the STP UFSAR Chapter 15 safety analysis are not impacted by removal of Control Rod 0-6. The impact of removing Control Rod 0-6 on each of the potentially impacted parameters is discussed below. The analysis supporting the evaluation of these impacted parameters was performed using NRC approved methodology described in TS 6.9.1.6 for the COLR. STPNOC has verified that the Unit 1 Cycle 20 COLR submitted to the NRC on November 12, 2015, (this document is not yet in NRC ADAMS; no accession number is available) remains unchanged as a result of Control Rod 0-6 being removed, with the exception of the revision number of a reference document, an added footnote describing removal of ROCA 0-6, and the COLR revision number.

Shutdown Margin The proposed change impacts the available shutdown margin. TS 3.1.1.1 states that the

required shutdown margin shall be within the COLR limit. Maintaining the shutdown margin within the limits specified by TS 3.1.1.1 ensures the safety analysis described in Chapter 15 of the UFSAR remains bounding. Section 2.3.1 of the COLR provides the limit for Modes 1 and 2.

An evaluation of the impact on the reduction of shutdown margin due to the removal of Control Rod 0-6 has been performed and the results are presented in Table 1 below. The shutdown margin is reduced from 2.42% Ap to 2.17% Ap, which remains bounded by the 1.3% Ap limit for Modes 1 and 2 specified in COLR Section 2.3.1. By maintaining the 1.3% Ap shutdown margin

Enclosure NOC-AE- 15003315 Page 7 of 22 limit, the safety analysis described in Chapter 15 of the UFSAR remains bounding with regards to shutdown margin for accidents initiated in Modes 1 and 2.

Table I Comparison of Effect on End-of-Life Shutdown Margin Cycle 19 Cycle 20 Cycle 20 RCCA in D-6 RCCA in D-6 No RCCA in D-6 Control Rod Worth, % Ap All Rods Inserted minus 69 .068 Worst Stuck Rod (N-I)

Less 10% 6.25 6.39 6.13 Control Rod Requirements, % Ap Reactivity Defects 3.60 3.60 3.59 Rod Insertion Allowance 0.39 0.37 0.37 Total Requirements, % Ap 3.99 3.97 3.96 Shutdown Margin, % Ap 2.26 2.42 2.17 Safety Analysis Limit, % Ap 1.30 1.30 1.30 Sections 2.3.2 and 2.3.3 and Figures 2 and 3 of the COLR provide the required shutdown margin limits as a function Of RCS critical boron concentration for Modes 3, 4, and 5. These figures are based on the shutdown margin required for the steam line break event from hot zero power (HZP) described in UFSAR Section 15.1.5 and the chemical and volume control system (CVCS) malfunction that results in a decrease in boron concentration in the reactor coolant system (RCS) described in UFSAR Section 15.4.6, respectively. By maintaining a shutdown margin of greater than 1.3% Ap, the steam line break event remains bounding. As discussed above, the removal of Control Rod D-6 does not result in exceeding the limit of 1.3% Ap. A key parameter for the CVCS malfunction event is shutdown margin. An evaluation of the effect on shutdown margin with Control Rod D-6 removed and the highest worth RCCA stuck out shows that the shutdown margin limits presented in Figures 2 and 3 of the COLR remain bounding.

Operationally, the required RCS shutdown margin boron concentrations for Modes 3, 4, and 5 will be higher with Control Rod D-6 removed in order to meet the COLR shutdown margin limits.

Table 2 below provides the minimum required shutdown boron concentration with all rods in (ARI) minus the most reactive stuck rod for beginning of cycle (BOC), middle of cycle (MOC) and end of cycle (EOC) conditions.

Enclosure NOC-AE- 15003315 Page 8 of 22 Table 2 Minimum Required Shutdown Boron Concentration with ARI Minus the Most Reactive Stuck Rod RCCA in D-6, ppm No RCCA in D-6, ppm 0

68 F [350 F 1567 °F 0

68 °F 350 °F f567 °F

[

BOC 1756 1711 1462 1767 1713 1469 MOC 1451 1322 938 1504 1379 966 EOC 718 505 48 750 559 63 TS 3.3.1, Table 3.3-1, Action 5, defines the action that must be taken when two extended range neutron flux monitors (Functional Unit 7 in Table 3.3-1) are not OPERABLE in Modes 3, 4, and 5; the analysis supporting this TS is discussed in UFSAR Section 15.4.6.2. To ensure these actions maintain the required shutdown margin requirements, the ratio of the minimum required shutdown margin boron concentration and the critical boron concentration with ARI minus the most reactive rod was evaluated to ensure that the value of 1.14 assumed in the safety analysis remains bounding. The results of this evaluation show that the ratio with Control Rod 0-6 installed is 1.179 and the value with Control Rod 0-6 removed is 1.178; both values are greater than the limit of 1.14.

Boron Worth The removal of Control Rod D-6 was also evaluated for impact on differential boron worth as a function of boron concentration in the ARI configuration. The removal of Control Rod D-6 increases the boron worth as function of boron concentration when all RCCAs are inserted into the core. This only impacts the CVCS malfunction that results in a decrease in boron concentration in the RCS described in UFSAR Section 15.4.6 for Modes 3, 4, and 5. The determination of the boron worth as a function of boron concentration is performed by conservatively assuming all RCCAs are out of the core. This provides the greatest boron worth as a function of boron concentration, which results in the greatest re'activity insertion for this event. Based on this conservative assumption, removal of Control Rod 0-6 has no impact on the boron worth as a function of boron concentration assumed in the analysis for this event.

Therefore, removal of Control Rod D-6 does not impact the results of the CVCS malfunction event described in UFSAR Section 15.4.6.

