ML20246A679
| ML20246A679 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, McGuire  |
| Issue date: | 06/30/1989 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20246A672 | List: |
| References | |
| NUDOCS 8907070060 | |
| Download: ML20246A679 (34) | |
Text
,
.d . ;c-
./
jf >
b -
DUKE POWER COMPANY
- o. MCGUIRE NUCLEAR STATION DESCRIPTION OF ADDITIONS TO AND DEVIATIONS FROM GENERIC EMERGENCY RESPONSE GUIDELINES June 1989 i
Nuclear Engineering Section Engineering Support Division Design Engineering Department i
i l
8907070060 890628 i PDR ADOCK 05000369 i F PDC 1
'd 4-
- ,(
TABLE OF CONTENTS
"- I. INTRODUCTION II. PLANT-SPECIFIC DESIGN DIFFERENCES III. GUIDELINE SET ADDITIONS AND DEVIATIONS IV. GUIDELINE TECHNICAL ADDITIONS AND DEVIATIONS I
l l
1 1
l l
l l
l l
o .
I. INTRODUCTION The purpose of this document is to describe and justify significant additions to or safety-significant deviations from the accident mitigation strategies in the generic Westinghouse Owners Group (WOG)
Emergency Response Guidelines (ERGS), when compared to the McGuire Nuclear Station plant-specific Emergency Procedure Guidelines (EPGs). The current ERG-HP Revision 1A is used as the standard for the original version of this document. However, as additional ERG revisions are released, the standard for comparison will become the latest version. The current McGuire EPGs in effect as of June 1, 1989 are used for the original version of this document. Should future EPG revisions impact this document it will be revised accordingly.
The EPGs are based on Revision 1 of the ERGS. Development of the EPGs was in part due to the plant-specific design differences between McGuire and the refe.rence Westinghouse plant which was the basis for the ERGS. In addition, development of the EPGs includes additions to, deletions from, and restructuring of the generic ERGS.
These enhancements or " deviations" were implemented in order to upgrade the comprehensiveness and usefulness of the station emergency procedures. The bases and justification for these deviations originated from the following:
o Preference for the ERG Revision 0 approach over l Revision 1 in some guidelines e Engineering evaluations e Operating philosophy e Operating experience l
e Experience with other vendor guidelines e Verification and validation activities The description and justification of the deviations that follow will demonstrate that the overall intent, structure, and format of the generic guidelines have been preserved and in many aspects improved.
1
_ _ . __ ______-________w
4 e The associated verification and validation activities have confirmed the correctness and appropriateness of the EPGs. A maintenance program has also been developed and implemented to address and incorporate potential EPG revisions as new information relevant to transient and accident mitigation and recovery is generated by the NSSS vendor, NRC, INPO, industry operating experience, and from within Duke Power Company.
The interpretation of what constitutes a " safety-significant deviation" from the NRC approved generic technical guidelines, is based on the description given in the Standard Review plan (NUREG-0800), Section 13.5.2, Appendix A, Section 3.3.2, Revision 0-July 1985. Although this guidance is fairly explicit, there remains a good amount of engineering judgement, and therefore a degree of subjectivity, when applying it. An attempt was made to focus on the accident mitigation strategies in the generic ERGS and ensure that any enhancements are identified. In addition, many enhancements that are not safety-significant deviations are identified even though they do not affect the mitigation strategies in the ERGS because: 1) plant-specific systems and equipment at McGuire do not exist in the generic ERG plant; 11) the configuration of the guideline set appears to be significantly changed, although the guidance remains the same. 'The former category is best described as additions, and the latter as human factors related changes. No attempt is made in this document to distinguish between additions and deviations. 4 Section II summarizes major design differences between McGuire and the ERG reference plant. These are presented as additions to the generic guidelines, and no bases for deviations are necessary. ERG guidance using systems and equipment that do not exict at McGuire are not a subject of this document.Section III providea an overview of the McGuire EPG configuration as compared to ERG-HP I Revision IA. Additions and deviations associated with the cverall l configuration of the guideline set are discussed.Section IV 2 j l _ - _ __ - -____ _____ _ __ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
d a
.i.
provides a description and justification for each addition or l safety-significant deviation not covered in Section III.
l.
For some historical perspective it is worth noting that an effort to l
document ERG /EPG safety-significant deviations was previously completed during.the Catawba Unit 1 original licensing time frame.
l l By letter dated June 18, 1984 Duke submitted a deviations document 1
i comparing the Catawba EPGs to ERG-HP Revision 1. By letter dated l
July 25, 1984 Duke submitted a deviations document comparing to ERG-HP BASIC. Subsequent correspondence and a meeting with the NRC staff on August 21, 1984 led to NRC approval of the deviations document in Supplements 4 & 6 to the Catawba Units 1 and 2 SER (NUREG-0954) dated December 1984.and May 1986. This constitutes NRC approval of the ERG /EPG deviations for Catawba. Due to the close-similarity between the Catawba and McGuire EPGs, at the time the Catawba deviations document was completed it was considered by Duke to be technically accurate for McGuire with only minor differences.
This McGuire document is similarly technically accurate for Catawba.
Due to the experience with the NRC review of the Catawba deviations document, in which only one deviation was withdrawn, it was decided by Duke at the time that a separate NRC review for McGuire was not technically useful. This document formally meets the McGuire-specific licensing requirements. It should be pointed out that the increase in the number of deviations in the McGuire document when-compared to the earlier Catawba document is due to a change in the definition of what constitutes a deviation.
3
e e II. PLANT SPECIFIC DESIGN DIFFERENCES The design of the McGuire Nuclear Station includes some features important to transient and accident mitigation and recovery which differ from the generic Westinghouse 4-loop plant which was utilized as the reference plant in the development of the generic ERGS. The major design differences are listed in this section and the utilization of these systems and equipment are discussed. Since the utilization in all cases is relatively straightforward and consistent with the plant licensing basis, or good operating practice, the additions to the generic ERGS associated with these design differences will only be discussed in this section.
Incorporation of these plant-specific design differences in the Catawba EPGs was approved by the NRC as stated in Supplement 4 to the Catawba SER, p. 13-8. In addition, two differences in the utilization of the guidelines are described.
