Text

Transmittal of Quarterly Update for Hatch, Units 1 and 2 Technical Specifications Bases
	ML041670569
Person / Time
Site:	Hatch
Issue date:	06/02/2004
From:	Randy Baker; Southern Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation
References
	NL-04-0981
	Download: ML041670569 (95)
	v • d • e

Intracompany Correspondence SOUTHERN A COMPANY Ener to Serve YourWorid" DATE: June 2, 2004 File: N/A Log: NL-04-0981 RE: Plant Edwin I. Hatch Units 1 and 2 TS Bases Distribution FROM: R.D. Baker TO: Manual Holders Attached you will find the quarterly update for the Units 1 and 2 Technical Specifications Bases. Included are the page change instructions as well as the actual replacement pages.

If you have any questions, please contact Ozzie Vidal at extension 7301.

OCV Attachments: Page Change instructions and Replacement Pages cc: SNC Document Services - RType: Hatch=CHAO2.002

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Edwin I. Hatch Unit 1 & Unit 2 Bases Revision Insertion Instructions June 4, 2004 UNIT 1 BASES - REVISION 40 Page Instruction Effective Page List (i - vi) Replace B 3.1-5 / -6 Replace B 3.1-23/-24 Replace B 3.3-31 -32 Replace B 3.3-73 / 4 Replace B 3.3-83 -84 Replace B 3.5-19 /-20 Replace B 3.6-25 / -26 Replace B 3.6-45 -46 Replace B 3.6-75 /-76 Replace B 3.6-83 -84 Replace B 3.6-89B -90 Replace B 3.7-1 L-2 Replace B 3.7-5 /-6 Replace B 3.7-23 / -24 Replace B 3.7-29 / -30 Replace B 3.7-37 / -38 Replace B 3.8-61 / -62 Replace B 3.9-23 / -24 Replace B 3.9-27 / -28 Replace UNIT 2 BASES - REVISION 46 Page Instruction Effective Page List (i- vi) Replace B 3.1-5 /-6 Replace B 3.1-23 /-24 Replace B 3.3-31 / -32 Replace B 3.3-73 I -74 Replace B 3.3-83 / -84 Replace B 3.3-1 59 / -160 Replace B 3.5-13 / -14 Replace B 3.5-19 /-20 Replace B 3.6-27 / -28 Replace B 3.6-45 / -46 Replace B 3.6-8 1 I -82 Replace B 3.6-89 / -90 Replace B 3.6-95 / -96 Replace B 3.7-1 / -2 Replace

Edwin I. Hatch Unit 1 & Unit 2 Bases Revision Insertion Instructions (Continued)

UNIT 2 BASES - REVISION 46 (continued)

Page Instruction B 3.7-5 / -6 Replace B 3.7-23 / -24 Replace B 3.7-29 / -30 Replace B 3.7-37 / -38 Replace B 3.8-61 / -62 Replace B 3.9-23 / -24 Replace B 3.9-27 / -28 Replace

HATCH UNIT 1 BASES EFFECTIVE PAGE LIST Unless noted otherwise, all of the pages in the HNP Unit 1 Bases are effective per Revision 0.

PAGE REVISION PAGE REVISION 37 iii 1 B 3.3-15 14 1 B 3.3-16 36 iv 1 B 3.3-17 36 V 33 B 3.3-18 14 vi 24 B 3.3-19 16 B 2.0-2 36 B 3.3-20 16 B 2.0-3 22 B 3.3-21 16 B 2.0-4 22 B 3.3-22 16 B 3.0-11 24 B 3.3-23 16 B 3.0-12 24 B 3.3-24 36 B 3.0-13 24 B 3.3-25 16 B 3.0-14 24 B 3.3-26 26 B 3.1-3 1 B 3.3-27 31 B 3.1-4 1 B 3.3-28 36 B 3.1-5 1 B 3.3-29 31 B 3.1-6 40 B 3.3-30 31 B 3.1-17 1 B 3.3-31 40 B 3.1-18 1 B 3.3-32 31 B 3.1-24 40 B 3.3-33 14 K> B3.1-34 17 B 3.3-37 15 B 3.1-39 28 B 3.3-38 15 B 3.1-40 28 B 3.3-39 15 B 3.1-44 33 B 3.3-40 28 B 3.1-45 33 B 3.3-41 28 B 3.2-1 37 B 3.3-42 14 B 3.2-2 37 B 3.3-43 28 B 3.2-3 37 B 3.3-44 14 B 3.2-4 37 B 3.3-47 23 B 3.2-6 36 B 3.3-48 14 B 3.2-7 36 B 3.3-49 31 B 3.2-9 37 B 3.3-50 28 B 3.2-10 37 B 3.3-51 28 B 3.2-11 37 B 3.3-52 31 B 3.2-12 37 B 3.3-53 36 B 3.3-3 36 B 3.3-54 36 B 3.3-4 30 B 3.3-56 36 B 3.3-5 35 B 3.3-57 31 B 3.3-6 16 B 3.3-58 31 B 3.3-7 36 B 3.3-59 1 B 3.3-8 16 B 3.3-60 1 B 3.3-9 16 B 3.3-61 32 B 3.3-10 16 B 3.3-62 1 B 3.3-11 16 B 3.3-63 1 B 3.3-12 16 B 3.3-64 8 B 3.3-13 14 B 3.3-65 1 B 3.3-14 14 B 3.3-66 1 HATCH UNIT 1 i REVISION 40

HATCH UNIT 1 BASES EFFECTIVE PAGE LIST (continued)

PAGE REVISION PAGE REVISION B 3.3-67 1 B 3.3-116 B 3.3-68 1 28 B 3.3-117 B 3.3-69 28 B 3.3-118 1 B 3.3-70 1 B 3.3-119 B 3.3-71 1 1 B 3.3-120 1I B 3.3-72 1 B 3.3-121 B 3.3-73 1 B 3.3-122 28 B 3.3-74 40 B 3.3-123 31 B 3.3-75 1 B 3.3-124 31 B 3.3-76 36 B 3.3-125 B 3.3-77 1 36 B 3.3-126 B 3.3-78 7 36 B 3.3-127 B 3.3-79 22 1 B 3.3-128 B 3.3-80 1 36 B 3.3-129 B 3.3-81 1 36 B 3.3-130 B 3.3-82 1 28 B 3.3-131 1 B 3.3-83 40 B 3.3-132 B 3.3-84 1 3 B 3.3-133 B 3.3-85 31 1 B 3.3-134 B 3.3-86 31 22 B 3.3-135 B 3.3-87 1 22 B 3.3-136 B 3.3-88 1 1 B 3.3-137 1 B 3.3-89 1 B 3.3-138 B 3.3-90 1 31 B 3.3-139 B 3.3-91 1 31 B 3.3-140 1 B 3.3-92 1 B 3.3-141 B 3.3-93 36 1 B 3.3-142 36 B 3.3-94 1 B 3.3-143 B 3.3-95 1 1 B 3.3-144 1 B 3.3-96 1 B 3.3-145 B 3.3-97 1 1 B 3.3-146 6 B 3.3-98 1 B 3.3-147 B 3.3-99 1 1 B 3.3-148 1 B 3.3-1 00 1 B 3.3-149 1 B 3.3-1 01 1 B 3.3-150 1 B 3.3-102 1 B 3.3-151 1 B 3.3-103 1 B 3.3-152 22 B 3.3-104 22 B 3.3-153 22 B 3.3-105 22 B 3.3-154 1 B 3.3-106 1 B 3.3-155 1 B 3.3-107 28 B 3.3-156 1 B 3.3-108 1 B 3.3-157 1 B 3.3-109 1 B 3.3-158 28 B 3.3-110 1 B 3.3-159 31 B 3.3-111 15 B 3.3-160 31 B 3.3-112 1 B 3.3-161 1 B 3.3-113 8 B 3.3-162 1 B 3.3-114 8 B 3.3-163 1 B 3.3-115 1 B 3.3-164 1 HATCH UNIT 1 Hl REVISION 40

HATCH UNIT 1 BASES EFFECTIVE PAGE LIST (continued)

PAGE REVISION PAGE REVISION B 3.3-165 1 B 3.4-28 15 B 3.3-166 1 B 3.4-29 15 B 3.3-167 1 B 3.4-30 15 B 3.3-168 31 B 3.4-31 15 B 3.3-169 31 B 3.4-32 15 B 3.3-170 31 B 3.4-33 15 B 3.3-171 1 B 3.4-34 B 3.3-172 1 1 B 3.4-35 1 B 3.3-173 1 B 3.4-36 B 3.3-174 1 1 B 3.4-37 1 B 3.3-175 13 B 3.4-38 B 3.3-176 1 1 B 3.4-39 12 B 3.3-177 31 B 3.4-40 B 3.3-178 12 31 B 3.4-41 12 B 3.3-179 1 B 3.4-42 B 3.3-180 12 1 B 3.4-43 12 B 3.3-181 B 3.4-44 12 B 3.3-182 1 B 3.4-45 23 B 3.3-183 1 B 3.4-46 12 B 3.3-184 1 B 3.4-47 12 B 3.3-185 B 3.4-48 23 B 3.3-186 11 B 3.4-49 23 B 3.3-187 1 B 3.4-50 23 B 3.3-188 1 B 3.5-2 16 B 3.3-189 I B 3.5-4 13 B 3.3-190 1 B 3.5-6 15 B 3.3-191 1 B 3.5-7 13 B 3.3-192 1 B 3.5-8 1 13 B 3.3-193 B 3.5-9 16 B 3.3-194 1 B 3.5-10 1 B 3.3-195 1 B 3.5-11 1 28 B 3.3-196 B 3.5-12 28 B 3.3-197 1 B 3.5-13 28 B 3.3-198 28 B 3.5-14 28 B 3.3-199 28 B 3.5-16 22 B 3.4-2 37 B 3.5-17 1 B 3.4-3 37 B 3.5-18 10 B 3.4-4 37 B 3.5-19 1 B 3.4-5 37 B 3.5-20 40 B 3.4-10 36 B 3.5-21 24 B 3.4-11 13 B 3.5-22 13 B 3.4-12 28 B 3.5-26 28 B 3.4-18 15 B 3.5-27 28 B 3.4-19 15 B 3.6-1 5 B 3.4-20 15 B 3.6-2 16 B 3.4-21 15 B 3.6-3 5 B 3.4-22 15 B 3.6-4 28 B 3.4-23 28 B 3.6-5 28 HATCH UNIT 1 ..

REVISION 40

HATCH UNIT 1 BASES EFFECTIVE PAGE LIST (continued)

PAGE REVISION PAGE REVISION B 3.6-7 16 B 3.6-63 1 B 3.6-11 5 B 3.6-64 B 3.6-12 1

5 B 3.6-65 B 3.6-17 5 B 3.6-66 1 B 3.6-18 B 3.6-67 1 B 3.6-19 B 3.6-68 32 B 3.6-20 B 3.6-69 1 1 1 B 3.6-21 1I B 3.6-70 B 3.6-22 B 3.6-71 1 B 3.6-23 1 B 3.6-72 1 B 3.6-24 28 B 3.6-73 1 B 3.6-25 28 B 3.6-74 4 B 3.6-26 40 B 3.6-75 1 B 3.6-27 16 B 3.6-76 34 B 3.6-28 40 B 3.6-77 40 B 3.6-29 B 3.6-78 1 9 1 B 3.6-30 9 B 3.6-79 B 3.6-31 1

9 B 3.6-80 1 B 3.6-32 I13 B 3.6-81 B 3.6-33 1

13 B 3.6-82 B 3.6-34 B 3.6-83 28 28 40 B 3.6-35 1 B 3.6-84 B 3.6-36 B 3.6-85 4 1 4 B 3.6-37 1 B 3.6-86 1 B 3.6-38 1 B 3.6-87 1 B 3.6-39 28 B 3.6-88 1 B 3.6-40 i 28 B 3.6-89 B 3.6-41 1 B 3.6-90 40 B 3.6-42 1 B 3.7-2 40 B 3.6-43 1 B 3.7-6 40 B 3.6-44 1 B 3.7-13 28 B 3.6-45 1 B 3.7-16 28 B 3.6-46 40 B 3.7-17 1 B 3.6-47 1 B 3.7-18 23 B 3.6-48 1 B 3.7-19 23 B 3.6-49 1 B 3.7-20 23 B 3.6-50 1 B 3.7-21 23 B 3.6-51 11 B 3.7-22 23 B 3.6-52 1 B 3.7-23 28 B 3.6-53 11 B 3.7-24 40 B 3.6-54 B 3.7-25 1 B 3.6-55 i 1 B 3.7-26 39 B 3.6-56 15 B 3.7-27 39 B 3.6-57 15 B 3.7-28 39 B 3.6-58 15 B 3.7-29 39 B 3.6-59 15 B 3.7-30 40 15 B 3.6-60 15 B 3.7-31 36 B 3.6-61 15 B 3.7-32 16 B 3.6-62 B 3.7-33 8 HATCH UNIT 1 iv REVISION 40

HATCH UNIT 1 BASES EFFECTIVE PAGE LIST (continued)

PAGE REVISION PAGE REVISION B 3.7-34 16 B 3.8-43 33 B 3.7-35 36 B 3.8-44 33 B 3.7-36 36 B 3.8-45 33 B 3.7-37 40 B 3.8-46 33 B 3.7-38 22 B 3.8-47 33 B 3.7-39 1 B 3.8-48 33 B 3.7-40 1 B 3.8-49 33 B 3.8-1 1 B 3.8-50 33 B 3.8-2 1 B 3.8-51 33 B 3.8-3 16 B 3.8-52 33 B 3.8-4 16 B 3.8-53 33 B 3.8-5 16 B 3.8-54 33 B 3.8-6 16 B 3.8-55 33 B 3.8-7 27 B 3.8-56 33 B 3.8-8 16 B 3.8-57 33 B 3.8-9 38 B 3.8-58 33 B 3.8-10 38 B 3.8-59 33 B 3.8-11 33 B 3.8-60 33 B 3.8-12 33 B 3.8-61 33 B3.8-13 33 B3.8-62 40 B3.8-14 33 B3.8-63 33 B3.8-15 .33 B 3.8-64

33 B 3.8-16 33 B 3.8-65 33 B 3.8-17 33 B 3.8-66 33 B 3.8-18 33 B 3.8-67 33 B 3.8-19 33 B 3.8-68 33 B 3.8-20 33 . B 3.8-69 33 B 3.8-21 33 B 3.8-70 33 B 3.8-22 33 B 3.8-71 33 B 3.8-23 33 B 3.8-72 33 B 3.8-24 33 B 3.8-73 33 B 3.8-25 33 B 3.8-74 33 B 3.8-26 33 B 3.8-75 33 B 3.8-27 33 B 3.8-76 33 B 3.8-28 33 B 3.8-77 33 B 3.8-29 33 B 3.8-78 33 B 3.8-30 33 B 3.8-79 33 B 3.8-31 33 B 3.8-80 33 B 3.8-32 33 B 3.8-81 33 B 3.8-33 33 B 3.8-82 33 B 3.8-34 33 B 3.8-83 33 B 3.8-35 33 B 3.8-84 33 B 3.8-36 33 B 3.8-85 33 B 3.8-37 33 B 3.9-7 23 B 3.8-38 33 B 3.9-16 22 B 3.8-39 33 B 3.9-17 22 B 3.8-40 33 B 3.9-19 22 B 3.8-41 33 B 3.9-20 1 B 3.8-42 33 B 3.9-21 1 HATCH UNIT 1 v REVISION 40

HATCH UNIT 1 BASES EFFECTIVE PAGE LIST (continued)

PAGE REVISION PAGE REVISION B 3.9-22 1 B 3.9-23 40 B 3.9-24 1 B 3.9-25 1 B 3.9-26 1 B 3.9-27 1 B 3.9-28 40 B 3.10-1 6 B 3.10-15 1 B 3.10-23 1 B 3.10-24 1 B 3.10-27 1 B 3.10-28 B 3.10-30 14 B 3.10-31 14 B 3.10-32 14 B 3.10-33 14 HATCH UNIT 1 vi REVISION 40

SDM B 3.1.1 BASES ACTIONS E.1. E.2, E.3. E.4. and E.5 (continued) information, to determine if the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, SRs may need to be performed to restore the component to OPERABLE status.