Rod Worth The removal of Control Rod D-6 has the effect of slightly increasing the rod worth of the adjacent RCCAs when all RCCAs are inserted. The worth of the most reactive stuck rod in an N-i configuration when considering Control Rod 0-6 inserted is 0.973 %Ap at core location F-8.

With Control Rod 0-6 not inserted, the worth of the most reactive stuck rod in an N-I configuration is 1 .071 %Ap at core location C-5.

The analysis for UFSAR events initiated from HZP, such as the spectrum of RCCA ejection accidents described in UFSAR Section 15.4.8 and uncontrolled RCCA bank withdrawal from a subcritical or low power startup conditions described in UFSAR Section 15.4.1, conservatively assumes the shutdown RCCAs are withdrawn from the core to maximize the reactivity insertion and power increase for these events. Therefore, since Control Rod 0-6 is in a shutdown bank,

Enclosure NOC-AE-1 5003315 Page 9 of 22 the removal of the control rod does not impact the results presented in the UFSAR for these events.

The removal of Control Rod D-6 will also impact the localized reactor core power distribution for events where a return to power with all control rods inserted can occur: specifically the steam line break event from zero power described in UFSAR Section 15.1.5. Section 5.3.14.4 of WCAP-9272 describes the reload core methodology for the steam line break event from zero power. 'The methodology first uses two-group three-dimensional neutronic calculations to determine ifthe reference transient analysis state points (reactor power level, inlet temperature, pressure, flow, and core boron concentration) reported in the RSAC remain bounding for the reload core. If the transient analysis state points are not bounding, the transient analysis is re-performed. A Departure from Nucleate Boiling (DNB) analysis is then performed using the power peaking factors for the reload core.

For the Unit 1 Cycle 20 core design with and without Control Rod D-6 removed, the transient state points as reported in the RSAC were found to be bounding. The neutronic analysis of the Unit 1 Cycle 20 evaluated the impact of (1) Control Rod D-6 inserted with the most reactive Control Rod (F-8) failing to insert into the core, and (2) Control Rod D-6 removed with the most reactive control rod (C-5) failing to insert into the core. A DNB analysis was then performed for both cases. The results of the analysis show that the DNB ratio was reduced from 3.011 for the analysis with Control Rod D-6 inserted to 1.811 with Control Rod 0-6 removed, which is still above the limit of 1.495. Therefore, the results of the steam line break from zero power with Control Rod 0-6 removed remains bounding.

Trip Reactivity The removal of Control Rod 0-6 reduces the trip reactivity as a function of rod insertion position, which reduces the trip reactivity as a function of time after the RCCAs begin to fall. The normalized trip reactivity as a function of RCCA insertion position and normalized trip reactivity as a function of time after the RCCAs begin to fall is presented in .UFSARFigures 15.0-4 and 15.0-5, respectively. An evaluation of the effects of the removal of Control Rod 0-6 shows that the trip reactivity as a function of RCCA insertion position and the resulting trip reactivity as a function of time after the RCCAs begin to fall remains bounding. Therefore, the removal of Control Rod 0-6 does not impact the trip reactivity assumed in UFSAR Chapter 15 events.

Table 3 below provides a comparison of the trip reactivity as a function of rod position for Unit 1 Cycle 20 with and without Control Rod 0-6 inserted.

Enclosure NOC-AE-1 5003315 Page 10 of 22 Table 3 Trip Reactivity Values Rod Position Unit 1 Cycle 20 Unit 1 Cycle 20 (Insertion fraction) RCCA in D-6 (%Ap) No RCCA in D-6 (% Ap) Limit (%Ap) 0.00 0.0000 0.0000 0.0000 0.03 0.0172 0.0172 0.0008 0.06 0.0375 0.0373 0.0045 0.10 0.0573 0.0573 0.0115 0.15 0.0735 0.0734 0.0216 0.20 0.0863 0.0862 0.0316 0.25 0.0999 0.0998 0.0481 0.30 0.1158 0.1157 0.0646 0.50 0.2415 0.2413 0.1350 0.70 0.7267 0.7250 0.4000 0.90 3.9600 3.9101 2.2000 Most Positive Moderator Density Coefficient The removal of Control Rod D-6 slightly impacts the most positive moderator density coefficient because this parameter is conservatively calculated assuming all RCCAs are inserted into the core. Figure 15.0-6 of the UFSAR presents the limit of 0.54 Aklgm/cc for this parameter. The value for this parameter is assumed in the following UFSAR Chapter 15 events:

Feedwater system malfunction causing an increase in feedwater flow (UFSAR Section 15.1.2);

Feedwater system pipe break (UFSAR Section 15.2.8);

Uncontrolled rod cluster bank withdrawal at power (UFSAR Section 15.4.2);

Startup of an inactive reactor coolant loop at an incorrect temperature (UFSAR Section 15.4.4); and

OVCS malfunction that increases RCS inventory (UFSAR Section 15.5.2).

The results of the evaluation for the most positive moderator density coefficient show that the limit assumed in the safety analysis remains bounding. The moderator density coefficient with Control Rod 0-6 installed is 0.3911 and the value with Control Rod D-6 removed is 0.3892; both values are less than the limit of 0.54. Therefore, the removal of Control Rod D-6 does not impact the results presented in the UFSAR for the above listed events.

Summary To summarize, the impact of the removal of Control Rod D-6 on the nuclear design and UFSAR Chapter 15 events has been evaluated using the NRC-approved methods described in TS 6.9.1.6. The results of the evaluation show that the power distribution within the reactor core during operation is not impacted. The effects on the available shutdown margin; boron worth when all RCCAs are inserted; rod worth of the adjacent RCCAs when all RCCAs are inserted; the trip reactivity as a function of time; and most positive moderator density coefficient have been evaluated and remain bounded by the safety analysis presented in UFSAR Chapter 15.