Ice Condenser and Containment Systems McGuire utilizes the ice condenser containment design rather than the standard dry containment used as the reference design in the ERGS. All generic guideline steps related to dry containment systems have been replaced with steps for the corresponding ice condenser systems.
The ice condenser counterpart of the dry containment spray system is the combined capability of the upper containment spray system and the residual heat removal auxiliary containment spray system.
Instructions concerning these systems in the EpGs are basically to initiate or verify actuation on high containment pressure and termination on low containment pressure. Containment spray flow is also reduced to conserve ECCS suction inventory for beyond design basis situations, provided that the containment pressure remains acceptably low.
The ice condenser design includes the Containment Air Return and Hydrogen Skimmer Fan system which circulates the post-accident 4
e ,
containment atmosphere through the ice condenser'in order to remove energy. These fans also serve to mix the containment atmosphere in order to prevent stagnant pockets of hydrogen. Proper actuation and performance'of this system is' ensured in the EPGs.
The Annulus Ventilation System filters leakage from the containment into the annulus between the containment vessel and the Reactor Building, and discharges it to the unit vent. Proper actuation and performance of this system is verified in the EPGs.
The Emergency Hydrogen Mitigation System (igniters), and the aforementioned hydrogen skimmer fans are utilized and guidance is provided to the operator concerning optimum methods of post-inadequate core cooling hydrogen mitigation. These methods are consistent with the licensing basis and good operating practice.
Standby Shutdown Facility The Standby Shutdown Facility (SSF) is utilized in the EPGs as a source of reactor coolant pump seal injection following a loss of all AC power. This capability provides an extra level of redundancy with respect to maintaining reactor coolant pump seal integrity which is not available in the generic reference plant. This capability is utilized to bypass the cooldown and depressurization sequence in the loss of all AC power EPG, provided that the Reactor Coolant System inventory confirms that pump scal integrity has been maintained.
Reactor Vessel Level Indication System The Reactor Vessel Level Indication System (RVLIS) for McGuire differs slightly from the generic RVLIS of the reference plant.
These differences (UHI-RVLIS) are described in the ERG Executive volume. The capabilities of the McGuire design are, however, essentially identical to the reference design. One deviation from j the generic utilization of the RVLIS results from these design differences. The dynamic head range in the generic RVLIS spans the reactor vessel. In the McGuire RVLIS it spans the bottom of the 5
f s
reactor vessel to the hot leg. As a result, in the unlikely situation that a vessel head void exists with one or more reactor coolant pumps running, the McGuire dynamic head range cannot provide an indication. Using the McGuire upper range for this purpose is overly complex since the indication is dependent on which RVLIS channel (A or B) is being monitored and which combination of pumps is operating. Due to the unlikeliness of this situation and the complexity of using the RVLIS, it is not included in the EPGs. This aspect was reviewed by the NRC during Catawba licensing.
Guideline Utilization The utilization of the EPG Function Restoration Guidelines (FRGs),
as initiated by alarm conditions in the Critical Safety Function (CSF) Status Trees, differs in two aspects from the generically recommended utilization. The first difference is that when an ORANGE or RED alarm condition occurs, the operating crew is permitted to implement the appropriate FRG in parallel with the procedure.in effect. In order for this parallel implementation to be used, the mitigation actions of the procedure in effect must not conflict with the FRG guidance, and manpower must be available to accommodate any increase in operator burden. The second difference is that should an off-normal condition result in a CSF alarm during unit heatup or cooldown, the operator can use the applicable FRG for guidance in responding to the abnormal symptoms. This can occur without the prerequisite generic entry conditions of a reactor trip, safety injection, or station blackout. These two utilization differences greatly enhance the usefulness of the FRGs, while maintaining the intent of restoring critical safety functions.
6
1 III. GUIDELINE SET ADDITIONS AND DEVIATIONS The EPG set configuration is based on Revision 1 to the generic ERGS and includes additions, deletions, and restructuring (all termed
" deviation") intended to enhance the quality of the guidelines. The following discussion will demonstrate the similarity of the EPGs and ERGS and identify and justify deviations.
e Table 1 illustrates a comparison of the Optimal Recovery Guidelines (ORGs).
o Table 2 illustrates a comparison of the Emergency Contingency Action Guidelines (ECAs) for the EPGs and Revision 1 of the generic ERGS.
e Table 3 illustrates a comparison of the Critical Safety Fbnction Status Trees.
e Table 4 illustrates a comparison of the Punction Restoration Guidelines (FRGs).
A discussion of each deviation follows. Those which were identified in the Catawba deviations document and were subsequently reviewed and approved by the NRC (Refer to Section I) are labeled "NRC APPROVED".
Deviation 1: The generic E-0, Reactor Trip or Safety Injection, guidaline differs from the EPG E-0, Safety Injection, in that the EPG is not entered unless a safety injection occurs or is required.
If a reactor trip has occurred without a safety injection, the EPG I
ES-0.1, Reactor Trip Response, is the entry point. This avoids the <
nuisance procedure transfer from ERG E-O to ES-0.I in the ERGS following a reactor trip. The technical guidance is unaffected.
Deviation 2: The generic ES-0.0, Rediagnosis, guideline has been deleted in the EPGs. Based on the verification and validation activities it was determined that this guideline is not useful. It is apparent that some use for ES-0.0 might have been warranted j during the ERG development process due to the relative unfamiliarity 7
p .- .
of the operating crew involved in the generic ERG validation program. Increasing familiarity with the EPGs will enable the operators to understand the overall intent. In addition, the critical Safety Function Status Trees and the Function Restoration Guidelines will ensure that the plant is maintained in a safe condition should the operating crew become temporarily disoriented in the Optimal Recovery Guidelines due to the more unlikely multiple l
failure scenarios. (NRC APPROVED)
Deviation 3: The generic ES-0.2 and ES-0.3 guidelines have been merged to reduce unnecessary guideline duplication. The technical content has been preserved. (NRC APPROVED)
Deviation 4: The generic ES-0.4 guideline has been deleted in the EPGs since it is only required for those plants not equipped with a RVLIS system. (NRC APPROVED)
Deviation 5: The EPGs include an additional guideline ES-0.3, SI Termination Following Spurious SI, to respond to indications of a spurious and unnecessary actuation of safety injection equipment.