Action must continue until all required components are OPERABLE.

SURVEILLANCE SR 3.1.1.1 REQUIREMENTS Adequate SDM must be verified to ensure that the reactor can be made subcritical from any initial operating condition. This can be accomplished via a test, an evaluation, or a combination of the two.

Adequate SDM is demonstrated by testing before or during the first startup after fuel movement or shuffling within the reactor pressure vessel, or control rod replacement. Control rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod or a control rod from another core location. Since core reactivity will vary during the cycle as a function of fuel depletion and poison bumup, the beginning of cycle (BOC) test must also account for changes in core reactivity during the cycle. Therefore, to obtain the SDM, the initial value must be changed by the value, WR', which is the difference between the calculated value of minimum SDM during the operating cycle and the calculated BOC SDM. If the value of R is positive (that is, BOC is the point in the cycle with the minimum SDM), no correction to the BOC measured value is required (Ref. 7). For the SDM demonstrations where the highest worth rod is determined solely on calculation, additional margin (0.10% Ak/k) must be added to the SDM limit of 0.28% Ak/k to account for uncertainties in the calculation of the highest worth control rod.

The SDM may be demonstrated during an in-sequence control rod withdrawal, in which the highest worth control rod is analytically determined, or during local criticals, where the highest worth control rod is determined by testing. Local critical tests require the withdrawal of out of sequence control rods. This testing would therefore require bypassing of the Rod Worth Minimizer to allow the out of sequence withdrawal, and therefore additional requirements must be met (see LCO 3.10.7, "Control Rod Testing - Operating").

(continued)

HATCH UNIT 1 B 3.1 -5 REVISION 1 l

-~

I SDM B 3.1.1 BASES SURVEILLANCE SR 3.1.1.1 (continued)

REQUIREMENTS The Frequency of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reaching criticality is allowed to provide a reasonable amount of time to perform the required calculations and have appropriate verification.

During MODE 5, adequate SDM is required to ensure that the reactor does not reach criticality during control rod withdrawals. An evaluation of each in-vessel fuel movement during fuel loading (including shuffling fuel within the core) is required to ensure adequate SDM is maintained during refueling. This evaluation ensures that the intermediate loading patterns are bounded by the safety analyses for the final core loading pattern. For example, bounding analyses that demonstrate adequate SDM for the most reactive configurations during the refueling may be performed to demonstrate acceptability of the entire fuel movement sequence. These bounding analyses include additional margins to the SDM limit to account for the associated uncertainties. Spiral offload/reload sequences inherently satisfy the SR, provided the fuel assemblies are reloaded in the same configuration analyzed for the new cycle. Removing fuel from the core will always result in an increase in SDM.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.

2. FSAR, Subsection 14.4.2.

3. NEDE-2401 1-P-A-US, "General Electric Standard Application for Reactor Fuel,' Supplement for United States, (revision specified in the COLR).

4. FSAR, Paragraph 14.3.3.3.

5. FSAR, Paragraph 14.3.3.4.

6. FSAR, Paragraph 3.6.5.2.

7. NEDE-2401 1-P-A, "General Electric Standard Application for Reactor Fuel," (revision specified in the COLR).

8. Technical Requirements Manual, Section 8.0.

9. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

HATCH UNIT 1 B 3.1 -6 REVISION 40

Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE SR 3.1.4.2 (continued)

REQUIREMENTS done on the CRDs at more frequent intervals in accordance with LCO 3.1.3 and LCO 3.1.5, 'Control Rod Scram Accumulators."

SR 3.1.4.3 When work that could affect the scram insertion time is performed on a control rod or the CRD System, testing must be done to demonstrate that each affected control rod retains adequate scram performance over the range of applicable reactor pressures from zero to the maximum permissible pressure. The scram testing must be performed once before declaring the control rod OPERABLE. The required scram time testing must demonstrate the affected control rod is still within acceptable limits. The limits for reactor pressures < 800 psig, required by footnote (b), are included in the Technical Requirements Manual (Ref. 7) and are established based on a high probability of meeting the acceptance criteria at reactor pressures : 800 psig. The limits for reactor pressures a 800 psig are found in Table 3.1.4-1. If testing demonstrates the affected control rod does not meet these limits, but is within the 7 second limit of Table 3.1.4-1, Note 2, the control rod can be declared OPERABLE and 'slow.'

Specific examples of work that could affect the scram times are (but are not limited to) the following: removal of any CRD for maintenance or modification; replacement of a control rod; and maintenance or modification of a scram solenoid pilot valve, scram valve, accumulator, isolation valve or check valve in the piping required for scram.

The Frequency of once prior to declaring the affected control rod OPERABLE is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY.

SR 3.1.4.4 When work that could affect the scram insertion time is performed on a control rod or CRD System, testing must be done to demonstrate each affected control rod is still within the limits of Table 3.1.4-1 with the reactor steam dome pressure 2 800 psig. Where work has been performed at high reactor pressure, the requirements of SR 3.1.4.3 and SR 3.1.4.4 can be satisfied with one test. However, for a control rod affected by work performed while shutdown, a zero pressure test and a high pressure test may be required. This testing ensures that, (continued)

HATCH UNIT 1 B 3.1 -23 REVISION 0

II Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE SR 3.1.4.4 (continued)

REQUIREMENTS prior to withdrawing the control rod for continued operation, the control rod scram performance is acceptable for operating reactor pressure conditions. Alternatively, a control rod scram test during hydrostatic pressure testing could also satisfy both criteria.

The Frequency of once prior to exceeding 40% RTP is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY.

This test is also used to demonstrate control rod OPERABILITY when 2 40% RTP after work that could affect the scram insertion time is performed on the CRD system.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. FSAR, Section 3.4.

3. FSAR, Appendix M.

4. FSAR, Sections 14.3 and 14.4.

5. NEDE-2401 1-P-A, "General Electric Standard Application for Reactor Fuel," (revision specified in the COLR).

6. Letter from R. F. Janecek (BWROG) to R. W. Starostecki (NRC), 'BWR Owners' Group Revised Reactivity Control Systems Technical Specifications", BWROG-8754, September 17, 1987.

7. Technical Requirements Manual, Table T5.0-1.

8. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23,1993.

HATCH UNIT 1 B 3.1-24 REVISION 40

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.17 (continued)

REQUIREMENTS Alternatively, the bypass setpoint may be adjusted to place the channel in a conservative condition (unbypass). If placed in the unbypass condition, this SR is met and the channel is considered OPERABLE.

The 24 month Frequency is based on a review of the surveillance test history and Reference 18.

REFERENCES 1. FSAR, Section 7.2.

2. FSAR, Chapter 14.

3. FSAR, Section 6.5.

4. FSAR, Appendix M.

5. FSAR, Subsection 14.3.3.

6. NEDO-23842, "Continuous Control Rod Withdrawal in the Startup Range," April 18,1978.

7. FSAR, Subsections 14.4.2 and 14.5.5.

8. P. Check (NRC) letter to G. Lainas (NRC), 'BWR Scram Discharge System Safety Evaluation," December 1, 1980.

9. NEDO-30851-P-A, 'Technical Specification Improvement Analyses for BWR Reactor Protection System," March 1988.

10. Technical Requirements Manual, Table T5.0-1.

11. NRC No.93-102, 'Final Policy Statement on Technical Specification Improvements," July 23, 1993.

12. NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," October 1995.

13. NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

(continued)

HATCH UNIT 1 I B 3.3-31 REVISION 40

II RPS Instrumentation B 3.3.1.1 BASES REFERENCES 14. NEDO-31960-A, Supplement 1, "BWR Owners' Group (continued) Long-Term Stability Solutions Licensing Methodology,"

November 1995.

15. NEDO-32465-A, 'BWR Owners' Group Long-Term Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications," March 1996.

16. NEDO-3241 OP-A, Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option IlIl Stability Trip Function,"

November 1997.

17. Letter, L.A. England (BWROG) to M.J. Virgilio, "BWR Owners' Group Guidelines for Stability Interim Corrective Action,"

June 6, 1994.

18. NRC Safety Evaluation Report for Amendment 232.

19. GE Letter NSA 02-250, "Plant Hatch IRM Technical Specifications," April 19, 2002.

20. NRC Safety Evaluation Report for AmendMent 234, Quarterly Surveillance Extension.

HATCH UNIT 1 B 3.3-32 REVISION 31

Remote Shutdown System B 3.3.3.2 BASES ACTIONS B.1 (continued)

If the Required Action and associated Completion Time of Condition A are not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Time is reasonable, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE The Surveillances are modified by a Note to indicate that when an REQUIREMENTS instrument channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. The Note is based upon a NRC Safety Evaluation Report (Ref. 1) which concluded that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability of monitoring required parameters, when necessary.

SR 3.3.3.2.1 Performance of the CHANNEL CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel against a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. As specified in the Surveillance, a CHANNEL CHECK is only required for those channels that are normally energized.

(continued)

HATCH UNIT 1 B 3.3-73 REVISION 1 l

1 ).

I Remote Shutdown System B 3.3.3.2 BASES SURVEILLANCE SR 3.3.3.2.1 (continued)

REQUIREMENTS The Frequency is based upon plant operating experience that demonstrates channel failure is rare.

SR 3.3.3.2.2 SR 3.3.3.2.2 verifies each required Remote Shutdown System transfer switch and control circuit performs the intended function. This verification is performed from the remote shutdown panel and locally, as appropriate. Operation of equipment from the remote shutdown panel is not necessary. The Surveillance can be satisfied by performance of a continuity check, or, in the case of the DG controls, the routine Surveillances of LCO 3.8.1 (since local control is utilized during the performance of some of the Surveillances of LCO 3.8.1).

This will ensure that if the control room becomes inaccessible, the plant can be placed and maintained in MODE 3 from the remote shutdown panel and the local control stations. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 24 month Frequency is based on a review of the surveillance test history and Reference 4.

SR 3.3.3.2.3 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies the channel responds to measured parameter values with the necessary range and accuracy.

The 24 month Frequency is based on a review of the surveillance test history and Reference 4.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 19.

2. Technical Requirements Manual, Table T6.0-1. I

3. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

4. NRC Safety Evaluation Report for Amendment 232.

HATCH UNIT 1 B 3.3-74 REVISION 40

EOC-RPT Instrumentation B 3.3.4.1 BASES SURVEILLANCE SR 3.3.4.1.5 (continued)

REQUIREMENTS STAGGERED TEST BASIS, is also based on a review of the surveillance test history and Reference 7.

SR 3.3.4.1.6 This SR ensures that the RPT breaker interruption time is provided to the EOC-RPT SYSTEM RESPONSE TIME test. Breaker interruption (i.e., trip) time is defined as breaker response time plus arc suppression time. Breaker response time is the time from application of voltage to the trip coil until the main contacts separate. Arc suppression time is the time from main contact separation until the complete suppression of the electrical arc across the open contacts.

Breaker response shall be verified by testing and added to the manufacturer's design arc suppression time to determine breaker interruption time. The breaker arc suppression time shall be validated by the performance of periodic contact gap measurements in accordance with plant procedures. The 60 month Frequency of the testing is based on the difficulty of performing the test and the reliability of the circuit breakers.

REFERENCES 1. FSAR, Section 7.17.

2. FSAR, Subsection 14.3.1.

3. Unit 2 FSAR, Paragraph 5.5.16.1 and Subsection 7.6.10.

4. GENE-770-06-1, "Bases For Changes To Surveillance Test Intervals And Allowed Out-Of-Service Times For Selected Instrumentation Technical Specifications," February 1991.

5. Technical Requirements Manual, Table T5.0-1.

6. NRC No.93-102, Final Policy Statement on Technical Specification Improvements," July 23, 1993.

7. NRC Safety Evaluation Report for Amendment 232.

8. NRC Safety Evaluation Report for Amendment 234, Quarterly Surveillance Extension.

HATCH UNIT 1 B 3.3-83 REVISION 40

ATWS-RPT Instrumentation B 3.3.4.2 B 3.3 INSTRUMENTATION B 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT)

Instrumentation BASES I i BACKGROUND The ATWS-RPT System initiates an RPT, adding negative reactivity, following events in which a scram does not (but should) occur, to lessen the effects of an ATWS event. Tripping the recirculation pumps adds negative reactivity from the increase in steam voiding in the core area as core flow decreases. When Reactor Vessel Water Level - ATWS-RPT Level or Reactor Steam Dome Pressure - High I setpoint is reached, the recirculation pump drive motor breakers trip.

The ATWS-RPT System (Ref. 1) includes sensors, relays, bypass capability, circuit breakers, and switches that are necessary to cause initiation of an RPT. The channels include electronic equipment (e.g.,

trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an ATWS-RPT signal to the trip logic.

The ATWS-RPT consists of two independent trip systems, with two channels of Reactor Steam Dome Pressure - High-and two channels of Reactor Vessel Water Level - ATWS-RPT Level in each trip system. Each ATWS-RPT trip system is a two-out-of-two logic for I

each Function. Thus, either two Reactor Water Level - ATWS-RPT Level or two Reactor Pressure - High signals are needed to trip a trip I i system. The outputs of the channels in a trip system are combined in a logic so that either trip system will trip both recirculation pumps (by tripping the respective drive motor breakers).

There is one drive motor breaker provided for each of the two recirculation pumps for a total of two breakers. The output of each trip system is provided to both recirculation pump breakers.

APPLICABLE The ATWS-RPT is not assumed in the safety analysis. The SAFETY ANALYSES, ATWS-RPT initiates an RPT to aid in preserving the integrity of the LCO, and fuel cladding following events in which a scram does not, but should, APPLICABILITY occur. Based on its contribution to the reduction of overall plant risk, I, however, the instrumentation meets Criterion 4 of the NRC Policy Statement (Ref. 3).

(continued)

HATCH UNIT 1 B 3.3-84 REVISION 3

ECCS - Shutdown B 3.5.2 BASES SURVEILLANCE SR 3.5.2.1 and SR 3.5.2.2 (continued)

REQUIREMENTS water level variations and instrument drift during the applicable MODES. Furthermore, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal suppression pool or CST water level condition.

SR 3.5.2.3. SR 3.5.2.5. and SR 3.5.2.6 The Bases provided for SR 3.5.1.1, SR 3.5.1.7, and SR 3.5.1.10 are applicable to SR 3.5.2.3, SR 3.5.2.5, and SR 3.5.2.6, respectively.

However, the LPCI flow rate requirement for SR 3.5.2.5 is based on a single pump, not the two pump flow rate requirement of SR 3.5.1.7.

SR 3.5.2.4 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

The 31 day Frequency is appropriate because the valves are operated under procedural control and the probability of their being mispositioned during this time period is low.

In MODES 4 and 5, the RHR System may operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Therefore, RHR valves that are required for LPCI subsystem operation may be aligned for decay heat removal. Therefore, this SR is modified by a Note that allows one LPCI subsystem of the RHR System to be considered OPERABLE for the ECCS function if all the required valves in the LPCI flow path can be manually realigned (remote or local) to allow injection into the RPV, and the system is not otherwise inoperable. This will ensure adequate core cooling if an inadvertent RPV draindown should occur.

(continued)

HATCH UNIT 1 B 3.5-19 REVISION 1 l

'l I ECCS - Shutdown B 3.5.2 BASES (continued) \

REFERENCES 1. NEDC-31376P, "E. I. Hatch Nuclear Plant Units 1 and 2 SAFERIGESTR-LOCA Loss-of-Coolant Accident Analysis,"

December 1986.

2. Technical Requirements Manual, Section 8.0. I

3. NRC No.93-102, NFinal Policy Statement on Technical Specification Improvements," July 23, 1993.

HATCH UNIT 1 B 3.5-20 REVISION 40

PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.8 (continued)

REQUIREMENTS

[nominal]. In addition, the EFCVs in the sample are representative of the various plant configurations, models, sizes, and operating environments. This ensures that any potentially common problem with a specific type of application of EFCV is detected at the earliest possible time.) This SR provides assurance that the instrumentation line EFCVs will perform as designed. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

The 24 month Frequency is based on a review of the surveillance test history and Reference 8. (The nominal 10 year interval is based on performance testing as discussed in NEDO-32977-A, "Excess Flow Check Valve Testing Relaxation' (Ref. 7). Furthermore, any EFCV failures will be evaluated to determine if additional testing in that test interval is warranted to ensure overall reliability is maintained.