Enclosure N OC-AE- 15003315 Page 11 of 22 Therefore, the removal of Control Rod D-6 does not impact the results presented in UFSAR Chapter 15. Table 4 presents a summary of the impact of removal of Control Rod D-6 on each Chapter 15 accident.

Table 4 Impact on UFSAR Chapter 15 Accident Analysis

UFSAR Description Comments 1 15.1.1 Feedwater System Malfunctions Causing a Bounded by 15.1.2.

Reduction in Feedwater Temperature Most positive moderator density coefficient 2 15.1.2 Feedwater System Malfunctions Causing antrpecivyrmisbodngAl ann Fedwaer Icrese lowother analysis parameters not impacted.

3Excessive Increase in Secondary Steam No impact. Core DNB limits are not 1513 Flow challenged by this event.

Inadvertent Opening of a Steam Generator Bounded by 15.1.5 4 15.1.4 Relief or Safety Valve Causing a Depressurization of the Main Steam System 5 15.1.5 Spectrum of Steam System Piping Failures Shutdown margin remains bounding. All Inside and Outside Containment other analysis parameters not impacted.

Steam Pressure Regulator Malfunction or Not applicable for STP 6 15.2.1 Failure that Results in Decreasing Steam Flow 7 15.2.2 Loss of External Electrical Load Bounded by 15.2.3 8 15.2.3 Turbine Trip Trip reactivity remains bounding. All other 8 analysis parameters not impacted.

Inadvertent Closure of Main Steam Bounded by 15.2.5 9 15.2.4 Isolation Valves Loss of Condenser Vacuum and Other Bounded by 15.2.3.

10 1.2.5 Events Causing Turbine Trip Loss of Non-Emergency AC Power to the Bounded by 15.3.2 for DNB and 15.2.7 for 11 5..6 Plant Auxiliaries (Loss-of-Offsite-Power) pressurizer overfill 12 15.2.7 Loss of Normal Feedwater Flow Analysis parameters not impacted Most positive moderator density coefficient 13 15.2.8 Feedwater System Pipe Break remains bounding. All other analysis parameters not impacted.

14 1.31 Partial Loss of Forced Reactor Coolant Bounded by 15.3.2

_______Flow 15 15.3.2 Complete Loss of Forced Reactor Coolant Trip reactivity remains bounding. All other Flow analysis parameters not impacted.

1 1533 Reactor Coolant Pump Shaft Seizure Trip reactivity remains bounding. All other (Locked Rotor) analysis parameters not impacted.

17 15.3.4 Reactor Coolant Pump Shaft Break Bounded by 15.3.3 Trip reactivity remains bounding. All other Uncontrolled Rod Cluster Control Assembly analysis parameters not impacted. Control 18 15.4.1 Bank Withdrawal from a Subcritidal or Low- Rod D-6 is in a shutdown bank which is Power Startup Condition assumed withdrawn from the core for this

______________________________analysis.

Enclosure NOC-AE- 15003315 Page 12 of 22 Table 4 Impact on UFSAR Chapter 15 Accident Analysis

UFSAR Description Comments Trip reactivity remains bounding. Most positive moderator density coefficient remains bounding. All other analysis Uncontrolled Rod Cluster Control Assembly prameterdos not impacted Cotrol RoduD-6 19 15.4.2 Bank Withdrawal at Powerreoadesntipcthasu d reactivity rate because it is in a shutdown bank which is withdrawn from the core at power conditions.

Trip reactivity and other analysis parameters 2 1543 Rod Cluster Control Assembly Mis- remain bounding. The change in reactivity operation (Dropped Rod) insertion due to the removal of Control Rod D-6 is bounded by the current analysis.

Trip reactivity remains bounding. Most 21 15.4A4 Startup of an Inactive Reactor Coolant positive moderator density coefficient Loop at an Incorrect Temperature remains bounding. All other analysis parameters not impacted.

Chemical and Volume Control System Required shutdown margin limits remain 1546 Malfunction that Results in a Decrease in bounding. Boron worth remains bounding.

22 Boron Concentration in the Reactor Coolant All other analysis parameters not impacted.

(Boron Dilution) ___________________

Inadertnt oadig o a uel ssebly No impact. Inadvertent loading is detected 23 15.4.7 Indetn odn faFe seby using incore instrumentation when the intoanosiionshutdown mprper banks are withdrawn.

Analysis parameters not impacted. The core with Control Rod D-6 removed would remain subcritical with the ejected rod and an 24Spctrm 1.4. ofRodCluterContol sseblyadditional RCCA with the highest worth not 24Spctru 1.4. ofRodCluser ontol Asemly inserted into the core. Shutdown RCCAs are Ejection Accidents withdrawn from the core to maximize the power increase due to these events therefore the removal of Control Rod D-6 has no impact.

Inadvertent Operation of the Emergency No impact. The UFSAR concludes "Spurious 25 15.5.1 Core Cooling System During Power SI without immediate reactor trip has no Operation effect on the RCS."

Chemical and Volume Control System Trip reactivity remains bounding. Most 26 15.5.2 Malfunction that Increases Reactor Coolant positive moderator density coefficient Inventory remains bounding. All other analysis parameters not impacted.

Inadvertent Opening of a Pressurizer Trip reactivity remains bounding. All other 2 1561 Safety or Relief Valve analysis parameters not impacted.