The recovery actions are designed to optimize the recovery from this specific event rather than any safety injection termination sequence. (NRC APPROVED)
Deviation 6: The generic E-1, Loss of Reactor or Secondary Coolant, has been somewhat modified in the EPGs. The EPG E-1, High Energy Line Break Inside Containment, differs from the generic guideline in that secondary breaks outside containment are covered in E-2, Steam Line Break Outside Containment (Refer to Deviation 7). This modification was implemented since the symptoms of a high energy line break inside containment are very explicit, and the verification of containment systems actuation and performance is
.only applicable for that type of transient. The generic approach to LOCA and steam line break mitigation has been preserved. The deviation serves to enhance operator response by addressing only applicable symptoms. (NRC APPROVED) l l
8 i
e o ;,
Deviation 7: 'The generic E-2, Faulted Steam Generator Isolation, has-been modified in the EPGs to cover only steam line breaks
outside containment. .The deviation is complementary with Deviation 6 discussed above. The EPG E-2, Steam Line Break outside Containment, preserves the generic approach'to steam line break mitigation. (NRC APPROVED)
Deviation 8: The EPGs include an additional guideline, ES-2.1, SI Termination Following Excessive Cooldown, which has been included to address the safety concerns associated with overcooling transients, (i.e., pressurized thermal shock, reactor vessel head voiding, etc.). The generic guidance for these concerns is included in the Function Restoration Guidelines. The basis for this additional guideline is to address these concerns during transient mitigation and recovery along the optimal recovery path, rather than relying solely on the Function Restoration Guidelines. This allows a preventive rather than a reactive approach to transient management.
The technical bases for ES-2.1 are consistent with the generic bases. (NRC APPROVED)
Deviation 9: The generic ECA-2.1, Uncontrolled Depressurization of All Steam Generators, has been mercad into EPG E-2, Steam Line Break Outside Containment, and other interfacing guidelines. The generic technical guidance has been preserved. (NRC APPROVED)
Deviation 10: The generic ECA-3.3, SGTR Without Pressurizer Pressure Control was deleted from the EPG set, as described on
- p. 119 of the ERG background document, based on a plant-specific PRA evaluation which determined that the probability of occurrence of the scenarios which this ERG is intended to mitigate is less than 1 x 10~ /yr. The extremely low probability is a valid basis for deleting it from the EPG set when considering that the ERGS were designed to mitigate sequences of probability greater than 1 x 10-8/yr. Refer to the ERG Executive volume, Introduction, p. 3.
I 9
i; e~ .*
Deviation J__1_: The generic Critical Safety Function (CSF) Status Tree F-0.1, Subcriticality, has been slightly modified in order to be compatible with the design basis of the Safety Parameter Display System (SPDS). The SPDS design requires that the Subcriticality CSF alarm indicate a GREEN condition (i.e., CSF satisfied) during normal operation. Since the generic F-0.1 is only intended to be applicable following reactor trip, the first question " REACTOR POWER
>5%" will generate'a RED condition during normal operation. In order to correct this incompatibility, an additional question
" REACTOR TRIP. REQUIRED" has been added, prior to checking the above question, in the EPG F-0.1. This modification preserves the intent of the generic F-0.1 and satisfies the SPDS design requirement. l (NRC APPROVED)
' Deviation 12: The generic CSF Status Tree F-0.4, Integrity, has been modified'to alarm on challenges to Reactor Coolant System (RCS) integrity due to_overpressurization considerations. This enhancement complements the existing alarms based on pressurized thermal shock and cold-overpressurization considerations. An ORANGE condition (i.e., a severe challenge to the CSF) is alarmed if RCS pressure exceeds 2400 psig (2250 psig if degraded containment environment instrument errors are in effect). The setpoint has been selected to indicate that a pressurization transient may challenge the pressurizer code safety valves, and thereby challenge the Integrity CSF. A new EPG, FR-P.3 Response to High Reactor Coolant System Pressure, provides the appropriate function recovery actions (Refer to Deviation 14). (NRC APPROVED) !
Deviation 13: The generic CSF Status Tree F-0.5, Containment, has been modified to explicitly monitor the containment hydrogen j J
concentration. If containment hydrogen concentration exceeds 0.5%, j an ORANGE condition is alarmed, and the operator is referred to the appropriate hydrogen mitigation guidance in FR-Z.1, Response to High j Containment Pressure. (NRC APPROVED) 10 1
1
Deviation 14: The EPGs include an additional guideline, FR-P.3, Response To High Reactor Coolant System Pressure, to enhance response to potential integrity challenges due to overpressurization. Entry conditions to FR-P.3 from the Integrity !
1 CSF Status Tree are discussed in Deviation 12. The objectives of FR-P.3 are to depressurize the RCS to terminate'the overpressure condition. Available systems are utilized in a manner consistent with design bases and good operating practice. (NRC APPROVED) i 11
, < TABLE 1 j' Comparison of EPGs and ERG Revision 1 Optimal Recovery Guidelines ERG Revision 1 EPGs E-0. Reactor Trip or Safety Injection (Deviation 1)
ES-0.0 'Rediagnosis (Deviation 2)
ES-0.1- . Reactor Trip Response (Deviation 1)
ES-0.2- Natural Circulation Cooldown (Deviation 3)
ES-0.3 Natural Circulation Cooldown with Steam (Deviation 3)
Void In Vessel (With RVLIS)
ES-0.4 Natural Circulation Cooldown with Steam (Deviation 4)
Void In Vessel (Without RVLIS)
[EPG ES-0.3, SI Termination Following (Deviation 5)
Spurious SI]
E-1 Loss of Reactor or Secondary Coolant (Deviation 6)
'ES-1.1 'SI Termination Same ES-1.2 Post-LOCA Cooldown and Depressurization Same
'ES-1.3 Transfer to Cold Leg Recirculation Same ES-1.4 Transfer to Hot Leg Recirculation Same E-2 Faulted Steam Generator Isolation (Deviation 7)
[EPG ES-2.1, SI Termination Following (Deviation 8)
Excessive Cooldown] ,,
I i
Same 2 E-3 Steam Generator Tube Rupture ES-3.1 Post-SGTR Cooldown Using Backfill Same as EPG ES-3.2 ES-3.2 Post-SGTR Cooldown Using Blowdown Same as EPG ES-3.3
]
ES-3.3 Post-SGTR Cooldown Using Steam Dump Same as EPG ES-3.1
'l l
l 12
,- y ,.,
.i b .