Operating experience has demonstrated that these components are highly reliable and that failures to isolate are very infrequent.

Therefore, testing of a representative sample was concluded to be acceptable from a reliability standpoint.) Any EFCV that fails to check flow during its surveillance test will be documented in the Hatch corrective action program as a surveillance test failure. The failure will be evaluated and corrected and, if the valve is repaired and not replaced, it will be added to the next cycle's surveillance.

SR 3.6.1.3.9 The TIP shear isolation valves are actuated by explosive charges. An in place functional test is not possible with this design. The explosive squib is removed and tested to provide assurance that the valves will actuate when required. The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. The Frequency of 24 months on a STAGGERED TEST BASIS is considered adequate given the administrative controls on replacement charges and the frequent checks of circuit continuity (SR 3.6.1.3.4). The 24 month Frequency is based on a review of the surveillance test history and Reference 8.

(continued)

HATCH UNIT 1 B 3.6-25 REVISION 28

I PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.10 REQUIREMENTS (continued) The analyses in References 1 and 3 are based on leakage that is less than the specified leakage rate. Leakage through each MSIV must be s 11.5 scfh when tested at > 28.0 psig.

The Frequency is required by the Primary Containment Leakage Rate Testing Program (Ref. 6).

SR 3.6.1.3.11 The valve seats of each 18 inch purge valve (supply and exhaust) having resilient material seats must be replaced every 24 months.

This will allow the opportunity for repair before gross leakage failure develops. The 24 month Frequency is based on a review of the surveillance test history and Reference 8.

SR 3.6.1.3.12 This SR provides assurance that the excess flow isolation dampers can close following an isolation signal. The 24 month Frequency is based on a review of the surveillance test history and Reference 8.

REFERENCES 1. FSAR, Section 14.4.

2. Technical Requirements Manual, Table T7.0-1. I

3. FSAR, Section 5.2.

4. 10 CFR 50, Appendix J, Option B.

5. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

6. Primary Containment Leakage Rate Testing Program.

7. NEDO-32977-A, "Excess Flow Check Valve Testing Relaxation."

8. NRC Safety Evaluation Report for Amendment 232.

HATCH UNIT 1 B 3.6-26 REVISION 40

Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.8 BASES SURVEILLANCE SR 3.6.1.8.1 (continued)

REQUIREMENTS Chamber-to-Drywell Vacuum Breaker Position Indication," as ACTIONS for inoperable closed position indicator channels.

If position indication is reliable (dual or open indication while torus-to-drywell differential pressure is steady at 0 psid), and indicates open, the alternate methods outlined in the TRM T3.6.1 ACTIONS can prove the indication to be in error and the vacuum breaker closed.

However, in this case the vacuum breaker is assumed open until otherwise proved to satisfy the leakage test, and this confirmation must be performed within the Technical Specification 3.6.1.8, Required Action B.1, Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

The 14 day Frequency is based on engineering judgment, is considered adequate in view of other indications of vacuum breaker status available to operations personnel, and has been shown to be acceptable through operating experience.

A Note is added to this SR which allows suppression chamber-to-drywell vacuum breakers opened in conjunction with the performance of a Surveillance to not be considered as failing this SR. These periods of opening vacuum breakers are controlled by plant procedures and do not represent inoperable vacuum breakers.

SR 3.6.1.8.2 Each required (i.e., required to be OPERABLE for opening) vacuum breaker must be cycled to ensure that it opens adequately to perform its design function and returns to the fully closed position. This ensures that the safety analysis assumptions are valid. The 31 day Frequency of this SR was developed, based on Inservice Testing Program requirements to perform valve testing at least once every 92 days. A 31 day Frequency was chosen to provide additional assurance that the vacuum breakers are OPERABLE, since they are located in a harsh environment (the suppression chamber airspace).

In addition, this functional test is required within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after a discharge of steam to the suppression chamber from the safety/relief valves.

SR 3.6.1.8.3 Verification of the vacuum breaker opening setpoint is necessary to ensure that the safety analysis assumption regarding vacuum breaker (continued)

HATCH UNIT 1 B 3.6-45 REVISION 1

[I Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.8 BASES 1/4 SURVEILLANCE SR 3.6.1.8.3 (continued)

REQUIREMENTS full open differential pressure of 0.5 psid is valid. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 24 month Frequency is based on a review of the surveillance test history and Reference 5. It is further justified because of other surveillances performed at shorter Frequencies that convey the proper functioning status of each vacuum breaker.

REFERENCES 1. FSAR, Section 5.2.

2. Unit 2 FSAR, Section 6.2.1.

3. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements,* July 23, 1993.

4. Technical Requirements Manual, TLCO 3.6.1.

5. NRC Safety Evaluation Report for Amendrient 232.

HATCH UNIT 1 B 3.6-46 REVISION 40

Secondary Containment B 3.6.4.1 BASES ACTIONS C.1. C.2, and C.3 (continued) case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.6.4.1.1 and SR 3.6.4.1.2 REQUIREMENTS Verifying that secondary containment equipment hatches and one access door in each access opening are closed ensures that the infiltration of outside air of such a magnitude as to prevent maintaining the desired negative pressure does not occur. Verifying that all such openings are closed provides adequate assurance that exfiltration from the secondary containment will not occur. SR 3.6.4.1.1 also requires equipment hatches to be sealed. In this application, the term "sealed' has no connotation of leak tightness. Maintaining secondary containment OPERABILITY requires verifying one door in the access opening is closed. An access opening contains one inner and one outer door. The intent is not to breach the secondary containment at any time when secondary containment is required. This is achieved by maintaining the inner or outer portion of the barrier closed at all times. However, all secondary containment access doors are normally kept closed, except when the access opening is being used for entry and exit or when maintenance is being performed on an access opening. When the secondary containment configuration excludes Zone I and/or Zone II, these SRs also include verifying the hatches and doors separating the common refueling floor zone from the reactor building(s). The 31 day Frequency for these SRs has been shown to be adequate, based on operating experience, and is considered adequate in view of the other indications of door and hatch status that are available to the operator.

SR 3.6.4.1.3 and SR 3.6.4.1.4 The Unit 1 and Unit 2 SGT Systems exhaust the secondary containment atmosphere to the environment through appropriate treatment equipment. To ensure that all fission products are treated, SR 3.6.4.1.3 verifies that the appropriate SGT System(s) will rapidly establish and maintain a negative pressure in the secondary containment. This is confirmed by demonstrating that the required SGT subsystem(s) will draw down the secondary containment to 2 0.20 inch of vacuum water gauge in < 120 seconds. This cannot be accomplished if the secondary containment boundary is not intact.

SR 3.6.4.1.4 demonstrates that the required SGT subsystem(s) can (continued)

HATCH UNIT 1 B 3.6-75 REVISION 34

II Secondary Containment B 3.6.4.1 BASES SURVEILLANCE SR 3.6.4.1.3 and SR 3.6.4.1.4 (continued)

REQUIREMENTS maintain 2 0.20 inch of vacuum water gauge for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at a flow rate s 4000 cfm for each SGT subsystem. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months test period allows secondary containment to be in thermal equilibrium at steady state conditions. Therefore, these two tests are used to ensure secondary containment boundary integrity. Since these SRs are secondary containment tests, they need not be performed with each SGT subsystem. The SGT subsystems are tested on a STAGGERED TEST BASIS, however, to ensure that in addition to the requirements of LCO 3.6.4.3, each SGT subsystem or combination of subsystems will perform this test. The number of SGT subsystems and the required combinations are dependent on the configuration of the secondary containment and are detailed in the Technical Requirements Manual (Ref. 3). The Note to SR 3.6.4.1.3 and SR 3.6.4.1.4 specifies that the number of required SGT subsystems be one less than the number required to meet LCO 3.6.4.3, NStandby Gas Treatment (SGT) System," for the given configuration. The 24 month Frequency, on a STAGGERED TEST BASIS, of SRs 3.6.4.1.3 and 3.6.4.1.4 is also based on a review of the surveillance test history and Reference 5.

REFERENCES 1. FSAR, Subsection 14.4.3.

2. FSAR, Subsection 14.4.4.

3. Technical Requirements Manual, Section 8.0.

4. NRC No.93-102, 'Final Policy Statement on Technical Specification Improvements,' July 23,1993.

5. NRC Safety Evaluation Report for Amendment 232.

HATCH UNIT 1 B 3.6-76 REVISION 40

SCIVs B 3.6.4.2 BASES (continued)

REFERENCES 1. FSAR, Subsection 14.3.3.

2. FSAR, Subsection 14.3.4.

3. Technical Requirements Manual, Section 8.0. I

4. NRC No.93-102, wFinal Policy Statement on Technical Specification Improvements,' July 23, 1993.

5. NRC Safety Evaluation Report for Amendment 232.

HATCH UNIT 1 B 3.6-83 REVISION 40

SGT System B 3.6.4.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.3 Standby Gas Treatment (SGT) System BASES BACKGROUND The SGT System is required by 10 CFR 50, Appendix A, GDC 41, "Containment Atmosphere Cleanup" (Ref. 1). The function of the SGT System is to ensure that radioactive materials that leak from the primary containment into the secondary containment following a Design Basis Accident (DBA) are filtered and adsorbed prior to exhausting to the environment.

The Unit 1 and Unit 2 SGT Systems each consists of two fully redundant subsystems, each with its own set of dampers, charcoal filter train, and controls. The Unit 1 SGT subsystems' ductwork is separate from the inlet to the filter train to the discharge of the fan.

The rest of the ductwork is common. The Unit 2 SGT subsystems' ductwork is separate except for the suction from the drywell and torus, which is common (however, this suction path is not required for subsystem OPERABILITY).

Each charcoal filter train consists of (components listed in order of the direction of the air flow):

a. A demister or moisture separator;

b. An electric heater;

c. A prefilter;

d. A high efficiency particulate air (HEPA) filter;

e. Two charcoal adsorbers for Unit 1 subsystems and one charcoal adsorber for Unit 2 subsystems;

f. A second HEPA filter; and

g. An axial vane fan for Unit 1 subsystems and a centrifugal fan for Unit 2 subsystems.

The sizing of the SGT Systems equipment and components is based on the results of an infiltration analysis, as well as an exfiltration analysis of the secondary containment. The internal pressure of the SGT Systems boundary region is maintained at a negative pressure when the system is in operation, to conservatively ensure zero (continued)

HATCH UNIT 1 B 3.6-84 REVISION 4

SGT System B 3.6.4.3 BASES ACTIONS F.1, F.2, and F.3 (continued) applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

Required Action F.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.6.4.3.1 REQUIREMENTS Operating each required Unit 1 and Unit 2 SGT subsystem for 2 10 continuous hours ensures that they are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on for

10 continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls and the redundancy available in the system.

SR 3.6.4.3.2 This SR verifies that the required Unit 1 and Unit 2 SGT filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies and additional information are discussed in detail in the VFTP.

(continued)

HATCH UNIT 1 B 3.6-89 REVISION 1 l

SGT System B 3.6.4.3 BASES SURVEILLANCE SR 3.6.4.3.3 REQUIREMENTS (continued) This SR verifies that each required Unit 1 and Unit 2 SGT subsystem starts on receipt of an actual or simulated initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.2.5 overlaps this SR to provide complete testing of the safety function. This Surveillance can be performed with the reactor at power. The 24 month Frequency is based on a review of the surveillance test history and Reference 6.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 41.

2. FSAR, Section 5.3.

3. Unit 2 FSAR, Subsection 6.2.3.

4. Technical Requirements Manual, Section 8.0. I

5. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

6. NRC Safety Evaluation Report for Amendment 232.

Ed /

HATCH UNIT 1 B 3.6-90 REVISION 40

RHRSW System B 3.7.1 B 3.7 PLANT SYSTEMS B 3.7.1 Residual Heat Removal Service Water (RHRSW) System BASES BACKGROUND The RHRSW System is designed to provide cooling water for the Residual Heat Removal (RHR) System heat exchangers, required for a safe reactor shutdown following a Design Basis Accident (DBA) or transient. The RHRSW System is operated whenever the RHR heat exchangers are required to operate in the shutdown cooling mode or in the suppression pool cooling or spray mode of the RHR System.

The RHRSW System consists of two independent and redundant subsystems. Each subsystem is made up of a header, two 4000 gpm pumps, a suction source, valves, piping, heat exchanger, and associated instrumentation. Either of the two subsystems is capable of providing the required cooling capacity with two pumps operating to maintain safe shutdown conditions. The two subsystems are separated from each other by normally closed motor operated cross tie valves, so that failure of one subsystem will not affect the OPERABILITY of the other subsystem. The RHRSW System is designed with sufficient redundancy so that no single active -

component failure can prevent it from achieving its.design function.

The RHRSW System is described in the FSAR, Section 10.6, Reference 1.

Cooling water is pumped by the RHRSW pumps from the Altamaha River through the tube side of the RHR heat exchangers, and discharges to the circulating water flume. A minimum flow line from the pump discharge to the intake structure prevents the pump from overheating when pumping against a closed discharge valve.

The system is initiated manually from the control room. If operating during a loss of coolant accident (LOCA) or a loss of offsite power (LOSP), the system is automatically tripped to allow the diesel generators to automatically power only that equipment necessary.

The system can be manually started any time the LOCA signal is manually overridden or clears. The system can be manually started any time after the LOSP signal is received.

(continued)

HATCH UNIT 1 B 3.7-1 REVISION 0

I RHRSW System B 3.7.1 BASES (continued) Iu APPLICABLE The RHRSW System removes heat from the suppression pool to limit SAFETY ANALYSES the suppression pool temperature and primary containment pressure following a LOCA. This ensures that the primary containment can perform its function of limiting the release of radioactive materials to the environment following a LOCA. The ability of the RHRSW System to support long term cooling of the reactor or primary containment is discussed in the FSAR, Section 10.6 and Subsection 14.4.3 (Refs. 1 and 2, respectively). These analyses explicitly assume that the RHRSW System will provide adequate cooling support to the equipment required for safe shutdown. These analyses include the evaluation of the long term primary containment response after a design basis LOCA.

The safety analyses for long term cooling were performed for various combinations of RHR System failures. The worst case single failure that would affect the performance of the RHRSW System is any failure that would disable one subsystem of the RHRSW System. As discussed in the FSAR, Subsection 14.4.3 (Ref. 2) for these analyses, I manual initiation of the OPERABLE RHRSW subsystem and the associated RHR System is assumed to occur 10 minutes after a DBA.

The RHRSW flow assumed in the analyses is 4000 gpm per pump with two pumps operating in one loop. In this case, the maximum suppression chamber water temperature and pressure are approximately 21 0F and 15 psig, respectively, well below the design temperature of 281 0F and maximum allowable pressure of 62 psig.

The RHRSW System satisfies Criterion 3 of the NRC Policy Statement (Ref. 3). I LCO Two RHRSW subsystems are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post accident heat loads, assuming the worst case single active failure occurs coincident with the loss of offsite power.

An RHRSW subsystem is considered OPERABLE when:

a. Two pumps are OPERABLE; and

b. An OPERABLE flow path is capable of taking suction from the intake structure and transferring the water to the RHR heat exchangers at the assumed flow rate. Additionally, the RHRSW cross tie valves (which allow the two RHRSW loops to be connected) must be closed so that failure of one subsystem will not affect the OPERABILITY of the other subsystems.

(continued)

HATCH UNIT 1 B 3.7-2 REVISION 40

RHRSW System B 3.7.1 BASES ACTIONS D.1 (continued)

With both RHRSW subsystems inoperable for reasons other than Condition B (e.g., both subsystems with inoperable flow paths, or one subsystem with an inoperable pump and one subsystem with an inoperable flow path), the RHRSW System is not capable of performing its intended function. At least one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time for restoring one RHRSW subsystem to OPERABLE status, is based on the Completion Times provided for the RHR suppression pool cooling and spray functions.