1562 Failure of Small Lines Carrying Primary Analysis parameters not impacted.

28 Coolant Outside Containment 29Stam 1.6.GnertorTubeRupureTrip reactivity remains bounding. All other 29Stam1.6.GnertorTubeRupureanalysis parameters not impacted.

Enclosure NOC-AE-1 5003315 Page 13 of 22 Table 4 Impact on UFSAR Chapter 15 Accident Analysis

UFSAR Description Comments Loss-of-Coolant Accident (Large and Small) Analysis parameters not impacted. Impact of 30 15.6.5 Resulting from a Spectrum of Postulated change in RCS liquid volume is negligible.

Piping Breaks Within the Reactor Coolant Pressure Boundary Radioactive Release From a Subsystem or Analysis parameters not impacted.

31 15.7 Component Trip reactivity remains bounding. All other 32 15.8 ATWS analysis parameters not impacted. All rods out MTC is not impacted by the removal of

_______ _____________________________ Control Rod D-6.

Note: UFSAR sections 15.4.5, A Malfunction or Failure of the Flow Controllerin a BWR Loop That Results in an Increased Reactor CoolantFlow Rate; and 15.6.4, Radiological Consequences of Main Steam Line FailureOutside Containment (BWR) apply to boiling water reactors and are not applicable to STP.

3.4. Field work required to remove Control Rod D-6 from service Control Rod D-6 will be removed from service by performing the following work items:

Unlatch the control rod drive shaft from the RCCA and CRDM and completely remove the drive shaft from the reactor vessel;

Install a flow restrictor on the top of the control rod guide tube to maintain proper reactor coolant flow in the upper internals;

Remove RCCA from the fuel assembly located in core location D-6;

Install a thimble plug in the fuel assembly located in core location D-6 to 'maintain proper

reactor coolant flow through the fuel assembly;

Remove D-6 display card from the Digital Rod Position Indication display;

Modify plant computer point for 0-6; and

Remove rod control system fuses for control power to the 0-6 CRDM.

The D-6 display card and the rod control system fuses are located in areas outside of the reactor containment building. The other components listed above are inside the RCS pressure boundary and the RCS pressure boundary itself is not affected. These changes have been reviewed and approved by STPNOC Engineering using site procedures for design changes and for reload safety evaluations. The original equipment manufacturer also reviewed these changes for acceptability.

3.5. Evaluation of potential desigqn impacts Thermal-hydraulic impacts In order to provide ROS flow characteristics through the 0-6 control rod upper guide tube that are equivalent to the original configuration (i.e., with a control rod installed), a flow restrictor will be installed at the top of the guide tube housing. The flow restrictor assembly is a passive device and has been used successfully if* other plants (e.g., Beaver Valley Units 1 and 2, DC Cook Units 1 and 2, Farley Unit 2, Salem Units 1 and 2) after a drive shaft has been removed from a guide tube location for part-length control rod deletion. The flow restrictor

Enclosure NOC-AE-1 5Q0331 5 Page 14 of 22 assembly is a readily-available part that is designed so that flow entering or exiting the upper plenum will be essentially unchanged compared to that of the original guide tube housing plate design with the control rod drive shaft in place. The flow restrictor for the guide tube hole is hydraulically equivalent to the previous RCS flow configuration.

A thimble plug assembly will be installed on the fuel assembly in core location D-6. The fuel assembly flow characteristics with the thimble plug installed are hydraulically equivalent to the fuel assembly with an RCCA installed. As stated in UFSAR Section 4.2.1.6, the thimble plug assembly will:

Accommodate the differential thermal expansion between the fuel assembly and core internals;

Maintain positive contact with the fuel assembly and the core internals; and

Limit the flow through each occupied thimble to acceptable design values.

The thermal-hydraulic reactor internal vessel evaluation is not impacted by removal of the control rod drive shaft and RCCA as long as the flow restrictor at the top of the guide tube housing and a thimble plug in the fuel assembly are installed. The core bypass flow will remain unaffected by the installation of the flow restrictor. The hydraulic equivalence between the D-6 upper guide tube with a control rod drive shaft installed versus a flow restrictor installed ensures that there will be no impact on rod drop times at other core locations and the current TS 3.1.3.4 rod drop time limits will continue to be met.

Seismic and structural impacts The installed flow restrictor is structurally adequate and meets the allowable ASME Code stress limits. Materials for the flow restrictor parts are designed and fabricated to meet the intent of ASME Code Subsection NG. It does not need to conform to the ASME Boiler and Pressure Vessel Code requirements because it is not a core support structure. Material properties were taken from Section II, Part A of the Code. The flow restrictor is classified as ANSI Safety Class Ill. Structural analysis of the flow restrictor showed that all of the calculated stresses are within the ASME Code allowable limits.

There is no potential interference of an installed guide tube flow restrictor and a thermal sleeve at core location D-6 because the guide funnel has a larger diameter. There is no impact on the functionality and structural integrity of the reactor vessel upper internals from removal of the control rod drive shaft and RCCA, installation of the flow restrictor, and use of a thimble plug in the fuel assembly in core location 0-6. Therefore, there is no impact on the current reactor vessel internals analyses.

The dynamic analysis (seismic and loss of coolant accident, discussed in UFSAR Section 3.9.1.4.8) of the D-6 CROM was performed using the Reactor Equipment System Model (RESM). Review of the RESM shows that it remains valid after removal of the Control Rod D-6 drive shaft and RCCA. Removal of the control rod drive shaft reduces the overall weight of the CRDM; however, given the magnitude of the CRDM weight, the overall impact of weight reduction is negligible. Therefore, the CRDM dynamic stress evaluation due to seismic and loss of coolant accident excitations in the current CROM Design Report remains valid with removal of the Control Rod D-6 drive shaft and RCCA.