TABLE 2 Comparison of EPGs and ERG Revision 1 Emergency-Contingency Action Guidelines ERG Revision 1 EPGs ECA-0.0' Loss of All AC Power Same ECA-0.1- Loss of All AC Power Recovery Without Same
.SI Required ECA-0.2 Loss of All AC Power Recovery With SI Required Same ECA-1.1 ' Loss of Emergency Coolant Recirculation Same ECA-1.2 =LOCA Outside Containment' Same ECA-2.1 Uncontrolled Depressurization (Deviation 9) of All Steam Generators ECA-3.1 .SGTR With Loss of Reactor Coolant - Same Subcooled Recovery Desired ECA-3.2 SGTR With Loss of Reactor Coolant - Same Saturated Recovery Desired ECA-3.3 SGTR Without Pressurizer Pressure control (Deviation 10) l' t.
13
_ _ __ _ ._ _ _ _ _ _ _ _________________ ____ _ _- _-__ _ _ ___ - - __ _ n
oc- . .,
f
' TABLE 3 Comparison of EPGs.and ERG Revision 1 I Critical Safety Function Status Trees i
ERG Revision 1 EPGs F-0.1 'Subcriticality (Deviation 11)
F-0.2 Core Cooling Same F-0.3 11 eat Sink Same
, , F- O '. 4 Integrity (Deviation 12)
-t F-0.5 Containment (Deviation 13) i-F-0.6- Inventory Same 14 l
L- __ _ - _ _ _
r . , u ,.
j l
TABLE 4 1 I
comparison of EPGs and ERG Revision 1 Ftnction Restoration Guidelines ERG Revision 1 EPGs FR-S.1 ' Response To Nuclear Power Generation /ATWS Same FR-S.2 -Response To Loss of Core Shutdown Same FR-C.1 Response To Inadequate Core Cooling Same i FR-C.2 Response To Degraded Core Cooling Same FR-C.3 Response To Saturated Core Cooling Same FR-H.1~ Response To Loss of Secondary Heat Sink Same FR-H.2 Response To Steam Generator Overpressure Same j FR-H.3 Response To Steam Generator High Level Same FR-H.4 Response To Loss of Normal Steam Release Same Capabilities FR-H.S Response To Steam Generator Low Level Same FR-P.'1 Response To Imminent Pressurized Thermal Same Shock Condition FR-P.2 Response To Anticipated Pressurized Thermal Same Shock Condition
[EPG FR-P.3, Response to High NC (RCS) System (Deviation 14)
Pressure]
FR-Z.1 Response To High Containment Pressure Same FR-Z.2 Response To Containment Flooding Same j FR-Z 3 Response To High Containment Radiation Level Same FR-I.1 Response To High Pressurizer Level Same FR-I.2 Response To Low Pressurizer Level Same {
1 t FR-I.3 Response To Voids In Reactor Vessel Same l 1
15 4
l
.. ___ -- - __ _____ - _ _ _ _ _ _ b
p-6 .
IV. GUIDELINE TECHNICAL ADDITIONS AND DEVIATIONS l This'section identifies and justifies safety-significant additions,-
deletions, and restructuring (all termed " deviation") between the generic ERG Revision 1A and the McGuire EPGs. Deviations due to plant-specific design differences and guideline set configuration differences are-detailed in Sections II and III, respectively, and t
are not included in the following summary. The items are listed in a sequence consistent with the guideline set as given in Tables 1-4.
Those which'were identified in the Catawba deviations document and were subsequently reviewed and approved by the NRC (Refer to Section I) are labeled "NRC APPROVED".
Deviation 15: In the EPG E-0,-Safety Injection, the main steam isolation valves are closed if the turbine fails to trip and cannot be manually tripped. This provides another method of terminating the overcooling that would occur. This action is not in the generic ERG.
Deviation 1 6: In the generic E-0, Reactor Trip Or Safety Injection, t guidance exists to restore the second emergency AC bus. This is not included in the EPG since the licensing basis only requires one train of safeguards, and it is not desirable to burden the operations staff with unnecessary actions at this early stage of the mitigation phase when time management is critical. Based on operator training the recovery of the second emergency AC bus is recognized as an important action that is to be undertaken as time and manpower become available.
Deviation 17: In the generic E-0, Reactor Trip Or Safety Injection, guidance for verifying SI flow is given in Step 15. This guidance has been moved forward to Step 7 in the EPG based on the importance and timeliness of this critical action.
16 l
6
- L Deviation 18: In the generic E-0, Reactor Trip or Safety Injection, guidance for tripping the reactor coolant pumps is given in Step 21.
This guidance has been moved forward to Step 8 in the EPG based on the importance and timeliness of this critical action.
Deviation 19: In the generic E-0, Reactor Trip Or Safety Injection, the sequence of transfers to other ERGS are based on diagnosis of:
- 1) faulted steam generators (SLB); 2) SGTR; and 3) LOCA. In the EpG E-0 the sequence of transfers to other EPGs are based on: 1) high energy line break in containment; 2) SLB outside containment; and 3)
SGTR. The EPG sequence' essentially accomplishes the same mitigation actions as the generic ERGS. However, the guideline wherein the specific steps are included may, in some cases, be different. (NRC APPROVED)
Deviation 20: In the generic E-0, Reactor Trip or Safety Injection, the SI termination criteria include an error-adjusted subcooled margin greater than zero. In the EPG this value is 50 F. This setpoint is preferred since this SI termination check is mainly applicable to spurious safety injection events. For these scenarios there is no incentive to keep Reactor Coolant System pressure very near saturation. A margin of 50 F provides two benefits. The first is some operating flexibility such that a small reduction in pressure will not result in losing subcooling and requiring a re-actuation of safety injection. The second benefit is the additional time available to the operator to observe the status of the unit, perform other actions, and foremost, to ensure that SI termination is appropriate. The only potential negative aspect would be the delay associated with waiting for the subcooled margin to increase from zero to 50 F. The consequences of this delay were evaluated and no negative aspects were identified. (NRC APPROVED)
Deviation 21: In the generic E-0, Reactor Trip Or Safety Iniection, Steps 32-36 are performed while symptoms e.re being re-checked since the event and appropriate transfer out of E-0 was not made on the first pass through the guideline. None of these steps include 17
h .
important actions based on the McGuire design, and none were included in the EPG. Therefore, the EPG focuses on making the determination of the correct transfer out of E-0 so that mitigation and recovery can continue. The operator is not distracted by procedure steps which require time and resources during a critical time frame.