The Required Action is modified by a Note indicating that the applicable Conditions of LCO 3.4.7 be entered and Required Actions taken if an inoperable RHRSW subsystem results in an inoperable RHR shutdown cooling subsystem. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.

E.1 and E.2 If the RHRSW subsystems cannot be not restored to OPERABLE status within the associated Completion Times, the unit must be placed in a MODE in which the LCO does not appiy. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.1.1 REQUIREMENTS Verifying the correct alignment for each manual, power operated, and automatic valve in each RHRSW subsystem flow path provides assurance that the proper flow paths will exist for RHRSW operation.

This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position, and yet considered in the correct position, provided it can be realigned to its accident position. This is acceptable because the RHRSW System is a manually initiated system. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being (continued)

HATCH UNIT 1 B 3.7-5 REVISION 0

II RHRSW System B 3.7.1 BASES SURVEILLANCE SR 3.7.1.1 (continued)

REQUIREMENTS mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions.

REFERENCES 1. FSAR, Section 10.6.

2. FSAR, Subsection 14.4.3.

3. NRC No.93-102, 'Final Policy Statement on Technical I Specification Improvements," July 23, 1993.

HATCH UNIT 1 B 3.7-6 REVISION 40

MCREC System B 3.7.4 BASES (continued)

SURVEILLANCE SR 3.7.4.1 REQUIREMENTS This SR verifies that a subsystem in a standby mode starts on demand and continues to operate. Standby systems should be checked periodically to ensure that they start and function properly.

As the environmental and normal operating conditions of this system are not severe, testing each subsystem once every 31 days provides an adequate check on this system. Since the MCREC System does not have heaters, each subsystem need only be operated for 2 15 minutes to demonstrate the function of the subsystem.

Furthermore, the 31 day Frequency is based on the known reliability of the equipment and the two subsystem redundancy available.

SR 3.7.4.2 This SR verifies that the required MCREC testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).

Specific test frequencies and additional information are discussed in detail in the VFTP.

SR 3.7.4.3 This SR verifies that on an actual or simulated initiation signal, each MCREC subsystem starts and operates. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.7.1.4 overlaps this SR to provide complete testing of the safety function. This Surveillance can be performed with the reactor at power. The 24 month Frequency is based on a review of the surveillance test history and Reference 9.

SR 3.7.4.4 This SR verifies the integrity of the control room enclosure and the assumed inleakage rates of potentially contaminated air. The control room positive pressure, with respect to potentially contaminated adjacent areas (the turbine building), is periodically tested to verify proper function of the MCREC System. During the pressurization mode of operation, the MCREC System is designed to slightly pressurize the control room 2 0.1 inches water gauge positive pressure with respect to the turbine building to prevent unfiltered inleakage, The MCREC System is designed to maintain this positive (continued)

HATCH UNIT 1 B 3.7-23 REVISION 28

II MCREC System B 3.7.4 BASES SURVEILLANCE SR 3.7.4.4 (continued)

REQUIREMENTS pressure at a flow rate of s 2750 cfm through the control room in the pressurization mode. This SR ensures the total flow rate meets the design analysis value of 2500 cfm +/- 10% and ensures the outside air flow rate is s 400 cfm. The 24 month Frequency, on a STAGGERED TEST BASIS, is based on a review of the surveillance test history and Reference 9.

REFERENCES 1. Unit 2 FSAR, Section 6.4.

2. Unit 2 FSAR, Section 9.4.1.

3. FSAR, Section 5.2.

4. FSAR, Chapter 14.

5. Unit 2 FSAR, Section 6.4.1.2.2.

6. Unit 2 FSAR, Table 15.1-28.

7. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

8. Technical Requirements Manual, Table T2.1-1. I

9. NRC Safety Evaluation Report for Amendment 232.

U--

HATCH UNIT 1 B 3.7-24 REVISION 40

Control Room AC System B 3.7.5 BASES ACTIONS E.1. E.2.1, E.2.2, and E.2.3 (continued)

During movement of irradiated fuel assemblies in the secondary containment, during CORE ALTERATIONS, or during OPDRVs, if any Required Action and associated Completion Time for Condition B or C is not met, the necessary OPERABLE control room AC subsystems may be placed immediately in operation. One operable control room AC subsystem is necessary if the outside air temperature is < 650F and the maximum outside air temperature in the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has been s 650F. If both of these conditions are not met, then two OPERABLE control room AC subsystems are necessary. This action ensures that the remaining subsystems are OPERABLE, that no failures that would prevent actuation will occur, and that any active failure will be readily detected.

An alternative to Required Action E.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.

If applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies in the secondary containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.

F.1 If three control room AC subsystems are inoperable in MODE 1, 2, or 3, the Control Room AC System may not be capable of performing the intended function. Therefore, LCO 3.0.3 must be entered immediately.

G.1. G.2. and G.3 The Required Actions of Condition G are modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not a sufficient reason to require a reactor shutdown.

(continued)

HATCH UNIT 1 B 3.7-29 REVISION 39 1

II Control Room AC System B 3.7.5 BASES ACTIONS G.1. G.2. and G.3 (continued)

During movement of irradiated fuel assemblies in the secondary containment, during CORE ALTERATIONS, or during OPDRVs, with three control room AC subsystems inoperable, action must be taken immediately to suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.

If applicable, CORE ALTERATIONS and handling of irradiated fuel in the secondary containment must be suspended immediately.

Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.

SURVEILLANCE SR 3.7.5.1 REQUIREMENTS This SR verifies that the heat removal capability of the system is sufficient to remove the control room heat load assumed in the safety analysis. The SR consists of a combination of testing and calculation.

The 24 month Frequency is appropriate since significant degradation of the Control Room AC System is not expected over this time period.

The 24 month Frequency is based on a review of the surveillance test history and Reference 4.

REFERENCES 1. Unit 2 FSAR, Sections 6.4 and 9.4.1.

2. NRC No.93-102, nFinal Policy Statement on Technical Specification Improvements," July 23, 1993.

3. Technical Requirements Manual, Table T2.1-1. I

4. NRC Safety Evaluation Report for Amendment 232.

<II HATCH UNIT 1 B 3.7-30 REVISION 40

Main Turbine Bypass System B 3.7.7 BASES SURVEILLANCE SR 3.7.7.3 (continued)

REQUIREMENTS The 24 month Frequency is based on a review of the surveillance test history and Reference 5.

REFERENCES 1. FSAR, Section 7.11.

2. FSAR, Section 14.3.2.1.

3. Technical Requirements Manual, Table T5.0-1. I

4. NRC No.93-102, t Final Policy Statement on Technical Specification Improvements," July 23, 1993.

5. NRC Safety Evaluation Report for Amendment 232.

HATCH UNIT 1 B 3.7-37 REVISION 40

[I Spent Fuel Storage Pool Water Level B 3.7.8 B 3.7 PLANT SYSTEMS I,-"

B 3.7.8 Spent Fuel Storage Pool Water Level BASES BACKGROUND The minimum water level in the spent fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident.

A general description of the spent fuel storage pool design is found in the FSAR, Section 10.3 (Ref. 1). The assumptions of the fuel handling accident in the spent fuel storage pool are found in Reference 2.

APPLICABLE The water level above the irradiated fuel assemblies is an explicit SAFETY ANALYSES assumption of the fuel handling accident; the point from which the water level is measured is shown in Figure B 3.5.2-1. A fuel handling accident in the spent fuel storage pool was evaluated (Ref. 2) and ensured that the radiological consequences (calculated whole body and thyroid doses at the exclusion area and low population zone boundaries) were well below the guideline doses of 10 CFR 100 (Ref. 3) and met the exposure guidelines of NUREG-0800 (Ref. 4). A fuel handling accident could release a fraction of the fission product inventory by breaching the fuel rod cladding as discussed in the Regulatory Guide 1.25 (Ref. 5).

The fuel handling accident is evaluated for the dropping of an irradiated fuel assembly onto the spent fuel storage pool racks (Ref. 2). The water level in the spent fuel storage pool provides for absorption of water soluble fission product gases and transport delays of soluble and insoluble gases that must pass through the water before being released to the secondary containment atmosphere.

This absorption and transport delay reduces the potential radioactivity of the release during a fuel handling accident.

The spent fuel storage pool water level satisfies Criterion 2 of the NRC Policy Statement (Ref. 6).

LCO The specified water level preserves the assumptions of the fuel handling accident analysis (Ref. 2). As such, it is the minimum required for fuel movement within the spent fuel storage pool.

(continued)

HATCH UNIT 1 B 3.7-38 REVISION 22

DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.8 REQUIREMENTS (continued) A battery performance discharge test is a constant current capacity test to detect any change in the capacity determined by the acceptance test. Initial conditions consistent with IEEE-450 need to be met prior to the performing of a battery performance discharge test. The test results reflect the overall effects of usage and age.

A battery modified performance discharge test is described in the Bases for SR 3.8.4.7. Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.4.8; however, only the modified performance discharge test may be used to satisfy SR 3.8.4.8, while satisfying the requirements of SR 3.8.4.7 at the same time.

The acceptance criteria for this Surveillance is consistent with IEEE-450 (Ref. 8) and IEEE-485 (Ref. 12). These references recommend that the battery be replaced if its capacity is below 80% of the manufacturer's rating. Although there may be ample capacity, the battery rate of deterioration is rapidly increasing.

The Frequency for this test is normally 60 months. If the battery shows degradation, or if the battery has reached 85% of its expected application service life and capacity is s 100% of the manufacturers rating, the Surveillance Frequency is reduced to 12 months. However, if the battery shows no degradation but has reached 85% of its expected application service life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity 2 100% of the manufacturer's rating. Degradation is indicated, according to IEEE-450 (Ref. 8), when the battery capacity drops by more than 10% of rated capacity from its capacity on the previous performance test or is more than 10% below the manufacturer's rating. All these Frequencies are consistent with the recommendations in IEEE-450 (Ref. 8).

This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required DC electrical power subsystem from service, perturb the electrical distribution system, and challenge safety systems. Credit may be taken for unplanned events that satisfy the Surveillance. The swing DG DC battery is exempted from this restriction, since it is required by both units' LCO 3.8.4 and cannot be performed in the manner required by the Note without resulting in a dual unit shutdown.

(continued)

HATCH UNIT 1 B 3.8-61 REVISION 33

DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.9 REQUIREMENTS (continued) With the exception of this Surveillance, all other Surveillances of this Specification (SR 3.8.4.1 through SR 3.8.4.8) are applied only to the Unit 1 DC sources. This Surveillance is provided to direct that the appropriate Surveillances for the required Unit 2 DC sources are governed by the Unit 2 Technical Specifications. Performance of the applicable Unit 2 Surveillances will satisfy both any Unit 2 requirements, as well as satisfying this Unit 1 SR.

The Frequency required by the applicable Unit 2 SR also governs performance of that SR for both Units.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 17.

2. Regulatory Guide 1.6.

3. IEEE Standard 308-1971.

4. FSAR, Section 8.5.

5. FSAR, Chapters 5 and 6.

6. FSAR, Chapter 14.

7. Regulatory Guide 1.93, December 1974.

8. IEEE Standard 450-1987.

9. Technical Requirements Manual, Section 9.0. I

10. Regulatory Guide 1.32, February 1977.

11. Regulatory Guide 1.129, December 1974.

12. IEEE Standard 485-1983.

13. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements,' July 23, 1993.

14. NRC Safety Evaluation Report for Amendment 232.

HATCH UNIT 1 B 3.8-62 REVISION 40 '

RHR - High Water Level B 3.9.7 I1- BASES (continued)

REFERENCES 1. 10 CFR 50, Appendix A, GDC 34.

2. Technical Requirements Manual, Section 8.0. I

3. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

HATCH UNIT 1 B 3.9-23 REVISION 40

II RHR - Low Water Level B 3.9.8 B 3.9 REFUELING OPERATIONS B 3.9.8 Residual Heat Removal (RHR) - Low Water Level BASES BACKGROUND The purpose of the RHR System in MODE 5 is to remove decay heat and sensible heat from the reactor coolant, as required by GDC 34 (Ref. 1). Each of the two shutdown cooling loops of the RHR System can provide the required decay heat removal. Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via the associated recirculation loop. The RHR heat exchangers transfer heat to the RHR Service Water System. The RHR shutdown cooling mode is manually controlled.

APPLICABLE With the unit in MODE 5, the RHR System is not required to SAFETY ANALYSES mitigate any events or accidents evaluated in the safety analyses.

The RHR System is required for removing decay heat to maintain the temperature of the reactor coolant.

The RHR System satisfies Criterion 4 of the NRC Policy Statement (Ref. 3).

LCO In MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level < 22 ft 1/8 inches above the RPV flange, two RHR shutdown cooling subsystems must be OPERABLE.

An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump and the associated heat exchanger, an RHRSW pump providing cooling to the heat exchanger with sufficient flow to maintain reactor coolant temperature in the desired range, valves, piping, instruments, and controls to ensure an OPERABLE flow path. The two required RHR shutdown cooling subsystems have a common suction source and are allowed to have a common heat exchanger and common discharge piping.

Since the piping and heat exchangers are passive components that are assumed not to fail, they are allowed to be common to both.

subsystems. Thus, to meet the LCO, both RHR pumps in one loop or one RHR pump in each of the two loops must be OPERABLE. If the I (continued)

HATCH UNIT B 3.9-24 3 REVISION 1

RHR - Low Water Level B 3.9.8 BASES ACTIONS B.1. B.2. and B.3 (continued)

2) sufficient standby gas treatment subsystem(s) are OPERABLE to maintain the secondary containment at a negative pressure with respect to the environment (dependent on secondary containment configuration, refer to Reference 2; single failure protection is not required while in this ACTION); and 3) secondary containment isolation capability is available in each associated secondary containment penetration flow path not isolated that is assumed to be isolated to mitigate radioactive releases (i.e., one secondary containment isolation valve and associated instrumentation are OPERABLE or other acceptable administrative controls to assure isolation capability. The administrative controls can consist of stationing a dedicated operator, who is in continuous communication with the control room, at the controls of the isolation device. In this way, the penetration can be rapidly isolated when a need for secondary containment isolation is indicated.). This may be performed as an administrative check, by examining logs or other information to determine whether the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In tlhis case, the Surveillance may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.

C.1 and C.2 If no RHR shutdown cooling subsystem is in operation, an alternate method of coolant circulation is required to be established within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The Completion Time is modified such that the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is applicable separately for each occurrence involving a loss of coolant circulation.

During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling subsystem), the reactor coolant temperature must be periodically monitored to ensure proper functioning of the alternate method. The once per hour Completion Time is deemed appropriate.

(continued)

HATCH UNIT 1 B 3.9-27 REVISION 1

II RHR - Low Water Level B 3.9.8 BASES (continued)

SURVEILLANCE SR 3.9.8.1 REQUIREMENTS This Surveillance demonstrates that one required RHR shutdown cooling subsystem is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystems in the control room.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 34.

2. Technical Requirements Manual, Section 8.0. I

3. NRC No.93-102, 0Final Policy Statement on Technical Specification Improvements," July 23,1993.

HATCH UNIT 1 B 3.9-28 REVISION 40

HATCH UNIT 2 BASES EFFECTIVE PAGE LIST Unless noted otherwise, all of the pages in the HNP Unit 2 Bases are effective per Revision 0.