Removal of the Control Rod 0-6 drive shaft and RCCA has a negligible effect on thermosiphoning inside the CRDM housing during normal operating conditions (i.e., water circulation inside the CRDM due to temperature differential between the outside and inside of the CRDM housing). Additionally, thermal transients caused by rod up and down motion, which

Enclosure NOC-AE- 15003315 Page 15 of 22 dominate the thermal response of the CRDM, are eliminated. Since Control Rod D-6 is in a shutdown bank, the up and down motions would be minimal during operation at power.

Therefore, the thermal stress evaluations in the current analysis (CRDM Design Report) and UFSAR Section 3.9.5 remain valid after the removal of the control rod drive shaft and RCCA.

Reactor coolant system water volume impact The change in reactor vessel metal volume due to removal of RCCA D-6 is estimated to be:

+ 0.024 gal flow restrictor addition

+ 0.197 gal thimble plug addition

- 2.08 gal 'control rod drive shaft removal

- 1.87 gal rod cluster control assembly removal

- 3.73 gal net change in metal volume (inverse change in RCS volume)

The total ROS water volume is approximately 100,000 gal; an increase of < 4 gallons results in a change of less than 0.004%, which is negligible.

Reactor vessel mass impacts Removal of the control rod drive shaft and RCCA will reduce the overall weight of the reactor vessel. The control rod drive shaft and RCCA have a combined weight of approximately 300 pounds and the weight of the reactor vessel head is approximately 350,000 pounds. The impact of the weight reduction (less than 0.09%) on the current reactor equipment system model used in the seismic and loss of coolant accident analyses (see UFSAR Section 3.9.5.3) of the reactor internals is negligible.

3.6. Adequate level of safety The evaluations of the impact on the safety analyses has demonstrated that requirements for reactivity control provided by control rods continue to be met. Therefore, the assumption in the STPNOC Probabilistic Risk Assessment model that control rod insertion will provide sufficient negative reactivity to shut down the reactor remains valid. Required TS surveillances on the control rods (currently tested on a monthly basis in accordance with TS Surveillance 4.1.3.1.2) will provide assurance that there are no common cause failures associated with removal of Control Rod D-6.

There will be a small reduction in the available shutdown margin; however, shutdown margin will be maintained within the limits provided in the COLR as required by TS 3.1.1.1. As shown in Table 1 (see Section 3.3), shutdown margin is maintained with substantial margin to the limit (0.87% Ap). The UFSAR Chapter 15 safety analyses remain bounding, thus providing assurance that integrity of the three fission product barriers (fuel cladding, RCS, and reactor containment) is maintained. Compliance with the TS will provide reasonable assurance that the proposed change does not endanger the health and safety of the public The proposed amendment has no impact to electrical grid stability.

Enclosure NOC-AE-1 5003315 Page 16 of 22 4.0 Regulatory evaluation 4.1. Applicable regulatory requirements/criteria Technical Specification 5.3.2, Control Rod Assemblies The Control Rod Assemblies section of TS is a Design Feature which is required per 10 CFR 50.36(c)(4). The proposed change does not eliminate the design feature (which is the full-length control rod assemblies); only the required number of control rod assemblies is changed. As outlined in the Technical Evaluation above, all safety analysis limits are met and the Unit 1 Cycle 20 core has been evaluated per the methodologies prescribed in TS 6.9.1.6.

10 CFR 50, Appendix A, General Design Criteria Criterion 4--Environmentaland dynamic effects design bases. Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associatedwith normal operation, maintenance, testing, and postulatedaccidents, including loss-of-coolant accidents. These structures, systems, and components shall be appropriatelyprotected against dynamic effects, including the effects of missiles, pipe whipping, and dischargingfluids, that may result from equipment failures and from events and conditions outside the nuclearpower unit. However, dynamic effects associatedwith postulatedpipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probabilityof fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.

The removal of Control Rod D-6 from the reactor vessel does not impact the response to accidents involving missiles or pipe breaks since the reactor shutdown function of the control rod system remains acceptable with 56 control rods. The design of the reactor vessel, reactor internals, and fuel assemblies to withstand the effects of missiles and pipe breaks is not impacted by the removal of Control Rod 0-6 since there is negligible impact to the thermal hydraulics of the reactor vessel and the internals. Also, there is no adverse impact to the physical design of the reactor vessel and the internals due to the removal of the control rod drive shaft and RCCA, the installation of the thimble plug in the D-6 fuel asSembly, and the addition of the guide tube flow restrictor.

Thus, the requirements of General Design Criteria (GDC) 4 are met with respect to the design of the system against the adverse effects of missile hazards inside the containment, pipe whipping and jets caused by broken pipes, and adverse environmental conditions resulting from high- and moderate-energy pipe breaks during normal plant operations, anticipated operational occurrences, and accident conditions.

Criterion 10O--Reactor design. The reactorcore and associatedcoolant, control, and protection systems shall be designed with appropriatemargin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipatedoperationaloccurrences.

This criterion is satisfied because the removal of Control Rod 0-6 does not impact the at-power core power distribution, shutdown margin is maintained, and the design and safety limits for the UFSAR Chapter 15 accidents remain satisfied.

Enclosure NOC-AE- 15003315 Page 17 of 22 Criterion 11l-Reactorinherent protection. The reactorcore and associated coolant systems shall be designed so that in the power operating range the net effect of the prompt inherent nuclear feedback characteristicstends to compensate for a rapid increase in reactivity.