Deviation 22: On the foldout for the generic E-0, Reactor Trip Or Safety Injection, is Item 3, Red Path Summary. The usefulness of this item has not been demonstrated and it has been deleted throughout the EPGs. A member of the control room operating crew will be continuously monitoring the CSF status trees under any abnormal condition. (NRC APPROVED)
Deviation 23: On the foldout for the EPG E-0, Safety Injection, and on many other EPG foldouts, a new item has been added. This item provides guidance for isolating the centrifugal charging pump (NV) miniflow line in order to optimize safety injection flow.
Reestablishing miniflow to prevent deadheading the pump is also j given. This item was judged to be very important for accident mitigation, and due to it being important at any time the charging pumps are in operation, it was best included on the foldout. This item is also an NRC commitment.
Deviation 24: On the foldout for EPG E-0, Safety Injection, and on AA many other EPG foldouts,-c:new item stating the generic characteristics of natural circulation has been added. This information was judged to be a valuable reminder that natural circulation should be closely monitored. Due to this information being broadly applicable and independent of the timing of events, it was best included on the foldout. This is sporadically included in the generic ERGS as " Attachment A."
Deviation 25: In the generic ES-0.1, Reactor Trip Response, Step 7 verifies that offsite power is available. This step is deleted throughout the EPGs since it is standard operating practice that 18
, 5 :.
l-l whenever offsite power has been interrupted, the operators will-l verify loading of the diesel generators and then initiate action to f- restore offsite power as time and resources become available. (NRC APPROVED)
Deviation 26: In the EPG ES-0.2, Natural Circulation Cooldown, the h cold shutdown boron concentration is increased to offset incomplete
!)
mixing in the Reactor Coolant System due to relatively stagnant conditions in the reactor vessel head and pressurizer. The upper head allowance is not in the generic ERG.
Deviation 27: In the EPG ES-0.2, Natural Circulation Cooldown, FR-I.3, Response To Void In Reactor Vessel, is referred to for mitigating a reactor vessel head void. This permits the reactor vessel head venting option to be considered rather than only the reactor coolant pump restart option permitted by the ERGS. RCS depressurization is continued while the venting operation is initiated.
Deviation 28: In the generic E-1, Loss Of Reactor Or Secondary Coolant, Step 2 checks if the reactor coolant pumps should be stopped. In the EPG this was moved to the foldout since it is
. applicable whenever the trip criterion is met.
Deviation 29: In the EPG E-1, High Energy Line Break Inside Containment, Step 2 verifies SI flow. This step was added since numerous transfers into E-1 are a result of degrading primary inventory, and SI re-actuation, if necessary, is of critical importance.
Deviation 30: In the EPG E-1, High Energy Line Break Inside Containment, Step 4 verifies that the equipment required to establish sump recirculation is available. In the generic ERG this does not occur until Step 12. The intent of moving this step forward in the EPG is to enable more time to restore recirculation 19
- n. 1 i.'
capability since it is critical to be able to perform that transition.
Deviation 31: In the EPG E-1, High Energy Line Break Inside Containment, Step 5 checks if a large break LOCA h1 occurred as indicated by low-head safety injection flow into the Reactor Coolant System. This can only occur if primary pressure is less than 200 psig. If this condition exists, then the next critical action is to transfer to the sump recirculation mode. Therefore, Steps 6-13 in the EPG (Steps 3-13 in the ERG) can be skipped since they are not applicable to a large break LOCA. This enhancement substantially streamlines the guidance. (NRC APPROVED)
Deviation 32: In the generic E-1, Loss Of Reactor Or Secondary Coolant, a combined loss of reactor or secondary coolant and a steam generator tube rupture result in a transfer to E-3 in Step 4. In the EPG this transfer is made only if the SI termination criteria are also met. This will result in the plant-specific guidelines handling the tube rupture aspects of this multiple transient within the LOCA guidelines (E-1 series) whereas the generic guidelines handle the LOCA aspects within the SGTR guidelines (E-3 series). In both situations the mitigation strategy is the same. The EPG approach is preferred since the LOCA aspects of this multiple transient are most important with respect to core cooling and the tube rupture aspects can be easily integrated into the LOCA mitigation strategy.
In the EpG E-1, High Energy Line Break Inside I Deviation 33:
containment, Step 11 checks for a feedwater line break as the cause of the abnormal containment conditions. This step is intended to assist in the diagnosis of the event, and provides confirmation to the operators.that they are in the appropriate procedure. This step ,
1 is not in the generic ERG. l l
l Deviation 34: In the generic E-1, Loss Of Reactor Or Secondary Coolant, Step 15 isolates the SI accumulators to prevent nitrogen I
20 l
I i.8 .
l 1q D
i injection fe: a large bioak LOCA. This action is not in the EPGs since there is no concern with nitrogen being injected into the Reactor Coolant System following large break LOCAs. In fact, the 1 FSAR Is0CA. analyses account for nitrogen injection. This generic step is a distraction.
Deviation 35: On the' foldout for the generic E-1, Loss Of Reactor Or Secondary Coolant, Items 3 (Secondary Integrity Criteria) and 4 l (E-3 Transition Criteria) were not included on the EPG due to insignificant perceived benefit.- Items 2 (SI Termination Criteria)
~
and 5 (Loss of Emergency Coolant Recirculation) were added to the EPG E-1 foldout due to the critical nature of these two ,
i considerations.