PAGE REVISION PAGE REVISION i 43 B 3.3-11 21 ii 35 B 3.3-12 21 iii 1 B 3.3-13 14 iv 1 B 3.3-14 14 v 39 B 3.3-15 14 vi 30 B 3.3-16 42 B 2.0-2 42 B 3.3-17 42 B 2.0-3 28 B 3.3-18 14 B 2.0-4 28 B 3.3-19 21 B 2.0-5 28 B 3.3-20 21 B 3.0-9 30 B 3.3-21 21 B 3.0-10 30 B 3.3-22 21 B 3.0-11 30 B 3.3-23 21 B 3.0-12 30 B 3.3-24 42 B 3.0-13 30 B 3.3-25 21 B 3.0-14 30 B 3.3-26 32 B 3.1-3 1 B 3.3-27 38 B 3.1-4 1 B 3.3-28 42 B 3.1-5 1 B 3.3-29 38 B 3.1-6 46 B 3.3-30 38 B 3.1-17 1 B 3.3-31 46 B 3.1-18 1 B 3.3-32 38 B 3.1-24 46 B 3.3-33 14 B 3.1-39 35 B 3.3-38 22 B 3.1-40 35 B 3.3-39 22 B 3.1-44 39 B 3.3-40 35 B 3.1-45 35 B 3.3-41 35 B 3.2-1 43 B 3.3-42 16 B 3.2-2 43 B 3.3-43 35 B 3.2-3 43 B 3.3-44 16 B 3.2-4 43 B 3.3-47 29 B 3.2-6 42 B 3.3-48 14 B 3.2-7 42 B 3.3-49 38 B 3.2-9 43 B 3.3-50 35 B 3.2-10 43 B 3.3-51 35 B 3.2-11 43 B 3.3-52 38 B 3.2-12 43 B 3.3-53 42 B 3.3-3 42 B 3.3-54 42 B 3.3-4 37 B 3.3-56 42 B 3.3-5 41 B 3.3-57 38 B 3.3-6 21 B 3.3-58 38 B 3.3-7 42 B 3.3-59 1 B 3.3-8 21 B 3.3-60 1 B 3.3-9 21 B 3.3-61 41 B 3.3-10 21 B 3.3-62 1 HATCH UNIT 2 i REVISION 46

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PAGE REVISION PAGE REVISION B 3.3-109 1 B 3.3-63 1 B 3.3-1 10 1 B 3.3-64 9 B 3.3-111 22 B 3.3-65 1 B 3.3-112 1 B 3.3-66 1 B 3.3-113 9 B 3.3-67 1 B 3.3-114 9 B 3.3-68 35 B 3.3-115 1 B 3.3-69 35 B 3.3-116 1 B 3.3-70 1 B 3.3-117 1 B 3.3-71 1 B 3.3-118 1 B 3.3-72 1 B 3.3-119 1 B 3.3-73 1 B 3.3-120 1 B 3.3-74 46 B 3.3-121 1 B 3.3-75 1 B 3.3-122 35 B 3.3-76 42 B 3.3-123 38 B 3.3-77 42 B 3.3-124 38 B 3.3-78 42 B 3.3-125 1 B 3.3-79 1 B 3.3-126 6 B 3.3-80 42 B 3.3-127 28 B 3.3-81 42 B 3.3-128 1 B 3.3-82 35 B 3.3-129 1 B 3.3-83 46 B 3.3-130 1 B 3.3-84 3 B 3.3-131 1 B 3.3-85 1 B 3.3-132 1 B 3.3-86 28 B 3.3-133 38 B 3.3-87 9 B 3.3-134 38 B 3.3-88 1 B 3.3-135 1 B 3.3-89 1 B 3.3-136 1 B 3.3-90 38 B 3.3-137 1 B 3.3-91 38 B 3.3-138 1 B 3.3-92 1 B 3.3-139 1 B 3.3-93 1 B 3.3-140 1 B 3.3-94 1 B 3.3-141 42 B 3.3-95 1 B 3.3-142 42 B 3.3-96 1 B 3.3-143 1 B 3.3-97 1 B 3.3-144 1 B 3.3-98 4 B 3.3-145 1 B 3.3-99 1 B 3.3-146 5 B 3.3-1 00 1 B 3.3-147 1 B 3.3-1 01 1 B 3.3-148 1 B 3.3-102 1 B 3.3-149 1 B 3.3-103 1 B 3.3-150 1 B 3.3-104 28 B 3.3-151 1 B 3.3-105 28 B 3.3-152 1 B 3.3-106 1 B 3.3-153 28 B 3.3-107 35 B 3.3-154 1 B 3.3-108 1 B 3.3-155 1 HATCH UNIT 2 ii REVISION 46

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PAGE REVISION PAGE REVISION B 3.3-156 1 B 3.4-5 43 B 3.3-157 1 B 3.4-10 42 B 3.3-158 35 B 3.4-11 15 B 3.3-159 38 B 3.4-12 35 B 3.3-160 46 B 3.4-18 22 B 3.3-161 38 B 3.4-19 22 B 3.3-162 35 B 3.4-20 22 B 3.3-163 35 B 3.4-21 22 B 3.3-164 35 B 3.4-22 22 B 3.3-165 35 B 3.4-23 35 B 3.3-166 35 B 3.4-28 22 B 3.3-167 35 B 3.4-29 22 B 3.3-168 35 B 3.4-30 22 B 3.3-169 38 B 3.4-31 22 B 3.3-170 38 B 3.4-32 22 B 3.3-171 38 B 3.4-33 22 B 3.3-172 35 B 3.4-34 1 B 3.3-173 35 B 3.4-35 1 B 3.3-174 35 B 3.4-36 1 B 3.3-175 35 B 3.4-37 1 B 3.3-176 35 B 3.4-38 1 B 3.3-177 35 B 3.4-39 14 B 3.3-178 38 B 3.4-40 14 B 3.3-179 38 B 3.4-41 14 B 3.3-180 35 B 3.4-42 14 B 3.3-1 81 35 B 3.4-43 28 B 3.3-182 35 B 3.4-44 28 B 3.3-183 35 B 3.4-45 29 B 3.3-184 35 B 3.4-46 14 B 3.3-185 38 B 3.4-47 14 B 3.3-186 38 B 3.4-48 29 B 3.3-187 35 B 3.4-49 29 B 3.3-188 35 B 3.4-50 29 B 3.3-189 35 B 3.5-2 20 B 3.3-190 35 B 3.5-3 20 B 3.3-191 35 B 3.5-4 13 B 3.3-192 35 B 3.5-6 22 B 3.3-193 35 B 3.5-7 20 B 3.3-194 35 B 3.5-8 20 B 3.3-195 35 B 3.5-9 20 B 3.3-196 35 B 3.5-10 20 B 3.3-197 35 B 3.5-11 35 B 3.3-198 35 B 3.5-12 35 B 3.3-199 35 B 3.5-13 35 B 3.3-200 35 B 3.5-14 46 B 3.4-2 43 B 3.5-16 28 B3.4-3 43 B3.5-17 1 B 3.4-4 43 B 3.5-18 11 HATCH UNIT 2 REVISION 46 iii

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PAGE REVISION PAGE REVISION B 3.5-19 1 B 3.6-51 12 B 3.5-20 46 B 3.6-52 1 B 3.5-21 30 B 3.6-53 1 B 3.5-22 5 B 3.6-54 1 B 3.5-26 35 B 3.6-55 1 B 3.5-27 35 B 3.6-56 22 B 3.6-1 7 B 3.6-57 22 B 3.6-2 21 B 3.6-58 22 B 3.6-3 7 B 3.6-59 1 B 3.6-4 35 B 3.6-60 22 B 3.6-5 35 B 3.6-61 22 B 3.6-7 21 B 3.6-62 22 B 3.6-11 7 B 3.6-63 1 B 3.6-12 7 B 3.6-64 1 B 3.6-17 7 B 3.6-65 1 B 3.6-18 7 B 3.6-66 3 B 3.6-19 7 B 3.6-67 35 B 3.6-20 1 B 3.6-68 35 B 3.6-21 1 B 3.6-69 1 B 3.6-22 1 B 3.6-70 1 B 3.6-23 35 B 3.6-71 1 B 3.6-24 35 B 3.6-72 1 B 3.6-25 35 B 3.6-73 1 B 3.6-26 35 B 3.6-74 1 B 3.6-27 46 B 3.6-75 1 B 3.6-28 21 B 3.6-76 1 B 3.6-29 1 B 3.6-77 1 B 3.6-30 10 B 3.6-78 1 B 3.6-31 1 B 3.6-79 4 B 3.6-32 1 B 3.6-80 4 B 3.6-33 13 B 3.6-81 40 B 3.6-34 15 B 3.6-82 46 B 3.6-35 35 B 3.6-83 1 B 3.6-36 1 B 3.6-84 1 B 3.6-37 1 B 3.6-85 1 B 3.6-38 1 B 3.6-86 1 B 3.6-39 1 B 3.6-87 1 B 3.6-40 35 B 3.6-88 35 B 3.6-41 1 B 3.6-89 46 B 3.6-42 1 B 3.6-90 4 B 3.6-43 1 B 3.6-91 4 B 3.6-44 19 B 3.6-92 1 B 3.6-45 35 B 3.6-93 1 B 3.6-46 46 B 3.6-94 1 B 3.6-47 1 B 3.6-95 1 B 3.6-48 1 B 3.6-96 46 B 3.6-49 1 B 3.7-2 46 B 3.6-50 1 B 3.7-6 46 HATCH UNIT 2 lV REVISION 46

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PAGE REVISION PAGE REVISION B 3.7-13 35 B 3.8-23 39 B 3.7-16 35 B 3.8-24 39 B 3.7-17 1 B 3.8-25 39 B 3.7-18 29 B 3.8-26 39 B 3.7-19 29 B 3.8-27 39 B 3.7-20 29 B 3.8-28 39 B 3.7-21 29 B 3.8-29 39 B 3.7-22 29 B 3.8-30 39 B 3.7-23 35 B 3.8-31 39 B 3.7-24 46 B 3.8-32 39 B 3.7-25 1 B 3.8-33 39 B 3.7-26 45 B 3.8-34 39 B 3.7-27 45 B 3.8-35 39 B 3.7-28 45 B 3.8-36 39 B 3.7-29 45 B 3.8-37 39 B 3.7-30 46 B 3.8-38 39 B 3.7-31 42 B 3.8-39 39 B 3.7-32 21 B 3.8-40 39 B 3.7-33 9 B 3.8-41 39 B 3.7-34 21 B 3.8-42 39 B 3.7-35 42 B 3.8-43 39 B 3.7-36 42 B 3.8-44 39 B 3.7-37 46 B 3.8-45 39 B 3.7-38 28 B 3.8-46 39 B 3.7-39 1 B 3.8-47 39 B 3.7-40 1 B 3.8-48 39 B 3.8-1 1 B 3.8-49 39 B 3.8-2 1 B 3.8-50 39 B 3.8-3 20 B 3.8-51 39 B 3.8-4 20 B 3.8-52 39 B 3.8-5 20 B 3.8-53 39 B 3.8-6 20 B 3.8-54 39 B 3.8-7 33 B 3.8-55 39 B 3.8-8 20 B 3.8-56 39 B 3.8-9 44 B 3.8-57 39 B 3.8-10 44 B 3.8-58 39 B 3.8-11 39 B 3.8-59 39 B 3.8-12 39 B 3.8-60 39 B 3.8-13 39 B 3.8-61 39 B 3.8-14 39 B 3.8-62 46 B 3.8-15 39 B 3.8-63 39 B 3.8-16 39 B 3.8-64 39 B 3.8-17 39 B 3.8-65 39 B 3.8-18 39 B 3.8-66 39 B 3.8-19 39 B 3.8-67 39 B 3.8-20 39 B 3.8-68 39 B 3.8-21 39 B 3.8-69 39 B 3.8-22 39 B 3.8-70 39 HATCH UNIT 2 v REVISION 46

HATCH UNIT 2 BASES EFFECTIVE PAGE LIST (continued)

PAGE REVISION PAGE REVISION B 3.8-71 39 B 3.8-72 39 B 3.8-73 39 B 3.8-74 39 B 3.8-75 39 B 3.8-76 39 B 3.8-77 39 B 3.8-78 39 B 3.8-79 39 B 3.8-80 39 B 3.8-81 39 B 3.8-82 39 B 3.8-83 39 B 3.8-84 39 B 3.8-85 39 B 3.9-7 29 B 3.9-16 28 B 3.9-17 28 B 3.9-19 28 B 3.9-20 28 B 3.9-21 1 B 3.9-22 1 B 3.9-23 46 B 3.9-24 1 B 3.9-25 1 B 3.9-26 1 B 3.9-27 1 B 3.9-28 46 B 3.10-1 5 B 3.10-15 1 B 3.10-23 1 B 3.10-24 1 B 3.10-27 1 B 3.10-30 14 B 3.10-31 14 B 3.10-32 14 B 3.10-33 1 HATCH UNIT 2 vi REVISION 46

SDM B 3.1.1 BASES ACTIONS E.1, E.2. E.3. E.4. and E.5 (continued) information, to determine if the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, SRs may need to be performed to restore the component to OPERABLE status.

Action must continue until all required components are OPERABLE.

SURVEILLANCE SR 3.1.1.1 REQUIREMENTS Adequate SDM must be verified to ensure that the reactor can be made subcritical from any initial operating condition. This can be accomplished via a test, an evaluation, or a combination of the two.

Adequate SDM is demonstrated by testing before or during the first startup after fuel movement or shuffling within the reactor pressure vessel, or control rod replacement. Control rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod or a control rod from another core location. Since core reactivity will vary during the cycle as a function of fuel depletion and poison burnup, the beginning of cycle (BOC) test must also account for changes in core reactivity during the cycle. Therefore, to obtain the SDM, the initial value must be changed by the value, 'R", which is the difference between the calculated value of minimum SDM during the operating cycle and the calculated BOC SDM. If the value of R is positive (that is, BOC is the point in the cycle with the minimum SDM), no correction to the BOC measured value is required (Ref. 7). For the SDM demonstrations where the highest worth rod is determined solely on calculation, additional margin (0.10% Ak/k) must be added to the SDM limit of 0.28% Ak/k to account for uncertainties in the calculation of the highest worth control rod.

The SDM may be demonstrated during an in-sequence control rod withdrawal, in which the highest worth control rod is analytically determined, or during local criticals, where the highest worth control rod is determined by testing. Local critical tests require the withdrawal of out of sequence control rods. This testing would therefore require bypassing of the Rod Worth Minimizer to allow the out of sequence withdrawal, and therefore additional requirements must be met (see LCO 3.10.7, "Control Rod Testing - Operating").

(continued)

HATCH UNIT 2 B 3.1-5 REVISION 1

SDM B 3.1.1 BASES SURVEILLANCE SR 3.1.1.1 (continued)

REQUIREMENTS The Frequency of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reaching criticality is allowed to provide a reasonable amount of time to perform the required calculations and have appropriate verification.

During MODE 5, adequate SDM is required to ernsure that the reactor does not reach criticality during control rod withdrawals. An evaluation of each in-vessel fuel movement during fuel loading (including shuffling fuel within the core) is required to ensure adequate SDM is maintained during refueling. This evaluation ensures that the intermediate loading patterns are bounded by the safety analyses for the final core loading pattern. For example, bounding analyses that demonstrate adequate SDM for the most reactive configurations during the refueling may be performed to demonstrate acceptability of the entire fuel movement sequence. These bounding analyses include additional margins to the SDM limit to account for the associated uncertainties. Spiral offload/reload sequences inherently satisfy the SR, provided the fuel assemblies are reloaded in the same configuration analyzed for the new cycle. Removing fuel from the core will always result in an increase in SDM.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.

2. FSAR, Subsection 15.1.38.

3. NEDE-2401 1-P-A-US, "General Electric Standard Application for Reactor Fuel," Supplement for United States, (revision specified in the COLR).

4. FSAR, Subsection 15.1.13.

5. FSAR, Subsection 15.1.14.

6. FSAR, Paragraph 4.3.2.4.1.

7. NEDE-2401 1-P-A, "General Electric Standard Application for Reactor Fuel," (revision specified in the COLR).

8. Technical Requirements Manual, Section 8.0.

9. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

HATCH UNIT 2 B 3.1 -6 REVISION 46

Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE SR 3.1.4.2 (continued)

REQUIREMENTS done on the CRDs at more frequent intervals in accordance with LCO 3.1.3 and LCO 3.1.5, "Control Rod Scram Accumulators."

SR 3.1.4.3 When work that could affect the scram insertion time is performed on a control rod or the CRD System, testing must be done to demonstrate that each affected control rod retains adequate scram performance over the range of applicable reactor pressures from zero to the maximum permissible pressure. The scram testing must be performed once before declaring the control rod OPERABLE. The required scram time testing must demonstrate the affected control rod is still within acceptable limits. The limits for reactor pressures < 800 psig, required by footnote (b), are included in the Technical Requirements Manual (Ref. 7) and are established based on a high probability of meeting the acceptance criteria at reactor pressures 2 800 psig. The limits for reactor pressures 2 800 psig are found in Table 3.1.4-1. If testing demonstrates the affected control rod does not meet these limits, but is within the 7 second limit of Table 3.1.4-1, Note 2, the control rod can be declared OPERABLE and "slow.'