This criterion is satisfied because removal of Control Rod 0-6 does not impact the ability to detect or control core power distribution and the at-power nuclear reactivity feedback coefficients (e.g., Doppler or Moderator Temperature Coefficients) remain unchanged because the effect of Control Rod D-6 removal is similar to the original condition with Control Rod 0-6 fully withdrawn from the reactor core during power operations.

Criterion 12--Suppression of reactorpower oscillations. The reactorcore and associated coolant, control, and protection systems shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.

Since Control Rod 0-6 is located in a shutdown bank which is withdrawn at power, the removal of this control rod will not result in power oscillations which can result in conditions exceeding specified acceptable fuel design limits.

Criterion23--Protection system failure modes. The protection system shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air), or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced.

The removal of Control Rod 0-6 from the reactor vessel does not impact the fail-safe function of the remaining 56 control rods which will reliably maintain an adequate reactor protection system. The mechanical removal of the control rod drive shaft and RCCA and installation of the thimble plug in the fuel assembly and flow restrictor in the guide tube do not have any mechanical impact on the function of the remaining 56 control rods. The electrical removal from service of Control Rod 0-6 involves pulling fuses to remove control power to the respective stationary, lift, and movable coils. The remaining 56 control rods are not impacted by this electrical change and will continue to meet their design function. The modification design change review process ensures that the plant modifications involve only Control Rod 0-6 and do not affect other control rods.

Thus, the requirements of GDC 23 are met by maintaining the capability to insert the control rods upon failure of the drive mechanisms or induced failure by an outside force (e.g., loss of electric power, instrumentation air, fire, radiation, extreme heat, pressure, cold, water, and steam).

Criterion25--Protection system requirements for reactivity control malfunctions. The protection system shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.

A Unit 1 Cycle 20 redesign reload safety evaluation was performed using the NRC-approved methods referenced in TS 6.9.1.6 and it was confirmed that the fuel design limits are not exceeded. The protection system (reactor trip function) remains fully capable of performing its function with 56 control rods and fuel design limits are not exceeded for analyzed malfunctions of the STP reactivity control systems.

Thus, the requirements of GDC 25 are met by ensuring that no fuel design limits are exceeded for any single malfunction or rod withdrawal accident.

Enclosure NOC-AE-1 5003315 Page 18 of 22 Criterion 26--Reactivity control system redundancy and capability. Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipatedoperationaloccurrences, and with appropriatemargin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.

This criterion is satisfied because removal of Control Rod D-6 does not impact the ability of the reactivity control systems to reliably control reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as a single stuck rod, specified acceptable fuel design limits are not exceeded. Evaluations of Control Rod 0-6 removal for Unit 1 Cycle 20 demonstrate that shutdown margin and safety analysis limits are met throughout the fuel cycle. The reactiv/ity control systems (e.g., rod control, reactor trip, RCS boron addition) continue to perform their design and safety functions with removal of Control Rod D-6.

Thus, the requirement of GDC 26 is met by demonstrating the ability to control reactivity changes to ensure that, under normal operation and anticipated operational occurrences with the appropriate margin for malfunction (such as stuck rods), no fuel design limits are exceeded and the reactor can be maintained subcritical under cold conditions.

Criterion 27--Combined reactivity control systems capability. The reactivity control systems shall be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controllingreactivity changes to assure that under postulatedaccident conditions and with appropriatemargin for stuck rods the capability to cool the core is maintained.

This criterion is satisfied because the removal of Control Rod D-6 does not impact the ability of the reactivity control systems to reliably control reactivity changes and that adequate shutdown margin is maintained when considering highest stuck rod worth. Evaluations of Control Rod D-6 removal for Unit 1 Cycle 20 demonstrate that shutdown margin and safety analysis limits are met throughout the fuel cycle.

Thus, the requirements of GDC 27 are met by demonstrating the ability to reliably control reactivity changes under accident conditions to ensure that no fuel design limits are exceeded and the capability to cool the core is maintained.

Criterion28--Reactivity limits. The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactorcoolant pressure boundary greaterthan limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactorpressure vessel internals to impairsignificantly the capability to cool the core. These postulated reactivity accidents shall include considerationof rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition.

This criterion is satisfied because removal of Control Rod 0-6 for Unit 1 Cycle 20 has been evaluated to ensure trip reactivity insertion rate, shutdown margin, and the safety analysis limits

Enclosure NOC-AE-1 5003315 Page 19 of 22 remain met for the rod ejection, rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition events for the entire fuel cycle.

Thus, the requirements of GDC 28 are satisfied by demonstrating the ability to reliably control the amount and rate of reactivity change to ensure that no reactivity accident will damage the reactor coolant pressure boundary or disturb the core or the core appurtenances such as to impair coolant flow.

Criterion29--Protection againstanticipatedoperationaloccurrences. The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipatedoperationaloccurrences.

The removal of Control Rod D-6 from the reactor vessel does not impact the ability of the reactivity control systems to reliably control reactivity changes and perform their safety functions. The mechanical removal of the control rod drive shaft and the RCCA and the installation of the thimble plug in the fuel assembly in core location D-6 and the flow restrictor in the guide tube do not have any mechanical impact on the function of the remaining 56 control rods. The electrical removal from service of Control Rod D-6 involves pulling fuses to remove control power to the respective stationary, lift, and movable coils of this particular control rod.

The remaining 56 control rods are not impacted by this electrical change and will continue to meet their design and safety functions. The modification to remove Control Rod D-6 from service will not adversely impact the capability of the remaining 56 control rods to reliably meet their safety functions in the event of anticipated operational occurrences.