Deviation 36: On the foldout for EPG ES-1.1, SI Termination Following High Energy Line Break Inside Containment, a reactor coolant pump. trip is required if the loss of subcooling trip criterion occurs and subcooling cannot be promptly restored. This foldout item is applicable during cooldown and depressurization sequences, and is intended to prevent pump damage and inventory depletion. (NRC APPROVED)
Deviation 37: In the generic ES-1.2, Post-LOCA Cooldown and Depressurization, a caution exists before Step 1 regarding transfer to cold leg recirculation. This caution does not exist in the EPG since this item is already on the foldout.
Deviation 38: In the EPG ES-1.2, post-LOCA Cooldown and Depressurization, the RVLIS reactor vessel head level is checked to see if a void exists prior to starting a reactor coolant pump. If a void exists then pressurizer level is increased to 50% to accommodate the removal of the void upon pump restart. This guidance is not in the generic ERG.
Deviation 39: In the EPG ES-1.1, Post-LOCA Cooldown and Depressurization, charging flow is throttled open and closed in 21
W ,'
[L * :, j p-l-
L order to smooth the pressure transient resulting from stopping safety injection pumps. This method is recognized in the ERG documentation as technically superior.
Deviation 40: In the EPG ES-1.2, Post-LOCA Cooldown and Deprescurization, steam generators are steamed if they exceed the primary' system pressure to prevent the steam generators from being a heat source, and to prevent dilution of tue primary boron concentration via backfilling if a ruptured tube exists. This action goes beyond the steaming that is initiated for cooldown purposes.
Deviation 41: In the EPG ES-1.3, Transfer to Cold Leg Recirculation, the containment sump level indication is checked to verify that sufficient sump inventory has accumulated to support the suction requirements of the SI pumps. This is not in the generic ERG.
Deviation 42: In the EPG P3-1.3, Transfer to Cold Leg Recirculation, the possibility that an RHR train has previously been aligned for normal RHR' mode is considered. This is not in the generic ERG.
Deviation 43: In the EPG ES-1,3, Transfer to Cold Leg Recirculation, following transfer from the injection mode to the recirculation mode, the operators are instructed to make up to the refueling water storage tank. This will establish a redundant ECCS suction inventory should sump recirculation be interrupted. (NRC APPROVED)
Deviation 44: In the EPG ES-2.1, SI Termination Following Excessive cooldown, the nuclear flux indication is checked to see if additional boration is needed to ensure that the reactor stays subcritical, prior to terminating safety injection. This is not in the generic ERGS.
22
. lt .
m Deviation 45: In the generic E-3, Steam Generator Tube Rupture, Step 1 checks if the reactor coolant pumps should be stopped. In the EPG this was moved to the foldout since it is applicable whenever the trip criterion is met.
Deviation 46: In the. generic E-3, Steam Generator Tube Rupture, Step 3 verifles that the steam generator blowdown is isolated. This guidance is not in the EPG since it has already been performed in E-0 following actuation of safety injection.
Deviation 47: In the generic E-3, Steam Generator Tube Rupture, the status of the reactor coolant pumps is not checked until Step 36.
The EPGs try to obtain two pumps operating in Step 7. The EPG approach is based on the desirability of establishing forced circulation, the normal and familiar operating mode, in parallel with initiating the RCS cooldown. (NRC APPROVED)
Deviation 48: In the generic E-3, Steam Generator Tube Rupture, Steps 8-12 are of low importance compared with proceeding with the cooldown. Since operator response time is at :a premium at this stage of the mitigation actions, these steps are not included in the EPG until later.
Deviation 49: In the generic E-3, Steam Generator Tube Rupture, the caution before Step 15 has been deleted in the EPGs. This caution unnecessarily delays the equalization of primary and ruptured steam generator pressures, and therefore contributes to greater tube leakage. Simultaneous cooldown and depressurization are emphasized in operator training.
Deviation 50: In the EPG E-3, Steam Generator Tube Rupture, a caution is included regarding which post-SGTR cooldown method to select. The backfill method is not recommended unless at least ene reactor coolant pump is available to provide boron mixing.
23 I
.---- -_-__- __. _ - - _ \
Di ? ,' . ) J 1 %
et>
Deviation 51: In the EPG ES-3.1, Post-SGTR Cooldown and Depressurization (the corresponding ERG is ES-3.3), the natural circulation cooldown rate of 50 F/hr is used if no reactor coolant pungs are in operation. The generic ERG cooldown rate is 100 F/hr.
The.EPG' guidance is preferred since it minimizes the potential for developing a reactor vessel head void. If a void develops, the EPG recommends discontinuing the depressurization until the void can be vented. The generic ERG mentions that a void may occur but provides no guidance.
Deviation 52: In the EPG ES-3.1, Post-SGTR Cooldown and Depressurization 'the
( corresponding ERG is ES-3.3), the ruptured steam generator and the primary are depressurized simultaneously while the cooldown is in progress. In the generic ERG the depressurization is initiated only after reaching 350 F, in order to simplify the operator. response. The EPG approach is preferred since
.it enables approaching the desired conditions sooner. The EPG approach is not~ considered overly complex by the McGuire operators.
i i
Deviation 53: In the EPG ES-3.2, SGTR Alternate Cooldown Using l,- Backfill (the corresponding ERG is ES-3.2), the cold shutdown boron l concentration is increased to compensate for boron dilution due to backfilling. This guidance does not exist in the generic ERG.
l l
Deviation 54: Same as Deviation 51 for EPG ES-3.2, Post-SGTR Alternate Cooldown Using Backfill (the ERG is ES-3.1).
I Deviation 55: Same as Deviation 51 for EPG ES-3.3, Post-SGTR Alternate Cooldown Using Blowdown (the ERG is ES-3.2).
l Deviation 56: Same as Deviation 52 for EPG ES-3.3, Post-SGTR I
Alternate Cooldown Using Blowdown (the ERG is ES-3.2).
i Deviation 57: In the EPG ECA-0, Loss of All AC Power, the target temperature for the Reactor Coolant System cooldown is also based on a cycle-specific calculation of the temperature at which reactoi-i 24 ]
L_______ _ _ - !
..:4. .'
L
.1 recriticality would occur. This is mentioned in the generic ERG l l'
documentation but is not included in the setpoint determination.
Deviation 58: In the EPG ECA-0, Loss of All AC Power, steam
. generator depressurization is constrained by ensuring that sufficient' steam pressure is maintained to power the turbine-driven auxiliary feedwater pump. This is not in the generic ERG.