Specific examples of work that could affect the scram times are (but are not limited to) the following: removal of any CRD for maintenance or modification; replacement of a control rod; and maintenance or modification of a scram solenoid pilot valve, scram valve, accumulator, isolation valve or check valve in the piping required for scram.

The Frequency of once prior to declaring the affected control rod OPERABLE is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY.

SR 3.1.4.4 When work that could affect the scram insertion time is performed on a control rod or CRD System, testing must be done to demonstrate each affected control rod is still within the limits of Table 3.1.4-1 with the reactor steam dome pressure 2 800 psig. Where work has been performed at high reactor pressure, the requirements of SR 3.1.4.3 and SR 3.1.4.4 can be satisfied with one test. However, for a control rod affected by work performed while shutdown, a zero pressure test and a high pressure test may be required. This testing ensures that, (continued)

HATCH UNIT 2 B 3.1-23 REVISION 0

Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE SR 3.1.4.4 (continued)

REQUIREMENTS prior to withdrawing the control rod for continued operation, the control rod scram performance is acceptable for operating reactor pressure conditions. Alternatively, a control rod scram test during hydrostatic pressure testing could also satisfy both criteria.

The Frequency of once prior to exceeding 40% RTP is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY.

This test is also used to demonstrate control rod OPERABILITY when 2 40% RTP after work that could affect the scram insertion time is performed on the CRD System.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. FSAR, Paragraph 4.2.3.2.

3. FSAR, Supplement 5A.4.3.

4. FSAR, Section 15.1.

5. NEDE-2401 1-P-A, 'General Electric Standard Application for Reactor Fuel,' (revision specified in the COLR).

6. Letter from R. F. Janecek (BWROG) to R. W. Starostecki (NRC), "BWR Owners' Group Revised Reactivity Control Systems Technical Specifications," BWROG-8754, September 17,1987.

7. Technical Requirements Manual, Table T5.0-1.

8. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements,' July 23,1993.

HATCH UNIT 2 B 3.1-24 REVISION 46

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.17 (continued)

REQUIREMENTS setpoints. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint. The actual Surveillance ensures that the OPRM Upscale Function is enabled (not bypassed) for the correct values of APRM Simulated Thermal Power and recirculation drive flow. Other Surveillances-ensure that the APRM Simulated Thermal Power and recirculation flow properly correlate with THERMAL POWER and core flow, respectively.

If any bypass setpoint is nonconservative (i.e., the OPRM Upscale Function is bypassed when APRM Simulated Thermal Power is a 25%

and recirculation drive flow is < 60% rated), then the affected channel is considered inoperable for the OPRM Upscale Function.

Alternatively, the bypass setpoint may be adjusted to place the channel in a conservative condition (unbypass). If placed in the unbypass condition, this SR is met and the channel is considered OPERABLE.

The 24 month Frequency is based on a review of the surveillance test history and Reference 20.

REFERENCES 1. FSAR, Section 7.2.

2. FSAR, Chapter 15.

3. FSAR, Subsection 6.3.3.

4. FSAR, Supplement 5A.

5. FSAR, Subsection 15.1.12.

6. NEDO-23842, "Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.

7. FSAR, Subsection 15.1.38.

8. P. Check (NRC) letter to G. Lainas (NRC), 'BWR Scram Discharge System Safety Evaluation," December 1, 1980.

9. NEDO-30851-P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System," March 1988.

10. Technical Requirements Manual, Table T5.0-1.

(continued)

HATCH UNIT 2 El 3.3-31 REVISION 46

RPS Instrumentation B 3.3.1.1 BASES REFERENCES 11. NRC No.93-102, "Final Policy Statement on Technical (continued) Specification Improvements," July 23,1993.

12. NEDO-32291, "System Analyses for Elimination of Selected Response Time Testing Requirements," January 1994.

13. NEDC-3241OP-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option IlIl Stability Trip Function," October 1995.

14. NEDO-31960-A, 'BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

15. NEDO-31960-A, Supplement 1, BWR Owners' Group Long-Term Stability Solutions Licensing Methodology,"

November 1995.

16. NEDO-32465-A, "BWR Owners' Group Long-Term Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications," March 1996.

17. NEDO-3241OP-A, Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option IlIl Stability Trip Function,"

November 1997.

18. Letter, L.A. England (BWROG) to M.J. Virgilio, "BWR Owners' Group Guidelines for Stability Interim Corrective Action,"

June 6, 1994.

19. NEDO-32291 -A, Supplement 1, "System Analyses for the Elimination of Selected Response Time Testing Requirements," October 1999.

20. NRC Safety Evaluation Report for Amendment 174.

21. GE Letter NSA 02-250, "Plant Hatch IRM Technical Specifications," April 19, 2002.

22. NRC Safety Evaluation Report for Amendment 176, Quarterly Surveillance Extension.

HATCH UNIT 2 B 3.3-32 REVISION 38

Remote Shutdown System B 3.3.3.2 BASES ACTIONS B.1 (continued)

If the Required Action and associated Completion Time of Condition A are not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Time is reasonable, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE The Surveillances are modified by a Note to indicate that when an REQUIREMENTS instrument channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. The Note is based upon a NRC Safety Evaluation Report (Reference 1) which concluded that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability of monitoring required parameters, when necessary.

SR 3.3.3.2.1 Performance of the CHANNEL CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. As specified in the Surveillance, a CHANNEL CHECK is only required for those channels that are normally energized.

(continued)

HATCH UNIT 2 B 3.3-73 REVISION 1 l

Remote Shutdown System B 3.3.3.2 BASES SURVEILLANCE SR 3.3.3.2.1 (continued)

REQUIREMENTS The Frequency is based upon plant operating experience that demonstrates channel failure is rare.

SR 3.3.3.2.2 SR 3.3.3.2.2 verifies each required Remote Shutdown System transfer switch and control circuit performs the intended function. This verification is performed from the remote shutdown panel and locally, as appropriate. Operation of equipment from the remote shutdown panel is not necessary. The Surveillance can be satisfied by performance of a continuity check, or in the case of the DG controls, the routine Surveillances of LCO 3.8.1 (since local control is utilized during the performance of some of the Surveillances of LCO 3.8.1).

This will ensure that if the control room becomes inaccessible, the plant can be placed and maintained in MODE 3 from the remote shutdown panel and the local control stations. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 24 month Frequency is based on a review of the surveillance test history and Reference 4.

SR 3.3.3.2.3 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies the channel responds to measured parameter values with the necessary range and accuracy.

The 24 month Frequency is based on a review of the surveillance test history and Reference 4.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 19.

2. Technical Requirements Manual, Table T6.0-1.

3. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

4. NRC Safety Evaluation Report for Amendment 174.

HATCH UNIT 2 B 3.3-74 REVISION 46

EOC-RPT Instrumentation B 3.3.4.1 BASES SURVEILLANCE SR 3.3.4.1.5 (continued)

REQUIREMENTS STAGGERED TEST BASIS, is also based on a review of the surveillance test history and Reference 7.

SR 3.3.4.1.6 This SR ensures that the RPT breaker interruption time is provided to the EOC-RPT SYSTEM RESPONSE TIME test. Breaker interruption (i.e., trip) time is defined as breaker response time plus arc suppression time. Breaker response time is the time from application of voltage to the trip coil until the main contacts separate. Arc suppression time is the time from main contact separation until the complete suppression of the electrical arc across the open contacts.

Breaker response shall be verified by testing and added to the manufacturer's design arc suppression time to determine breaker interruption time. The breaker arc suppression time shall be validated by the performance of periodic contact gap measurements in accordance with plant procedures. The 60 month Frequency of the testing is based on the difficulty of performing the test and the reliability of the circuit breakers.

REFERENCES 1. FSAR, Subsection 7.6.10.

2. FSAR, Subsections 15.1.1, 15.1.2, and 15.1.3.

3. FSAR, Paragraph 5.5.1 6.1 and Subsection 7.6.10.

4. GENE-770-06-1, "Bases For Changes To Surveillance Test Intervals And Allowed Out-Of-Service Times For Selected Instrumentation Technical Specifications," February 1991.

5. Technical Requirements Manual, Table T5.0-1.

6. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

7. NRC Safety Evaluation Report for Amendment 174.

8. NRC Safety Evaluation Report for Amendment 176, Quarterly Surveillance Extension.

HATCH UNIT 2 B 3.3-83 REVISION 46

ATWS-RPT Instrumentation B 3.3.4.2 B 3.3 INSTRUMENTATION B 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT)

Instrumentation BASES BACKGROUND The ATWS-RPT System initiates an RPT, adding negative reactivity, following events in which a scram does not (but should) occur, to lessen the effects of an ATWS event. Tripping the recirculation pumps adds negative reactivity from the increase in steam voiding in the core area as core flow decreases. When Reactor Vessel Water Level - ATWS-RPT Level or Reactor Steam Dome Pressure - High I setpoint is reached, the recirculation pump drive motor breakers trip.

The ATWS-RPT System (Ref. 1) includes sensors, relays, bypass capability, circuit breakers, and switches that are necessary to cause initiation of an RPT. The channels include electronic equipment (e.g.,

trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an ATWS-RPT signal to the trip logic.

The ATWS-RPT consists of two independent trip systems, with two channels of Reactor Steam Dome Pressure - High and two channels of Reactor Vessel Water Level - ATWS-RPT Level in each trip system. Each ATWS-RPT trip system is a two-out-of-two logic for each Function. Thus, either two Reactor Water Level - ATWS-RPT I Level or two Reactor Pressure - High signals are needed to trip a trip system. The outputs of the channels in a trip system are combined in a logic so that either trip system will trip both recirculation pumps (by tripping the respective drive motor breakers).

There is one drive motor breaker provided for each of the two recirculation pumps for a total of two breakers. The output of each trip system is provided to both recirculation pump breakers.

APPLICABLE The ATWS-RPT is not assumed in the safety analysis. The SAFETY ANALYSES ATWS-RPT initiates an RPT to aid in preserving the integrity of the LCO, and fuel cladding following events in which a scram does not, but should, APPLICABILITY occur. Based on its contribution to the reduction of overall plant risk, however, the instrumentation meets Criterion 4 of the NRC Policy Statement (Ref. 3).

(continued)

HATCH UNIT 2 B 3.3-84 REVISION 3

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES SURVEILLANCE SR 3.3.6.1.2 (continued)

REQUIREMENTS function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.

The 92 day on an ALTERNATE TEST BASIS Frequency is based on a review of the surveillance test history, drift analysis of the associated trip units (if applicable), and Reference 10.

SR 3.3.6.1.3, SR 3.3.6.1.4. and SR 3.3.6.1.5 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the plant specific setpoint methodology.

The 92 day on an ALTERNATE TEST BASIS Frequency of SR 3.3.6.1.3 is based on a review of the surveillance test history, drift analysis of the associated pressure (or vacuum) switches (if applicable), and Reference 10. The 184 day Frequency of SR 3.3.6.1.4 and the 24 month Frequency of SR 3.3.6.1.5 are based on a review of the surveillance test history, drift analysis of the associated instrumentation (if applicable), and Reference 9.

SR 3.3.6.1.6 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel.

The system functional testing performed on PCIVs in LCO 3.6.1.3 overlaps this Surveillance to provide complete testing of the assumed safety function. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 24 month Frequencyis based on a review of the surveillance test history and Reference 9.

SR 3.3.6.1.7 This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident (continued)

HATCH UNIT 2 B 3.3-159 REVISION 38

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES SURVEILLANCE SR 3.3.6.1.7 (continued)

REQUIREMENTS analysis. The instrument response times must be added to the PCIV closure times to obtain the ISOLATION SYSTEM RESPONSE TIME.ISOLATION SYSTEM RESPONSE TIME acceptance criteria are included in Reference 6. This test may be performed in one measurement, or in overlapping segments, with verification that all components are tested.

A Note to the Surveillance states that channel sensors are excluded from ISOLATION SYSTEM RESPONSE TIME testing. The exclusion of the channel sensors is supported by Reference 8 which indicates that the sensors' response times are a small fraction of the total response time. Even if the sensors experienced response time degradation, they would be expected to respond in the microsecond to millisecond range until complete failure.

ISOLATION SYSTEM RESPONSE TIME tests are conducted on a 24 month STAGGERED TEST BASIS. This Frequency is consistent with the typical industry refueling cycle and is based upon plant operating experience that shows that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences.

The 24 month Frequency, on a STAGGERED TEST BASIS, is also based on a review of the surveillance test history and Reference 9.

REFERENCES 1. FSAR, Section 6.3.

2. FSAR, Chapter 15.

3. FSAR, Paragraph 4.2.3.4.2.

4. NEDC-31677P-A, "Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation,"

July 1990.

5. NEDC-30851 P-A Supplement 2, "Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation," March 1989.

6. Technical Requirements Manual, Table T5.0-1.

7. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

(continued)

HATCH~~~ UNT2B3316_EIIN4 HATCH UNIT 2 B 3.3-160 REVISION 46

ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.12 REQUIREMENTS (continued) The pneumatic actuator of each ADS valve is stroked to verify that the pilot disc rod lifts when the actuator strokes. Pilot rod lift is determined by measurement of rod travel. The total amount of lift of the pilot rod from the valve closed position to the open position shall meet criteria established by the S/RV supplier. SRs 3.5.1.1 1 and 3.3.5.1.5 overlap this SR to provide testing of the S/RV relief mode function. Additional functional testing is performed by tests required by the ASME OM Code (Ref. 17).

The 24 month Frequency is based on a review of the surveillance test history and Reference 18.

SR 3.5.1.13 This SR ensures that the ECCS RESPONSE TIMES are less than or equal to the maximum values assumed in the accident analysis.

Response time testing acceptance criteria are included in Reference 14. A Note to the Surveillance states that the instrumentation portion of the response time may be assumed from established limits. The exclusion of the instrumentation from the response time surveillance is supported by Reference 15, which concludes that instrumentation will continue to respond in the microsecond to millisecond range prior to complete failure.

The 24 month Frequency is based on the need to perform the I Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 24 month Frequency is based on a review of the surveillance test history and Reference 18.

REFERENCES 1. FSAR, Paragraph 6.3.2.2.3.

2. FSAR, Paragraph 6.3.2.2.4.

3. FSAR, Paragraph 6.3.2.2.1.

4. FSAR, Paragraph 6.3.2.2.2.

5. FSAR, Subsection 15.1.39.

6. FSAR, Subsection 15.1.40.

(continued)

HATCH UNIT 2 B 3.5-1 3 REVISION 35

ECCS - Operating B 3.5.1 BASES REFERENCES 7. FSAR, Subsection 15.1.33.

(continued)

8. 10 CFR 50, Appendix K.

9. FSAR, Subsection 6.3.3.

10. NEDC-31376P, "E.I. Hatch Nuclear Plant Units 1 and 2 SAFERIGESTR-LOCA Loss-of-Coolant Analysis," December 1986.

11. 10 CFR 50.46.

12. Memorandum from R. L. Baer (NRC) to V. Stello, Jr. (NRC),

'Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.

13. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

14; Technical Requirements Manual, Table T5.0-1.

15. NEDO-32291, NSystem Analyses for Elimination of Selected Response Time Testing Requirements," January 1994.

16. NEDC-32041 P, 'Safety Review for Edwin I. Hatch Nuclear Power Plant Units 1 and 2 Updated Safety/Relief Valve Performance Requirements," April 1996.

17. ASME, OM Code - 1995, "Code for Operation and Maintenance of Nuclear Power Plants," Appendix I.

18. NRC Safety Evaluation Report for Amendment 174.

HATCH UNIT 2 B 3.5-14 'REVISION 46

ECCS - Shutdown B 3.5.2 BASES SURVEILLANCE SR 3.5.2.1 and SR 3.5.2.2 (continued)

REQUIREMENTS water level variations and instrument drift during the applicable MODES. Furthermore, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal suppression pool or CST water level condition.