Thus, the requirements of GDC 29 are met by demonstrating a high probability of control rod insertion under anticipated operational occurrences.

Other The requirements of 10 CFR 50.62(c)(3) concerning an alternate rod injection system as stated in Standard Review Plan 4.6 are for boiling water reactors and do not apply to the STP design

which is a pressurized water reactor.

4.2. Precedence A similar request for a Technical Specification change to allow the removal of the center control rod assembly for Arkansas Nuclear One - Unit 1 (ANO-1) (NRC Legacy ADAMS accession number 8607280077) was approved by the NRC (Amendment 103) as documented in a letter from Guy S. Vissing (NRC) to Gene Campbell (ANO-1) dated November 14, 1986 (NRC Legacy ADAMS accession number unavailable). The STPNOC proposed license amendment is not for removal of a center control rod; however, the precedence does evaluate the impact on shutdown margin, reactor core thermal-hydraulics, and structural integrity of the RCS. (See References 6.52 and 6.53.)

4.3. No siginificant hazards consideration determination STP Nuclear Operating Company (STPNOC) is proposing an amendment to Unit I Technical Specification (TS) 5.3.2, Control Rod Assemblies, to require Unit 1 Cycle 20 to contain 56 full-length control rods with no full-length control rod in core location D-6. Currently, TS 5.3.2 requires the core to contain 57 full-length control rods.

STPNOC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below.

Enclosure N OC-AE- 15003315 Page 20 of 22

1) Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No. Removal of Control Rod D-6 for Unit 1 Cycle 20 will be performed using approved plant procedures. The change in the probability and consequences of accidents previously evaluated in the Updated Final Safety Analysis Report (UFSAR) has been evaluated and found not to be significant. An evaluation of the impact on the safety analysis shows that the current safety analysis remains bounding. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2) Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No. Removal of Control Rod D-6 for Unit 1 Cycle 20 does not create any new failure modes and the design function and operation of SSCs is not changed. No new operator actions are created. The modification to remove Control Rod D-6 ensures that Reactor Coolant System flowrate through the reactor vessel remains unchanged. Reactivity control and insertion characteristics continue to meet all design and safety functions and plant equipment will continue to meet applicable design and safety requirements. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3) Does the proposed change involve a significant reduction in.a margin of safety?

Response: No. Removal of Control Rod 0-6 for Unit 1 Cycle 20 does not exceed or alter a UFSAR design basis or safety limit. Therefore, the proposed change does not significantly reduce a margin of safety.

Based on the above, STPNOC concludes that the proposed amendment does not involve a significant hazards consideration under the Standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.4. Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0) Environmental consideration A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, Or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Enclosure NOC-AE- 15003315 Page 21 of 22 6.0 References 6.1. Technical Specification 3.1.1.1, Shutdown Margin.

6.2. Technical 'Specification 3.1 .3.1, Movable ControlAssemblies, Group Height.

6.3. Technical Specification 3.1.3.4, Rod Drop Time.

6.4. Technical Specification 3.1.3.5, Shutdown Rod Insertion Limit.

6.5. Technical Specification 3.3.1, Reactor Trip System Instrumentation.

6.6. Technical Specification 5.3.2, Control Rod Assemblies.

6.7. Technical Specification 6.9.1.6, Core Operating Limits Report (COLR).

6.8. UFSAR Figure 15.0-4, Minimum Trip Reactivity Versus Rod Position.

6.9. UFSAR Figure 15.0-5, Normalized RCCA Negative Reactivity Insertion Versus Time.

6.10. UFSAR Figure 15.0-6, Moderator Density Coefficient.

6.11. U FSAR Section 3.9.1.4.8, Evaluation of the Control Rod Drive Mechanisms.

6.12. UFSAR Section 3.9.5, Reactor Pressure Vessel Internals.

6.13. UFSAR Section 3.9.5.3, Reactor Pressure Vessel Internals, Design Loading Categories.

6.14. UFSAR Section 4.2.1.6, Incore Control Components.

6.15. UFSAR Section 4.2.2.3.1, Rod Cluster ControlAssembly.

6.16. U FSAR Section 4.3.2.4.12, Rod Cluster ControlAssemblies.

6.17. U FSAR Section 15.1.1, Feedwater System Malfunctions Causing a Reduction in Feedwater Temperature.

6.18. UFSAR Section 15.1.2, FeedwaterSystem Malfunctions Causing an Increase in FeedwaterFlow.

6.19. UFSAR Section 15.1.3, Excessive Increase in Secondary Steam Flow.

6.20. UFSAR Section 15.1.4, Inadvertent Opening of a Steam GeneratorRelief or Safety Valve Causing a Depressurizationof the Main Steam System.

6.21. UFSAR Section 15.1.5, Spectrum of Steam System Piping Failures Inside and Outside Containment.

6.22. UFSAR Section 15.2.1, Steam PressureRegulatorMalfunction or Failure that Results in DecreasingSteam Flow.

6.23. UFSAR Section 15.2.2, Loss, of External ElectricalLoad.

6.24. UFSAR Section 15.2.3, Turbine Trip.

6.25. UFSAR Section 15.2.4, Inadvertent Closure of Main Steam Isolation Valves.

6.26. UFSAR Section 15.2.5, Loss of Condenser Vacuum and Other Events Causing a Turbine Trip.

6.27. UFSAR Section 15.2.6, Loss of NonemergencyAC Power to the PlantAuxiliaries (Loss of Offsite Power).

6.28. UFSAR Section 15.2.7, Loss of Normal FeedwaterFlow.

6.29. UFSAR Section 15.2.8, FeedwaterSystem Pipe Break.

6.30. UFSAR Section 15.3.1, PartialLoss of ForcedReactor Coolant Flow.

6.31. UFSAR Section 15.3.2, Complete Loss of Forced Reactor Coolant Flow.

Enclosure NOC-AE- 15003315 Page 22 of 22 6.32. UFSAR Section 15.3.3, Reactor Coolant Pump Shaft Seizure (Locked Rotor).