Deviation 59: In the EPG ECA-1 1, Loss of Emergency Coolant Recirculation; Step 2' attempts to reestablish the injection mode if the recirculation mode is~1ost. This is intended to provide some emergency coolant while the recirculation mode is recovered. This option is not in the generic ERG.
Deviation 60: In the generic ERG ECA-1.1, Loss of Emergency Coolant Recirculation, Step 11 attempts to maintain one reactor coolant pump in operation. The EPG stops all reactor coolant pumps in Step 4.
The EPG method is preferred since the inventory depletion that can result following a SBLOCA with continued pump operation will be avoided. Since inventory conservation is critical, due to the potential for safety injection to be lost for some period of time, the pumps-off approach is better. Also, operation of the pumps under saturated conditions is an extreme action which should be used only during inadequate core cooling mitigation.
Deviation 61: In the generic ERG ECA-1.1, Loss of Emergency Coolant Recirculation, safety injection flow and RCS inventory are controlled in the following ways. Step 9 establishes one train of safety injection to minimize suction inventory depletion. In Step 12 safety injection is terminated if the core exit is subcooled and RVLIS indicates the core is covered (pumps off) or that the RCS void fraction is less than 25% (1 pump on). If the safety injection termination criteria are not met, then the safety injection flow is reduced to match decay heat. In the EPG, if the available suction inventory reaches a critically low value, or if the RCS is saturated, then the safety injection flow is reduced to match decay 25
q-t ,
L L
i
- l. '
heat Otherwise, safoty injection flow is reduced to maintain RCS j inventory stable as indicated by RVLIS and pressurizer level. SI l flow is not terminated in the EPGs, since it is likely to require reinitiation.
l Deviation 62: In the EPG ECA-1.1, Loss of Emergency Coolant [
Recirculation, Step 7 verifios that an adequate steam generator heat ;
sink is being maintained. Since the steam generators are a part of l the mitigation strategy, in particular to support the reflux boiling mode of heat transfer, this step is needed. This step is not in the generic ERG.
Deviation 63: In the EPG ECA-1.1, Loss of Emergency Coolant Recirculation, the scenario where the capability for high pressure recirculation has been lost has been specifically addressed. The
. mitigation strategy is to pump containment sump water back to the refueling water storage tank using an RHR pump, and then realign the safety injection pumps for the injection mode. An obvious constraint on this method is the potential dose consequences, which is noted in the guidance. This is not in the generic ERG.
Deviation 64: In the EPG ECA-1.2, LOCA Outside Containment, Step 3 !
blocks the automatic switchover of the RHR pump suction from the refueling water storage tank to the containment sump if the minimum ,
sump level for recirculation is unavailable. This is not in the generic ERG. !
l Deviation 65: In the generic ERG ECA-3.1, SGTR With loss of Reactor l Coolant - Subcooled Recovery Desired, transition to ECA-3.2 occurs $
when the available ECCS suction inventory is low or if the ruptured steam generator level is offscale high. In the EPG, the first transition criteria is used, but the second is only used if the RCS pressure is high enough that steam generator overfill can lead to lifting the main stcam code safety valves. A transition to ECA-3.2 is alc0 made in the EPGs if the RCS is saturated.
26
V: ., it ,
4 l'
Deviation 66: In the EPG ECA-3.1, SGTR With Continuous NC System Leakage: Subcooled Recovery, the scenario where a steam leak inside containment on a ruptured steam generator is diagnosed, the event is recognized as a LOCA and a transfer to E-1 High Energy Line Break Inside Containment, is made. The scenario is then mitigated as a LOCA. This guidance is not in the generic ERG.
Deviation 67: In the EPG ECA-3,1, SGTR With Continuous NC System Leakage: Subcooled Recovery, if no reactor coolant pumps are in operation, then the ruptured steam generator (s) is steamed to the condenser in order to avoid a stagnant hot primary loop (s). A stagnant loop would eventually flash and inhibit depressurization.
Steaming is only permitted until the RHR mode can be established.
The steam generator may not be steamed if the level is offscale high. This guidance is not in the generic ERG.
Deviation'68: In the generic ERG ECA-3.1, SGTR With Loss of Reactor Coolant-Subcooled Recovery Desired, Step 24 requires RCS depressurization to maintain subcooling less than 10 F. In the EPG, Step 7 requires maintaining subcooling less than 50 F. The 50 F setpoint is preferred since controlling pressure within a 0-10 F band is operationally very restrictive, and very difficult with a PORV providing pressure control. Furthermore, the benefit of the lower pressure resulting from the tighter subcooling limit does not offset the operational burden.
Deviation 69: In the generic ECA-3.1, SGTR With Loss of Reactor Coolant - Subcooled Recovery Desired, Step 5 terminates containment spray. This step does not exist in the EPG since the scenario of concern is a SGTR plus LOCA, and this scenario is not mitigated in ECA-3.1 in the EPGs (Refer to Deviation 32).
Deviation 70: In the generic ECA-3.1, SGTR With Loss of Reactor Coolant - Subcooled Recovery Desired, the SI accumulators are isolated prior to injecting if the RCS is subcooled. If the RCS is not subcooled they are allowed to inject and then isolated prior to 27
- ..a<.
nitrogen injection. In the EPG the accumulators are isolated below 1000 psig as in a normal unit cooldown provided that the SI termination criteria are met. If the RCS is not subcooled then a transition to ECA-3.2 is made.
Deviation 71: In the EPG ECA-3.2, SGTR With Continuous NC System Leakage: Saturated Recovery, Step 1 calls for a rapid cooldown of the RCS to 500 F. The purpose of this action is to enable depressurizing the primary to a low enough pressure so that the main steam line code safety valves will not be challenged. This guddance is not in the generic ERG.
Deviation 72: In the generic ERG ECA-3.2, SGTR With Loss of Reactor Coolant - Saturated Recovery Desired, Step 18 depressurizes the RCS to minLmize subcooling. This is done in Step 6 of the EPG in order to minimize inventory loss earlier.