SR 3.5.2.3. SR 3.5.2.5. and SR 3.5.2.6 The Bases provided for SR 3.5.1.1, SR 3.5.1.7, and SR 3.5.1.10 are applicable to SR 3.5.2.3, SR 3.5.2.5, and SR 3.5.2.6, respectively.

However, the LPCI flow rate requirement for SR 3.5.2.5 is based on a single pump, not the two pump flow rate requirement of SR 3.5.1.7.

SR 3.5.2.4 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

The 31 day Frequency is appropriate because the valves are operated under procedural control and the probability of their being mispositioned during this time period is low.

In MODES 4 and 5, the RHR System may operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Therefore, RHR valves that are required for LPCI subsystem operation may be aligned for decay heat removal. Therefore, this SR is modified by a Note that allows one LPCI subsystem of the RHR System to be considered OPERABLE for the ECCS function if all the required valves in the LPCI flow path can be manually realigned (remote or local) to allow injection into the RPV, and the system is not otherwise inoperable. This will ensure adequate core cooling if an inadvertent RPV draindown should occur.

(continued)

HATCH UNIT 2 B 3.5-1 9 REVISION 1 I

ECCS - Shutdown B 3.5.2 BASES (continued)

REFERENCES 1. NEDC-31376P, -E.I. Hatch Nuclear Plant Units 1 and 2 SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis,"

December 1986.

2. Technical Requirements Manual, Section 8.0. I

3. NRC No.93-102, Final Policy Statement on Technical Specification Improvements,N July 23, 1993.

HATCH UNIT 2 B 3.5-20 REVISION 46

PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.13 REQUIREMENTS (continued) This SR provides assurance that the excess flow isolation dampers can close following an isolation signal. The 24 month Frequency is based on a review of the surveillance test history and Reference 9.

REFERENCES 1. FSAR, Chapter 15.

2. Technical Requirements Manual, Table T7.0-1. I

3. FSAR, Subsection 15.1.39.

4. FSAR, Section 6.2.

5. 10 CFR 50, Appendix J, Option B.

6. NRC No.93-102, 'Final Policy Statement on Technical Specification Improvements," July 23, 1993.

7. Primary Containment Leakage Rate Testing Program.

8. NEDO-32977-A, "Excess Flow Check Valve Testing Relaxation."

9. NRC Safety Evaluation Report for Amendment 174.

HATCH UNIT 2 B 3.6-27 REVISION 46

Drywell Pressure B 3.6.1.4 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.4 Drywell Pressure BASES BACKGROUND The drywell pressure is limited during normal operations to preserve the initial conditions assumed in the accident analysis for a Design Basis Accident (DBA) or loss of coolant accident (LOCA).

APPLICABLE Primary containment performance is evaluated for the entire SAFETY ANALYSES spectrum of break sizes for postulated LOCAs (Ref. 1). Among the inputs to the DBA is the initial primary containment internal pressure (Ref. 1). Analyses assume an initial drywell pressure of 1.75 psig.

This limitation ensures that the safety analysis remains valid by maintaining the expected initial conditions and ensures that the peak LOCA drywell internal pressure does not exceed the maximum allowable of 62 psig.

The maximum calculated drywell pressure occurs during the reactor blowdown phase of the DBA, which assumes an instantaneous recirculation line break. The calculated peak drywell pressure for this limiting event is 46.9 psig (Ref. 1).

Drywell pressure satisfies Criterion 2 of the NRC Policy Statement (Ref. 2).

LCO In the event of a DBA, with an initial drywell pressure < 1.75 psig, the resultant peak drywell accident pressure will be maintained below the drywell design pressure.

APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, maintaining drywell pressure within limits is not required in MODE 4 or 5.

(continued)

HATCH UNIT 2 B 3.6-28 REVISION 21

Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.8 BASES SURVEILLANCE SR 3.6.1.8.1 (continued)

REQUIREMENTS Vacuum Breaker Position Indication," as ACTIONS for inoperable closed position indicator channels. In this case the vacuum breaker is assumed open until otherwise proved to satisfy the leakage test, and this confirmation must be performed within the Technical Specification 3.6.1.8, Required Action B.1 Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

The 14 day Frequency is based on engineering judgment, is considered adequate in view of other indications of vacuum breaker status available to operations personnel, and has been shown to be acceptable through operating experience.

A Note is added to this SR which allows suppression chamber-to-drywell vacuum breakers opened in conjunction with the performance of a Surveillance to not be considered as failing this SR. These periods of opening vacuum breakers are controlled by plant procedures and do not represent inoperable vacuum breakers.

SR 3.6.1.8.2 Each required (i.e., required to be OPERABLE for.opening) vacuum breaker must be cycled to ensure that it opens adequately to perform its design function and returns to the fully closed position. This ensures that the safety analysis assumptions are valid. The 31 day Frequency of this SR was developed, based on Inservice Testing Program requirements to perform valve testing at least once every 92 days. A 31 day Frequency was chosen to provide additional assurance that the vacuum breakers are OPERABLE, since they are located in a harsh environment (the suppression chamber airspace).

In addition, this functional test is required within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after a discharge of steam to the suppression chamber from the safety/relief valves.

SR 3.6.1.8.3 Verification of the vacuum breaker opening setpoint is necessary to ensure that the safety analysis assumption regarding vacuum breaker full open differential pressure of 0.5 psid is valid. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 24 month Frequency is based on a review of the surveillance test history and Reference 4. It is further justified (continued)

HATCH UNIT 2 B 3.6-45 REVISION 35

Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.8 BASES SURVEILLANCE SR 3.6.1.8.3 (continued)

REQUIREMENTS because of other surveillances performed at shorter Frequencies that convey the proper functioning status of each vacuum breaker.

REFERENCES 1. FSAR, Subsection 6.2.1.

2. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

3. Technical Requirements Manual, TLCO 3.6.1. I

4. NRC Safety Evaluation Report for Amendment 174.

I ,

HATCH UNIT 2 B 3.6-46 REVISION 46

Secondary Containment B 3.6.4.1 BASES ACTIONS C.1, C.2, and C.3 (continued) inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.6.4.1.1 and SR 3.6.4.1.2 REQUIREMENTS Verifying that secondary containment equipment hatches and one access door in each access opening are closed ensures that the infiltration of outside air of such a magnitude as to prevent maintaining the desired negative pressure does not occur. Verifying that all such openings are closed provides adequate assurance that exfiltration from the secondary containment will not occur. SR 3.6.4.1.1 also requires equipment hatches to be sealed. In this application, the term "sealed" has no connotation of leak tightness. Maintaining secondary containment OPERABILITY requires verifying one door in the access opening is closed. An access opening contains one inner and one outer door. The intent is not to breach the secondary containment at any time when secondary containment is required. This is achieved by maintaining the inner or outer portion of the barrier closed at all times. However, all secondary containment access doors are normally kept closed, except when the access opening is being used for entry and exit or when maintenance is being performed on an access opening. When the secondary containment configuration excludes Zone I and/or Zone II, these SRs also include verifying the hatches and doors separating the common refueling floor zone from the reactor building(s). The 31 day Frequency for these SRs has been shown to be adequate, based on operating experience, and is considered adequate in view of the other indications of door and hatch status that are available to the operator.

SR 3.6.4.1.3 and SR 3.6.4.1.4 The Unit 1 and Unit 2 SGT Systems exhausts the secondary containment atmosphere to the environment through appropriate treatment equipment. To ensure that all fission products are treated, SR 3.6.4.1.3 verifies that the appropriate SGT System(s) will rapidly establish and maintain a negative pressure in the secondary containment. This is confirmed by demonstrating that the required SGT subsystem(s) will draw down the secondary containment to 2 0.20 inch of vacuum water gauge in s 120 seconds. This cannot be accomplished if the secondary containment boundary is not intact.

SR 3.6.4.1.4 demonstrates that the required SGT subsystem(s) can (continued)

HATCH UNIT 2 B 3.6-81 REVISION 40

Secondary Containment B 3.6.4.1 BASES _

SURVEILLANCE SR 3.6.4.1.3 and SR 3.6.4.1.4 (continued)

REQUIREMENTS maintain 2 0.20 inch of vacuum water gauge for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at a flow rate s 4000 cfm for each SGT subsystem. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months test period allows secondary containment to be in thermal equilibrium at steady state conditions. Therefore, these two tests are used to ensure secondary containment boundary integrity. Since these SRs are secondary containment tests, they need not be performed with each SGT subsystem. The SGT subsystems are tested on a STAGGERED TEST BASIS, however, to ensure that in addition to the requirements of LCO 3.6.4.3, each SGT subsystem or combination of subsystems will perform this test. The number of SGT subsystems and the required combinations are dependent on the configuration of the secondary containment and are detailed in the Technical Requirements Manual (Ref. 3). The Note to SR 3.6.4.1.3 and SR 3.6.4.1.4 specifies that the number of required SGT subsystems be one less than the number required to meet LCO 3.6.4.3, Standby Gas Treatment (SGT) System," for the given configuration. The 24 month Frequency, on a STAGGERED TEST BASIS, of SRs 3.6.4.1.3 and 3.6.4.1.4 is also based on a review of the surveillance test history and Reference 5.

REFERENCES 1. FSAR, Subsection 15.1.39.

2. FSAR, Subsection 15.1.41.

3. Technical Requirements Manual, Section 8.0.

4. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

5. NRC Safety Evaluation Report for Amendment 174.

HATCH UNIT 2 B 3.6-82 REVISION 46

SCIVs B 3.6.4.2 K> BASES (continued)

REFERENCES . 1. FSAR, Subsection 15.1.39.

2. FSAR, Subsection 15.1.41.

3. Technical Requirements Manual, Section 8.0. I

4. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

5. NRC Safety Evaluation Report for Amendment 174.

HATCH UNIT 2 B 3.6-89 REVISION 46

SGT System B 3.6.4.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.3 Standby Gas Treatment (SGT) System BASES BACKGROUND The SGT System is required by 10 CFR 50, Appendix A, GDC 41, "Containment Atmosphere Cleanup" (Ref. 1). The function of the SGT System is to ensure that radioactive materials that leak from the primary containment into the secondary containment following a Design Basis Accident (DBA) are filtered and adsorbed prior to exhausting to the environment.

The Unit 1 and Unit 2 SGT Systems each consists of two fully redundant subsystems, each with its own set of dampers, charcoal filter train, and controls. The Unit 1 SGT subsystems' ductwork is separate from the inlet to the filter train to the discharge of the fan.

The rest of the ductwork is common. The Unit 2 SGT subsystems' ductwork is separate except for the suction from the drywell and torus, which is common (however, this suction path is not required for subsystem OPERABILITY).

Each charcoal filter train consists of (components listed in order of the direction of the air flow):

a. A demister or moisture separator;

b. An electric heater;

c. A prefilter;

d. A high efficiency particulate air (HEPA) filter;

e. Two charcoal adsorbers for Unit 1 subsystems and one charcoal adsorber for Unit 2 subsystems;

f. A second HEPA filter; and

g. An axial vane fan for Unit 1 subsystems and a centrifugal fan for Unit 2 subsystems.

The sizing of the SGT Systems equipment and components is based on the results of an infiltration analysis, as well as an exfiltration analysis of the secondary containment. The internal pressure of the SGT Systems boundary region is maintained at a negative pressure when the system is in operation, to conservatively ensure zero (continued)

HATCH UNIT 2 B 3.6-90 REVISION 4

SGT System B 3.6.4.3 BASES ACTIONS F.1, F.2. and F.3 (continued)

When two or more required SGT subsystems are inoperable, if applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies in secondary containment must immediately be suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

Required Action F.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.6.4.3.1 REQUIREMENTS Operating each required Unit 1 and Unit 2 SGT subsystem for 2 10 continuous hours ensures that they are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on for 2 10 continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls and the redundancy available in the system.

SR 3.6.4.3.2 This SR verifies that the required Unit 1 and Unit 2 SGT filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies and additional information are discussed in detail in the VFTP.

(continued)

HATCH UNIT 2 B 3.6-95 REVISION 1 1

SGT System B 3.6.4.3 BASES SURVEILLANCE SR 3.6.4.3.3 REQUIREMENTS (continued) This SR verifies that each required Unit 1 and Unit 2 SGT subsystem starts on receipt of an actual or simulated initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.2.5 overlaps this SR to provide complete testing of the safety function. This Surveillance can be performed with the reactor at power. The 24 month Frequency is based on a review of the surveillance test history and Reference 8.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 41.

2. Unit 1 FSAR, Section 5.3.

3. FSAR, Subsection 6.2.3.

4. FSAR, Subsection 15.1.39.

5. FSAR, Subsection 15.1.41.

6. Technical Requirements Manual, Section 8.0 I

7. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

8. NRC Safety Evaluation Report for Amendment 174.

HATCH UNIT 2 B 3.6-96 REVJISION 46

RHRSW System B 3.7.1 B 3.7 PLANT SYSTEMS B 3.7.1 Residual Heat Removal Service Water (RHRSW) System BASES BACKGROUND The RHRSW System is designed to provide cooling water for the Residual Heat Removal (RHR) System heat exchangers, required for a safe reactor shutdown following a Design Basis Accident (DBA) or transient. The RHRSW System is operated whenever the RHR heat exchangers are required to operate in the shutdown cooling mode or in the suppression pool cooling or spray mode of the RHR System.

The RHRSW System consists of two independent and redundant subsystems. Each subsystem is made up of a header, two 4000 gpm pumps, a suction source, valves, piping, heat exchanger, and associated instrumentation. Either of the two subsystems is capable of providing the required cooling capacity with two pumps operating to maintain safe shutdown conditions. The two subsystems are separated from each other by normally closed motor operated cross tie valves, so that failure of one subsystem will not affect the OPERABILITY of the other subsystem. The RHRSW System is designed with sufficient redundancy so that no single active component failure can prevent it from achieving its design function.

The RHRSW System is described in the FSAR, Subsection 9.2.7, Reference 1.

Cooling water is pumped by the RHRSW pumps from the Altamaha River through the tube side of the RHR heat exchangers, and discharges to the circulating water flume. A minimum flow line from the pump discharge to the intake structure prevents the pump from overheating when pumping against a closed discharge valve.

The system is initiated manually from the control room. If operating during a loss of coolant accident (LOCA) or a loss of offsite power (LOSP), the system is automatically tripped to allow the diesel generators to automatically power only that equipment necessary.

The system can be manually started any time the LOCA signal is manually overridden or clears. The system can be manually started any time after the LOSP signal is received.

(continued)

HATCH UNIT 2 B 3.7-1 REVISION 0

RHRSW System B 3.7.1 BASES (continued)

APPLICABLE The RHRSW System removes heat from the suppression pool to limit SAFETY ANALYSES the suppression pool temperature and primary containment pressure following a LOCA. This ensures that the primary containment can perform its function of limiting the release of radioactive materials to the environment following a LOCA. The ability of the RHRSW System to support long term cooling of the reactor or primary containment is discussed in the FSAR, Subsections 9.2.7, 7.4.5, and Chapter 15 (Refs. 1, 2 and 3, respectively). These analyses explicitly assume that the RHRSW System will provide adequate cooling support to the I

equipment required for safe shutdown. These analyses include the evaluation of the long term primary containment response after a design basis LOCA.

The safety analyses for long term cooling were performed for various combinations of RHR System failures. The worst case single failure that would affect the performance of the RHRSW System is any failure that would disable one subsystem of the RHRSW System. As discussed in the FSAR, Paragraph 6.2.1.4.3 (Ref. 4) for these analyses, manual initiation of the OPERABLE RHRSW subsystem and the associated RHR System is assumed to occur 10 minutes after a DBA. The RHRSW flow assumed in the analyses is 4000 gpm per pump with two pumps operating in one loop. In this case, the maximum suppression chamber water temperature and pressure are approximately 21 0F and 37 psig, respectively, well below the design temperature of 3400 F and maximum allowable pressure of 62 psig.

The RHRSW System satisfies Criterion 3 of the NRC Policy Statement (Ref. 5).

LCO Two RHRSW subsystems are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post accident heat loads, assuming the worst case single active failure occurs coincident with the loss of offsite power.