6.33. UFSAR Section 15.3.4, Reactor Coolant Pump Shaft Break.

6.34. UFSAR Section-15.4.1, Uncontrolled Rod Cluster ControlAssembly Bank Withdrawal from a Subcritical or Low Power Startup Condition.

6.35. UFSAR Section 15.4.2, Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power.

6.36. UFSAR Section 15.4.3, Rod Cluster Control Assembly Misoperation.

6.37. UFSAR Section 15.4.4, Startup of an Inactive Reactor Coolant Loop at an Incorrect Temperature.

6.38. UFSAR Section 15.4.5, A Malfunction or Failure of the Flow Controllerin a BWR Loop That Results in an Increased Reactor Coolant Flow Rate.

6.39. UFSAR Section 15.4.6, Chemical and Volume Control System Malfunction That Results in a Decrease in Boron Concentration in the Reactor Coolant.

6.40. UFSAR Section 15.4.7, Inadvertent Loading of a FuelAssembly into an Improper Position.

6.41. UFSAR Section 15.4.8, Spectrum of Rod ClusterControlAssembly Ejection Accidents 6.42. UFSAR Section 15.5.1, Inadvertent Operation of ECCS During Power Operation.

6.43. UFSAR Section 15.5.2, Chemical and Volume Control System Malfunction that Increases Reactor Coolant lnventory.

6.44. U FSAR Section 15.6.1, Inadvertent Opening of a PressurizerSafety or Relief Valve.

6.45. UFSAR Section 15.6.2, Failureof Small Lines Carrying Primary Coolant Outside Containment.

6.46. UFSAR Section 15.6.3, Steam Generator Tube Rupture.

6.47. UFSAR Section 15.6.4, Radiological Consequences of Main Steam Line Failure Outside Containment (BWR).

6.48. UFSAR Section 15.6.5, Loss of Coolant Accidents.

6.49. UFSAR Section 15.7, Radioactive Release From a Subsystem or Component.

6.50. UFSAR Section 15.8, Anticipated Transients Without Scram.

6.51. Unit 1 Cycle 20 Core Operating Limits Report, Revision 0.

6.52. Letter from John F. Stolz (ANO - 1) to NRC; "Request for a Technical Specification Change to Allow the Removal of Center Control Rod Assembly" (1CAN078609); dated July 18, 1986, NRC Legacy ADAMS accession number 8607280077.

6.53. Letter from Guy S. Vissing (NRC) to Gene Campbell (ANO-1); Amendment No. 103 to Facility Operating License No. DPR-51 to allow the removal of the center control rod assembly from Arkansas Nuclear One; dated November 14, 1986; NRC Legacy ADAMS accession number unavailable.

6.54. WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985.

Attachment 1 Technical Specification Markup

DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 193 fuel assemblies. Each fuel assembly shall consist of a matrix of zircaloy, ZIRLOTM or Optimized ZIRLOTM Clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide as fuel material. Limited substitutions of zirconium alloy, ZIRLOTM or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff-approved codes and methods, and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.

fl[enadd superscript "*"-

CONTROL ROD ASSEMBLIES for a footnote 5.3.2 The core shall contain 57 fl-legth control rod assemblies. The full-length control rod assemblies shall contain a nominal 158.9 inches of absorber material. The absorber material within each assembly shall be silver-indium-cadmium or hafnium. Mixtures of hafnium and silver-indium-cadmium are not permitted within a bank. All control rods shall be clad with stainless steel tubing.

5.4 (NOT USED) 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological towers shall be located as shown on Figure 5.1-1.

5.6 FUEL STORAGE 5.6.1 CRITICALITY 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:

INSERT:

The Unit I Cycle 20 core shall contain 56 full-length control rod assemblies with no full-length control rod assembly installed in core location D-6.

SOUTH TEXAS - UNITS 1 & 2 5-6 Unit 1 - Amendment No. 2,!0,!.,43, Unit 2 - Amendment No. 2,.6,32,50 51,76,79,85, 91, 186

Attachment 2 Clean Technical Specification Page

DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3,1 The reactor core shall contain 193 fuel assemblies. Each fuel assembly shall consist of a matrix of zircaloy, ZIRLOTM or Optimized ZIRLOTM clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide as fuel material. Limited substitutions of zirconium alloy, ZIRLOTM or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff-approved codes and methods, and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.

CONTROL.ROD ASSEMBLIES 5.3.2 The core shall contain 57* full-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 158.9 inches of absorber material. The absorber material within each assembly shall be silver-indium-cadmium or hafnium. Mixtures of hafnium and silver-indium-cadmium are not permitted within a bank. All control rods shall be clad with stainless steel tubing.

5.4 (NOT USED) 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological towers shall be located as shown on Figure 5.1-1.

5.6 FUEL STORAGE 5.6.1 CRITICALITY 5.6.1 .1 The spent fuel storage racks are designed and shall be maintained with:

The Unit 1 Cycle 20 core shall contain 56 full-length control rod assemblies with no full-lengthI control rod assembly installed in core location D-6.

SOUTH TEXAS - UNITS 1 & 2 5-6 Unit 1 - Amendment No. 2,04,* ,*

61,5,9,9,9, 101 18,x.

Unit 2 - Amendment No. 2,*,T32,~50 51,76,79,85, 91, 186