Deviation 73: In the EPG ECA-3.2, SGTR With Continuous NC System Leakage: Saturated Recovery, the SI accumulators are isolated below 1000 psig as in a normal unit cooldown, provided that the SI termination criteria are met.' It is not conceivable for the events mitigated in this EPG that these criteria will not be met prior to nitrogen injection from the accumulators.
Deviation 74: In the EPG Critical Safety Function Status Tree F.O.2, Core Cooling, the RVLIS setpoints have been modified. The pumps off setpoint (ERG setpoint 2) has been increased from a collapsed liquid level of 3 1/2 feet to 6 feet above the bottom of the core. The pumps on setpoints (ERG setpoints 3-6) have been increased in terms of inventory from 50% to 25% void fraction.
These setpoints are preferred in order to provide additional margin to inadequate core cooling and to allow additional response time for operator action.
(NRC APPROVED) 1
)
28
I D3viation 75: In the EPG Critical Safety Function Status Tree F-0.4, Integrity, temperature setpoints have been conservatively increased to include an allowance for the reactor vessel downcomer temperature being colder tnan the indicated cold leg temperature, and to provide an earlier indication of an overcooling event in progress. (NRC APPROVED)
Deviation 76: In the generic ERG FR-C.1, Response to Inadequate Core Cooling, depressurization of the RCS by opening pressurizer PORVs is not undertaken unless core exit thermocouple temperatures are greater than 1200 F. The EPGs implement this action if core exit thermocouple temperatures are greater than 700 F and increasing. This enhancement avoids waiting for the core to heat up if the mitigation actions already performed have not been successful. (NRC APPROVED)
Deviation 77: In the EPG FR-C.3, Response To Saturated Core Conditions, the possibility that a failed open pressurizer spray valve is the cause is checked. The valve is closed or the reactor coolant pumps supplying spray are tripped. This is not in the generic ERG.
Deviation 78: In the EPG FR-H.1, Response to Loss of Secondary Heat Sink, the need to initiate feed-and-bleed is checked in Step 2, whereas in the generic ERG this occurs in Step 10. This action is considered sooner in the EPG due to its urgency. In addition, the feed-and-bleed initiation criteria are slightly modified. The recommended generic criterion of low steam generator level in N-1 loops is modified by also requiring RCS temperatures to be increasing. Without RCS tem.peratures increasing, there is no loss of heat sink and initiating feed and bleed would be too drastic.
Feed and bleed is also immediately initiated if the RCS is saturated due to a loss of heat sink, regardless of steam generator conditions.
l 29
Deviation 79: In the EPG FR-H.1, Response to Loss of Secondary Heat Sink, one reactor coolant pump is left operating (or restarted) to promote primary-to-secondary heat transfer, provided that one steam generator has at least 15% indicated wide range level. The pump is operated in the loop with the highest steam generator level. This action is intended to obtain the benefits of forced circulation-- l specifically, heat transfer, thermal mixing, and pressurizer spray capability are assured. All pumps are tripped prior to initiating feed and bloed per the generic guidance. (NRC APPROVED)
Deviation 80: In the EPG, FR-il.1, Response to Loss of Secondary Heat, following initiation of feed and bleed cooling, RCS
' temperature is monitored to assess the success of core cooling in this mode. The need for additional or less feed and/or bleed capacity is determined. This is not in the generic ERG.
(NRC APPROVED)
Deviation 81: In the EPG FR-P.1, Response to Imminent Pressurized Thermal Shock Conditions, Step 2 checks if the RCS is subcooled. If the RCS is not subcooled then most of the guidance associated with terminating SI and depressurizing the RCS is not applicable and it is skipped. This substantially streamlines use of the guidance.
This is not in the generic ERG.
Deviation 82: In the generic FR-P.1, Response to Imminent Pressurized Thermal Shock Condition, Step 3 closes any pressurizer PORVs that are open below the normal actuation setpoint. RCS depressurization is subsequently performed in Step 15. In the EPGs the RCS is depressurized beginning in Step 5, and the PORVs are closed in Step 7 only if they are not in use for RCS depressurization. The generic guidance can result in pressurizing j the RCS, which is undesired. The EPr ,tidance focuses on depressurizing much earlier. The target subcooling margin in the ,
ERG is 10 F, whereas in the EPG it is 100 F. The 10 F target is operationally very restrictive, in particular if the PORV is in use.
RCS voiding is more likely. The EPG setpoint remains conservative 30
~
e + ,
with respect to the allowable subcooling of 200 F during the soak period.
Deviation 83: In the generic FR-P.1, Response to Imminent Pressurized Thermal Shock Condition, the SI termination criteria assume that all reactor coolant pumps are off, which is not necessarily the case. The generic RVLIS setpoint of top of the core is considered very extreme, and difficult to comprehend in terms of the consequences of SI termination under these conditions. The EPG employs standard SI termination criteria. When considering the 200 F allowable subcooling during the soak period, the generic SI termination criteria are unwarranted.
Deviation 84: In the EPG FR-P.1, Response to Imminent Pressurized Thermal Shock Conditions, the desirability of maintaining or reestablishing forced circulation for thermal mixing is addressed.
This is not included in the generic ERG unless the subcooled margin is less than 50 F.
Devjation 85: In the generic FR-I.3, Response to Voids in Reactor vessel, the guideline is not used if safety injection is in operation. This restriction is not included in the EPGs. The use of this guideline in the EPGs is structured in such a way that it supplements the mitigation strategies in progress,- time and resources permitting.
Deviation 86: In the EPG FR-I.3, Response to Void in Reactor vessel, the void mitigation approach that utilizes repressurization is not permitted if a LOCA, SGTR, or PTS condition exists.
Repressurizing under such circumstances goes against the overall mitigation strategy. This is not in the generic ERG.
Deviation 87: In the EPG FR-I.3, Response to Void in Reactor Vessel, the measures employed to control venting of voids in order to manage the release of hydrogen into containment are not used under specific conditions. Since hydrogen can only be generated 31
-)
- > _. o ; # 1, :
l following a LOCA, controlled venting is only implemented if the pressurizer relief tank has ruptured or if containment pressure has exceeded the high-high setpoint of 3 psig. These two criteria identify that a LOCA has occurred. This is not in the generic ERG.
(NRC APPROVED) l W
32 j