An RHRSW subsystem is considered OPERABLE when:

a. Two pumps are OPERABLE; and

b. An OPERABLE flow path is capable of taking suction from the intake structure and transferring the water to the RHR heat exchangers at the assumed flow rate. Additionally, the RHRSW cross tie valves (which allow the two RHRSW loops to be connected) must be closed so that failure of one subsystem will not affect the OPERABILITY of the other subsystems.

(continued)

HATCH UNIT 2 B 3.7-2 REVISION 46

RHRSW System B 3.7.1 BASES ACTIONS D.1 (continued)

With both RHRSW subsystems inoperable for reasons other than Condition B (e.g., both subsystems with inoperable flow paths, or one subsystem with an inoperable pump and one subsystem with an inoperable flow path), the RHRSW System is not capable of performing its intended function. At least one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time for restoring one RHRSW subsystem to OPERABLE status, is based on the Completion Times provided for the RHR suppression pool cooling and spray functions.

The Required Action is modified by a Note indicating that the applicable Conditions of LCO 3.4.7 be entered and Required Actions taken if an inoperable RHRSW subsystem results in an inoperable RHR shutdown cooling subsystem. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.

E.1 and E.2 If the RHRSW subsystems cannot be not restored to OPERABLE status within the associated Completion Times, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.1.1 REQUIREMENTS Verifying the correct alignment for each manual, power operated, and automatic valve in each RHRSW subsystem flow path provides assurance that the proper flow paths will exist for RHRSW operation.

This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position, and yet considered in the correct position, provided it can be realigned to its accident position. This is acceptable because the RHRSW System is a manually initiated system. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being (continued)

HATCH UNIT 2 B 3.7-5 REVISION 0

RHRSW System B 3.7.1 BASES SURVEILLANCE SR 3.7.1.1 (continued)

REQUIREMENTS mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions.

REFERENCES 1. FSAR, Subsection 9.2.7.

2. FSAR, Subsection 7.4.5. I

3. FSAR, Chapter 15.

4. FSAR, Paragraph 6.2.1.4.3.

5. NRC No.93-102, Final Policy Statement on Technical Specification Improvements," July 23, 1993.

HATCH UNIT 2 B 3.7-6 REVISION 46

MCREC System B 3.7.4 BASES (continued)

SURVEILLANCE SR 3.7.4.1 REQUIREMENTS This SR verifies that a subsystem in a standby mode starts on demand and continues to operate. Standby systems should be checked periodically to ensure that they start and function properly.

As the environmental and normal operating conditions of this system are not severe, testing each subsystem once evpry 31 days provides an adequate check on this system. Since the MCREC System does not have heaters, each subsystem need only be operated for 2 15 minutes to demonstrate the function of the subsystem.

Furthermore, the 31 day Frequency is based on the known reliability of the equipment and the two subsystem redundancy available.

SR 3.7.4.2 This SR verifies that the required MCREC testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).

Specific test frequencies and additional information are discussed in detail in the VFTP.

SR 3.7.4.3 This SR verifies that on an actual or simulated initiation signal, each MCREC subsystem starts and operates. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.7.1.4 overlaps this SR to provide complete testing of the safety function. This Surveillance can be performed with the reactor at power. The 24 month Frequency is based on a review of the surveillance test history and Reference 9.

SR 3.7.4.4 This SR verifies the integrity of the control room enclosure and the assumed inleakage rates of potentially contaminated air. The control room positive pressure, with respect to potentially contaminated adjacent areas (the turbine building), is periodically tested to verify proper function of the MCREC System. During the pressurization mode of operation, the MCREC System is designed to slightly pressurize the control room > 0.1 inches water gauge positive pressure with respect to the turbine building to prevent unfiltered inleakage. The MCREC System is designed to maintain this positive (continued)

HATCH UNIT 2 B 3.7-23 REVISION 35

- 5.1 -

MCREC System B 3.7.4 BASES _

SURVEILLANCE SR 3.7.4.4 (continued)

REQUIREMENTS pressure at a flow rate of s 2750 cfm through the control room in the pressurization mode. This SR ensures the total flow rate meets the design analysis value of 2500 cfm +/- 10% and ensures the outside air flow rate is < 400 cfm. The 24 month Frequency, on a STAGGERED TEST BASIS, is based on a review of the surveillance test history and Reference 9.

REFERENCES 1. FSAR, Section 6.4.

2. FSAR, Section 9.4.1.

3. FSAR, Chapter 6.

4. FSAR, Chapter 15.

5. FSAR, Section 6.4.1.2.2.

6. FSAR, Table 15.1-28.

7. NRC No.93-102, "Final Policy Statement an Technical Specification Improvements," July 23, 1993.

8. Technical Requirements Manual, Table T2.1-1. I

9. NRC Safety Evaluation Report for Amendment 174.

HATCH UNIT 2 B 3.7-24 REVISION 46

Control Room AC System B 3.7.5 BASES ACTIONS E.1. E.2.1. E.2.2, and E.2.3 (continued)

During movement of irradiated fuel assemblies in the secondary containment, during CORE ALTERATIONS, or during OPDRVs, if any Required Action and associated Completion Time for Condition B or C is not met, the necessary OPERABLE control room AC subsystems may be placed immediately in operation. One operable control room AC subsystem is necessary if the outside air temperature is s 650 F and the maximum outside air temperature in the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has been < 65 0F. If both of these conditions are not met, then two OPERABLE control room AC subsystems are necessary. This action ensures that the remaining subsystems are OPERABLE, that no failures that would prevent actuation will occur, and that any active failure will be readily detected.

An alternative to Required Action E.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.

If applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies in the secondary containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.

F.1 If three control room AC subsystems are inoperable in MODE 1, 2, or 3, the Control Room AC System may not be capable of performing the intended function. Therefore, LCO 3.0.3 must be entered immediately.

G.1, G.2. and G.3 The Required Actions of Condition G are modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not a sufficient reason to require a reactor shutdown.

(continued)

HATCH UNIT 2 B 3.7-29 REVISION 45 1

Control Room AC System B 3.7.5 BASES ACTIONS G.1, G.2. and G.3 (continued)

During movement of irradiated fuel assemblies in the secondary containment, during CORE ALTERATIONS, or during OPDRVs, with three control room AC subsystems inoperable, action must be taken immediately to suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.

If applicable, CORE ALTERATIONS and handling of irradiated fuel in the secondary containment must be suspended immediately.

Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.

SURVEILLANCE SR 3.7.5.1 REQUIREMENTS

,This SR verifies that the heat removal capability of the system is sufficient to remove the control room heat load assumed in the safety analysis. The SR consists of a combination of testing and calculation.

The 24 month Frequency is appropriate since significant degradation of the Control Room AC System is not expected over this time period.

The 24 month Frequency is based on a review of the surveillance test history and Reference 4.

REFERENCES 1. FSAR, Sections 6.4 and 9.4.1.

2. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

3. Technical Requirements Manual, Table T2.1-1. I

4. NRC Safety Evaluation Report for Amendment 174.

HATCH UNIT 2 B 3.7-30 REVISION 46

Main Turbine Bypass System B 3.7.7 BASES SURVEILLANCE SR 3.7.7.3 (continued)

REQUIREMENTS The 24 month Frequency is based on a review of the surveillance test history and Reference 5.

REFERENCES 1. FSAR, Section 7.7.4.

2. FSAR, Section 15.1.7.

3. Technical Requirements Manual, Table T5.0-1. I

4. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

5. NRC Safety Evaluation Report for Amendment 174.

HATCH UNIT 2 B 3.7-37 REVISION 46

Spent Fuel Storage Pool Water Level B 3.7.8 B 3.7 PLANT SYSTEMS B 3.7.8 Spent Fuel Storage Pool Water Level BASES BACKGROUND The minimum water level in the spent fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident.

A general description of the spent fuel storage pool design is found in the FSAR, Section 9.1.2 (Ref. 1). The assumptions of the fuel handling accident in the spent fuel storage pool are found in Reference 2.

APPLICABLE The water level above the irradiated fuel assemblies is an explicit SAFETY ANALYSES assumption of the fuel handling accident; the point from which the water level is measured is shown in Figure B 3.5.2-1. A fuel handling I accident in the spent fuel storage pool was evaluated (Ref. 2) and ensured that the radiological consequences (calculated whole body and thyroid doses at the exclusion area and low population zone boundaries) were well below the guideline doses of 10 CFR 100 (Ref. 3) and met the exposure guidelines of NUREG-0800 (Ref. 4).

A fuel handling accident could release a fraction of the fission product inventory by breaching the fuel rod cladding as discussed in the Regulatory Guide 1.25 (Ref. 5).

The fuel handling accident is evaluated for the dropping of an irradiated fuel assembly onto the spent fuel storage pool racks (Ref. 2). The water level in the spent fuel storage pool provides for absorption of water soluble fission product gases and transport delays of soluble and insoluble gases that must pass through the water before being released to the secondary containment atmosphere.

This absorption and transport delay reduces the potential radioactivity of the release during a fuel handling accident.

The spent fuel storage pool water level satisfies Criterion 2 of the NRC Policy Statement (Ref. 6).

LCO The specified water level preserves the assumptions of the fuel handling accident analysis (Ref. 2). As such, it is the minimum required for fuel movement within the spent fuel storage pool.

(continued)

HATCH UNIT 2 B 3.7-38 REVISION 28

DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.8 (continued)

REQUIREMENTS acceptance test. Initial conditions consistent with IEEE 450 need to be met prior to the performing of a battery performance discharge test. The test results reflect the overall effects of usage and age.

A battery modified performance discharge test is described in the Bases for SR 3.8.4.7. Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.4.8; however, only the modified performance discharge test may be used to satisfy SR 3.8.4.8, while satisfying the requirements of SR 3.8.4.7 at the same time.

The acceptance criteria for this Surveillance is consistent with IEEE-450 (Ref. 8) and IEEE-485 (Ref. 12). These references recommend that the battery be replaced if its capacity is below 80% of the manufacturer's rating. Although there may be ample capacity, the battery rate of deterioration is rapidly increasing.

The Frequency for this test is normally 60 months. If the battery shows degradation, or if the battery has reached 85% of its expected application service life and capacity is s 100% of the manufacturers rating, the Surveillance Frequencyis reduced to 12 months. However, if the battery shows no degradation but has reached 85% of its expected application service life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity 2 100% of the manufacturer's rating. Degradation is indicated, according to IEEE-450 (Ref. 8), when the battery capacity drops by more than 10% of rated capacity from its capacity on the previous performance test or is more than 10% below the manufacturer's rating. All these Frequencies are consistent with the recommendations in IEEE-450 (Ref. 8).

This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required DC electrical power subsystem from service, perturb the electrical distribution system, and challenge safety systems. Credit may be taken for unplanned events that satisfy the Surveillance. The swing DG DC battery is exempted from this restriction, since it is required by both units' LCO 3.8.4 and cannot be performed in the manner required by the Note without resulting in a dual unit shutdown.

(continued)

HATCH UNIT 2 B 3.8-61 REVISION 39

DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.9 REQUIREMENTS (continued) With the exception of this Surveillance, all other Surveillances of this Specification (SR 3.8.4.1 through SR 3.8.4.8) are applied only to the Unit 2 DC sources. This Surveillance is provided to direct that the appropriate Surveillances for the required Unit 1 DC sources are governed by the Unit 1 Technical Specifications. Performance of the applicable Unit 1 Surveillances will satisfy both any Unit 1 requirements, as well as satisfying this Unit 2 SR.

The Frequency required by the applicable Unit 1 SR also governs performance of that SR for both Units.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 17.

2. Regulatory Guide 1.6.

3. IEEE Standard 308-1971.

4. FSAR, Paragraphs 8.3.2.1.1 and 8.3.2.1.2.

5. FSAR, Chapter 6.

6. FSAR, Chapter 15.

7. Regulatory Guide 1.93, December 1974.

8. IEEE Standard 450-1987.

9. Technical Requirements Manual, Section 9.0. I

10. Regulatory Guide 1.32, February 1977.

11. Regulatory Guide 1.129, December 1974.

12. IEEE Standard 485-1983.

13. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements,' July 23, 1993.

14. NRC Safety Evaluation Report for Amendment 174.

HATCH UNIT 2 B 3.8-62 REVISION 46

RHR - High Water Level B 3.9.7 K)~ BASES (continued)

REFERENCES 1. 10 CFR 50, Appendix A, GDC 34.

2. Technical Requirements Manual, Section 8.0. I

3. NRC No.93-102, 'Final Policy Statement on Technical Specification Improvements," July 23, 1993.

HATCH UNIT 2 B 3.9-23 REVISION 46

_I1 RHR - Low Water Level B 3.9.8 B 3.9 REFUELING OPERATIONS B 3.9.8 Residual Heat Removal (RHR) - Low Water Level BASES BACKGROUND The purpose of the RHR System in MODE 5 is to remove decay heat and sensible heat from the reactor coolant, as required by GOC 34 (Ref. 1). Each of the two shutdown cooling loops of the RHR System can provide the required decay heat removal. Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via the associated recirculation loop. The RHR heat exchangers transfer heat to the RHR Service Water System. The RHR shutdown cooling mode is manually controlled.

APPLICABLE With the unit in MODE 5, the RHR System is not required to SAFETY ANALYSES mitigate any events or accidents evaluated in the safety analyses.

The RHR System is required for removing decay heat to maintain the temperature of the reactor coolant.

The RHR System satisfies Criterion 4 of the NRC Policy Statement (Ref. 3).

LCO In MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level < 22 ft 1/8 inches above the RPV flange, two RHR shutdown cooling subsystems must be OPERABLE.

An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump and the associated heat exchanger, an RHRSW pump providing cooling to the heat exchanger with sufficient flow to maintain reactor coolant temperature in the desired range, valves, piping, instruments, and controls to ensure an OPERABLE flow path. The two required RHR shutdown cooling subsystems have a common suction source and are allowed to have a common heat exchanger and common discharge piping.

Since the piping and heat exchangers are passive components that are assumed not to fail, they are allowed to be common to both subsystems. Thus, to meet the LCO, both RHR pumps in one loop or one RHR pump in each of the two loops must be OPERABLE. If the I (continued)

HATCH UNIT 2 B 3.9-24 REVISION 1

RHR - Low Water Level B 3.9.8 BASES ACTIONS B.1. B.2, and B.3 (continued)

2) sufficient standby gas treatment subsystem(s) are OPERABLE to maintain the secondary containment at a negative pressure with respect to the environment (dependent on secondary containment configuration, refer to Reference 2; single failure protection is not required while in this ACTION); and 3) secondary containment isolation capability is available in each associated secondary containment penetration flow path not isolated that is assumed to be isolated to mitigate radioactive releases (i.e., one secondary containment isolation valve and associated instrumentation are OPERABLE or other acceptable administrative controls to assure isolation capability. The administrative controls can consist of stationing a dedicated operator, who is in continuous communication with the control room, at the controls of the isolation device. In this way, the penetration can be rapidly isolated when a need for secondary containment isolation is indicated.). This may be performed as an administrative check, by examining logs or other information to determine whether the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, the Surveillance may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.

C.1 and C.2 If no RHR shutdown cooling subsystem is in operation, an alternate method of coolant circulation is required to be established within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The Completion Time is modified such that the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is applicable separately for each occurrence involving a loss of coolant circulation.

During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling subsystem), the reactor coolant temperature must be periodically monitored to ensure proper functioning of the alternate method. The once per hour Completion Time is deemed appropriate.

(continued)

HATCH UNIT 2 B 3.9-27 REVISION 1

RHR - Low Water Level B 3.9.8 BASES (continued)

SURVEILLANCE SR 3.9.8.1 REQUIREMENTS This Surveillance demonstrates that one required RHR shutdown cooling subsystem is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystems in the control room.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 34.

2. Technical Requirements Manual, Section 8.0. I

3. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

HATCH UNIT 2 B 3.9-28 REVISION 46

NL-04-0981, Transmittal of Quarterly Update for Hatch, Units 1 and 2 Technical Specifications Bases

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NL-04-0981, Transmittal of Quarterly Update for Hatch, Units 1 and 2 Technical Specifications Bases

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